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June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 3.0 -Design of Structures, Systems and Components Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:   
June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 3.0 -Design of Structures, Systems and Components Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:   
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* NOATHWEST llEDICAl lSOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components This page intentionally left blank.
* NOATHWEST llEDICAl lSOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components This page intentionally left blank.
::.**.*.*.* ; .. ;* .. NWMI ........... NOITHWUTMEDICAllSOTOHS Rev Date 0 6/29/2015 1 6/26/2017 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components REVISIO N HISTOR Y Reason for Revision Revised By Initial Application Not required Incorporate changes based on responses to C. Haass NRC Requests for Additional Information   
::.**.*.*.* ; .. ;* .. NWMI ........... NOITHWUTMEDICAllSOTOHS Rev Date 0 6/29/2015 1 6/26/2017 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components REVISIO N HISTOR Y Reason for Revision Revised By Initial Application Not required Incorporate changes based on responses to C. Haass NRC Requests for Additional Information   
.... ; .. NWMI ::.**.*.*.* .*.******** . * .. ! . llOIT1tWUT lllllOICAl tsOTOf'U NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Th is page i n ten t iona ll y l eft bl ank.   
.... ; .. NWMI ::.**.*.*.* .*.******** . * .. ! . llOIT1tWUT lllllOICAl tsOTOf'U NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Th is page i n ten t iona ll y l eft bl ank.   
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* 10 CFR 20 , "Standards for Protection Against Radiation" 10 CFR 30 , "Rules of General Applicability to Domestic Licensing of Byproduct Material" 10 CFR 50 , "Domestic Licensing of Production and Utilization Facilities" 10 CFR 70 , "Domestic Licensing of Special Nuclear Material" 10 CFR 71 , "Energy: Packaging and Transportation of Radioactive Material" 10 CFR 73 , "Physical Protection of Plants and Materials" I 0 CFR 74 , "Material Control and Accounting of Special Nuclear Material" 10 CFR 851 , "Worker Safety and Health Program" 21 CFR 210 , " Current Good Manufacturing Practice in Manufacturing , Processing, Packaging , or Holding of Drug s' 21 CFR 211 , " Current Good Manufacturing Practice for Finished Pharmaceuticals" 29 CFR 1910 , " Occupational Safety and Health Standards" 40 CFR 61 , "National Emissions Standards for Hazardous Air Pollutants (NESHAP)" 40 CFR 63 , "NESHAP for Source Categories" 40 CFR 141, " National Primary Drinking Water Regulations" 3.1.3 U.S. Nuclear Regulatory Commission Table 3-3 lists the NRC design inputs for the RPF identified in NWMI-DRD-2013-030. The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference. Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) CF Ra Docket Number: NRC-2011-0135 (NRC, 2012) Title Final Int e rim Staff Guidanc e Au g m e nting NU REG-153 7, " Guid e lin e s for Pr e parin g and R e vi e wing Appli c ations for th e Li ce n s ing of N on-Pow e r R e a c t o r s," Part s 1 and 2, for Li ce n s ing Radioi s otop e Produ c tion Fa c iliti es and Aqu e ous Homog e n e ous R e a c tor s NRC Regulatory Guides -Power Reactors (Division  
* 10 CFR 20 , "Standards for Protection Against Radiation" 10 CFR 30 , "Rules of General Applicability to Domestic Licensing of Byproduct Material" 10 CFR 50 , "Domestic Licensing of Production and Utilization Facilities" 10 CFR 70 , "Domestic Licensing of Special Nuclear Material" 10 CFR 71 , "Energy: Packaging and Transportation of Radioactive Material" 10 CFR 73 , "Physical Protection of Plants and Materials" I 0 CFR 74 , "Material Control and Accounting of Special Nuclear Material" 10 CFR 851 , "Worker Safety and Health Program" 21 CFR 210 , " Current Good Manufacturing Practice in Manufacturing , Processing, Packaging , or Holding of Drug s' 21 CFR 211 , " Current Good Manufacturing Practice for Finished Pharmaceuticals" 29 CFR 1910 , " Occupational Safety and Health Standards" 40 CFR 61 , "National Emissions Standards for Hazardous Air Pollutants (NESHAP)" 40 CFR 63 , "NESHAP for Source Categories" 40 CFR 141, " National Primary Drinking Water Regulations" 3.1.3 U.S. Nuclear Regulatory Commission Table 3-3 lists the NRC design inputs for the RPF identified in NWMI-DRD-2013-030. The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference. Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) CF Ra Docket Number: NRC-2011-0135 (NRC, 2012) Title Final Int e rim Staff Guidanc e Au g m e nting NU REG-153 7, " Guid e lin e s for Pr e parin g and R e vi e wing Appli c ations for th e Li ce n s ing of N on-Pow e r R e a c t o r s," Part s 1 and 2, for Li ce n s ing Radioi s otop e Produ c tion Fa c iliti es and Aqu e ous Homog e n e ous R e a c tor s NRC Regulatory Guides -Power Reactors (Division
: 1) Regulatory Guide I .53 A ppli c ation of th e Singl e-Failur e Crit e rion to Saf ety S ys t e m s , 2003 (R201 I) Regulatory Guide 1.60 D e sign Response Sp e ctra for S e ismi c Design of Nuclear Pow e r Plants, 2014 Regulatory Guide 1.76 D es i g n Ba s i s Tornado and Tornado Mis s il es fo r N uclear P owe r Plants , 2 007 Regulatory Guide 1.97 Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, 2006 (R2013) Regulatory Guide I. I 00 S e ismi c Qualifi c ation of El ec tri c al and A c tiv e M e chani c al E quipm e nt and Fun c tional Qualifi c ation of A c tiv e M ec hani c al Equipm e nt for N ucl e ar P owe r Plant s, 2009 Regulatory Guide 1.152 Criteria for Use of Computers in Safety Systems of Nuclear Power Plants , 201 l Regulatory Guide 1.166 Pr e-Earthquak e Planning and Imm e diat e Nuclear Pow e r Plant Op e rator P os t E arthquak e A c tions , 1997 Regulatory Guide 1.167 Restart of a Nuclear Power Plant Shut down by a Seismic Event, 1997 Regulatory Guide I .208 P e rforman ce Ba se d Appr o a c h t o D e fin e th e Sit e-Sp ec ifi c Earthquak e Ground Motion , 2007 NRC Regulatory Guides -Fuels And Materials Facilities (Division  
: 1) Regulatory Guide I .53 A ppli c ation of th e Singl e-Failur e Crit e rion to Saf ety S ys t e m s , 2003 (R201 I) Regulatory Guide 1.60 D e sign Response Sp e ctra for S e ismi c Design of Nuclear Pow e r Plants, 2014 Regulatory Guide 1.76 D es i g n Ba s i s Tornado and Tornado Mis s il es fo r N uclear P owe r Plants , 2 007 Regulatory Guide 1.97 Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, 2006 (R2013) Regulatory Guide I. I 00 S e ismi c Qualifi c ation of El ec tri c al and A c tiv e M e chani c al E quipm e nt and Fun c tional Qualifi c ation of A c tiv e M ec hani c al Equipm e nt for N ucl e ar P owe r Plant s, 2009 Regulatory Guide 1.152 Criteria for Use of Computers in Safety Systems of Nuclear Power Plants , 201 l Regulatory Guide 1.166 Pr e-Earthquak e Planning and Imm e diat e Nuclear Pow e r Plant Op e rator P os t E arthquak e A c tions , 1997 Regulatory Guide 1.167 Restart of a Nuclear Power Plant Shut down by a Seismic Event, 1997 Regulatory Guide I .208 P e rforman ce Ba se d Appr o a c h t o D e fin e th e Sit e-Sp ec ifi c Earthquak e Ground Motion , 2007 NRC Regulatory Guides -Fuels And Materials Facilities (Division
: 3) Regulatory Guide 3 .3 Quali ty A s s uran ce Program R e quir e m e nt s for Fu el R e pro cess in g Plants and for Plu to nium Pr ocess in g and Fu e l Fabri c ati o n Plants , 1974 (R2013) 3-8   
: 3) Regulatory Guide 3 .3 Quali ty A s s uran ce Program R e quir e m e nt s for Fu el R e pro cess in g Plants and for Plu to nium Pr ocess in g and Fu e l Fabri c ati o n Plants , 1974 (R2013) 3-8   
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* NORTKW(STMEDtcAl.ISOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) Title Regulatory Guide 3.6 Content of Technical Specification for Fuel Reprocessing Plants, l 973 (R2013) Regulatory Guide 3. l 0 Liquid Wast e Tr e atm e nt S y st em D es ign Guid e for Plutonium Proc e ssing and Fu e l Fabrication Plant s, l 973 (R2013) Regulatory Guide 3.18 Confinement Barriers and Systems for Fuel Reprocessing Plants, 1974 (R2013) R eg ulatory Guide 3.2 0 Pro cess Offgas S ys tems for Fu e l R e pro ce ssing Plants, 1974 (R20l3) Regulatory Guide 3.71 Nuclear Criticality Safety Standards for Fuels and Materials Facilities, 2010 NRC Regulatory Guides -Materials and Plant Protection (Division  
* NORTKW(STMEDtcAl.ISOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) Title Regulatory Guide 3.6 Content of Technical Specification for Fuel Reprocessing Plants, l 973 (R2013) Regulatory Guide 3. l 0 Liquid Wast e Tr e atm e nt S y st em D es ign Guid e for Plutonium Proc e ssing and Fu e l Fabrication Plant s, l 973 (R2013) Regulatory Guide 3.18 Confinement Barriers and Systems for Fuel Reprocessing Plants, 1974 (R2013) R eg ulatory Guide 3.2 0 Pro cess Offgas S ys tems for Fu e l R e pro ce ssing Plants, 1974 (R20l3) Regulatory Guide 3.71 Nuclear Criticality Safety Standards for Fuels and Materials Facilities, 2010 NRC Regulatory Guides -Materials and Plant Protection (Division
: 5) Regulatory Guide 5.7 Entry/Exit Control for Protect e d Areas, Vital Areas , and Material Access Areas, May 1980 (R20l0) Regulatory Guide 5.12 Genera l Use of Lo c ks in th e Prot ect ion and Contro l of Facilitie s and Sp ec ial N ucl e ar Mat er ial s, 1973 (R20l0) Regulatory Guide 5.27 Special Nuclear Material Doorway Monitors, 1974 Regulatory Guide 5.44 P e rim e ter Intrusi o n Alarm Syst e m s, 1997 (R2010) Regulatory Guide 5.57 Shipping and Receiving Control of Strategic Special Nuclear Material, 1980 Regulatory Guide 5.65 Vital Area Ac cess Contro l, Prot ec tion of Ph ysical S ec urity Equipment, and K ey and Lo c k Contro l s, 1986 (R2010) Regulatory Guide 5. 71 Cyber Security Programs for Nuclear Facilities, 2010 NUREG-0700, Human-System Interface Design Review Guidelines NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWREdition Section 2.3.1 Section 2.3.2 Section 3.3.l Section 3.3.2 Section 3.7.1 Section 3.7.2 " Re g ional Climatology
: 5) Regulatory Guide 5.7 Entry/Exit Control for Protect e d Areas, Vital Areas , and Material Access Areas, May 1980 (R20l0) Regulatory Guide 5.12 Genera l Use of Lo c ks in th e Prot ect ion and Contro l of Facilitie s and Sp ec ial N ucl e ar Mat er ial s, 1973 (R20l0) Regulatory Guide 5.27 Special Nuclear Material Doorway Monitors, 1974 Regulatory Guide 5.44 P e rim e ter Intrusi o n Alarm Syst e m s, 1997 (R2010) Regulatory Guide 5.57 Shipping and Receiving Control of Strategic Special Nuclear Material, 1980 Regulatory Guide 5.65 Vital Area Ac cess Contro l, Prot ec tion of Ph ysical S ec urity Equipment, and K ey and Lo c k Contro l s, 1986 (R2010) Regulatory Guide 5. 71 Cyber Security Programs for Nuclear Facilities, 2010 NUREG-0700, Human-System Interface Design Review Guidelines NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWREdition Section 2.3.1 Section 2.3.2 Section 3.3.l Section 3.3.2 Section 3.7.1 Section 3.7.2 " Re g ional Climatology
," Rev. 3 , March 2007 "Local Climatology," Rev. 3, March 2007 " Wind Loading," R ev. 3 , March 2007 "Tornado Loading," Rev. 3, March 2007 " Seismic De sign Parameters," March 20 07 "Seismic System Analysis," Rev. 4 , September 2013 Sec tion 3.7.3 " Seismic Subsystem Analysis," Rev. 4 , September 2013 NUREG-1513, Integrated Safety Analysis Guidance Document NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility Part 3, Appendix D "Natural Hazard Phenomena" NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors -Format and Content, Part 1 NUREG/CR-4604, Statistical Methods for Nuclear Material Management NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook Process hazard analysis "Development of Quantitative Risk Analyses" NUREG/CR-6463, Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems -Final Report NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology a Co mplet e references are provided in Sect ion 3.6. 3-9   
," Rev. 3 , March 2007 "Local Climatology," Rev. 3, March 2007 " Wind Loading," R ev. 3 , March 2007 "Tornado Loading," Rev. 3, March 2007 " Seismic De sign Parameters," March 20 07 "Seismic System Analysis," Rev. 4 , September 2013 Sec tion 3.7.3 " Seismic Subsystem Analysis," Rev. 4 , September 2013 NUREG-1513, Integrated Safety Analysis Guidance Document NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility Part 3, Appendix D "Natural Hazard Phenomena" NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors -Format and Content, Part 1 NUREG/CR-4604, Statistical Methods for Nuclear Material Management NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook Process hazard analysis "Development of Quantitative Risk Analyses" NUREG/CR-6463, Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems -Final Report NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology a Co mplet e references are provided in Sect ion 3.6. 3-9   
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.;.-.;* .. NWMI ..**.. ..* **.* ........ *.* .
.;.-.;* .. NWMI ..**.. ..* **.* ........ *.* .
* NOllTHWHTMlOICAllSOTOPH Document number 3 IEEE 902 IEEE 946 IEEE 1012 IEEE 1015 IEEE 1023 IEEE 102 8 IEEE 1046 IE E E 1050 IEEE 1100 IEEE 12 8 9 IEEE 1584 ANSI/IE EE C2 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Guide for Maintenance, Operation , and Safety of Industrial and Commercial Power Systems (Yellow Book), 1998 G e n e rating S tation s, 2004 Standard Criteria for Softwar e Verification and Validation, 2012 R ec omm e nd ed Pra c ti ce A ppl y in g L ow-Vo lta ge Ci r c uit Br ea k e r s Use d in Indu s trial and C omm e r c ial Po we r S ys t e m s (Blu e Book), 2006 (C2007) Guide for the Application of Human Factors Engineering to S y stems, Equipment, and Facilities of Nuclear Power Generating Stations, 2004 (R2010) Standard f o r So ftwar e R e views a nd Audit s , 200 8 Application Guide for Distributed Digital Control and Monitoring for Pow e r Plants, 1991 (Rl996) Gu id e f o r In str um e ntati o n and Co ntr o l E quipm e nt Gr o u n ding in G e n e r at in g Stations , 2004 Recommend e d Practi ce for Powering and Grounding El e ctronic Equipment (Emerald Book), 2005 Guid e for the A ppli c ati o n of Human Fa c t o rs E n g in ee ring i n th e D es i g n of Co mput e r-B ased M o nit or ing a nd C o ntr o l Di s pl ays for N ucl e ar P owe r Ge n e ratin g Stati o n s, I 99 8 (R2004) IEEE Guid e for Performing Arc-Flash Hazard Calculations , 2002 201 2 N ati o n a l E l ec tri c al Saf ety Cod e (NESC), 20 I 2 Illuminating Engineering Society of North America (IES) IES-20 I I Th e Li g htin g Handbo o k , 20 I I ANSl/IES RP-1-12 IES RP-7 American National Standard Practice for Offic e Lighting , 20 I 2 A m e ri c an Na ti o nal Standar d Pra c ti ce fo r Li g htin g Indu s trial Fa c ilit ies , 1991 (W2001) International Society of Automation (ISA) ANSl/ISA-5. 1-2009 ISA-5.3-1983 ISA-5.4-199 I ISA-5.5-1985 ANS l/ISA-5.06.01-2007 ANSI/ISA 7.0.01-1996 AN SI/ISA-12.0 I .01-2013 ISA-18.1-1979 ISA-TR20.00
* NOllTHWHTMlOICAllSOTOPH Document number 3 IEEE 902 IEEE 946 IEEE 1012 IEEE 1015 IEEE 1023 IEEE 102 8 IEEE 1046 IE E E 1050 IEEE 1100 IEEE 12 8 9 IEEE 1584 ANSI/IE EE C2 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Guide for Maintenance, Operation , and Safety of Industrial and Commercial Power Systems (Yellow Book), 1998 G e n e rating S tation s, 2004 Standard Criteria for Softwar e Verification and Validation, 2012 R ec omm e nd ed Pra c ti ce A ppl y in g L ow-Vo lta ge Ci r c uit Br ea k e r s Use d in Indu s trial and C omm e r c ial Po we r S ys t e m s (Blu e Book), 2006 (C2007) Guide for the Application of Human Factors Engineering to S y stems, Equipment, and Facilities of Nuclear Power Generating Stations, 2004 (R2010) Standard f o r So ftwar e R e views a nd Audit s , 200 8 Application Guide for Distributed Digital Control and Monitoring for Pow e r Plants, 1991 (Rl996) Gu id e f o r In str um e ntati o n and Co ntr o l E quipm e nt Gr o u n ding in G e n e r at in g Stations , 2004 Recommend e d Practi ce for Powering and Grounding El e ctronic Equipment (Emerald Book), 2005 Guid e for the A ppli c ati o n of Human Fa c t o rs E n g in ee ring i n th e D es i g n of Co mput e r-B ased M o nit or ing a nd C o ntr o l Di s pl ays for N ucl e ar P owe r Ge n e ratin g Stati o n s, I 99 8 (R2004) IEEE Guid e for Performing Arc-Flash Hazard Calculations , 2002 201 2 N ati o n a l E l ec tri c al Saf ety Cod e (NESC), 20 I 2 Illuminating Engineering Society of North America (IES) IES-20 I I Th e Li g htin g Handbo o k , 20 I I ANSl/IES RP-1-12 IES RP-7 American National Standard Practice for Offic e Lighting , 20 I 2 A m e ri c an Na ti o nal Standar d Pra c ti ce fo r Li g htin g Indu s trial Fa c ilit ies , 1991 (W2001) International Society of Automation (ISA) ANSl/ISA-5. 1-2009 ISA-5.3-1983 ISA-5.4-199 I ISA-5.5-1985 ANS l/ISA-5.06.01-2007 ANSI/ISA 7.0.01-1996 AN SI/ISA-12.0 I .01-2013 ISA-18.1-1979 ISA-TR20.00
.01-2007 In s trum e ntati o n S y mbol s and Id e ntifi c ati o n , 2009 Graphic Symbols for Distributed Control/Shared Displa y Instrumentation , Logic, and Computer System s , 1983 In s trum e nt L oo p Dia g ram s , 19 9 1 Graphic Symbols for Process Displays, 1985 Fun c ti o nal R e quir e m e nt s D oc um e ntati o n for Co ntrol So f tw are A ppli c a t i o n s, 2007 Quality Standard for Instrument Air D efi niti o n s and Inf o rmation P e rt a inin g t o El ec tri c al Equi p m e nt in Ha z a r d o u s (C la ss ifi e d) L oc ation s, 2013 Annunciator Sequences and Specifications, 1979 (R2004) Sp ec ifi c ati o n Fo rm s f o r Pr ocess Mea s u re m e nt and C o nt ro l In s trum e nt s Part 1: Ge n e ral Con s id e rations U pd a t e d with 27 n ew s p ec ifi c ati o n form s in 2004-2 006 and updat e d w i t h 11 n ew s p ec ifi cat i o n form s in 200 7 3-18
.01-2007 In s trum e ntati o n S y mbol s and Id e ntifi c ati o n , 2009 Graphic Symbols for Distributed Control/Shared Displa y Instrumentation , Logic, and Computer System s , 1983 In s trum e nt L oo p Dia g ram s , 19 9 1 Graphic Symbols for Process Displays, 1985 Fun c ti o nal R e quir e m e nt s D oc um e ntati o n for Co ntrol So f tw are A ppli c a t i o n s, 2007 Quality Standard for Instrument Air D efi niti o n s and Inf o rmation P e rt a inin g t o El ec tri c al Equi p m e nt in Ha z a r d o u s (C la ss ifi e d) L oc ation s, 2013 Annunciator Sequences and Specifications, 1979 (R2004) Sp ec ifi c ati o n Fo rm s f o r Pr ocess Mea s u re m e nt and C o nt ro l In s trum e nt s Part 1: Ge n e ral Con s id e rations U pd a t e d with 27 n ew s p ec ifi c ati o n form s in 2004-2 006 and updat e d w i t h 11 n ew s p ec ifi cat i o n form s in 200 7 3-18
::.**.*.*. .; ... ... NWMI ........ *.* 0 ." NORTKWlST MlDtCAI. ISOTOrt:S Document numbera ISA-RP60.1-1990 ISA-67.01.01-2002 ANSUISA-67.04.01-2006 I SA-RP67.04.02-20 10 ANSl/ISA-75
::.**.*.*. .; ... ... NWMI ........ *.* 0 ." NORTKWlST MlDtCAI. ISOTOrt:S Document numbera ISA-RP60.1-1990 ISA-67.01.01-2002 ANSUISA-67.04.01-2006 I SA-RP67.04.02-20 10 ANSl/ISA-75
.05.01-2000 ANSl/ISA-82.03-1988 ISA-TR84.00.04-2011 ISA-TR84.00.09-2013 ISA-TR9 l .00.02-2003 ANSU IS A-TR 99.00.0l-2007 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Control Cent e r Faciliti es, 1990 Transducer and Transmitt er Installation for Nuclear Safety App li ca ti ons, 20 0 2 (R2 00 7) Setpoints for Nuclear Safety-Related Instrumentation, 2006 (R2011) Met h odo l ogies for t h e D e t e rmination of Setpoints for Nuclear Safety-R elated In s trum e ntation, 20 10 Control Valve Terminolo gy, 2000 (R2005) Safety Standard for E l ectrica l a nd E l ec tr o ni c T es t , M e asuring , Con t ro lli ng, a nd R e l a t ed E qu ipment, 19 88 Part 1 Guideline for th e Impl e m e ntation of ANSJIISA-84.00.01-2004  
.05.01-2000 ANSl/ISA-82.03-1988 ISA-TR84.00.04-2011 ISA-TR84.00.09-2013 ISA-TR9 l .00.02-2003 ANSU IS A-TR 99.00.0l-2007 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Control Cent e r Faciliti es, 1990 Transducer and Transmitt er Installation for Nuclear Safety App li ca ti ons, 20 0 2 (R2 00 7) Setpoints for Nuclear Safety-Related Instrumentation, 2006 (R2011) Met h odo l ogies for t h e D e t e rmination of Setpoints for Nuclear Safety-R elated In s trum e ntation, 20 10 Control Valve Terminolo gy, 2000 (R2005) Safety Standard for E l ectrica l a nd E l ec tr o ni c T es t , M e asuring , Con t ro lli ng, a nd R e l a t ed E qu ipment, 19 88 Part 1 Guideline for th e Impl e m e ntation of ANSJIISA-84.00.01-2004
(!EC 61511), 2011 Secur i ty Countermeasures R e l ate d to Safety I nstrumented Syste m s (S I S), 2013 Criticality Classification Guideline for In str um e ntation , 2003 Security Te c hnolo gies for In dustr i a l Automatio n a nd Control Systems, 2007 International Atomic Energy Agency (IAEA) IAEA-TECDOC-1250 IAEA-TECDOC-134 7 I AEA-T EC DO C-14 30 Se i sm i c Desi gn Considerations of Nuclear Fuel Cycle Facilities, 2 001 Consideration of External Events in the D esign of Nuclear Facilities Other Than Nuclear Pow e r Plants , With Emphasis on Earthquakes, 2003 Radioi sotope Handling Fac ili ties and Automation of R adioisotope Production , 2 004 International Code Council (ICC) IBC 2 012 IFC 2012 IMC 2 01 2 Int e rn a tional Building Co d e, 20 12 Int er national Fire Code, 2012 Int e rnational Mec h anica l Code, 2 01 2 International Code Council Evaluation Service (ICC-ES) I CC-ES AC 1 56 "Ac ceptance Cr it e ria for Se i sm ic Cert ifi cation by S hak e-T ab le Testing of No n s tructu ra l Co mponen ts," 20 I 0 National Electrical Contractors Association (NECA) NECA 1 NECA90 NECA 100 NECA 101 NECA/AA 104 NECA/NEMA 105 NECA 111 Standard Pra c ti ce of Good Workma n s hip in E l ec trical Constructio n , 20 10 R eco mmend e d Practi ce for Commissioning Building Electrical Sy s tem s (ANSI), 2009 Sym b o l s fo r Electrica l Construction Drawings (ANSI), 2013 Standard for Installing St ee l Conduits (Rigid, IMC, EMT) (ANSI), 2013 Standard for I nsta lling A l uminu m Buildin g Wire and Cab l e (ANS I), 2012 Standard for Installing M eta l Cable Tra y Systems (ANSI), 2007 Sta nda rd for I nsta llin g Nonmeta lli c R aceways (RNC, ENT, LFNC) (ANSI), 2003 3-19 "NWMI ...... ..* ... ........... * *. *
(!EC 61511), 2011 Secur i ty Countermeasures R e l ate d to Safety I nstrumented Syste m s (S I S), 2013 Criticality Classification Guideline for In str um e ntation , 2003 Security Te c hnolo gies for In dustr i a l Automatio n a nd Control Systems, 2007 International Atomic Energy Agency (IAEA) IAEA-TECDOC-1250 IAEA-TECDOC-134 7 I AEA-T EC DO C-14 30 Se i sm i c Desi gn Considerations of Nuclear Fuel Cycle Facilities, 2 001 Consideration of External Events in the D esign of Nuclear Facilities Other Than Nuclear Pow e r Plants , With Emphasis on Earthquakes, 2003 Radioi sotope Handling Fac ili ties and Automation of R adioisotope Production , 2 004 International Code Council (ICC) IBC 2 012 IFC 2012 IMC 2 01 2 Int e rn a tional Building Co d e, 20 12 Int er national Fire Code, 2012 Int e rnational Mec h anica l Code, 2 01 2 International Code Council Evaluation Service (ICC-ES) I CC-ES AC 1 56 "Ac ceptance Cr it e ria for Se i sm ic Cert ifi cation by S hak e-T ab le Testing of No n s tructu ra l Co mponen ts," 20 I 0 National Electrical Contractors Association (NECA) NECA 1 NECA90 NECA 100 NECA 101 NECA/AA 104 NECA/NEMA 105 NECA 111 Standard Pra c ti ce of Good Workma n s hip in E l ec trical Constructio n , 20 10 R eco mmend e d Practi ce for Commissioning Building Electrical Sy s tem s (ANSI), 2009 Sym b o l s fo r Electrica l Construction Drawings (ANSI), 2013 Standard for Installing St ee l Conduits (Rigid, IMC, EMT) (ANSI), 2013 Standard for I nsta lling A l uminu m Buildin g Wire and Cab l e (ANS I), 2012 Standard for Installing M eta l Cable Tra y Systems (ANSI), 2007 Sta nda rd for I nsta llin g Nonmeta lli c R aceways (RNC, ENT, LFNC) (ANSI), 2003 3-19 "NWMI ...... ..* ... ........... * *. *
* NOllTHWllT lllOtCAl ISDTOPH Document number 3 NECA 120 NECA 202 NECA 230 NECA/FOA 301 NECA331 NECA400 NECA402 NECA/EGSA 404 NECA407 NECA408 NECA409 NECA 410 NECA 411 NECA420 NECA430 NECA/IESNA 500 NECA/IESNA 501 NECA/IESNA 502 NECA/BICSI 568 NECA/NCSCB 600 NECA/NEMA 605 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Standard for Installing Armored Cable (Type AC) and Metal-Clad Cable (Type MC) (ANSI), 2013 Sta ndard for In sta llin g and Maintaining Indu str ial H eat Tracing Systems (ANS I), 2013 Standard for Selecting, Installing, and Maintaining Electri c Motors and Motor Controllers (ANSI), 2010 Standard for In sta llin g and Testing Fiber Optics, 2009 Standard for Building and S ervice Entrance Grounding and Bonding, 2009 Sta ndard for In sta llin g and Maintaining Switchboards (ANS I), 2007 Standard for Installing and Maintaining Motor Control Centers (ANSI), 2007 Standard for In sta llin g Generator Sets (ANSI), 2014 Re c ommended Practice for Installing and Maintaining Pane/boards (ANSI), 2009 Sta nd ard for Installing and Mai ntainin g Bu sways (ANS I), 2 009 Standard for Installing and Maintaining Dry-Type Transformers (ANSI), 2009 Standard for In sta llin g and Maintaining Liquid-Filled Transformers (ANS I), 2013 Standard for In s talling and Maintaining Uninterruptibl e Power Supplies (UPS) (ANSI), 2006 Standard for Fuse Applicat i ons (ANS I), 2014 Standard for Installing Medium-Voltage Metal-Clad Switchgear (ANSI), 2006 R ecommended Practic e for In sta llin g Ind oor Lighting S y stems (ANS I), 2006 Recommended Practice for Installing Exterior Lighting S ys tems (ANSI), 2006 R ecomme nd ed Practice for Installing Indu strial Lighting S y stems (ANS I), 2006 Standard for Installing Building Telecommunications Cabling (ANSI), 2006 R eco mm e nd ed Practice for In sta llin g and Maintaining Medium-Voltage Ca bl e (ANS I), 2014 Installing Underground Nonmetallic Utility Duct (ANSI), 2005 National Electrical Manufacturers Association (NEMA) NEMAMG-1 Motors and Generators, 2009 InterNational Electrical Testing Association (NET A) ANSI/NETA ATS-2013 ANSI/NETA ETT-2010 ANSI/NETA MTS-2011 Standard for Acceptance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2013 Standard for Certification of E l ec tri ca l Testing T ec hni c ian s, 2010 Maintenance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2011 National Fire Protection Association (NFPA) NFPA 1 FPA2 Fire Code, 2015 Hydrogen Technologies Code, 20 11 3-20   
* NOllTHWllT lllOtCAl ISDTOPH Document number 3 NECA 120 NECA 202 NECA 230 NECA/FOA 301 NECA331 NECA400 NECA402 NECA/EGSA 404 NECA407 NECA408 NECA409 NECA 410 NECA 411 NECA420 NECA430 NECA/IESNA 500 NECA/IESNA 501 NECA/IESNA 502 NECA/BICSI 568 NECA/NCSCB 600 NECA/NEMA 605 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Standard for Installing Armored Cable (Type AC) and Metal-Clad Cable (Type MC) (ANSI), 2013 Sta ndard for In sta llin g and Maintaining Indu str ial H eat Tracing Systems (ANS I), 2013 Standard for Selecting, Installing, and Maintaining Electri c Motors and Motor Controllers (ANSI), 2010 Standard for In sta llin g and Testing Fiber Optics, 2009 Standard for Building and S ervice Entrance Grounding and Bonding, 2009 Sta ndard for In sta llin g and Maintaining Switchboards (ANS I), 2007 Standard for Installing and Maintaining Motor Control Centers (ANSI), 2007 Standard for In sta llin g Generator Sets (ANSI), 2014 Re c ommended Practice for Installing and Maintaining Pane/boards (ANSI), 2009 Sta nd ard for Installing and Mai ntainin g Bu sways (ANS I), 2 009 Standard for Installing and Maintaining Dry-Type Transformers (ANSI), 2009 Standard for In sta llin g and Maintaining Liquid-Filled Transformers (ANS I), 2013 Standard for In s talling and Maintaining Uninterruptibl e Power Supplies (UPS) (ANSI), 2006 Standard for Fuse Applicat i ons (ANS I), 2014 Standard for Installing Medium-Voltage Metal-Clad Switchgear (ANSI), 2006 R ecommended Practic e for In sta llin g Ind oor Lighting S y stems (ANS I), 2006 Recommended Practice for Installing Exterior Lighting S ys tems (ANSI), 2006 R ecomme nd ed Practice for Installing Indu strial Lighting S y stems (ANS I), 2006 Standard for Installing Building Telecommunications Cabling (ANSI), 2006 R eco mm e nd ed Practice for In sta llin g and Maintaining Medium-Voltage Ca bl e (ANS I), 2014 Installing Underground Nonmetallic Utility Duct (ANSI), 2005 National Electrical Manufacturers Association (NEMA) NEMAMG-1 Motors and Generators, 2009 InterNational Electrical Testing Association (NET A) ANSI/NETA ATS-2013 ANSI/NETA ETT-2010 ANSI/NETA MTS-2011 Standard for Acceptance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2013 Standard for Certification of E l ec tri ca l Testing T ec hni c ian s, 2010 Maintenance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2011 National Fire Protection Association (NFPA) NFPA 1 FPA2 Fire Code, 2015 Hydrogen Technologies Code, 20 11 3-20   
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* 0 NOITifMST MEDICAL ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Structural loads are due to the following: * * *
* 0 NOITifMST MEDICAL ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Structural loads are due to the following: * * *
* Self-weight of building materials and SSC s Occupancy and normal use of the RPF Off-normal conditions and accidents Natural phenomena hazards Section 3.1 de s cribes the s tructural discipline s ource requirements for these crit e ria. Structural load c riteria are summar i zed below. Site-specific natural phenomena hazard criteria a re based on the phy s ical location of the RPF given in Chapter 2.0 , Sections 2.3 and 2.5. 3.2.3.1.1 Dead Loads Dead loads consist of the weight of all materials of construction comprising the building , including walls , floors , roofs , ceilings , confinement doors , stairways , built-in partitions , wall and floor finishes , and cladding. Dead loads also consist of the weight of fixed equipment , including the weight of cranes. The density of all interconnections (e.g., heating , ventilation , and air conditioning  
* Self-weight of building materials and SSC s Occupancy and normal use of the RPF Off-normal conditions and accidents Natural phenomena hazards Section 3.1 de s cribes the s tructural discipline s ource requirements for these crit e ria. Structural load c riteria are summar i zed below. Site-specific natural phenomena hazard criteria a re based on the phy s ical location of the RPF given in Chapter 2.0 , Sections 2.3 and 2.5. 3.2.3.1.1 Dead Loads Dead loads consist of the weight of all materials of construction comprising the building , including walls , floors , roofs , ceilings , confinement doors , stairways , built-in partitions , wall and floor finishes , and cladding. Dead loads also consist of the weight of fixed equipment , including the weight of cranes. The density of all interconnections (e.g., heating , ventilation , and air conditioning
[HV AC] ductwork , conduits , cable trays , and piping) between equipment will be conservatively estimated and included in the final design for dead load for fixtures attached to ceiling s or anchored to floors in the RPF. 3.2.3.1.2 Lateral Earth and Ground Water Pressure Loads Lateral earth and ground w ater pressure loads are lateral pressures due to the weight of adjacent soil and groundwater , respecti v ely. The design lateral earth load is a function of the composition of the soil. The Discovery Ridge Phase I En v ironmental Assessment (Terracon , 2011 a) indicates that the soils present are clayey gravels consistent with the Unified Soil Classification "GC." In addition , the assessment indicates that expansi v e soils are pre s ent. Chapter 2.0 , Section 2.5.3 present s additional on-s ite soil information. The de s ign lateral earth pre ss ure load for the RPF i s based on ASCE 7 , Table 3.2.1 , and ha s been augmented to account for the expansi v e s oils (e.g., s urcharge load is applied to account for the weight of the facility above the soils a djacent to the tank hot cell). The design groundwater depth is estimated to be approximately  
[HV AC] ductwork , conduits , cable trays , and piping) between equipment will be conservatively estimated and included in the final design for dead load for fixtures attached to ceiling s or anchored to floors in the RPF. 3.2.3.1.2 Lateral Earth and Ground Water Pressure Loads Lateral earth and ground w ater pressure loads are lateral pressures due to the weight of adjacent soil and groundwater , respecti v ely. The design lateral earth load is a function of the composition of the soil. The Discovery Ridge Phase I En v ironmental Assessment (Terracon , 2011 a) indicates that the soils present are clayey gravels consistent with the Unified Soil Classification "GC." In addition , the assessment indicates that expansi v e soils are pre s ent. Chapter 2.0 , Section 2.5.3 present s additional on-s ite soil information. The de s ign lateral earth pre ss ure load for the RPF i s based on ASCE 7 , Table 3.2.1 , and ha s been augmented to account for the expansi v e s oils (e.g., s urcharge load is applied to account for the weight of the facility above the soils a djacent to the tank hot cell). The design groundwater depth is estimated to be approximately  


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* HOITHWEST MEDtCAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Per NUREG-1520, Section 3.2.3.4(1)(c), and ASCE 7 , Chapter 5, flood loads will be based on the water level of the 100-year flood (one percent probability of exceedance per year). The facility has been determined to be above both the 100-year and the 500-year flood plain. Chapter 2 , Section 2.4.3, provides additional detail for flood protection measures.
* HOITHWEST MEDtCAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Per NUREG-1520, Section 3.2.3.4(1)(c), and ASCE 7 , Chapter 5, flood loads will be based on the water level of the 100-year flood (one percent probability of exceedance per year). The facility has been determined to be above both the 100-year and the 500-year flood plain. Chapter 2 , Section 2.4.3, provides additional detail for flood protection measures.
Postulated flooding from component failures in the building compartments will be prevented from adversely affecting plant safety or posing any hazard to the public. Exterior or access openings and penetrations into the RPF will be above the maximum postulated flooding level. Therefore , flood loads are considered highly unlikely and are not considered design loads. 3.3.1.1.2 Flooding from Inadvertent Discharge of Fire Protection System Water Design of fire suppression systems using water (e.g., automatic sprinklers, hose stations) includes elements such as the grading and channeling of floors , raising of equipment mounts above floors , shelving and floor drains , and other passive means. These features will ensure sufficient capacity for gravity-dri ven collection and drainage of the maximum water discharge rate and duration to avoid localized flooding and resulting water damage to equipment within the area. In addition, particularly sensitive systems and components, whether electrical, optical , mechanical and/or chemical, are typically protected within enclosures designed for the anticipated adverse environmental conditions resulting from these types of water discharges. If critical for safety, these water-sensitive systems and components will be installed within the appropriate severe environment-rated enclosures in accordance with the relevant industry standard(s) (e.g., Nationa l Electrical Manufacturers Association  
Postulated flooding from component failures in the building compartments will be prevented from adversely affecting plant safety or posing any hazard to the public. Exterior or access openings and penetrations into the RPF will be above the maximum postulated flooding level. Therefore , flood loads are considered highly unlikely and are not considered design loads. 3.3.1.1.2 Flooding from Inadvertent Discharge of Fire Protection System Water Design of fire suppression systems using water (e.g., automatic sprinklers, hose stations) includes elements such as the grading and channeling of floors , raising of equipment mounts above floors , shelving and floor drains , and other passive means. These features will ensure sufficient capacity for gravity-dri ven collection and drainage of the maximum water discharge rate and duration to avoid localized flooding and resulting water damage to equipment within the area. In addition, particularly sensitive systems and components, whether electrical, optical , mechanical and/or chemical, are typically protected within enclosures designed for the anticipated adverse environmental conditions resulting from these types of water discharges. If critical for safety, these water-sensitive systems and components will be installed within the appropriate severe environment-rated enclosures in accordance with the relevant industry standard(s) (e.g., Nationa l Electrical Manufacturers Association
[NEMA] enclosure standards).
[NEMA] enclosure standards).
Selection of specific fire suppression systems for facility locations will be guided by recommendations in relevant industry standards (e.g., NFPA 801, Standard for Fire Protection for Faciliti e s Handling Radioactive Materials) and will depend on the level of fire hazards at those lo cations , as determined from the final facility and process systems designs. These final detailed designs will include any facility design elements and sensitive equipment protection measures deemed necessary for addressing the maximum inadvertent rate and duration of water discharges from the fire protection systems. The final comprehensive facility design , along with commitments to design codes , standards , and other referenced documents (including any exceptions or exemptions to the identified requirements), will be identified and provided as part of the Operating License Application.
Selection of specific fire suppression systems for facility locations will be guided by recommendations in relevant industry standards (e.g., NFPA 801, Standard for Fire Protection for Faciliti e s Handling Radioactive Materials) and will depend on the level of fire hazards at those lo cations , as determined from the final facility and process systems designs. These final detailed designs will include any facility design elements and sensitive equipment protection measures deemed necessary for addressing the maximum inadvertent rate and duration of water discharges from the fire protection systems. The final comprehensive facility design , along with commitments to design codes , standards , and other referenced documents (including any exceptions or exemptions to the identified requirements), will be identified and provided as part of the Operating License Application.
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Instrumentation Seismic recording instrumentation will be triaxial digital systems that record accelerations versus time accurately for periods between 0 and 10 sec. Recorders will have rechargeable batteries such that if there is a loss of power , recording will still occur. All instrumentation will be housed in appropriate weather and creature-proofed enclosures.
Instrumentation Seismic recording instrumentation will be triaxial digital systems that record accelerations versus time accurately for periods between 0 and 10 sec. Recorders will have rechargeable batteries such that if there is a loss of power , recording will still occur. All instrumentation will be housed in appropriate weather and creature-proofed enclosures.
As a minimum , one recorder should be located in the free-field mounted on rock or competent ground generally representative of the site. In addition , at sites classified as Seismic Design Category D , E , or Fin accordance with ASCE 7 , Chapter 11 , using Occupancy Category IV , recorders will be located and attached to the foundations and roof s of the RPF and in the control room. The systems will have the capability to produce motion time histories. Response spectra will be computed separately.
As a minimum , one recorder should be located in the free-field mounted on rock or competent ground generally representative of the site. In addition , at sites classified as Seismic Design Category D , E , or Fin accordance with ASCE 7 , Chapter 11 , using Occupancy Category IV , recorders will be located and attached to the foundations and roof s of the RPF and in the control room. The systems will have the capability to produce motion time histories. Response spectra will be computed separately.
The purpose of the instrumentation is to (1) permit a comparison of measured responses of C-1 structures and selected components with predetermined results of analyses that predict when damage might occur, (2) permit facility operators to understand the possible extent of damage within the facility immediately following an earthquake, and (3) be able to determine when an safe-shutdown earthquake event has occurred that would require the emptying of the tank(s) for inspection as specified in NFPA 59A , Standard for the Production , Storage , and H and ling of Liquefied Natural Gas , Section 4.1.3.6( c ). 3-41
The purpose of the instrumentation is to (1) permit a comparison of measured responses of C-1 structures and selected components with predetermined results of analyses that predict when damage might occur, (2) permit facility operators to understand the possible extent of damage within the facility immediately following an earthquake, and (3) be able to determine when an safe-shutdown earthquake event has occurred that would require the emptying of the tank(s) for inspection as specified in NFPA 59A , Standard for the Production , Storage , and H and ling of Liquefied Natural Gas , Section 4.1.3.6( c ). 3-41
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* NORTHW(n 11(01CA1.ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Seismic instrumentation for the RPF site is not an IROFS; it provides no safety function and is therefore not "safety-related." Although the seismic recorders have no safety function , they must be designed to withstand any credible level of shaking to ensure that the ground motion would be recorded in the highly unlikely event of an earthquake. This capability requires verification of adequate capacity from the manufacturer (e.g., prior shake table tests of their product line), provision of adequate anchorage (e.g., manufacturer-provided anchor specifications to ensure accurate recordings), and a check for seismic interaction hazards such as water spray or falling fixtures.
* NORTHW(n 11(01CA1.ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Seismic instrumentation for the RPF site is not an IROFS; it provides no safety function and is therefore not "safety-related." Although the seismic recorders have no safety function , they must be designed to withstand any credible level of shaking to ensure that the ground motion would be recorded in the highly unlikely event of an earthquake. This capability requires verification of adequate capacity from the manufacturer (e.g., prior shake table tests of their product line), provision of adequate anchorage (e.g., manufacturer-provided anchor specifications to ensure accurate recordings), and a check for seismic interaction hazards such as water spray or falling fixtures.
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* Receive liquid waste that is divided into high-dose source terms and low-dose source terms to lag storage * * * * *
* Receive liquid waste that is divided into high-dose source terms and low-dose source terms to lag storage * * * * *
* Transfer remotely loaded drums with high-activity solid waste via a solid waste drum transit system to a waste encapsulation cell Encapsulate solid waste drums Load drums with solidification agent and low-dose liquid waste Load high-integrity containers with solidification agent and high-dose liquid waste Handle and load a waste shipping cask with radiological waste drums/containers Safety-related functions: Maintain subcriticality conditions through mass limits Prevent spread of contamination to manned areas of the facility that could result in personnel exposure to radioactive materials or toxic chemicals Provide shielding, distance , or other means to minimize personnel exposure to penetrating radiation Design Basis Values *
* Transfer remotely loaded drums with high-activity solid waste via a solid waste drum transit system to a waste encapsulation cell Encapsulate solid waste drums Load drums with solidification agent and low-dose liquid waste Load high-integrity containers with solidification agent and high-dose liquid waste Handle and load a waste shipping cask with radiological waste drums/containers Safety-related functions: Maintain subcriticality conditions through mass limits Prevent spread of contamination to manned areas of the facility that could result in personnel exposure to radioactive materials or toxic chemicals Provide shielding, distance , or other means to minimize personnel exposure to penetrating radiation Design Basis Values *
* Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs 30-year design life with the exception of common replaceable parts (e.g., pumps) 3-58
* Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs 30-year design life with the exception of common replaceable parts (e.g., pumps) 3-58
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* NCMmlWESTMEDK:AllSOTDPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components 3.5.2.7.7 Criticality Accident Alarm System Chapter 6.0 , Section 6.3.3.1 , and Chapter 7.0 , Section 7.3.7 , provide descriptions of the criticality accident alarm system. Design Basis Functions
* NCMmlWESTMEDK:AllSOTDPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components 3.5.2.7.7 Criticality Accident Alarm System Chapter 6.0 , Section 6.3.3.1 , and Chapter 7.0 , Section 7.3.7 , provide descriptions of the criticality accident alarm system. Design Basis Functions
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* Provide confinement of airborne radioactive materials by providing for the rapid , automatic closure of isolation dampers within confinement zones for various accident conditions Provide conditioned air to ensure suitable environmental conditions for personnel and equipment inRPF Design Basis Values
* Provide confinement of airborne radioactive materials by providing for the rapid , automatic closure of isolation dampers within confinement zones for various accident conditions Provide conditioned air to ensure suitable environmental conditions for personnel and equipment inRPF Design Basis Values
* Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs * * *
* Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs * * *
* Provide an integrated leak rate for confinement boundaries that meets the requirements of accident analy s es that complies with 10 CFR 10.61 30-year design life Maintain occupied space at 24 degrees Celsius (0 C) (75 degrees Fahrenheit  
* Provide an integrated leak rate for confinement boundaries that meets the requirements of accident analy s es that complies with 10 CFR 10.61 30-year design life Maintain occupied space at 24 degrees Celsius (0 C) (75 degrees Fahrenheit
[°F]) (summer) and 22°C (72°F) (winter), with active ventilation to support workers and equipment Maintain air quality that complies with 10 CFR 20 dose limits for normal operations and shutdown 3.5.2.7.13 Fire Protection System Chapter 9.0 , Section 9.3 provides a detailed description of the RPF fire protection system. Design Basis Functions
[°F]) (summer) and 22°C (72°F) (winter), with active ventilation to support workers and equipment Maintain air quality that complies with 10 CFR 20 dose limits for normal operations and shutdown 3.5.2.7.13 Fire Protection System Chapter 9.0 , Section 9.3 provides a detailed description of the RPF fire protection system. Design Basis Functions
* Provide detection and suppression of fire s *  
* Provide detection and suppression of fire s *  
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* Provide standard gas bottles , with capacity of approximately 8 , 495 L (300 cubic feet [ft 3]) 3.5.2.7.17 Process Chilled Water System Chapter 9.0 , Section 9.7.1 provides a detailed description of the RPF chilled water system. Design Basis Functions  
* Provide standard gas bottles , with capacity of approximately 8 , 495 L (300 cubic feet [ft 3]) 3.5.2.7.17 Process Chilled Water System Chapter 9.0 , Section 9.7.1 provides a detailed description of the RPF chilled water system. Design Basis Functions  
*
*
* Provide process chilled water loop for three secondary loops heat exchangers One large geometry secondary loop in hot cell One criticality-safe geometry secondary loop in hot cell One criticality-safe geometry secondary loop in target fabrication area Provide monitoring of chilled water loops for loss of primary containment 3-62
* Provide process chilled water loop for three secondary loops heat exchangers One large geometry secondary loop in hot cell One criticality-safe geometry secondary loop in hot cell One criticality-safe geometry secondary loop in target fabrication area Provide monitoring of chilled water loops for loss of primary containment 3-62
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* Provide cover gas to prevent flammable conditions in secondary loops Design Basis Values
* Provide cover gas to prevent flammable conditions in secondary loops Design Basis Values
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June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 6.0 -Engineered Safety Features Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:   
June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 6.0 -Engineered Safety Features Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:   
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.; ... ; .. NWMI ..*... ..* .... ........ *.*  " "NDlrTHWHTMBHCAllSOTOPH Rev Date 0 6/29/2015 1 6/26/2017 NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features REVISIO N HISTORY Reason for Revision Revised By Initial Application Not required Incorporate changes based on responses to C. Haass NRC Requests for Additional I nformation
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An upset during the dissolver operation (e.g., loss of vacuum pump operation) will result in closing Valve 1 and opening Valve 2 to contain dissolver offgas within the dissolver and off gas vessels. Due to the short duration of dissolver operation , dissolution is assumed to go to completion independent of an off gas system upset. The pressure relief tank will contain the offgas as dissolution is completed. Valves 3 , 4 , and 5 are provided for upset recovery.
An upset during the dissolver operation (e.g., loss of vacuum pump operation) will result in closing Valve 1 and opening Valve 2 to contain dissolver offgas within the dissolver and off gas vessels. Due to the short duration of dissolver operation , dissolution is assumed to go to completion independent of an off gas system upset. The pressure relief tank will contain the offgas as dissolution is completed. Valves 3 , 4 , and 5 are provided for upset recovery.
After correction of the upset cause , gases collected in the pressure relief tank will be routed to the downstream treatment unit operations via Valve 3 or returned to a caustic s crubber via Valve 4. Liquid condensed in the pressure relief tank as a result of activation will be routed to the dissolver offgas liquid waste collection tank via Valve 5 for disposal.
After correction of the upset cause , gases collected in the pressure relief tank will be routed to the downstream treatment unit operations via Valve 3 or returned to a caustic s crubber via Valve 4. Liquid condensed in the pressure relief tank as a result of activation will be routed to the dissolver offgas liquid waste collection tank via Valve 5 for disposal.
6-15
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::.**.*.*.* .. ;.-.;* .. NWMI ......... *.* . NOATHWESTMEOICAUSOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Accident Mitigated
::.**.*.*.* .. ;.-.;* .. NWMI ......... *.* . NOATHWESTMEOICAUSOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Accident Mitigated
* Irradiated target dissolver off gas system malfunctions, including loss of power during target dissolution operations System Components Pressure relief valves Pressure relief tank (DS-TK-500)
* Irradiated target dissolver off gas system malfunctions, including loss of power during target dissolution operations System Components Pressure relief valves Pressure relief tank (DS-TK-500)
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* Provide an off gas release height for ventilation gases consistent with the stack height used as input to mitigated dose consequence evaluations. Design Basis The Zone I exhaust stack height is 23 m (75 ft). Test Requirements The above analysis is based on information developed for the Construction Permit Application.
* Provide an off gas release height for ventilation gases consistent with the stack height used as input to mitigated dose consequence evaluations. Design Basis The Zone I exhaust stack height is 23 m (75 ft). Test Requirements The above analysis is based on information developed for the Construction Permit Application.
Additional detailed information on test requirements will be developed for the Operating License Application.
Additional detailed information on test requirements will be developed for the Operating License Application.
6-20 NWMI ..**.. .. ... **** .. .. .. * *. * ! 0 NOltTifWEST MEDtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2.1.7.8 IROFS CS-09, Double Wall Piping IROFS CS-09 , "Double Wall Piping," is identified by the accident analyses in Chapter 13.0. This IROFS has both a confinement and nuclear criticality prevention function. As a PEC, the piping system conveying fissile solution between credited confinement locations will be provided with a double-wall barrier to contain any spills that may occur from the primary confinement  
6-20 NWMI ..**.. .. ... **** .. .. .. * *. * ! 0 NOltTifWEST MEDtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2.1.7.8 IROFS CS-09, Double Wall Piping IROFS CS-09 , "Double Wall Piping," is identified by the accident analyses in Chapter 13.0. This IROFS has both a confinement and nuclear criticality prevention function. As a PEC, the piping system conveying fissile solution between credited confinement locations will be provided with a double-wall barrier to contain any spills that may occur from the primary confinement
[Proprietary Information]
[Proprietary Information]
Figure 6-7. Proposed Location of Double-Wall Piping (Example) piping. This IROFS will be used at those locations that pass through the facility , where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection purposes. Figure 6-7 provides an example location where IROFS CS-09 will be applied (e.g., the transfer line between the recycle uranium decay tanks and the [Proprietary Information]).
Figure 6-7. Proposed Location of Double-Wall Piping (Example) piping. This IROFS will be used at those locations that pass through the facility , where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection purposes. Figure 6-7 provides an example location where IROFS CS-09 will be applied (e.g., the transfer line between the recycle uranium decay tanks and the [Proprietary Information]).
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Section 6.3.1.2 summarizes IROFS related to preventing a nuclear criticality identified by the accident analyses described in Chapter I 3.0. 6.3.1.1 Preliminary Criticality Safety Evaluat ion s A series of calculations were performed to support the Construction Permit Application investigating parameters associated with prevention of nuclear criticality in the current equipment configuration of major process systems. The calculations are described in the following documents: * * * *
Section 6.3.1.2 summarizes IROFS related to preventing a nuclear criticality identified by the accident analyses described in Chapter I 3.0. 6.3.1.1 Preliminary Criticality Safety Evaluat ion s A series of calculations were performed to support the Construction Permit Application investigating parameters associated with prevention of nuclear criticality in the current equipment configuration of major process systems. The calculations are described in the following documents: * * * *
* NWMI-2015-CRITCALC-001 , Singl e Param e t e r Sub c riti c al Limit s for 20 wt% 235 U-U ranium M e tal , Uranium Oxid e, and Homog e nou s Wat e r Mixtures NWMI-2015-CRITCALC-002 , I r radiat e d Tar ge t Low-Enri c h e d Uranium Mat e rial Di ss olution NWMI-2015-CRITCALC-003 , 55-Gallon Drum Arra ys NWMI-2015-CRITCALC-005 , Targ e t Fabri ca tion Tanks , W e t Pro cesses, and Storag e NWMI-2015-CRITCALC-006 , Tank Hot Ce ll Calculations were performed using the MCNP 6.1 code (LA-CP-13-00634 , MC N P6 Use r Manual). Validation of the MCNP 6.1 code used in the calculations i s described in [Proprietary Information].
* NWMI-2015-CRITCALC-001 , Singl e Param e t e r Sub c riti c al Limit s for 20 wt% 235 U-U ranium M e tal , Uranium Oxid e, and Homog e nou s Wat e r Mixtures NWMI-2015-CRITCALC-002 , I r radiat e d Tar ge t Low-Enri c h e d Uranium Mat e rial Di ss olution NWMI-2015-CRITCALC-003 , 55-Gallon Drum Arra ys NWMI-2015-CRITCALC-005 , Targ e t Fabri ca tion Tanks , W e t Pro cesses, and Storag e NWMI-2015-CRITCALC-006 , Tank Hot Ce ll Calculations were performed using the MCNP 6.1 code (LA-CP-13-00634 , MC N P6 Use r Manual). Validation of the MCNP 6.1 code used in the calculations i s described in [Proprietary Information].
The validation report documents the methodology and results for the bias and bias uncertainty values calculated for homogeneous and heterogeneous uranium systems for the MCNP 6.1 code system. The bias is expressed as USLs calculated using a facility-specific  
The validation report documents the methodology and results for the bias and bias uncertainty values calculated for homogeneous and heterogeneous uranium systems for the MCNP 6.1 code system. The bias is expressed as USLs calculated using a facility-specific
[Proprietary Information].
[Proprietary Information].
The primary focus of the validation was to determine the bias and bias uncertainty for intermediate-enriched uranium (IEU) systems. However , sufficient experiments for low-enriched uranium (LEU) and high-enriched uranium were included to demon s trate that there is no v ariation in the USL with varying enrichment.
The primary focus of the validation was to determine the bias and bias uncertainty for intermediate-enriched uranium (IEU) systems. However , sufficient experiments for low-enriched uranium (LEU) and high-enriched uranium were included to demon s trate that there is no v ariation in the USL with varying enrichment.
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The analyst must document any extrapolation beyond the validation area of applicability , and justification must be documented for no adjustments to the margin of subcriticality when extrapolating.
The analyst must document any extrapolation beyond the validation area of applicability , and justification must be documented for no adjustments to the margin of subcriticality when extrapolating.
Table 6-4. Area of Applicability Summary Parameter Fissile materi a l Fissile material form H/235 U ratio Average neutron energy causing fission E nrichment Moderating materials R eflect in g materials Absorber materials Geometry
Table 6-4. Area of Applicability Summary Parameter Fissile materi a l Fissile material form H/235 U ratio Average neutron energy causing fission E nrichment Moderating materials R eflect in g materials Absorber materials Geometry
* Source: [Proprietary In format i o n]. ANECF = average neutr on energy causing fission. Area of Applicability  
* Source: [Proprietary In format i o n]. ANECF = average neutr on energy causing fission. Area of Applicability
[Proprietary Inform ation] [Proprietary Information]  
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The RPF was divided into 13 activity groups for de ve lopment of preliminary CS Es of the activities and associated equipment.
The RPF was divided into 13 activity groups for de ve lopment of preliminary CS Es of the activities and associated equipment.
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.; ... ; .. NWMI .... ** . ..* *.. ..... .... .. ' ! *. * ! ." NORTHWEST llEDtcAL tSOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features
.; ... ; .. NWMI .... ** . ..* *.. ..... .... .. ' ! *. * ! ." NORTHWEST llEDtcAL tSOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features
* NWMI-2015-CSE-13 , Analytical Laborato ry (Tab le 6-13) The CSEs will be updated for final design and the Operating License Application. Criticality controls are selected based on the following order of preference:
* NWMI-2015-CSE-13 , Analytical Laborato ry (Tab le 6-13) The CSEs will be updated for final design and the Operating License Application. Criticality controls are selected based on the following order of preference:
Passive engineered controls Active engineered contro ls Enhanced administrative controls Administrative controls Note that a number of features li sted in the preliminary CSEs are duplicated in multiple activity groups (e.g., the floor of cells is verified to be flat , with no collect ion points deeper than 3.5 centimeters  
Passive engineered controls Active engineered contro ls Enhanced administrative controls Administrative controls Note that a number of features li sted in the preliminary CSEs are duplicated in multiple activity groups (e.g., the floor of cells is verified to be flat , with no collect ion points deeper than 3.5 centimeters
[cm]). Duplications are included in the current listings to c learly identify minor dimension variations that may exist in the defined features for different activity groups. Table 6-6. [Proprietary Information]
[cm]). Duplications are included in the current listings to c learly identify minor dimension variations that may exist in the defined features for different activity groups. Table 6-6. [Proprietary Information]
Double-Contingency Controls Identifier" CSE-0I-PDF1 [Proprietary Information]
Double-Contingency Controls Identifier" CSE-0I-PDF1 [Proprietary Information]
CSE-0 l-PDF2 [Proprietary Information]
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CSE-01-ACI  
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Double-Contingency Controls (2 pages) Identifier*
Double-Contingency Controls (2 pages) Identifier*
Feature description and basis CSE-02-PDF l [Proprietary Information]
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[Proprietary Inform a tion] = [Proprietary Inform a tion] 6-40   
[Proprietary Inform a tion] = [Proprietary Inform a tion] 6-40   
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* NOITMWUT M£DICAl ISOTOPU [Propri e t ary Information]
* NOITMWUT M£DICAl ISOTOPU [Propri e t ary Information]
NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-8. [Proprietary Information]
NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-8. [Proprietary Information]
Double-Contingency Controls (2 pages) Identifier" CSE-03-PDFI  
Double-Contingency Controls (2 pages) Identifier" CSE-03-PDFI
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CSE-03-PDF2  
CSE-03-PDF2
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C S E-03-PDF5 [Propri e t ary Inform a tion] CSE-03-PDF6  
C S E-03-PDF5 [Propri e t ary Inform a tion] CSE-03-PDF6
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C S E-0 3-PDF 11 [Propri etary In fo rmation] CS E-03-PDF12  
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I X Mo a [Propri etary In fo rm a ti o n]. ion exc han ge. = m o l y bd e num. Feature description and basis [Pro pri e tary In fo rm a ti o n] [P ro pri e t a ry Inform a ti o n]. 6-41 NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]
I X Mo a [Propri etary In fo rm a ti o n]. ion exc han ge. = m o l y bd e num. Feature description and basis [Pro pri e tary In fo rm a ti o n] [P ro pri e t a ry Inform a ti o n]. 6-41 NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]
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CSE-04-PD FS" [Propri etary In formation] CSE-04-PDF6*  
CSE-04-PD FS" [Propri etary In formation] CSE-04-PDF6*
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CSE-04-PD F7* [Propri etary In fo rm a tion] CSE-04-PDFS* [Proprietary Information]
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CSE-04-AC6" [Proprietary In format ion] CSE-04-AC7" [Proprietary Information]
CSE-04-AC6" [Proprietary In format ion] CSE-04-AC7" [Proprietary Information]
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CSE-05-PD F3b [Pr o pri etary Inform at ion] CSE-05-PDF4b  
CSE-05-PD F3b [Pr o pri etary Inform at ion] CSE-05-PDF4b
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CSE-05-PD F7 h [Propri etary Inform at ion] CSE-05-PDF8b  
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.. ;.-.;* .. NWMI *:.**.*.* . .......... *:* ' *. ! * . NOmtWEn MEOK:Al lSOTOHS NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-9. [Proprietary Information]
.. ;.-.;* .. NWMI *:.**.*.* . .......... *:* ' *. ! * . NOmtWEn MEOK:Al lSOTOHS NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-9. [Proprietary Information]
Double-Contingency Controls (8 pages) Identifier CSE-05-AEFlb  
Double-Contingency Controls (8 pages) Identifier CSE-05-AEFlb
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CSE-0 6-A C3 c [Proprietary Information]
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* NOllTNWUT llEDtCAl tsOTOPH NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-10. [Proprietary Information]
* NOllTNWUT llEDtCAl tsOTOPH NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-10. [Proprietary Information]
Double-Contingency Controls (2 pages) ldentifiera Feature description and basis C S E-0 8-PDFl [Propri et a ry Inform a tion] CSE-08-PDF2  
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C S E-0 8-PD F 9 [Propri e t ary Inform a tion] CSE-08-[Proprietary Information]
C S E-0 8-PD F 9 [Propri e t ary Inform a tion] CSE-08-[Proprietary Information]
PDFlO CSE-0 8-[Propri e t ary I nfo rm a ti o n] PDF!! CS E-08-[Proprietary Information]
PDFlO CSE-0 8-[Propri e t ary I nfo rm a ti o n] PDF!! CS E-08-[Proprietary Information]
PDF12 CSE-0 8-AE F I [Propri e t ary Inform a tion] CSE-08-ACI  
PDF12 CSE-0 8-AE F I [Propri e t ary Inform a tion] CSE-08-ACI
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CSE-0 8-AC2 [Propri e t ary Inform a tion] * [P ro pri etary I nfo rm a t io n] DBE = des i gn b as i s earthq u ake. IX io n exc h a n ge. 6-51   
CSE-0 8-AC2 [Propri e t ary Inform a tion] * [P ro pri etary I nfo rm a t io n] DBE = des i gn b as i s earthq u ake. IX io n exc h a n ge. 6-51   
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PDF l b CSE-10-AEflb [Proprietary Information]
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CSE-I 1-AEF l (Propri e t a ry Information]
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* [Pr o pri etary In fo rm a ti o n] D B E HE P A des i gn b as i s ea rth q u a k e. = hi g h-e ffi cie n cy p a rti c ul ate a i r. Feature description and basis Mo NO x 6-56 m o l y bd e num. nitroge n ox id e.   
* [Pr o pri etary In fo rm a ti o n] D B E HE P A des i gn b as i s ea rth q u a k e. = hi g h-e ffi cie n cy p a rti c ul ate a i r. Feature description and basis Mo NO x 6-56 m o l y bd e num. nitroge n ox id e.   
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..... ;* .. *NWMI ...... ...* ... ........ *.* . ' *: ! ." NomtWfST MEDICAL ISOTOH:S NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Solid uranium will be handled outside of processing systems during: * * * *
..... ;* .. *NWMI ...... ...* ... ........ *.* . ' *: ! ." NomtWfST MEDICAL ISOTOH:S NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Solid uranium will be handled outside of processing systems during: * * * *
* Receipt and processing of fresh uranium (and presumably shipment of spent uranium back to the supplier)  
* Receipt and processing of fresh uranium (and presumably shipment of spent uranium back to the supplier)
[Proprietary Information]
[Proprietary Information]
Fabrication of targets using [Proprietary Information]
Fabrication of targets using [Proprietary Information]
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If the IROFS fails , accidental nuclear criticality is possible without additional control. System Components As a PEC , fixed interaction control fixtures or workstations will be provided for holding or processing approved containers containing approved quantities of uranium metal , [Proprietary Information], batches of targets , and batches of samples. Functional Requirements The fixtures are designed to hold only the approved container or batch and are fixed with 2-ft edge spacing from all other fissile material containers , workstations , or fissile solution tanks, vessels, and ion exchange (IX) columns. Where LEU target material is handled in open containers , the design will prevent spills from readily spreading to an adjacent workstation or storage location. Design Basis Final workstation and fixture spacing will be determined in final design when all process upsets are evaluated.
If the IROFS fails , accidental nuclear criticality is possible without additional control. System Components As a PEC , fixed interaction control fixtures or workstations will be provided for holding or processing approved containers containing approved quantities of uranium metal , [Proprietary Information], batches of targets , and batches of samples. Functional Requirements The fixtures are designed to hold only the approved container or batch and are fixed with 2-ft edge spacing from all other fissile material containers , workstations , or fissile solution tanks, vessels, and ion exchange (IX) columns. Where LEU target material is handled in open containers , the design will prevent spills from readily spreading to an adjacent workstation or storage location. Design Basis Final workstation and fixture spacing will be determined in final design when all process upsets are evaluated.
Workstations with interaction controls include the following (not an all-inclusive listing):  
Workstations with interaction controls include the following (not an all-inclusive listing):  
* * * [Proprietary Information]  
* * * [Proprietary Information]
[Proprietary Information]
[Proprietary Information]
Target basket fixture that provides safe spacing of a batch of targets from one another in the target receipt cell Test Requirements The above analysis is based on information developed for the Construction Permit Application.
Target basket fixture that provides safe spacing of a batch of targets from one another in the target receipt cell Test Requirements The above analysis is based on information developed for the Construction Permit Application.
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NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features I 0 CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register, as amended. I 0 CFR 20.1201, "Occ upational Dose Limits for Adults," Code of Federal Regulations, Office of the Federal Register, as amended. IO CFR 20.1301, "Dose Limits for Individual Members of the Public," Code of Federal Regulations, Office of the Federal Register, as amended. 10 CFR 50.59, "Changes, Tests , and Experiments," Code of Federal Regulations , Office of the Federal Register, as amended. IO CFR 70.61, "Performance Requirements," Code of Federal Regulations, Office of the Federal Register, as amended. ANSI/ ANS-8.1, Nuclear Criticality Safety in Operations with Fissionable Material Outside of Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2014. ANSI/ ANS-8.3, Criticality Accident Alarm System, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois , I 997 (Reaffirmed in 2012). ANSI/ ANS-8. 7, Nuclear Criticality Safety in the Storage of Fissile Materia l s, American National Standards Institute/American Nuclear Society, La Grange Park, Illinoi s, 1998 (Reaffirmed in 2007). ANSI/ ANS-8.10, Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement, American National Standards Institute/American Nuclear Society, La Grange Park, Illinoi s, 2015. ANSI/ANS-8.19, Administrative Practices for Nuclear Criticality Safety, American National Standards Institute/American Nuclear Society, La Grange Park , Illinois, 2014. ANSI/ ANS-8.20, Nuclear Criticality Safety Training, American National Standards Institute/ American Nuclear Society, La Grange Park , Illinois, 1991 (Reaffirmed in 2005). ANSI/ANS-8.22 , Nuclear Criticality Saf e ty Based on Limiting and Controlling Moderators , American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 1997 (Reaffirmed in 2011 ). ANSI/ ANS-8.23, Nuclear Criticality Accident Emergency Planning and Response , American National Standards Institute/American Nuclear Society, La Grange Park , Illinoi s, 2007 (Reaffirmed in 2012). ANSI/ANS-8.24, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, American National Standards Institute/ American Nuclear Society , La Grange Park , Illinois, 2007 (Reaffirmed in 2012). ANSI/ANS-8.26, Criticality Safety Engineer Training and Qualification Program , American National Standards Institute/ American Nuclear Society , La Grange Park, Illinoi s, 200 7 (Reaffirmed in 2012). ANSI/ ANS-15 .1 , The Development of Technical Specifications for Research Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2013. ANS I Nl 3.1, Sampling and Monitoring Releases of Airborne Radioacti ve Substances from the Stacks and Ducts of Nuclear Facilities, American Nuclear Society, La Grange Park, Illinois, 2011. 6-72   
NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features I 0 CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register, as amended. I 0 CFR 20.1201, "Occ upational Dose Limits for Adults," Code of Federal Regulations, Office of the Federal Register, as amended. IO CFR 20.1301, "Dose Limits for Individual Members of the Public," Code of Federal Regulations, Office of the Federal Register, as amended. 10 CFR 50.59, "Changes, Tests , and Experiments," Code of Federal Regulations , Office of the Federal Register, as amended. IO CFR 70.61, "Performance Requirements," Code of Federal Regulations, Office of the Federal Register, as amended. ANSI/ ANS-8.1, Nuclear Criticality Safety in Operations with Fissionable Material Outside of Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2014. ANSI/ ANS-8.3, Criticality Accident Alarm System, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois , I 997 (Reaffirmed in 2012). ANSI/ ANS-8. 7, Nuclear Criticality Safety in the Storage of Fissile Materia l s, American National Standards Institute/American Nuclear Society, La Grange Park, Illinoi s, 1998 (Reaffirmed in 2007). ANSI/ ANS-8.10, Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement, American National Standards Institute/American Nuclear Society, La Grange Park, Illinoi s, 2015. ANSI/ANS-8.19, Administrative Practices for Nuclear Criticality Safety, American National Standards Institute/American Nuclear Society, La Grange Park , Illinois, 2014. ANSI/ ANS-8.20, Nuclear Criticality Safety Training, American National Standards Institute/ American Nuclear Society, La Grange Park , Illinois, 1991 (Reaffirmed in 2005). ANSI/ANS-8.22 , Nuclear Criticality Saf e ty Based on Limiting and Controlling Moderators , American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 1997 (Reaffirmed in 2011 ). ANSI/ ANS-8.23, Nuclear Criticality Accident Emergency Planning and Response , American National Standards Institute/American Nuclear Society, La Grange Park , Illinoi s, 2007 (Reaffirmed in 2012). ANSI/ANS-8.24, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, American National Standards Institute/ American Nuclear Society , La Grange Park , Illinois, 2007 (Reaffirmed in 2012). ANSI/ANS-8.26, Criticality Safety Engineer Training and Qualification Program , American National Standards Institute/ American Nuclear Society , La Grange Park, Illinoi s, 200 7 (Reaffirmed in 2012). ANSI/ ANS-15 .1 , The Development of Technical Specifications for Research Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2013. ANS I Nl 3.1, Sampling and Monitoring Releases of Airborne Radioacti ve Substances from the Stacks and Ducts of Nuclear Facilities, American Nuclear Society, La Grange Park, Illinois, 2011. 6-72   
.:;.-.;* .. NWMI ............ ............ *
.:;.-.;* .. NWMI ............ ............ *
* NOITifWEn MEIUCAL. 1$0TOP£S NWMl-2 0 13-02 1, R e v. 1 Ch apter 6.0 -E n gineered Safety Featu r e s ASME AG-1, Code on Nuclear Air and Gas Treatment, American Society of Mec h anica l Engineers, New York , New York, 2003. LA-CP-13-00634, MCNP6 User Manual , Rev. 0, Los A l amos Nationa l Laboratory, Los A l amos , New Mexico, May 2013. NRC , 2012, Fina l Interim Staff Guidance Augmenting NUREG-1537, "Guidelines for Preparing and R ev iewing Applications for the Licensing of Non-Power Reactors , " Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors, Docket Number: NRC-201 1-0135, U.S. Nuc l ear Regu l atory Commission, Washington , D.C., Octo b er 30, 20 1 2. NUREG-1520 , Standard R e view Plan for the Review of a Licens e Application for a Fuel Cycle Facility , Rev. 1, U.S. Nuclear Regulatory Commission , Office of Nuc l ear Material Safety and Safeguards , Washington , D.C., May 2010. NUREG-153 7 , Guidelines for Preparing and Revi ewi ng Applications for the Lic e nsing of Non-Power Reactors -Format and Content , Part 1 , U.S. Nuc l ear Regulatory Commission, Office of Nuclear Reactor Regulation , Was h ington , D.C., February 1996. NUREG/CR-4604 I PNL-5849, Statistical Methods for Nuclear Material Manag e ment , Pacific Northwest Laboratory, Richland, Wash i ngto n , December , 1988. NUREG/CR-6698, Guide for Validation of Nuclear Criticality Saf ety Calculational Methodology , U.S. Nuclear Regulatory Commission, Office of Nuclear Materia l Safety and Safeguards, Washington , D.C., January 2001. [Proprietary Information]  
* NOITifWEn MEIUCAL. 1$0TOP£S NWMl-2 0 13-02 1, R e v. 1 Ch apter 6.0 -E n gineered Safety Featu r e s ASME AG-1, Code on Nuclear Air and Gas Treatment, American Society of Mec h anica l Engineers, New York , New York, 2003. LA-CP-13-00634, MCNP6 User Manual , Rev. 0, Los A l amos Nationa l Laboratory, Los A l amos , New Mexico, May 2013. NRC , 2012, Fina l Interim Staff Guidance Augmenting NUREG-1537, "Guidelines for Preparing and R ev iewing Applications for the Licensing of Non-Power Reactors , " Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors, Docket Number: NRC-201 1-0135, U.S. Nuc l ear Regu l atory Commission, Washington , D.C., Octo b er 30, 20 1 2. NUREG-1520 , Standard R e view Plan for the Review of a Licens e Application for a Fuel Cycle Facility , Rev. 1, U.S. Nuclear Regulatory Commission , Office of Nuc l ear Material Safety and Safeguards , Washington , D.C., May 2010. NUREG-153 7 , Guidelines for Preparing and Revi ewi ng Applications for the Lic e nsing of Non-Power Reactors -Format and Content , Part 1 , U.S. Nuc l ear Regulatory Commission, Office of Nuclear Reactor Regulation , Was h ington , D.C., February 1996. NUREG/CR-4604 I PNL-5849, Statistical Methods for Nuclear Material Manag e ment , Pacific Northwest Laboratory, Richland, Wash i ngto n , December , 1988. NUREG/CR-6698, Guide for Validation of Nuclear Criticality Saf ety Calculational Methodology , U.S. Nuclear Regulatory Commission, Office of Nuclear Materia l Safety and Safeguards, Washington , D.C., January 2001. [Proprietary Information]
[Proprietary Information]
[Proprietary Information]
NWMI-2015-SD D-013, S yste m Design D esc ription for V e ntilation, Rev. A, Northwest Medical Isoto p es , LLC, Corva ll is , Oregon , 2015. NWMI-2015-CRITCALC-001, Sing l e Param e ter Sub c riti c al Limits for 20 wt% 235 U-Uranium Metal , Uranium Oxide , and Homogenous Water Mixtures , Rev. A , Northwest Medica l Isotopes , LLC , Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-002, Irradiated Target Low-Enriched Uranium Material Dissolution , Rev. A Northwest Medical Isotopes , LLC , Corva ll is , Oregon , 2015. NWMI-2015-CRITCALC-003, 55-Gallon Drum Arrays, Rev. A Northwest Medical Isotopes, LLC, Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-005 , Target Fabrication Tanks , W e t Pro cesse s , and Storage , Rev. A , Northwest Medical Isotopes , LLC , Corvallis, Oregon , 2015. NWMI-2015-CRITCALC-006, Tank Hot Ce ll , Rev. A , Northwest Medica l Isotopes , LLC , Corva ll is , Oregon, 2015. NWM I-2015-CSE-001, NWMI Preliminary Criticality Safety Evaluation:
NWMI-2015-SD D-013, S yste m Design D esc ription for V e ntilation, Rev. A, Northwest Medical Isoto p es , LLC, Corva ll is , Oregon , 2015. NWMI-2015-CRITCALC-001, Sing l e Param e ter Sub c riti c al Limits for 20 wt% 235 U-Uranium Metal , Uranium Oxide , and Homogenous Water Mixtures , Rev. A , Northwest Medica l Isotopes , LLC , Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-002, Irradiated Target Low-Enriched Uranium Material Dissolution , Rev. A Northwest Medical Isotopes , LLC , Corva ll is , Oregon , 2015. NWMI-2015-CRITCALC-003, 55-Gallon Drum Arrays, Rev. A Northwest Medical Isotopes, LLC, Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-005 , Target Fabrication Tanks , W e t Pro cesse s , and Storage , Rev. A , Northwest Medical Isotopes , LLC , Corvallis, Oregon , 2015. NWMI-2015-CRITCALC-006, Tank Hot Ce ll , Rev. A , Northwest Medica l Isotopes , LLC , Corva ll is , Oregon, 2015. NWM I-2015-CSE-001, NWMI Preliminary Criticality Safety Evaluation:
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==SUMMARY==
==SUMMARY==
DESCRIPTION The Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF) preliminary instrumentation and control (I&C) configuration includes the special nuclear material (SNM) preparation and handling processes (e.g., target fabrication , and uranium recovery and recycle), radioisotope extraction and purification processes (e.g., target receipt and disassembly , target dissolution , molybdenum  
DESCRIPTION The Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF) preliminary instrumentation and control (I&C) configuration includes the special nuclear material (SNM) preparation and handling processes (e.g., target fabrication , and uranium recovery and recycle), radioisotope extraction and purification processes (e.g., target receipt and disassembly , target dissolution , molybdenum
[Mo] recovery and purification , and waste handling), process utility systems , criticality accident alarm system (CAAS), and sy s tems associated with radiation monitoring.
[Mo] recovery and purification , and waste handling), process utility systems , criticality accident alarm system (CAAS), and sy s tems associated with radiation monitoring.
The SNM processes will be enclosed predominately by hot cells and glovebox designs except for the target fabrication area. The facility process control (FPC) system will provide monitoring and control of the process systems within the RPF. In addition, the FPC system will provide monitoring of related components within the RPF. The process strategy for the RPF involves the use of batch or batch processes with relatively simple control steps. The building management system (BMS) (a subset of the FPC system) will monitor the RPF ventilation system and mechanical utility systems. The BMS primary functions will be to monitor the facility ventilation system and monitor and control (turn on and off) the mechanical utility systems. Engineered safety feature (ESF) systems will operate on actuation of an alarm setpoint reached for a specific monitoring instrument/device. For redundancy , this will be in addition to the FPC system or BMS ability to actuate ESF as needed. Each ESF safety function will use hard-wired analog controls/interlocks to protect workers, the public , and environment.
The SNM processes will be enclosed predominately by hot cells and glovebox designs except for the target fabrication area. The facility process control (FPC) system will provide monitoring and control of the process systems within the RPF. In addition, the FPC system will provide monitoring of related components within the RPF. The process strategy for the RPF involves the use of batch or batch processes with relatively simple control steps. The building management system (BMS) (a subset of the FPC system) will monitor the RPF ventilation system and mechanical utility systems. The BMS primary functions will be to monitor the facility ventilation system and monitor and control (turn on and off) the mechanical utility systems. Engineered safety feature (ESF) systems will operate on actuation of an alarm setpoint reached for a specific monitoring instrument/device. For redundancy , this will be in addition to the FPC system or BMS ability to actuate ESF as needed. Each ESF safety function will use hard-wired analog controls/interlocks to protect workers, the public , and environment.
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* Double-contingency principle  
* Double-contingency principle  
-Process designs should incorporate sufficient factors of safety to require at least two unlikely, independent , and concurrent changes in process conditions before a criticality accident is possible (baseline design criteria of 10 CFR 70.64 , "Requirements for New Facilities or New Processes at Existing Facilities
-Process designs should incorporate sufficient factors of safety to require at least two unlikely, independent , and concurrent changes in process conditions before a criticality accident is possible (baseline design criteria of 10 CFR 70.64 , "Requirements for New Facilities or New Processes at Existing Facilities
," paragraph  
," paragraph
[9]). The safety program will ensure that each IROFS will be available and reliable to perform its intended function when needed and in the context of the performance requirements of this section (10 CFR 70.61, " Performance Requirements
[9]). The safety program will ensure that each IROFS will be available and reliable to perform its intended function when needed and in the context of the performance requirements of this section (10 CFR 70.61, " Performance Requirements
," paragraph  
," paragraph
[e]). The FPC system trip and alarm annunciation are protective functions and will be part of the overall protection and safety monitoring systems for the RPF. The specific equipment design basis for the instrumentation and equipment used for the FPC system trip and alarming functions is discussed in Section 7.2.2. The following discussion relates to the design basis used for monitoring specific signal values for RPF trips and alarms , requirements for performance , requirements for specific modes of operation of the RPF and the FPC system , and the general design criteria noted in Table 7-1. 7.2.4.1.1 Safety Functions Corresponding Protective or Mitigative Actions for Design Basis Events IEEE 603-2009, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations (Sections 4a and 4b). The results of the integrated safety analysis (ISA) for the RPF structures , systems, and components (SSC) are discussed in Chapter 13.0 , " Accident Analysis." Conditions that require monitoring and the subsequent action to be taken are described in Chapter 13.0. 7-16   
[e]). The FPC system trip and alarm annunciation are protective functions and will be part of the overall protection and safety monitoring systems for the RPF. The specific equipment design basis for the instrumentation and equipment used for the FPC system trip and alarming functions is discussed in Section 7.2.2. The following discussion relates to the design basis used for monitoring specific signal values for RPF trips and alarms , requirements for performance , requirements for specific modes of operation of the RPF and the FPC system , and the general design criteria noted in Table 7-1. 7.2.4.1.1 Safety Functions Corresponding Protective or Mitigative Actions for Design Basis Events IEEE 603-2009, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations (Sections 4a and 4b). The results of the integrated safety analysis (ISA) for the RPF structures , systems, and components (SSC) are discussed in Chapter 13.0 , " Accident Analysis." Conditions that require monitoring and the subsequent action to be taken are described in Chapter 13.0. 7-16   
.;.-.;* .. NWMI ..*... ..* .... ..... .... .. ' !*. * ! .*
.;.-.;* .. NWMI ..*... ..* .... ..... .... .. ' !*. * ! .*
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* * * *  
* * * *  
* * * * *
* * * * *
* Fresh uranium receipt and dissolution Nitrate extraction Ac id-d eficient uranyl nitrate (ADUN) concentration  
* Fresh uranium receipt and dissolution Nitrate extraction Ac id-d eficient uranyl nitrate (ADUN) concentration
[Proprietary Information]  
[Proprietary Information]
[Propr ietary Information]  
[Propr ietary Information]
[Propri etary Information]
[Propri etary Information]
Target fa bric a tion waste Target assem bl y [Proprietary Information]
Target fa bric a tion waste Target assem bl y [Proprietary Information]
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* 7.3.3.1 Cask receipt Target receipt Target disassembly Design Criteria Design criteria for the target receipt and disassembly I&C system s are described in Section 7.2. 7.3.3.2 Design Basis and Safety Requirements The design basis and safety requirements for the target receipt and disassembly I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.3.3.3 System Description The target receipt and disassembly I&C system will be defined in the Operating License Application.
* 7.3.3.1 Cask receipt Target receipt Target disassembly Design Criteria Design criteria for the target receipt and disassembly I&C system s are described in Section 7.2. 7.3.3.2 Design Basis and Safety Requirements The design basis and safety requirements for the target receipt and disassembly I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.3.3.3 System Description The target receipt and disassembly I&C system will be defined in the Operating License Application.
The strategy and associated parameters for the l&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations , which sets the requirements for the process monitoring and contro l equipment, and the associated instrumentation.
The strategy and associated parameters for the l&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations , which sets the requirements for the process monitoring and contro l equipment, and the associated instrumentation.
7-32
7-32
::.**.*.*.* ..... .. NWMI ........... *. ! ! ." NORTHWEST Mt:DacA.L ISOT01'£S NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Normal operating functions will be performed remotely using the FPC system RMI in the truck bay , cask preparation airlock , and the operating gallery. Redundant control functions will be provided in the control room. In addition, the implementation of IROFS CS-14 , CS-15 , CS-20 , CS-27, and RS-10 interlocks for this system are under development.
::.**.*.*.* ..... .. NWMI ........... *. ! ! ." NORTHWEST Mt:DacA.L ISOT01'£S NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Normal operating functions will be performed remotely using the FPC system RMI in the truck bay , cask preparation airlock , and the operating gallery. Redundant control functions will be provided in the control room. In addition, the implementation of IROFS CS-14 , CS-15 , CS-20 , CS-27, and RS-10 interlocks for this system are under development.
Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments, control logic and elements , indication , alarm, and control features will be developed for the Operating License Application.
Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments, control logic and elements , indication , alarm, and control features will be developed for the Operating License Application.
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* Valve position S olid wa s t e
* Valve position S olid wa s t e
* Ac tuat e gr out mix e r (M)
* Ac tuat e gr out mix e r (M)
* Pr ess ur e e n c ap s ul a tion High-dose waste decay Hi g h-do se was t e h a ndlin g TB D = t o b e d e t e nnin e d. TBD TBD 7-41 TBD TBD Primary control location C ontrol ro o m Low dose solidification room C ontr o l room Control room L ow d ose s olidific a tion room Low dose solidification room L ow do se s olidific a tion room Low dose solidification room Low do se so lidifi cat i on ro o m
* Pr ess ur e e n c ap s ul a tion High-dose waste decay Hi g h-do se was t e h a ndlin g TB D = t o b e d e t e nnin e d. TBD TBD 7-41 TBD TBD Primary control location C ontrol ro o m Low dose solidification room C ontr o l room Control room L ow d ose s olidific a tion room Low dose solidification room L ow do se s olidific a tion room Low dose solidification room Low do se so lidifi cat i on ro o m
::.**.*.*. .; ... NWMI ........ *.* ' ! : . NOlllfWHT MEDtCAl ISOTOPf.S NWM l-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-12 provides a preliminary listing of the waste handling system interlocks and permissive signals that have been identified.
::.**.*.*. .; ... NWMI ........ *.* ' ! : . NOlllfWHT MEDtCAl ISOTOPf.S NWM l-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-12 provides a preliminary listing of the waste handling system interlocks and permissive signals that have been identified.
These devices will be further developed and detailed information will be provided in the Operating License Application.
These devices will be further developed and detailed information will be provided in the Operating License Application.
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* NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems The alarms will consist of the following ca pabiliti es: * * * "A lert li g ht" will illuminate when the radiation level exceeds preset limit s with an adjustable se tpoint " High alarm red light" will illuminate when radiation level s exceed a predetermined alarm setpoint "Fai lure alarm" will sou nd when either the power or the channel's electronics fail The visual alarms will be accompanied by a simultaneous audible alarm annunciator at the selected detector locations and in the control room. The annunciator windows for the monitors will be located in the control room. The alarm can be manually re set when the alarm conditions are corrected.
* NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems The alarms will consist of the following ca pabiliti es: * * * "A lert li g ht" will illuminate when the radiation level exceeds preset limit s with an adjustable se tpoint " High alarm red light" will illuminate when radiation level s exceed a predetermined alarm setpoint "Fai lure alarm" will sou nd when either the power or the channel's electronics fail The visual alarms will be accompanied by a simultaneous audible alarm annunciator at the selected detector locations and in the control room. The annunciator windows for the monitors will be located in the control room. The alarm can be manually re set when the alarm conditions are corrected.
The local a larm horn s and warning lights will remain on until the radiation le ve l is belo w the pre se nt le ve l. Additional CAM requirements and locations are described in Chapter 11.0. 7.6.3.2 Stack Release Monitoring The exhaust stacks will be provided with continuous monitors for noble gases , particulate , and iodine. The stack monitoring system design ba sis is to continuously monitor the radioactive stack relea ses. Additional information will be provided in the Operating License Application. Airborne exposure pathway monitoring is described in Chapter 11.0. 7.6.4 System Performance Analysis and Conclusions The system performance analysis and conclusions for eac h process system will be provided in the Operating License Application.
The local a larm horn s and warning lights will remain on until the radiation le ve l is belo w the pre se nt le ve l. Additional CAM requirements and locations are described in Chapter 11.0. 7.6.3.2 Stack Release Monitoring The exhaust stacks will be provided with continuous monitors for noble gases , particulate , and iodine. The stack monitoring system design ba sis is to continuously monitor the radioactive stack relea ses. Additional information will be provided in the Operating License Application. Airborne exposure pathway monitoring is described in Chapter 11.0. 7.6.4 System Performance Analysis and Conclusions The system performance analysis and conclusions for eac h process system will be provided in the Operating License Application.
The overall I&C syste m performance analysis i s provided in Section 7.2. 7-49
The overall I&C syste m performance analysis i s provided in Section 7.2. 7-49
:;.**.*.* .. .. .. NWMI ........ *.* .  "NOflT H WHT MEDtCALISOTOPU  
:;.**.*.* .. .. .. NWMI ........ *.* .  "NOflT H WHT MEDtCALISOTOPU  


Revision as of 17:06, 26 April 2019

NWMI-2013-021, Revision 1, Construction Permit Application for Radioisotope Production Facility, Chapters 3.0, 6.0, 7.0, 8.0, 9.0 and 13.0, Attachment 3
ML17193A428
Person / Time
Site: Northwest Medical Isotopes
Issue date: 06/30/2017
From:
Northwest Medical Isotopes
To:
Office of Nuclear Reactor Regulation
Shared Package
ML17193A418 List:
References
NWMI-LTR-2017-007 NWMI-2013-021, Rev. 1
Download: ML17193A428 (243)
v • d • e


Text

.. ;.:: .. NWMI ..... . *.* .. *:. !* NORTHWEST MEDICAL ISOTOPES ATTACHMENT 3 Northwest Medical Isotopes, LLC Docket No. 50-609 Construction Permit Application for Radioisotope Production Facility Chapters 3.0, 6.0, 7.0, 8.0, 9.0 , and 13.0 (Document No. NWMl-2013-021, Rev. 1, June 2017) Public Version Information is being provided via hard copy

  • * * * * * * * * ****** * * ** ** * ** * ** * * * ** * ** ** ** * * . *. *. * . NORTHWEST MEDICAL ISOTOPES *
  • Chapter 3.0 -Design of Structures, Systems, and Components Prepared by: Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 June 2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, Oregon 97330 This page intentionally left blank.

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  • NORTHWEST MEDtcAL ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Chapter 3.0 -Design of Structures, Systems, and Components Construction Permit Application for Radioisotope Production Facility NWMl-2013-021 , Rev. 1 Date Published:

June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 3.0 -Design of Structures, Systems and Components Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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.;.-.;* .. NWMI ...*.. ... .... ........... * * . NOITHWEST MlDICAl ISOTOPCS NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components CONTENTS 3.0 DESIGN OF STRUCTURES , SYSTEMS, AND COMPONENTS

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........................... 3-1 3.1 Design Criteria .......................................

............................................................................ 3-4 3. l.1 Radioisotope Production Facility Structures, Systems, and Components

............ 3-4 3.1.2 Code of Federal Regulations

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...... 3-8 3.1.3 U.S. Nuclear Regulatory Commission

................................................................. 3-8 3.1.4 Other Federal Regulations , Guidelines , and Standards

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3-10 3.1.5 Local Government Documents

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3-10 3.1.6 Discovery Ridge/University of Missouri ........................................................... 3-11 3 .1. 7 Codes and Standards

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... 3-12 3.2 Meteorological Damage ............................

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3-24 3.2.1 Combinations of Loads ................

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3-25 3.2.1.1 Nuclear Safety-R elated Structures , Systems , and Components

........ 3-26 3 .2.1.2 Commercial and Nuclear Non-Safety-Related Structures, Systems, and Components

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3-26 3 .2.2 Combinations for Serviceability Based Acceptance Criteria .............................

3-27 3.2.3 Norma l Loads .......................................

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.............. 3-27 3.2.4 Wind Loading ................................

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......... 3-30 3.2.4.1 Wind Load ....................................................

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3-30 3.2.4.2 Tornado Loading ...........................................................................

.... 3-30 3.2.4.3 Effect of Fai lure of Structures, Systems , or Components Not Designed for Tornado Loads .........................................................

...... 3-32 3.2.5 Rain , Snow, and Ice Loading ................................................

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3-32 3.2.5.1 Rain Loads ..............................

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3-32 3.2.5.2 Snow Load ..........................

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... 3-33 3.2.5.3 Atmospheric Ice Load ......................

................................................. 3-34 3.2.6 Operating Thermal/Self-Straining Loads ..............................................

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3-34 3.2.7 Operating Pipe Reaction Loads .......................................

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............ 3-34 3.2.8 External Hazards ...................

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3-34 3.3 Water Damage .......................

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3-35 3.3.1 Flood Protection

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....... 3-35 3.3.1.1 Flood Protection Measures for Structures, Systems, and Components

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.. 3-35 3.3.1.2 Flood Protection from Externa l Sources ....................

........................ 3-36 3.3.1.3 Compartment Flooding from Fire Protection Discharge

.................... 3-37 3. 3 .1 .4 Compartment Flooding from Postulated Component Failures ........... 3-3 7 3.3.1.5 Permanent Dewatering System .........................

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............. 3-37 3.3.1.6 Structural Design for Flooding ..................................

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3-37 3.4 Seismic Damage .....................................................

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..... 3-38 3.4.1 Seismic Input. ..........

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................ 3-38 3.4.1.1 Design Response Spectra ..........................................

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.. 3-38 3.4.1.2 Method of Analysis ..........

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.............................. 3-39 3 .4.2 Seismic Qualification of Subsystems and Equipment...

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.... 3-40 3.4.2.1 Qualification by Analysis .........................

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3-40 3.4.2.2 Qualification by Testing ...........

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....................... 3-41 3.4.3 Seismic Instrumentation

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.............. 3-41 3.4.3.1 Location and Description

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...... 3-42 3.4.3.2 Operability and Characteristics

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3.5 Systems

and Components

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........ 3-43 3.5.1 General Design Basis Information

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....................................................... 3-43 3.5.1.1 Classification of Systems and Components Important to Safety ........ 3-4 3 3.5.1.2 C l assification Definitions

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3-43 3.5.1.3 Nuclear Safety C l assifications for Structures, Systems, and Components

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......... 3-44 3.5.2 Radioisotope Production Facility .............

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......................................... 3-47 3.5.2.1 System Classification

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3-5 2 3.5.2.2 C l assification of Systems and Components Important to Safety ...... 3-52 3.5.2.3 De sign Basis Functions, Va lu es, and Criteria .....................

.............. 3-54 3.5.2.4 System Functions/Safety Functions

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................. 3-54 3.5.2.5 Systems and Compo n ents ...............................................

.................. 3-54 3.5.2.6 Qualification Methods ..................................................

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3-55 3.5.2. 7 Radioi soto pe Production Facility Specific Sy s tem Design Basis Functions and Values ...........................................

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........ 3-55 3.6 References

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  • NOITNWEST llEDICAl tSGTOPCS NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-1. Table 3-2. Table 3-3. Table 3-4. Table 3-5. Table 3-6. Table 3-7. Table 3-8. Table 3-9. Table 3-10. Table 3-11. Table 3-12. Table 3-13. Table 3-14. Table 3-15. Table 3-16. Table 3-17. Table 3-18. Table 3-19. Table 3-20. Table 3-21. Table 3-22. Table 3-23. Table 3-24. Table 3-25. TABLES List of Sy s tem and Associated Systems and Construction Permit Application Crosswalk (2 pages) ......................

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............................... 3-4 Summary of Items Relied on for Safety Identified by Accident Analyses (3 page s) ....................................................

....................................................................... 3-5 Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) .............................. 3-8 Other Federal Regulation s, Guideline s, and Standards

.................................................. 3-10 Local Go v ernment Document s (2 pa g e s) .......................................................

................ 3-1 0 Discovery Ridge/University of Missouri Requirements

................................................ 3-11 Design Codes and Standards (12 page s) ........................................................................ 3-12 Load Symbol Definitions (2 pages) ..............................................................

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.. 3-24 Load Combinations for Strength Based Acceptance Criteria , Nuclear Safety-Related ............

............................................................................................................... 3-26 Load Combinations for Strength Base Acceptance Criteria , Commercial

.................... 3-27 Load Combinations for Serviceabilit y Ba s ed Acceptance C riteria ................................ 3-27 Lateral Earth Pressure Loads ......................................................................................... 3-2 8 Floor Live Loads ............................................................................................................ 3-29 Crane Lo a d Criteria ....................................................................................................... 3-29 Wind Loading Criteria .....................

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.... 3-30 Design-B as i s Tornado Field Characteri s tics .................................................................. 3-31 Design-Ba s is Tornado Mi s sile Spectrum ....................................................................... 3-3 2 Rain Load C riteria ......................................................................................................... 3-3 3 Snow Load C riteria ........................................................................................................ 3-33 Extreme Winter Precipitation Load Criteria ....................................................

.............. 3-34 Atmospheric Ice Load Criteria ....................................................................................... 3-34 De s ign Criteria Requirements ( 4 page s) ...........

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............ 3-4 7 Sy s tem Cl ass ification s .......................................................................................

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3-5 2 System S a fety and Seismic Cla s sification and Associated Quality Level Group (2 pages) ......................................................................................................................... 3-5 2 Likelihood Index Limit Guidelines ................................................................................ 3-5 3 3-iii

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  • NOATHWfSTMEDtCAllSOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components TERMS Acronyms and Abbreviations 99 Mo molybdenum-99 AASHTO American Association of State Highway and Transportation Officials ACGIH American Conference on Governmental Industrial Hygienists ACI American Concrete Institute AHRI Air Conditioning, Heating and Refrigeration Institute AISC American Institute of Steel Construction ALARA as low as reasonably achievable AMCA Air Movement and Control Association ANS American Nuclear Society ANSI American National Standards Institute ASCE American Society of Civil Engineers ASHRAE American Society of Heating , Refrigeration , and Air-Conditioning Engineers ASME American Society of Mechanical Engineers ASNT American Society for Nondestructive Testing ASTM American Society for Testing and Materials A WS American Welding Society BMS building management system CDC Centers for Disease Control and Prevention CFR Code of Federal Regulations CRR Collected Rules and Regulations CSR Missouri Code of State Regulations Discovery Ridge Discovery Ridge Research Park DBE design basis event DBEQ design basis earthquake DOE U.S. Department of Energy EIA Electronic Industries Alliance ESF engineered safety feature FEMA Federal Emergency Management Agency FPC facility process control FSAR final safety analysis report H 2 hydrogen gas HR hydrometeorological report HY AC heating , ventilation , and air conditioning I&C instrumentation and control IAEA International Atomic Energy Agency IBC International Building Code ICC International Code Council ICC-ES International Code Council Evaluation Service IEEE Institute of Electrical and Electronics Engineers IES Illuminating Engineering Society IFC International Fire Code IROFS items relied on for safety ISA International Society of Automation ISG Interim Staff Guidance IX ion exchange LEU low enriched uranium MDNR Missouri Department of Natural Resources Mo molybdenum 3-iv

..**.. ; .. ;. NWMI ..* ... ........... !' 0 llOITNW£ST tsOTOPU MO DOT MRI MU NECA NEMA NEP NESHAP NETA NFPA NIOSH NOAA NRC NS NSR NWMI NWS PMF PMP PMWP QA QA PP RCA RPF SEP SMACNA SNM SR SSC TIA U.S. UL UPS USGS Units o c O f µ cm cm 2 ft ft2 ft 3 g gal hp hr m. in.2 kg kip NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Missouri Department of Transportation mean recurrence interval University of Missouri National Electrical Contractors Association National Electrical Manufacturers A s sociation normal electrical power National Emissions Standard s for Hazardous Air Pollutants InterNational Electrical Te s ting Association National Fire Protection Association National Institute for Occupational Safety and Health National Oceanic and Atmospheric Administration U.S. Nuclear Regulatory Commission non-seismic non-safety-related Northwest Medical Isotopes , LLC National Weather Service probable maximum flood probable maximum precipitation probable ma x imum winter precipitation q u a lit y as s ura n ce q u a li t y assura n ce p ro gram p l a n radiologically controlled area Radioisotope Production Facility standby electrical power Sheet Metal and Air Conditioning Contractors National Association special nuclear material safety related structure s, s ystems and components Telecommunications Indu s try A s sociation United State s Underwriters Laboratory uninterruptible power s upply U.S. Geological Survey degrees Celsius degrees Fahrenheit micron centimeter s quare centimeters feet square feet cubic feet acceleration of gravity gallon horsepower hour inch square inch kilogram thousand pounds-force 3-v

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  • NOITHWUT MEDICAi. ISOTOf'll km kW L lb lbf m m 2 rru mi2 rrun MT rad sec kilometer kilowatt liter pound pound-force meter square meter mile square mile minute metric ton NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components absorbed radiation dose second 3-vi
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  • NORTHWEST MEDfCAl ISOTOP(S N WMl-2013-021, R ev. 1 Chapter 3.0 -Des i gn of Structures, Syste m s a n d Components

3.0 DESIGN

OF STRUCTURES, SYSTEMS, AND COMPONENTS This chapter identifies and describes the principal architectural and engineering design criteria for the facility structures , systems and components (SSC) for the Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF). The information presented emphasizes the safety and protective functions and related design features that help provide defense-in-depth against the uncontrolled re l ease of radioactive material to the environment.

The bases for the design criteria for some of the systems discussed in this chapter are developed in other chapters of the Construction Permit App l ication and are appropriately cross-referenced , when required. NWMI's RPF design is based on applicable standards , guides , codes , and criteria and provides reasonable assurance that the RPF SSCs , including electromechanical systems , are: * * * * * *

  • Built and will function as designed and required by the analyses in Chapter 13. 0 , "Accident Analysis" Built to have acceptable protection of the public health and safety and environment from radiologica l risks (e.g., radioactive materials , exposure) resulting from operations Protected against potential meteorological damage Protected against potential hydrologica l (water) damage Protected against seismic damage Provided surveillance activities and technical specifications required to respond to or mitigate consequences of seismic damage Based on technical specifications developed to ensure that safety-related functions of electromechanical systems and components will be operable and protect the health and safety of workers , the public , and environment The design of the RPF and SSCs are based on defense-in-depth practices. The NRC defines design-in-depth as the following:

An approach to d es igning and op e rating nucl e ar faciliti es that prevent s and mitigate s ac c id e nt s that rel e a se radiation or ha z ardou s materials. Th e k ey i s creating multipl e ind e p e ndent and r e dundant la ye r s of d e f e n se to c omp e n s at e for potential human and m ec hanical failur es s o that no s ingl e la y er , no matter how r obu s t , i s e x clu s iv e l y r e li e d upon. D efe n se in d e pth i nclud es th e u se of access control s, phy s i c al barri e r s, r e dundant and div e rs e k ey s af e ty funct i on s , and e m e rg e n cy r es pon se m e asur es. Defense-in-depth is a de s ign phi l osophy , applied from the outset and through completion of the design , that is based on providing successive levels of protection such that health and safety are not wholly dependent on any single element of the design , construction , maintenance , or operation of the facility. The net effect of incorporating defense-in

-depth pract i ces is a conservatively designed faci l ity and systems that exhibit higher to l erances to fa il ures and external challenges. The risk insights obtained through performance of accident ana l ysis can then be used to supplement the final design by focusing attention on the prevention and mitigation of the higher risk potential accidents.

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.. ... ; .. NWMI ...... ..* .... ..... .... .. * "* . NORTMWEST MUHCAL lSOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components This application to the U.S. Nuclear Regulatory Commi ss ion (NRC) seeks to obtain a license for a production facil i ty under Title l 0 , Code of F e d e ral R e gulation s (CFR), Part 50 (10 CFR 50), " Domestic Licensing of Production and Utilization Facilities." Embedded in the 10 CFR 50-licensed facility will be several activities subject to 10 CFR 70 , " Domestic Lic e nsing of Special Nuclear Material," to receive , possess , use , and transfer special nuclear material (SNM) and 10 CFR 30 , " Rule s of General Applicability to Domestic Licensing of Byproduct Material ," to proces s and transport molybdenum-99 (99 Mo) for medical application

s. This 10 CFR 50 license application for the RPF follow s the guidance in NUREG-1537 , Guid e lin es for Pr e paring and R ev i e wing A ppli c ation s for th e Li ce n si n g of Non Po we r R e a c t o r s -F o rmat and Cont e nt , that encompas s e s activities regulated under different NRC requirements (e.g., 10 C FR 70 and I 0 CFR 30), in accordance w ith 10 CFR 50.31 , " Combining Applications," and 10 CFR 50.32 , " Elimination o f Repetition

." The NRC has determined that a radioisotope separation and processing facility , which also conducts separation of SNM , will be considered a production facility and as such , will be s ubject to licensing under 10 CFR 50. The operation of the NWMI RPF will primarily be focused on the di s a ss embly of irradiated low-enriched uranium (LEU) targets , separation and purification of fi s sion product 99 Mo , and the recycle of LEU that is licen s ed under I 0 CFR 50. RPF operations will also include the fabrication of LEU targets , which will be licensed under 10 CFR 70. These targets will be shipped to NWMI's network of re s earch or test reactors for irradiation (con s idered a c onnected action) and returned to the RPF for proce ss ing. The LEU u s ed for the production of LEU tar g et material s will be obtained from the U.S. Department of Energy (DOE) and from LEU reclaimed from processing the irradiated targets. NWMI's licensing approach for the RPF defines the following unit processe s and areas that fall under the following NRC regulation s: * *

  • 10 CFR 50 , " Dome s tic Licensing of Production and Utilization Facilities" Target receipt and disassembly system Target dissolution system Molybdenum (Mo) recovery and purification system Uranium reco v ery and recycle system Waste management system Associated laboratory and support area s 10 CFR 70 , " Domestic Licensing of Special Nuclear Material" Target fabrication s ystem F r esh LEU (from DOE) receipt area A ss ociated laboratory and support areas 10 CFR 30 , " Rules of General Applicability to Domestic Licensing of Byproduct Material" Any byproduct materials produced or extracted in the RPF Design information for the complete range of normal operating conditions for various facility systems i s provided throughout the Construction Permit Application , and includes the following.
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  • RPF-specific design criteria (e.g., codes and standards , NRC guidelines) for SSCs are provided in Sections 3.1. NRC general design criteria and associated applicability to the RPF SSCs are addressed in Section 3.5. RPF description is presented in Chapter 4.0 , " R a dioisotope Production Facility Description

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  • Postulated initi ating events and credible accidents that form the design basis for the SSCs are discussed in Chapter 13.0. Potential ha zards and credib le accidents that cou ld be e ncount ered in the RPF during operations in volving SNM , irradiate d a nd unirrradiated , Mo recovery and purification , uranium recovery and recycle , waste management , and/or the use of hazardous chemica l s relative to these radiochemical processes that form the bases for the SSCs l ocated in the RPF , are discussed in C h apter 13.0. Design redundancy of SSCs to protect against unsafe conditions with respect to sing l e failures of engineered safety features (ESF) and contro l systems are described in Chapter 6.0, " Engineere d Safety Features ," and Chapter 7.0, "Instrumentation and Contro l System ," respectively. ESFs are described in C h apter 6.0 , and the administrative contro l s are discussed in Chapter 14.0 , "Technica l Specifications." Quality stan d ards comme n surate wit h the safety functions and potential risks that were used in the design of the SSCs are described in Table 3-7 (Section 3.1.7). Hydrological design bases d escribing the most severe predicted hydrological events during the lif e of the facility are provided in Chapte r 2.0 , " Site Characteristics

, Section 2.4. Design criteria for facility SSCs to withstand the most severe predicted hydrological events during the l ifetime of the facility are provided in Section 3.3. Seismic design bases for t h e facility are provided in Chapter 2.0 , Section 2.5. Seismic design criteria for the facility SSCs are prov id ed in Sectio n 3.4. Analyses co n cerning function , reliability , and maintainability of SSCs are described throughout th e Construction Permit Application. Meteorological de s ign bases describing the most severe weather extreme s predicted to occur during the life of the faci li ty are provided in Chapter 2.0, Section 2.3. Design criteria for faci li ty SSCs to withs t and the most severe weather extremes predicted to occur during t h e li fe of the facility are provided in Section 3.2. Potential condition s or other item s t h at wi ll be probable s ubjects of technical specifications associated wit h the RPF s tru ctures and design features are discussed in Chapter 14.0. 3-3

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  • NOkTNWlST lllDICAl tsOTDIO 3.1 DESIGN CRITERIA NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Section 3.1 describes the design criteria applied to the RPF and SSCs within the facility.

The principal design criteria for a production facility establish the necessary design, fabrication , construction, testing , and performance requirements for SSCs important to safety (i.e., those that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of workers and the public). The systems associated with the RPF are identified.

Those items relied on for safety (IROFS) are identified in Chapters 6.0 and 13.0. Requirements are derived from:

  • Code of Federal Regulations
  • U.S. Nuclear Regulatory Commission
  • Federal regulations , guidelines , and standards Local government regulations and requirements
  • Discovery Ridge Research Park (Discovery Ridge) covenants University of Missouri System (MU) requirements
  • Other codes and standards

3.1.1 Radioisotope

Production Facility Structures, Systems, and Components Table 3-1 lists the RPF systems and identifies the RPF material accountability area and the Construction Permit Application reference chapter that provides the associated detailed system descriptions.

Table 3-1. List of System and Associated Systems and Construction Permit Application Crosswalk (2 pages) Primary structure and associated systems . Construction Permit Application reference (primary references)

Radioisotope Production Facility (RPF -primary structure) 10 CFR 70" Target fabrication IO CFR sob Target receipt and disassembly Target dissolution Molybdenum recovery and purification Uranium recovery and recycle Waste handling Criticality accident alarm Radiation monitoring Normal electrical power Standby electrical power Process vessel ventilation Facility ventilation Fire protection Plant and instrument air Emergency purge gas Gas supply Chapter 4.0, Sections 4.1.3.1 and 4.4 Chapter 4.0, Section 4.1.3.2, 4.3.2, and 4.3.3 Chapter 4.0 , Sections 4.1.3.3 and 4.3.4 Chapter 4.0, Sections 4.1.3.4 and 4.3.5 Chapter 4.0 , Sections 4.1.3.5 and 4.3.6 Chapter 4.0, Section 4.1.3.6; Chapter 9.0, Section 9. 7.2 Ch a pter 6.0 , Section 6.3.3.1; Chapter 7.0 , Section 7.3.7 Chapter 7 .0, Section 7 .6; Chapter 11.0, Section 11.1.4 3-4 Chapter 8.0 , Section 8.1 Chapter 8.0, Section 8.2 Chapter 9.0 , Section 9.1 Chapter 9.0, Section 9.1 Chapter 9.0, Section 9.3 Chapter 9.0, Section 9.7.1 Chapter 6.0 , Section 6.2.1.7 .5 Chapter 9.0, Section 9. 7.1

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  • NOR'THWHT MEDtcAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-1. List of System and Associated Systems and Construction Permit Application Crosswalk (2 pages) Primary structure and associated systems Process chilled water Facility chilled water Facility heated water Process stream Demineralized wate r Chemical supply Biological shield Facility process control Construction Permit Application reference (primary references)

Chapter 9.0, Section 9.7.1 Chapter 9.0, Section 9.7. l C hapter 9 .0 , Section 9. 7. I Chapter 9.0, Section 9.7.l C hapter9.0 , Section 9.7.1 Chapter 9.0 , Section 9.7.4 C hapter 4.0 , Section 4.2 Chapter 7.0, Section 7.2.3

  • 10 CF R 70, " Dom est ic Licensing of S pe c i a l N ucl ea r Material ," Co d e of F ede ral R egu l ations, Office of th e Federal R eg i ste r , as a mended. b 10 CFR 50 , " Dome s tic Licensing of Production a nd U tili zat ion Fac iliti es," Code of F e deral R egu lati o n s, Office of th e Fe d eral R eg i ste r , as amended. In addition to Table 3-2 , NWMI-2015-LIST-003 , NWM I Radioi soto p e Produ ction Facility Ma ster Equipment Li st, provides a s ummary of the RPF systems , components, and equipment used in the RPF design. Table 3-2 provides a summary of the IROFSs identified by the accident analyses in Chapter 13.0, and a crosswalk to where the IROFSs are described in the Construction Permit Application.

Chapter 13.0 also provides the as s ociated detailed descriptions.

Table 3-2 also identifies whether the IROFS are considered ESFs or administrative controls.

Additional IROFS may be identified (or the current IROFS modified) during the RPF final design and development of the Operating Licen se Application. Table 3-2. Summary ofltems Relied on for Safety Identified by Accident Analyses (3 pages) IROFS Construction Permit Application designator Descriptor ESF AC crosswalk (primary references)

RS-01 RS-02 RS-03 RS-04 RS-05 RS-06 RS-07 RS-08 RS-09 Hot ce ll liquid co nfin e ment boundary Reserved*

Hot cell secondary confinement bound a ry Hot cell shielding boundary R eserve d* Reserved*

R ese rved*

Sample and analysis of low-dose waste tank dose rate prior to transfer outside the hot cell shielded boundary Prim a ry off gas r e lief sys tem 3-5 C h a pt er 6.0 , Sections 6.2.1.1 -6.2.1.6 Chapter 13.0, Section 13.2.2.8 Chapter 6.0, Sections 6.2.1.1 -6.2.1.6 C hapter 13.0 , Sections 13.2.2.8, 13.2.3.8 Chapter 6.0, Sections 6.2.1.1 -6.2.1.6 Chapter 13 .0 , Sections 13 .2.2.8 , 13 .2.4.8 Chapter 13.0, Section 13.2.7.1 C hapter 6.0 , Section 6.2.1.7 Chapter 13.0 , Section 13.2.3.8

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  • NORTHWEST MEOM:Al tsOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-2. Summary of Items Relied on for Safety Identified by Accident Analyses (3 pages) IROFS Construction Permit Application designator Descriptor ESF AC crosswalk (primary references)

RS-10 Active radiation monitoring and isolation of ./ Chapter 6.0, Section 6.2. I. 7 low-dose waste transfer Chapter 13.0, Section 13.2.7.1 RS-1 I Reserved" RS-12 Cask containment sampling prior to closure ./ Chapter 13.0, Section 13.2.7.l lid removal RS-13 Cask local ventilation during closure lid ./ Chapter 6.0 , Section 6.2.1.7 removal and docking preparations Chapter 13.0 , Section 13.2.7.1 RS-14 Reserved" RS-15 Cask docking port enabling sensor Chapter 6.0 , Section 6.2.1.7 Chapter 1 3.0 , Section 13.2.7.1 CS-01 Reserved" CS-02 Mass and batch handling limits for uranium Chapter 13.0 , Section 13.2.7.2 metal , uranium oxides , target s, and laboratory sample outside process sy s t e ms CS-03 Interaction control spacing provided by ./ Chapter 13.0, Section 13.2.7.2 administrative control CS-04 Interaction control spacing pro v ided by ./ Chapter 6.0 , Section 6.3.1.2 pa s sively de s ign e d fixtures a nd work s tation Chapter 13.0 , Section 13.2.7.2 placement CS-05 Container batch volume limit ./ Chapter 13.0, Section 13.2.7.2 CS-06 Pencil tank , ves s el , or piping safe geometr y ./ Chapter 6.0 , Section 6.3 .1.2 confinement u s ing the diameter of tanks , Chapter 13.0 , Section 13.2.4.8 vessels , or piping CS-07 Pencil tank and vessel spacing control using ./ Chapter 6.0, Section 6.3. I .2 fixed interaction spacing of individual tanks Chapter 13.0, Section 13.2.2.8 or vessels CS-08 Floor and sump geometry control of slab ./ Chapter 6.0 , Section 6.3.1.2 depth , sump diameter or depth for floor spill Chapter 13.0 , Section 1 3.2.2.8 containment berms CS-09 Double-wall piping ./ Chapter 6.0, Section 6.2. l. 7 Chapter 13.0, Section 13.2.2.8 CS-10 Closed safe geometry heatin g or cooling loop ./ Ch a pter 6.0, Section 6.3.1.2 with monitoring and alarm Chapter 13.0 , Section 13.2.4.8 CS-I I Simple overflow to normally empty safe ./ Chapter 6.0 , Section 6.3. I .2 geometry tank with level alarm Chapter 13.0, Section 13.2.7.2 CS-12 Condensing pot or s e al pot in ventil a tion vent ./ Chapter 6.0 , Section 6.3.1.2 line Chapter 13.0 , Section 13.2.7.2 CS-13 Simple overflow to normally empty safe ./ Chapter 6.0, Section 6.3. l.2 geometry floor with level alarm in the hot cell Chapter 13.0, Section 13.2.7.2 containment boundary CS-14 Active di s charge monitoring and isol a tion ./ Chapter 6.0 , Section 6.3.1.2 Chapter 13.0 , Section 13.2.7.2 3-6

..... .. NWMI ::.**.*.*.* ........... . . *****. * *. *

  • NORTlfWESl MEDICAi. lSOTOP£S NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-2. S ummary o f It e m s Re li e d on for Safety Id e ntifi ed by A ccid e nt A nal yses (3 pa ge s) IROFS Construction Permit Application designator Descriptor ESF AC crosswalk (primary references)

CS-1 5 Indepen d ent active d isc h arge mo n itoring and ,/ Chap t er 6.0 , Se ct ion 6.3.1.2 iso l ati o n C h apter 13.0 , Sec t ion 1 3.2.7.2 CS-1 6 Sa mplin g a nd a n a l ys i s o f u ra nium m ass or ,/ C hapt e r 1 3.0 , Se ction 1 3.2.7.2 co ncent ra ti on prior to di s ch a r ge o r dispo sa l CS-17 In d epen d e n t sam pling a nd ana l ysis of ,/ C h apte r 1 3.0 , Sectio n 13.2.7.2 urani um co n ce n tra ti on p rio r to d isc h arge or disposa l CS-1 8 B ac kflow pr eve nti on d ev i ce ,/ Cha pt e r 6.0 , Sect ion s 6.2.1.7 a nd 6.3.1.2 C h a pt e r 1 3.0 , Se ct i on 1 3.2.4.8 CS-1 9 Safe-geometry day ta n ks ,/ C h a p te r 6.0 , Sec t ion 6.3.1.2 C h a p te r 13.0 , Sec t io n 13.2.4.8 C S-2 0 Eva p o rator o r co n ce ntrat o r con d e n sa t e ,/ C h a pt er 6.0, Se cti o n 6.3.1.2 m o nitor i ng C h a pt e r 1 3.0 , Se ction 1 3.2.4.8 CS-2 1 Vis u a l inspection of access ib le surfaces fo r ,/ C h apter 13.0 , Sec t io n 1 3.2.7.2 foreign d e b ris CS-2 2 Gra m es tim a t or s u rvey of a cc ess ib le s urfa ces ,/ C h a pt e r 1 3.0 , Se cti o n 1 3.2.7.2 for ga mma a cti v it y CS-23 Non d est ru ctive assay of i t ems with ,/ C h a p te r 1 3.0, Sectio n 13.2.7.2 inaccessi b le s urfaces CS-2 4 Ind e p e nd e nt n o nd es tru c ti ve a ssay o f it e m s ,/ C h a pt e r 1 3.0, S ec ti o n 1 3.2.7.2 w ith ina ccess ible s urfa ces CS-25 Target h o u sing weighi n g p rior to d isposa l ,/ C h apter 13.0 , Section 1 3.2.7.2 CS-26 Proc ess i n g co mp o n e n t safe vo l um e ,/ C h a pt e r 6.0 , Sec t i on 6.3 .1.2 co nfin e m e nt C hapt e r 1 3.0 , Sec t i o n 1 3.2.7.2 CS-2 7 Closed beating or co o li n g l oop wit h ,/ C h apter 6.0 , Sec t ion 6.3.1.2 mo n itori ng a n d alarm C h a p te r 13.0 , Sec t ion 13.2.4.8 F S-0 1 E nh a n ce d lift p roce d u r e ,/ C h a pt e r 1 3.0 , Sec t i on 13.2.2.8 a nd 13.2.7.1 FS-0 2 Ove r hea d crane s ,/ Chapter 13.0 , Sec ti on 13.2.7.3 FS-0 3 P rocess vesse l e m e r ge n cy pu rge sys t e m ,/ C h a pt e r 6.0 , S e ct i o n 6.2.1.7 C h a pt e r 1 3.0 , S ec ti o n 1 3.2.7.3 FS-0 4 I rr a d iated t arget cask l ifting fixture ,/ C h ap t er 6.0 , Sec t i on 6.2.1. 7 Cha p ter 1 3.0 , Section 13.2.6.5 F S-05 Ex h a u s t s ta ck h eig ht ,/ C h a pt e r 6.0 , S ec ti o n 6.2.1. 7 C h a pt e r 1 3.0 , S ec tion 1 3.2.7.3

  • R eserve d -IR OFS d es i gna t o r c urr e ntl y un ass i gne d. AC a dmin is tr a ti ve co n tro l. IR OFS ite m s r e li e d o n fo r safe t y. ESF = e n g in ee r e d safe t y featu r e. 3-7
.;. NWMI ...*.. ..* .... .*.* .. *.*. * "NDmlWESTMEDICALISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components 3.1.2 Code of Federal Regulations NWMI-DRD-2013-030, NWMI Radioisotop e Produ c tion Facility D e sign R e quir e ments Do c ument, summarizes the CFR design inputs (in whole or in part) for the RPF , which include the following
  • * * * * *
  • * * * * * *
  • 10 CFR 20 , "Standards for Protection Against Radiation" 10 CFR 30 , "Rules of General Applicability to Domestic Licensing of Byproduct Material" 10 CFR 50 , "Domestic Licensing of Production and Utilization Facilities" 10 CFR 70 , "Domestic Licensing of Special Nuclear Material" 10 CFR 71 , "Energy: Packaging and Transportation of Radioactive Material" 10 CFR 73 , "Physical Protection of Plants and Materials" I 0 CFR 74 , "Material Control and Accounting of Special Nuclear Material" 10 CFR 851 , "Worker Safety and Health Program" 21 CFR 210 , " Current Good Manufacturing Practice in Manufacturing , Processing, Packaging , or Holding of Drug s' 21 CFR 211 , " Current Good Manufacturing Practice for Finished Pharmaceuticals" 29 CFR 1910 , " Occupational Safety and Health Standards" 40 CFR 61 , "National Emissions Standards for Hazardous Air Pollutants (NESHAP)" 40 CFR 63 , "NESHAP for Source Categories" 40 CFR 141, " National Primary Drinking Water Regulations" 3.1.3 U.S. Nuclear Regulatory Commission Table 3-3 lists the NRC design inputs for the RPF identified in NWMI-DRD-2013-030. The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference. Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) CF Ra Docket Number: NRC-2011-0135 (NRC, 2012) Title Final Int e rim Staff Guidanc e Au g m e nting NU REG-153 7, " Guid e lin e s for Pr e parin g and R e vi e wing Appli c ations for th e Li ce n s ing of N on-Pow e r R e a c t o r s," Part s 1 and 2, for Li ce n s ing Radioi s otop e Produ c tion Fa c iliti es and Aqu e ous Homog e n e ous R e a c tor s NRC Regulatory Guides -Power Reactors (Division
1) Regulatory Guide I .53 A ppli c ation of th e Singl e-Failur e Crit e rion to Saf ety S ys t e m s , 2003 (R201 I) Regulatory Guide 1.60 D e sign Response Sp e ctra for S e ismi c Design of Nuclear Pow e r Plants, 2014 Regulatory Guide 1.76 D es i g n Ba s i s Tornado and Tornado Mis s il es fo r N uclear P owe r Plants , 2 007 Regulatory Guide 1.97 Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, 2006 (R2013) Regulatory Guide I. I 00 S e ismi c Qualifi c ation of El ec tri c al and A c tiv e M e chani c al E quipm e nt and Fun c tional Qualifi c ation of A c tiv e M ec hani c al Equipm e nt for N ucl e ar P owe r Plant s, 2009 Regulatory Guide 1.152 Criteria for Use of Computers in Safety Systems of Nuclear Power Plants , 201 l Regulatory Guide 1.166 Pr e-Earthquak e Planning and Imm e diat e Nuclear Pow e r Plant Op e rator P os t E arthquak e A c tions , 1997 Regulatory Guide 1.167 Restart of a Nuclear Power Plant Shut down by a Seismic Event, 1997 Regulatory Guide I .208 P e rforman ce Ba se d Appr o a c h t o D e fin e th e Sit e-Sp ec ifi c Earthquak e Ground Motion , 2007 NRC Regulatory Guides -Fuels And Materials Facilities (Division
3) Regulatory Guide 3 .3 Quali ty A s s uran ce Program R e quir e m e nt s for Fu el R e pro cess in g Plants and for Plu to nium Pr ocess in g and Fu e l Fabri c ati o n Plants , 1974 (R2013) 3-8

......... *.* .; ... ; .. NWMI ...........

  • NORTKW(STMEDtcAl.ISOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-3. Relevant U.S. Nuclear Regulatory Commission Guidance (2 pages) Title Regulatory Guide 3.6 Content of Technical Specification for Fuel Reprocessing Plants, l 973 (R2013) Regulatory Guide 3. l 0 Liquid Wast e Tr e atm e nt S y st em D es ign Guid e for Plutonium Proc e ssing and Fu e l Fabrication Plant s, l 973 (R2013) Regulatory Guide 3.18 Confinement Barriers and Systems for Fuel Reprocessing Plants, 1974 (R2013) R eg ulatory Guide 3.2 0 Pro cess Offgas S ys tems for Fu e l R e pro ce ssing Plants, 1974 (R20l3) Regulatory Guide 3.71 Nuclear Criticality Safety Standards for Fuels and Materials Facilities, 2010 NRC Regulatory Guides -Materials and Plant Protection (Division
5) Regulatory Guide 5.7 Entry/Exit Control for Protect e d Areas, Vital Areas , and Material Access Areas, May 1980 (R20l0) Regulatory Guide 5.12 Genera l Use of Lo c ks in th e Prot ect ion and Contro l of Facilitie s and Sp ec ial N ucl e ar Mat er ial s, 1973 (R20l0) Regulatory Guide 5.27 Special Nuclear Material Doorway Monitors, 1974 Regulatory Guide 5.44 P e rim e ter Intrusi o n Alarm Syst e m s, 1997 (R2010) Regulatory Guide 5.57 Shipping and Receiving Control of Strategic Special Nuclear Material, 1980 Regulatory Guide 5.65 Vital Area Ac cess Contro l, Prot ec tion of Ph ysical S ec urity Equipment, and K ey and Lo c k Contro l s, 1986 (R2010) Regulatory Guide 5. 71 Cyber Security Programs for Nuclear Facilities, 2010 NUREG-0700, Human-System Interface Design Review Guidelines NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWREdition Section 2.3.1 Section 2.3.2 Section 3.3.l Section 3.3.2 Section 3.7.1 Section 3.7.2 " Re g ional Climatology

," Rev. 3 , March 2007 "Local Climatology," Rev. 3, March 2007 " Wind Loading," R ev. 3 , March 2007 "Tornado Loading," Rev. 3, March 2007 " Seismic De sign Parameters," March 20 07 "Seismic System Analysis," Rev. 4 , September 2013 Sec tion 3.7.3 " Seismic Subsystem Analysis," Rev. 4 , September 2013 NUREG-1513, Integrated Safety Analysis Guidance Document NUREG-1520, Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility Part 3, Appendix D "Natural Hazard Phenomena" NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors -Format and Content, Part 1 NUREG/CR-4604, Statistical Methods for Nuclear Material Management NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook Process hazard analysis "Development of Quantitative Risk Analyses" NUREG/CR-6463, Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems -Final Report NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology a Co mplet e references are provided in Sect ion 3.6. 3-9

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  • NCMTNWUTM£DICAL ISOTOf'lS NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components

3.1.4 Other

Federal Regulations, Guidelines, and Standards Table 3-4 lists other Federal design inputs for the RPF (NWMl-DRD-2013-030).

The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference.

Table 3-4. Other Federal Regulations, Guidelines , and Standards Referencea Title Federal Emergency Management Agency (FEMA) N I A "Nationa l Flood In s urance Program , Flood Insurance Rate Map , Boone County, Missouri and Incorporated Areas" National Oceanic and Atmospheric Administration (NOAA) Hydrometeorological Probabl e Maximum Precipitation Est imat es , U nit ed States East of th e 105th Meridian Report No. 51 Hydrometeorological Application of Probable Maximum Precipitation Estimates, United States East of the 105th Report No. 52 Meridian Hydrometeorological Seasonal Variati o n of JO-Squar e-Mil e Probabl e Maximum Pr ecipita tion Estimates , U nit e d Report No. 53 States Eas t of the 105th Meridian U.S. Geological Survey (USGS) N I A "2 00 8 U.S. G eo l ogical Survey National Seismic H aza rd Map s" Open-File Report Documentation for the 2008 Update of the United States National Seismic Hazard Maps 2008-1128 Centers for Disease Control and Prevention (CDC) NIOSH 2003-136 Guidance for Filtration and Air-Cleaning Systems to Protect Building Environments from Airborne Chemical, Biological , and Radiological Attacks

  • Co mpl e t e references a r e provided in Section 3.6 C D C FEMA NIOSH Ce nter s for Di sease C ontrol and Pr eve ntion. NOAA Federa l E m e r ge ncy Man age m e nt Agency. Nat ional In s titut e for Occup at i o n a l Safety and USGS H ea l t h. 3.1.5 Local Government Documents Na tion a l Oc ea nic a nd Atmospheric Administration.

U.S. Geological S urv ey. Table 3-5 lists the design inputs for the RPF from the State of Missouri , City of Colum bia , and Boone County government sources (NWMI-DRD-2013

-030). The RPF system design descriptions identify the specific requirements for that system produced b y each applicable reference.

Table 3-5. Local Government Documents (2 pages) Referencea Title Missouri Code of State Regulations (CSR), Title 10 10 CSR 10-6.01 Ambient Air Quality Standar d s Missouri CSR, Title 20 20 CSR 2030-2.040(1) Evaluation Criteria for Building De s ign Missouri Department of Transportation (MODOT) Standards and Specifications Missouri Department of Natural Resources (MDNR) Missouri State Adopted International Code Council (ICC) Building Code Set 2012 Boone County Building Code City of Columbia, Missouri, Code of Ordinances Article II -Building and Fire Codes Section 6-16 , Adopted Building Code 3-10

..... .. NWMI ...... ..* **: ........ *.* * ." NOmlWUT llllEOtCAl ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-5. Local Government Documents (2 pages) Referencea Section 6-17 , Amendments Se ction 9-2 1 Section 9-22 Building Code Fi re Co d e Fire Code C S R I C C a Co mpl ete refe r e n ces a r e prov id e d in Sec ti o n 3.6 C ode of S t at e R e gulat i o n s. = In te rn a ti o n a l C od e Co un c il. 3.1.6 Discovery Ridge/University of Missouri MDN R MO D OT Title Misso uri D e p a rtm e nt of Na tu ral R e so ur ces. Mi ss ouri D e p a rtm e n t of Tra n s p ort a tio n. T a bl e 3-6 li s t s th e MU sys t em r e quir e m e nt s and Di sc o very Rid ge co v e n a nt s d es i gn input s for the RP F id e ntified in NWMI-DRD-2 013-030. Th e RPF s y s tem d es ign d escri pti o n s identi fy the s p ec ifi c r e quir e m e nt s fo r th a t sys t em p ro du ced b y eac h a ppli ca bl e r efe r e n ce. Table 3-6. Discover y Ridge/University of Missouri Requirements Requirements Reference section/requirementa Civ il D es i g n a n d co n s tru ct i on of t he c i v il system i s r eg ul a t ed by t he N R C as re qui red b y D iscovery Rid ge/MU. Collected Rules and Regulations (CRR) St ru c tu ra l C RR Sect i o n 70.060.I , " Co d es and Sta nd a rd s" -A dop ts I CC co d es University of Missouri, Consultant Procedures and Design Guidelines E l ect ri ca l Sec ti o n 2.4.2, " Buildin g Co d es a nd S t a n da rd s fo r U ni vers i ty Fac iliti es" HV AC CPDG Di v ision 23 , " Heatin g, Ventilatin g, and Air-Conditioning (HV AC)" In st rum e n ta t ion Sec tion 2.4.2 , " Bu i ldin g Co d es an d Sta nd ar d s fo r U ni versity Fac iliti es" a nd Co n tro l s Plannin g CPDG Se c t io n 2.4 , "Planning , D e sign a nd Contra c t Do c um e nt De v elopment Guidelines for Ma s ter Con s truction Deli v ery Method" Pl umbing C PDG Di v i s i o n 22 , " Plumbin g" Proce ss S e ction 2.4.2 , " Building C ode s and St a ndard s for Uni ve r s ity Faciliti es" U niv e rsit y of Mis s ouri , Facilit ies Mana ge m e nt Poli c y and Proc e dure s Manual E lectrical Chapter 2 , " De s ign and C onstruction P o licy" I n st rum e nt a ti o n C h a pt e r 2 , " Des i g n a nd Co n st ru c ti o n P o l icy" a nd Co nt ro l s Structural Section 3.A , Refers to CRR 70.060 for the Ba s ic Building Code Section 3.0 , Refers to the University Building Adopted Codes for currently adopted code s U ni v ersity Building Adopted Codes IMC-2012 Int e rnational M ec hani c al Co d e Struc tu ra l A d o pt s IBC 2 0 12 a Co mpl et e refe r e n ces a r e pro v id e d in Sec ti on 3.6 CRR C o ll e c t ed Rul es and R e g ul at i o n s. IBC In terna ti o n a l B u i l ding C o d e. I C C In te rn at i o n a l C ode C o u nci l. M U NRC 3-11 U ni v e r s i ty o f M i ssouri. U.S. N u c l e ar R egu l atory Co mmi s s i on.

...... ; ... NWMI :;.**.*.*.* ....... :.* . * . NORTHWUT MEDttAl tsOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components

3.1.7 Codes

and Standards Table 3-7 lists design inputs for the RPF identified in NWMI-DRD-2013-030.

The RPF system design descriptions identify the specific requirements for that system produced by each applicable reference. The Construction Pennit Application and associated preliminary design documents identify codes , standards , and other referenced documents that may be applicable to the RPF. The specific RPF design codes , standards , and other referenced documents , including exceptions or exemptions to the identified requirements , will be finalized in the RPF final design and provided to the NRC. In addition, the codes, standards, and referenced documents for the RPF safety SSCs that are needed to demonstrate compliance with regulatory requirements will be identified and committed to in the Operating License Application. Table 3-7. Design Codes and Standards (12 pages) Document number* Document title American Concrete Institute (ACI) ACI 349 Code R e qu i r e m e nt s.fo r Nuclea r Safety-R e l a t ed Co n c r e t e S t ruc tur es an d Co mm e n tc 11 y, 20 1 3 American Institute of Steel Construction (AISC) ANSI/ AISC N690 S p ec ifi ca ti on for Safety-R e l ated Stee l St ructures for N u clear F a c iliti es, 20 1 2 Air Movement and Control Association (AMCA) AMCA Publication 201 Fans and Syst e ms , 2002 (R2011) AMCA Publication 203 ANSI/AMCA 210 AMCA Publication 211 AMCA Publication 311 Fi e ld P e rforman ce M e asur e m e nt of Fan S y st e ms , 1990 (R2011) Laboratory Methods for Testing Fans for Aerodynamic Performance Rating, 2007 C e rtifi e d Ratin gs Pr o gram -Produ c t Ratin g Manual f o r Fan Air P e rformanc e, 2013 Certified Ratings Program -Product Rating Manualfor Fan Sound Performance, 2006 (R2010) American Conference on Governmental Industrial Hygienists (ACGIH) ACGIH 2097 Industrial Ventilation:

A Manual of Recommended Practice for Design, 2013 American National Standards Institute (ANSI) ANSl/ITSDF B56.1 Safety Standard for Low Lift and High Lift Trucks ANSI/IEEE C2 ANSI C84.1 ANSI N 13 s erie s ANSI N13.1 ANSI N323D ANSl/AIHA/ASSE Z9.5 ANSI/NEMA Z535.1 ANSI/NEMA Z535.2 ANSI/NEMA Z535.3 ANSl/NEMA Z535.4 2012 Nati o nal El e ctri c al Saf ety Cod e (NE SC), 2012 American National Standard for Electric Power Systems and Equipment-Voltage Ratings (60 Hertz), 2011 Addresses radiation monitoring equipment Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities 2011 Am e ri c an N ati o nal Standard.for Install e d Radiation Pr o t ec tion Instrum e ntation , 2002 Laboratory Ventilation, 2012 Saf e ty Color s, 2006 (R2011) Environmental and Facility Safety Signs, 2011 Criteria f o r Saf e ty Symbol s, 2011 Product Safety Signs and Labels, 2011 3-12

...

..... ........ *.* * !*.*!* NOmlWESTlllDICAllSOTOr£1 Document number* NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title ANSI/AMCA 204 Balanc e Quality and Vibration L eve ls for Fans , 2005 (R2012) ANSI/AMCA 210 Laboratory Methods of Testing Fans for Aerodynamic Performance Rating , 2007 ANSI/AHRJ Standard 390 P erfo rmanc e Rating of Single Package Vertical Air-Conditioners and H eat Pump s, 2003 ANSI/ AHRJ Standard 410 Forced-Circulation Air-Cooling and Air-Heating Coils, 2001 ANSI/AHRJ Standard 430 Performance Rating of Central Station Air-Handling Units , 2009 ANSI/ AHRJ Standard 850 Performance Rating of Commercial and Industrial Air Filter Equipment, 2013 ANSI/HI 3.1-3.5 Rotary Pump s, 2008 ANSI N42. l 7B American National Standard Performance Specifications for Health Physics Instrumentation

-Occupational Airborne Radioactivity Monitoring Instrum entation, 1989 ANSI N42. l 8 Specification and P e rforman ce of On-Sit e In stru m en tati o n for Continuousl y Monitoring Radioa ctiv ity in Effluents, 2004 ANSI/IEEE N320 American National Standard P erformance Specifications for R eactor Emergency Radiological Monitoring Instrumentation, 1979 American Nuclear Society (ANS) ANSI/ ANS-2.3 ANSI/ ANS-2.26 ANSI/ ANS-2.27 ANSI/ ANS-2.29 ANSI/ ANS-6.4 ANSI/ ANS-6.4.2 ANSI/ ANS-8.1 ANSI/ANS-8.3 ANSI/ANS-8.7 ANSI/ANS-8.10 ANSI/ ANS-8.19 ANSI/ ANS-8.20 ANSI/ ANS-8.21 ANSI/ ANS-8.24 ANSI/ ANS-10.4 Estimating Tornado, Hurri cane, and Extreme Straight Line Wind Characteristics at Nuclear Facility Sites, 2011 Categorization of Nuclear Facility Structures, Systems, and Components for Seismic D es i g n , 2004 (R2010) Criteria for Inv estigations of Nuclear Facility Sites for Seismic Hazard Assessments, 2008 Probabilistic Seismic Ha zard Ana l ys is , 2008 Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants, 2006 Specification for Radiation Shielding Mat e rials , 2006 Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, 1998 (R2007) (W2014) Critica ll y Accident Alarm Syst e m , I 997 (R2012) Nuclear Criticality Safety in the Storage of Fissile Materials, 1998 (R2007) Criteri a for Nuclear Criticality Control in Operations wit h Shielding and Confinement, 1983 (R2005) Administrative Practices for Nuclear Criticality Safety, 1996 (R2014) Nuclear Criticality Saf ety Trainin g, 1991 (R2005) Use of Fixed Neutron Absorbers in Nuclear Facilities Outsid e R eactors, 1995 (R201 l) Va lidation of Neutro n Tran spo rt Methods for Nuclea r Criticality Safety Ca l cu lation s, 2007 (R20 1 2) Verification and Validation of Non-Safety-Related Scientific and Engineering Computer Programs for the Nuclear Industry , 2008 3-13

..... .. NWMI ..**.. ... .... ........ *.* . NOllTHWHTMEOICAllSOTO,H Document number 3 ANSI/ANS-10.5 ANSI/ ANS-15.17 ANSI/ANS-40.37 ANSI/ANS-55.l ANSI/ANS-55.4 ANSI/ ANS-55.6 ANSI/ ANS-58.3 ANSI/ ANS-58.8 ANSI/ ANS-59.3 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Accom m odating User Needs in Co mput er Program Development, 2006 (R2011) Fire Protection Program Criteria for R esearch R eactors, 1981 (R2000) (W20 10) Mobile Low-Level R adioactive Waste Pr ocessi n g Syste m s, 2009 Solid Radi oactive Waste Processing System for Light Water Cooled R eactor Plants, 1992 (R2009) Gaseous Radioa ctive Waste Processing Systems for Li ght Water R eactor Plants , 1993 (R2007) Liquid Radi oactive Waste Processing Syst e m for Light Water R eactor Plants, 1993 (R2 007) Physical Protection for Nuclear Safety-R e lat ed Systems and Co mpon ents, 1992 (R2008) Time R esponse Design Criteria for Safety-Related Operator Actions, 1994 (R2008) Nuclear Safety Cr it er ia for Co ntrol Air Systems, 1992 (R2002) (W2012) Design Guides for Radioactive Material Handling Facilities and Equipment, Remote Systems Technology Division, 1988, Air Conditioning, Heating and Refrigeration Institute (ARRI) ANSI/ARRI Standard 365 Performance Rating of Commercial and Industrial U nita ry Air-Condit i o nin g Condensing Units, 2009 ANSI/ARRI Standard 410 Forced-Circu lati on Air-Conditioning and Air-Heating Coils , 2001 A m er ican Soc iet y of Civil E ngineer s (ASCE) ASCE4 ASCE 7 ASCE43 ASCE Manual of Practice 37 Seismic Analysis of Safety-Related Nuclear Structures and Commentary, 2000 Minimum Design Loads for Buildings and Other Structures, 2005 (R2010) Seismic Design Crite ria for Structures, Systems, and Components in Nuclear Facilities, 2005 Design and Construct ion of Sa nita ry and Storm Sewers, 1969 American Society of Heating, Refrigeration and Air-Conditioning Engineers (ASHRAE) ANSI/ ASHRAE Standard 15 ANSI/ ASHRA E 51-07 ANSI/ ASHRAE Standard 52.2 ANSI/ ASHRAE Standard 55 ANSI/ ASHRAE Standard 62.l ASHRAE Standard 70 ANSI/ ASHRAE/IES Sta nd a rd 90.1 ANSI/ ASHRAE 110 Safety Standard for R efrigeration Systems, 20 1 3 Laboratory Methods of T esting Fans for Certified Aerodynam i c Performance Ratin g, 2007 Method for Testing General Ventilation Air Clean in g Devices for R emoval Efficiency by Particle Size, 2007 Thermal Environmental Conditions for Human O cc upan cy, 2013 Venti l ation for Acceptab l e Indoor A ir Quality, 2010 Method of Testing the Performance of Air Outlets and Air Inlets, 2011 Energy Standard for Building s Except Low-Rise R esidentia l Buildin gs, 20 10 Method of Testing P erformance of Laborato ry Fume Hoods, 1995 3-14

.; ... NWMI ::.**.*.*. ........... * ! *: !

  • NORTHWEST MlOtCAl tsOTOPlS Document number 3 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title ANSI/ ASHRAE 111 Measurement , Testing, Adjusting and Balan ci ng of Buildin g H ea ting , Ventilation , Air-Cond i tioning and R ef rig e ration Systems, 2008 American Society of Mechanical Engineers (ASME) ASME A I 7. I Saf e ty Code for El ev ators and Esca lator s, 201 3 ASME AG-1 Code on Nuclear Air and Gas Treatment , 2012 ASME Bl6.5 Pip e Flanges and Flang e d Fittings:

NPW !Ii through 24 , 2003 ASME B20. l Safety Standard for Conve yors and R e lat ed Equipment, 2012 ASME B30. l 7 Ov e rh eard and Gantry Cran es (Top Running Bridge, Singl e Girder , Underhung H o i st), 2006 ASME B30.20 B elow-the-H ook Lifting Devices, 2013 ASME B31.3 Pro ce ss Pipin g, 20 14 ASME B3 l .9 Building Services Piping , 2011/2014 ASME B3 l. l 2 H y drog e n Pipin g and Pip e lin es, 20 14 ASME B40.100 Pressure Gauges and Gauge Attachments, 2013 ASME B40.200 Th e rmometers, Dir ec t R eading and R e mot e R ead in g, 2013 ASME Boiler and Pres su re Section VIII Division 1 , 20I0/2 013 Vessel Code Section IX ASME HST-1 P e rformance Standard for El ec tric C hain Hoists , 2012 ASME N509 Nuclear Power Plant Air-Cleaning Units and Components, 2002 (R2008) ASME 510 T es ting of Nuclear A ir-Tr ea tm e nt S ys t e m s, 2007 ASME NQA-1 Quality Assurance R equirements for Nuclear Facility Applicatio n s, 2008 with NQA-1 a-2009 a ddenda ASME QME-1 Qua l ification of Active M ec hani c a l Equipment Use d in Nuclear Power Plants , 2012 American Society for Nondestructive Testing (ASNT) SNT-TC-I A R ec ommend e d Practice No. SNT-T C-JA: P e rsonne l Qualifi c ation and Certifica ti on in No nd es tru c ti ve T es ting , 2011 American Society for Testing and Materials (ASTM) ASTM C I 055 ASTM Cl217 ASTM C l5 33 ASTM Cl554 ASTM C1572 ASTM Cl615 ASTM Cl661 Standard Guid e for H e at ed S yste m Surfa ce Condit ion s that Produ ce Contact Burn Injuri es, 2003 (20 14) Standard Guide for Design of Equipment for Processing Nuclear and Radioa ctive Materials, 2000 Standard Guid e for Gen e ral D es i gn Consideratio n s for H ot Ce ll Equipment , 20 15 Standard Guide for Materials Handling Equipment for Hot Cells, 2011 Standard Guid e for Dry L e ad Glass and Oil-Fill ed L e ad Gla ss Radiation Shi e ldin g Window Compon e nt s for R e mot e l y Op erated Fa c ilities , 2010 Standard Guide for Mechanical Driv e Systems for R emo t e Operation in Hot Cell Facilities, 2010 Standa rd Guid e for Viewing S ys t e m s for R e mot e l y Op e rat e d Fa c iliti es, 2013 3-1 5

........... ;.-.; .. NWMI ........ *. ' !*. . NOkTHWEST MEOtCAL ISOTOPH Document numbera ASTM E493 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Standard Practic e for L e aks Using the Mass Spectromet e r Leak Detector in the Insid e-Out Testing Mod e, 2011 ASTM F 14 71 Standard T est M e thod for Air C l e an in g P e rforman ce of Hi g h-Effi c i e n cy Parti c ulat e Air-F i lt e r S y stem , 2 00 9 American Welding Society (A WS) AWS B 2.l/B2.1M AWS Dl.l/ Dl.IM AWS Dl.3/Dl.3M AWS Dl.6/Dl.6M AWS D 9.l/ D 9.1M AWS QCl Sp ec ifi c ation for W e ldin g Pr oce dur e and Performan ce Qualifi c ation , 2009 Structural Welding Cod e -Steel, 2010 St ru ctu r a l W e lding Cod e -Sh ee t S t ee l , 2008 Structural Welding Code -Stainless St ee l , 2007 Sh ee t Metal W e l d in g Cod e, 2006 Standard/or AWS Certification of Welding In s p ec tors , 2007 Centers for Disease Control and Prevention (CDC) -National Institute for Occupational Safety and Health (NIOSH) DHHS (NIOSH) Publication Guidance for Filtration and Air Cleaning Systems to Prot ect Building Environments No. 2003-136 from Airborne Chemical, Biologi ca l , and Radiologi c al Attacks, 2003 Electronic Industries Alliance (EIA)/Telecommunications Industr y Association (TIA) ANSI/TIA-568-C. l ANSI/TIA-568-C.2 ANSl/TIA-568-C.3 ANSI/TIA-5 69 ANSI!fIA-606 ANSI/TIA-607 ANSI/TIA-758-A International Code Council ICC All7.l IECC IMC IPC Commercial Building T e lecommunications Cabling Standard , 2012 Balan ce d Twist e d-Pair T e l eco mmuni c at i on s Cab lin g and Components Standards , 2014 Opti c al Fib e r Cabling and Components Standard, 2011 T e l ec ommuni c ations Pathwa ys and Spa ces, 20 1 3 Ad mini st rati on Standard for Commercial T e l eco mmuni catio n s Infra s tru ctu r e, 2012 Comm e rcial Building Grounding (Earthing) and B o n din g Requir e m e nts for T e le c ommuni c ations, 2013 Customer-Owned Outsid e Plant T e l ec ommunications Infrastructure Standard, 2004 Accessible and Usable Buildings and Fa c iliti es Standard , 2009 2 0 1 2 In t e rnational En e r gy C o ns e rvation Cod e, May 2011 20 12 Int e rnational Mechanical Code, June 2011 int e rnational Plumbing Cod e, Apr il 20 11 Institute of Electrical and Electronics Engineers (IEEE) IEEE 7-4.3.2 IEEE 141 IEEE 14 2 Sta nd ard Cr i ter i a fo r Digital Computer s in Saf e ty Syst e m s of Nuclear Pow e r G e n e rating Stations , 2 00 3 Recommended Practice for Electric Power Distribution for Industrial Plants (Red Book), 1993 (R1999) R e commend e d Practi ce for Grounding of Industrial and Comm e rcia l Power S y stems (Gr ee n Bo o k), 2007 3-16

..... ; .. NWMI ..**.. ..* .... ........ *.* .. *****. *. , . *

  • NOltntWEST MEDtCAL ISOTOPES Document number 3 IEEE 241 IEEE 242 IEEE 308 IEEE 315 IEEE 323 IEEE 336 IEEE 338 IEEE 344 IEEE 379 IEEE 384 IEEE 399 IEEE 446 IEEE 493 IEEE 497 IEEE 519 IEEE 535 IEEE 577 IEEE 603 IEEE 650 IEEE 739 IEEE 828 IEEE 829 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Re commended Practice for Electric Power Systems in Commercial Buildings (Gray Book), 1990 (Rl997) R e commend e d Practice for Prot ec tion and Coordination of Indu s trial and Comm e rcial Power Syst e ms (Buff Book), 2001 Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations, 2012 Graph i c Symbols for El ec tri c al and El ec tronics Diagram s, 1975 (Rl 993) Standard for QualifYing Class IE Equipment for Nuclear Power Generating Stations, 2003 Recomm e nd e d Practic e for Installation , In s p e ction, and T e sting for Class IE Pow e r, Instrum e ntation, and Control Equipment at Nucl e ar Fa c iliti e s, 20 10 Standard for Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems, 2012 Recommend e d Practic e for S e ismic Qualification of Class IE Equipment for N ucl e ar Po we r G e nerating Stations, 2013 Standard Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems, 2014 Standard Crit e ria/or Ind e pend e n c e of Cla ss IE Equipm e nt and Circuits , 2008 Recommended Practice for Power Systems Analysis (Brown Book), 1997 R ec ommend e d Practic e for Em e rgency and Standby Pow e r S y stems for Industrial and Commercial Appli c ations (Orange Book), 1 995 (R2000) Recommended Practice for the Design of R eliable Industrial and Commercial Power Systems (Gold Book), 2007 Standard Crit e ria for A cc ident Monitoring Instrum e ntation for Nucl e ar Pow e r Gen e rating Stations , 2010 Recommended Practice and R equirements for Harmonic Control in Electrical Power Systems, 2014 Standard for Qualification of Class IE Lead Storag e Batt e ri e s for Nuclear Power Generating Stations , 20 1 3 Standard R equirements for R eliability Analysis in the Design and Operation of Safety Systems for Nuclear Facilities, 2012 Standard Crit e ria for Safety S y st e ms for N uclear Power G e nerating Stations, 2009 Standard for Qualification of Class IE Static Batt ery Chargers and Inverters for Nuclear Pow er Generating Stations, 2006 Recommended Practice for En e rgy Manag e ment in Industrial and Comm e r c ial Facilities (Bronz e Book), 1995 (R2000) Standard for Configuration Management in Systems and Software Engineering, 2012 Standard for Software and S y st e m Test Docum e ntation , 2008 3-17

.;.-.;* .. NWMI ..**.. ..* **.* ........ *.* .

  • NOllTHWHTMlOICAllSOTOPH Document number 3 IEEE 902 IEEE 946 IEEE 1012 IEEE 1015 IEEE 1023 IEEE 102 8 IEEE 1046 IE E E 1050 IEEE 1100 IEEE 12 8 9 IEEE 1584 ANSI/IE EE C2 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Guide for Maintenance, Operation , and Safety of Industrial and Commercial Power Systems (Yellow Book), 1998 G e n e rating S tation s, 2004 Standard Criteria for Softwar e Verification and Validation, 2012 R ec omm e nd ed Pra c ti ce A ppl y in g L ow-Vo lta ge Ci r c uit Br ea k e r s Use d in Indu s trial and C omm e r c ial Po we r S ys t e m s (Blu e Book), 2006 (C2007) Guide for the Application of Human Factors Engineering to S y stems, Equipment, and Facilities of Nuclear Power Generating Stations, 2004 (R2010) Standard f o r So ftwar e R e views a nd Audit s , 200 8 Application Guide for Distributed Digital Control and Monitoring for Pow e r Plants, 1991 (Rl996) Gu id e f o r In str um e ntati o n and Co ntr o l E quipm e nt Gr o u n ding in G e n e r at in g Stations , 2004 Recommend e d Practi ce for Powering and Grounding El e ctronic Equipment (Emerald Book), 2005 Guid e for the A ppli c ati o n of Human Fa c t o rs E n g in ee ring i n th e D es i g n of Co mput e r-B ased M o nit or ing a nd C o ntr o l Di s pl ays for N ucl e ar P owe r Ge n e ratin g Stati o n s, I 99 8 (R2004) IEEE Guid e for Performing Arc-Flash Hazard Calculations , 2002 201 2 N ati o n a l E l ec tri c al Saf ety Cod e (NESC), 20 I 2 Illuminating Engineering Society of North America (IES) IES-20 I I Th e Li g htin g Handbo o k , 20 I I ANSl/IES RP-1-12 IES RP-7 American National Standard Practice for Offic e Lighting , 20 I 2 A m e ri c an Na ti o nal Standar d Pra c ti ce fo r Li g htin g Indu s trial Fa c ilit ies , 1991 (W2001) International Society of Automation (ISA) ANSl/ISA-5. 1-2009 ISA-5.3-1983 ISA-5.4-199 I ISA-5.5-1985 ANS l/ISA-5.06.01-2007 ANSI/ISA 7.0.01-1996 AN SI/ISA-12.0 I .01-2013 ISA-18.1-1979 ISA-TR20.00

.01-2007 In s trum e ntati o n S y mbol s and Id e ntifi c ati o n , 2009 Graphic Symbols for Distributed Control/Shared Displa y Instrumentation , Logic, and Computer System s , 1983 In s trum e nt L oo p Dia g ram s , 19 9 1 Graphic Symbols for Process Displays, 1985 Fun c ti o nal R e quir e m e nt s D oc um e ntati o n for Co ntrol So f tw are A ppli c a t i o n s, 2007 Quality Standard for Instrument Air D efi niti o n s and Inf o rmation P e rt a inin g t o El ec tri c al Equi p m e nt in Ha z a r d o u s (C la ss ifi e d) L oc ation s, 2013 Annunciator Sequences and Specifications, 1979 (R2004) Sp ec ifi c ati o n Fo rm s f o r Pr ocess Mea s u re m e nt and C o nt ro l In s trum e nt s Part 1: Ge n e ral Con s id e rations U pd a t e d with 27 n ew s p ec ifi c ati o n form s in 2004-2 006 and updat e d w i t h 11 n ew s p ec ifi cat i o n form s in 200 7 3-18

.**.*.*. .; ... ... NWMI ........ *.* 0 ." NORTKWlST MlDtCAI. ISOTOrt:S Document numbera ISA-RP60.1-1990 ISA-67.01.01-2002 ANSUISA-67.04.01-2006 I SA-RP67.04.02-20 10 ANSl/ISA-75

.05.01-2000 ANSl/ISA-82.03-1988 ISA-TR84.00.04-2011 ISA-TR84.00.09-2013 ISA-TR9 l .00.02-2003 ANSU IS A-TR 99.00.0l-2007 NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Control Cent e r Faciliti es, 1990 Transducer and Transmitt er Installation for Nuclear Safety App li ca ti ons, 20 0 2 (R2 00 7) Setpoints for Nuclear Safety-Related Instrumentation, 2006 (R2011) Met h odo l ogies for t h e D e t e rmination of Setpoints for Nuclear Safety-R elated In s trum e ntation, 20 10 Control Valve Terminolo gy, 2000 (R2005) Safety Standard for E l ectrica l a nd E l ec tr o ni c T es t , M e asuring , Con t ro lli ng, a nd R e l a t ed E qu ipment, 19 88 Part 1 Guideline for th e Impl e m e ntation of ANSJIISA-84.00.01-2004

(!EC 61511), 2011 Secur i ty Countermeasures R e l ate d to Safety I nstrumented Syste m s (S I S), 2013 Criticality Classification Guideline for In str um e ntation , 2003 Security Te c hnolo gies for In dustr i a l Automatio n a nd Control Systems, 2007 International Atomic Energy Agency (IAEA) IAEA-TECDOC-1250 IAEA-TECDOC-134 7 I AEA-T EC DO C-14 30 Se i sm i c Desi gn Considerations of Nuclear Fuel Cycle Facilities, 2 001 Consideration of External Events in the D esign of Nuclear Facilities Other Than Nuclear Pow e r Plants , With Emphasis on Earthquakes, 2003 Radioi sotope Handling Fac ili ties and Automation of R adioisotope Production , 2 004 International Code Council (ICC) IBC 2 012 IFC 2012 IMC 2 01 2 Int e rn a tional Building Co d e, 20 12 Int er national Fire Code, 2012 Int e rnational Mec h anica l Code, 2 01 2 International Code Council Evaluation Service (ICC-ES) I CC-ES AC 1 56 "Ac ceptance Cr it e ria for Se i sm ic Cert ifi cation by S hak e-T ab le Testing of No n s tructu ra l Co mponen ts," 20 I 0 National Electrical Contractors Association (NECA) NECA 1 NECA90 NECA 100 NECA 101 NECA/AA 104 NECA/NEMA 105 NECA 111 Standard Pra c ti ce of Good Workma n s hip in E l ec trical Constructio n , 20 10 R eco mmend e d Practi ce for Commissioning Building Electrical Sy s tem s (ANSI), 2009 Sym b o l s fo r Electrica l Construction Drawings (ANSI), 2013 Standard for Installing St ee l Conduits (Rigid, IMC, EMT) (ANSI), 2013 Standard for I nsta lling A l uminu m Buildin g Wire and Cab l e (ANS I), 2012 Standard for Installing M eta l Cable Tra y Systems (ANSI), 2007 Sta nda rd for I nsta llin g Nonmeta lli c R aceways (RNC, ENT, LFNC) (ANSI), 2003 3-19 "NWMI ...... ..* ... ........... * *. *

  • NOllTHWllT lllOtCAl ISDTOPH Document number 3 NECA 120 NECA 202 NECA 230 NECA/FOA 301 NECA331 NECA400 NECA402 NECA/EGSA 404 NECA407 NECA408 NECA409 NECA 410 NECA 411 NECA420 NECA430 NECA/IESNA 500 NECA/IESNA 501 NECA/IESNA 502 NECA/BICSI 568 NECA/NCSCB 600 NECA/NEMA 605 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Standard for Installing Armored Cable (Type AC) and Metal-Clad Cable (Type MC) (ANSI), 2013 Sta ndard for In sta llin g and Maintaining Indu str ial H eat Tracing Systems (ANS I), 2013 Standard for Selecting, Installing, and Maintaining Electri c Motors and Motor Controllers (ANSI), 2010 Standard for In sta llin g and Testing Fiber Optics, 2009 Standard for Building and S ervice Entrance Grounding and Bonding, 2009 Sta ndard for In sta llin g and Maintaining Switchboards (ANS I), 2007 Standard for Installing and Maintaining Motor Control Centers (ANSI), 2007 Standard for In sta llin g Generator Sets (ANSI), 2014 Re c ommended Practice for Installing and Maintaining Pane/boards (ANSI), 2009 Sta nd ard for Installing and Mai ntainin g Bu sways (ANS I), 2 009 Standard for Installing and Maintaining Dry-Type Transformers (ANSI), 2009 Standard for In sta llin g and Maintaining Liquid-Filled Transformers (ANS I), 2013 Standard for In s talling and Maintaining Uninterruptibl e Power Supplies (UPS) (ANSI), 2006 Standard for Fuse Applicat i ons (ANS I), 2014 Standard for Installing Medium-Voltage Metal-Clad Switchgear (ANSI), 2006 R ecommended Practic e for In sta llin g Ind oor Lighting S y stems (ANS I), 2006 Recommended Practice for Installing Exterior Lighting S ys tems (ANSI), 2006 R ecomme nd ed Practice for Installing Indu strial Lighting S y stems (ANS I), 2006 Standard for Installing Building Telecommunications Cabling (ANSI), 2006 R eco mm e nd ed Practice for In sta llin g and Maintaining Medium-Voltage Ca bl e (ANS I), 2014 Installing Underground Nonmetallic Utility Duct (ANSI), 2005 National Electrical Manufacturers Association (NEMA) NEMAMG-1 Motors and Generators, 2009 InterNational Electrical Testing Association (NET A) ANSI/NETA ATS-2013 ANSI/NETA ETT-2010 ANSI/NETA MTS-2011 Standard for Acceptance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2013 Standard for Certification of E l ec tri ca l Testing T ec hni c ian s, 2010 Maintenance Testing Specifications for Electrical Power Distribution Equipment and Systems, 2011 National Fire Protection Association (NFPA) NFPA 1 FPA2 Fire Code, 2015 Hydrogen Technologies Code, 20 11 3-20

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  • NORTHWEST MEOtCAl ISOTOPES Document numbera NFPA4 NFPA 10 NFPA 13 NFPA 14 NFPA20 NFPA 22 NFPA 24 NFPA 25 NFPA 30 NFPA 37 NFPA45 NFPA 55 NFPA68 NFPA 69 NFPA 70 NFPA 70B NFPA 70E NFPA 72 NFPA 75 NFPA 79 NFPA 80 NFPA 80A NFPA 86 NFPA 86C NFPA90A NFPA 90B NFPA 91 NFPA 92 NFPA92A NFPA 92B NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Standard for Integrated Fire Protection and Life Safety System Testing, 2015 Standard for Portable Fir e Extinguishers , 2013 Standard for the Installation of Sprinkler Systems, 2013 Standard for the Installation of Standpip e and Hose S y stems , 2013 Standard for the Installation of Stationary Pumps for Fire Protection, 2013 Standard for Wat e r Tanks for Privat e Fir e Prot ec tion, 2013 Standard for the Installation of Private Fire Service Mains and Their Appurtenances, 2013 Standard for th e Inspection, T es ting, and Maint e nan ce of Water-Bas e d Fir e Protection S ys t e ms, 2014 Flammable and Combustible Liquids Code, 2015 Standard for th e Installation and Us e of Stationary Combustion Engin es and Ga s Turbin e s , 2015 Standard on Fire Protection for Laboratories Us ing Chemicals, 2015 C o mpr e s s ed Gases and C ry o ge nic F l uid s Cod e, 2013 Standard on Explosion Protection by Deflagration Venting, 2013 Standard on E x plosion Pr e vention Systems , 2014 National Electrical Code (NEC), 2014 R ec omm e nd e d Pra c ti ce for El ec trical Equipm e nt Maint e nan c e , 2013 Standard for Electrical Safety in the Workplace, 2015 National Fir e Alarm and Signaling Cod e, 2013 Standard for the Fire Protection of Information Technology Equipment, 2013 El e ctri c al Standard for Indu s trial Machin ery, 2015 Standard for Fire Doors and Other Opening Protectives, 2013 Re c omm e nd ed Pra c tic e for Pr o t e ction of Buildings from E x t e rior Fir e Expo s ure s, 2012 Standard for Ovens and Furnaces, 2015 Standard for Industrial Furna ce s Using a Special Processing Atmosph e r e, 1999 Standard for the Installation of Air-Conditioning and Ventilating System, 2015 Standard for th e Installation of Warm Air H e atin g and Air-Conditionin g S y stems , 2015 Standard for Exhaust Systems for Air Conveying of Vapors, Gases, Mists, and Noncombustible Particulate Solids, 2015 Standard for Smoke Control S y stems, 2012 Standard for Smoke-Control Systems Utilizing Barri ers and Pressure Differences, 2009 Standard for Smoke Managem e nt Systems in Malls , At r ia, and Large Spac e s , 2009 3-21

..... ;. NWMI ...... ..* .... ........... 0 *. * ' NORTlfWUT MEDICAL ISOTOPES Document numbera NFPA IOIB NFPA 105 NFPA 110 NFPA Ill NFPA 170 NFPA 204 NFPA 220 NFPA 221 NFPA 262 NFPA 297 NFPA 329 NFPA 400 NFPA496 NFPA 497 NFPA 704 NFPA 730 NFPA 731 NFPA 780 NFPA 791 NFPA 801 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-7. Design Codes and Standards (12 pages) Document title Code for Means of Egress for Buildings and Structures, 2002 (W-Next Edition) Stan dard for the Installation of Smoke Door Assemblies and Other Opening Protectives , 20 13 Standard for Emergency and Standby Power Systems, 2013 Stan dard on Stored Electrical E n ergy Emergency and Standby Pow e r Systems, 2013 Standard for Fire Safety and Emergency Symbols, 2012 Standard for Smoke and Heat Venting , 2012 Standard on Types of Building Construction, 2015 Standard for High Challenge Fire Walls, Fire Walls, and Fire Barri er Walls, 2015 Standard Method of Test for Flam e Travel and Smoke of Wires and Cables for Use in Air-Handling Spaces, 2015 Guide on Principles and Pra c tices for Communications Systems, 1995 Recommended Practic e for Handling R e leases of Flammabl e and Combustible Liquids and Gases, 2015 Hazardous Materials Code, 2013 Standard for Purged and Pr ess urized Enclosures for Electrical Equipment, 2013 Re c ommend e d Practice for th e Classification of Flammabl e liquids, Gases , or Vapors and of Ha za rdou s (Classifi e d) locations for Electrical In s tallation s in C h e mi ca l Process Areas , 2012 Standard System for the Identification of the Hazards of Materials for Emergency Response, 2012 Guide for Pr e mis es Security , 20 14 Standard for the Installation of Electronic Premises Security Systems, 2015 Standard for the In sta llation of Lightning Protection S ys t e ms, 2014 Recommended Practice and Procedures for Unlabeled Electrical Equipment Evaluation, 201 Standard for Fir e Prot ec tion for Facilities Handling Radi oactive Materials, 2014 Sheet Metal and Air Conditioning Contractors National Association (SMACNA) National Oceanic and Atmospheric Administration (NOAA) NOAA Atlas 14 SMACNA 1143 SMACNA 1520 SMACNA 1922 SMACNA 1966 SMACNA-2006 Precipitation-Frequenc y Atlas of the United States, Vol. 8 Version 2.0 , 2013 HVAC Air Duct Leakag e T est, 1985 Round Industrial Duct Construction Standard, 1999 R ectangu lar Industrial Du ct Co n st ru ctio n Standard, 2004 HVAC Duct Construction Standard -Metal and Flexible, 2006 HVAC Syst e ms Duct D esign, 2006 3-22

......... *.* .; ... .. NWMI ....... :.* 0

  • NORlltWEST MEDtCAl ISOTOPES NWMl-201 3-021 , Rev. 1 Chapter 3.0 -Des i gn o f S t ructures , Systems and Component s Ta ble 3-7. D esig n Co d es a nd S tandard s (1 2 pa ges) Document number 3 Document title ANSl/SMACNA 00 1-2008 S e ismi c R e strai n t Manual: Guid e li n e s for M ec hani c a l S ys t e ms, 2 0 08 U.S. Wea th er Bu rea u Technica l Paper N o. 40 Rainfa ll Fr e qu e nc y At l as of th e Un i t e d Stat es for Durations from 30 Minut e s to 24 Hour s and R e tur n Period s from 1 to 1 00 Y e ars , 1963 U nd erw rit e r s La bor a tor y , I n c. (UL) Fe d e r a l S p e cifi ca ti o n s UL 18 1 UL 499 UL 55 5 UL 5 8 6 UL900 UL 1 9 9 5 Sta n dard for Fa c tory-Made A ir Du c t s and C onne c tor s, 2013 Sta nd a rd for E l ec t r i c H e a ti n g Ap plian ces, 20 14 S t andard for Fi re Damper s, 2006 S t a nd a rd f o r H ig h Effic i e n cy, P a r tic u l at e, Air Fi lt er U n i t s , 2 009 S t andard for Air Filt e r Uni ts, 2004 H eat ing a nd Coo lin g Equ i pme n t, 2 011
  • Co mpl ete refere n ces a r e p rov id e d in Sect i o n 3.6 ACG JH A m e ri ca n Co n fe r e n ce o n Gove rn men t al I AEA Int e rn a t i ona l A tomic E n e r gy A ge n cy. I nd u s tr i a l H yg i e ni sts. I CC In te rn a tion a l Co d e Co un c i l. AC I A m e r i ca n Co n c r ete In s titut e. I CC-ES In te rn at i o n a l Co d e Co un c il Eva l u a ti o n Se rvi ce. AH RI Air Co n dit i onin g , H ea ting a nd R efr i gera t ion I EEE In sti tu te of E l ec t r i ca l a nd E l ec t ro ni cs E n g in ee r s. I ns titut e. IES Illumin a ting E n g in ee rin g S o c i ety. A I SC A m e r ica n In s titu te of S tee l Co n struct i o n. I SA In te rn a ti o n a l Soc i ety of A ut o m a ti o n. AMCA A ir M ove m e nt and Co ntr o l Associatio
n. NECA Nationa l E l ec tric a l Co nt rac t o r s Assoc i at i o n. ANS A m e ri ca n N u clear Soc i e t y. NEMA Na ti o n a l E l ec tric a l Ma nu fac tu re r s Ass o c i a ti o n. ANS I Ame ri ca n Na ti o n a l Sta nd a rd s In st i t ut e. NETA Int erNat i o n a l E l ec tri ca l Tes t i n g Assoc i a tion. ASCE Ame ri ca n Soc i ety of C i v i l E n g i neers. NF P A Natio n a l F ir e Pro tect i o n Ass o c ia t i o n. AS H RAE A m e ri ca n Soc i ety of H ea tin g , R efrige r a ti o n N I OS H Nat i o n a l In s t i tut e for Occ up a ti o n a l Safety a nd a nd A i r-Co n diti o ni ng E n g in ee r s. H ea l t h. AS M E Am e ri c an Soc i ety of Mec h a n i ca l E n gi n ee r s. NOAA Natio n a l Ocea n i c a nd A tm os p h e ri c ASNT Ame ri c an Soc i ety fo r No nd es tru ct i ve A d minis trat i o n Tes t i n g. SMACNA S h eet M eta l an d Ai r Co ndit io nin g Co n trac t o r s AST M A m e ri c an Soc i ety for T es t i ng a nd Ma t e ri a l s. Nat i o n a l Assoc i a ti o n. A W S A m er i c an W e ldin g Soc i e ty. TI A T e l eco mmuni ca ti o n s Indu s try Associa t i o n. CDC Ce n ters fo r Di sease Co nt ro l a n d P reve nti o n. UL U nd erwr it e r s La b oratory. E I A E l ec tr o ni c In d u stries A ll i a n ce. 3-2 3

...... .. NWMI ...... ..* .... ........ *.* . NOmlWESTMUHCAl.ISOTOP£1 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components

3.2 METEOROLOGICAL

DAMAGE RPF meteorological accidents with radiological consequences are evaluated in NWMI-2015-SAFETY-01 l , Eva luation of Natural Ph enomeno n and Man-Made Events on Safety Features and It e ms R e li ed on for Safety. The basis for the structural design of the RPF is described in NWMI-2013-043 , NWM! Radioisotope Production Facility Structural D esign Ba s i s. Updates and development of technical specifications associated with the meteorological design of the RPF SSCs will be provided in Chapter 14.0 as part of the Operating License Application.

The demands on structural elements due to applied loads are evaluated using the criteria and methodology discussed below. The effect of each load case is determined separately, and total demand is determined by combining the load effects using the load combinations for evaluating strength and evaluating the serv iceability criteria given below. Four categories of load cases are used: normal , severe environmental, extreme environmental, and abnormal load s. The definition of each load is the following: * * *

  • Normal loads are loads that are expected to be encountered during normal plant operations and shutdown, and load due to natural hazard phenomena likely to be encountered during the service life of the facility.

Severe environmental loads are loads that may be encountered infrequently during the service life of the facility. Extreme environmental loads are loads that are credible but are highly improbable to occur during the service life of the facility.

Abnormal loads are loads generated by a postulated high-energy pipe break accident used as a de s ign basis. Definitions of load case symbols are provided in Table 3-8. Table 3-8. Load Symbol Definitions (2 pages) Symbol Definition Norma l Load Cases D Dead loads due to the weight of the structural elements , fixed-position equipment, and other permanent appurtenant items; weight of crane trolley and bridge F Load due to fluids with well-defined pre ss ur es a nd maximum heights H Load due to lateral earth pressure, groundwater pressure , or pressure of bulk materials L Live load due to occupancy and moveabl e equipment, includin g impact Lr Roof live load C cr Rated capacity of crane (will include the ma x imum wheel load s of the crane and the vertical, lateral , and l ongitudinal forces induced by th e moving crane) S Snow load as stipu lated in ASCE 7" for risk category IV facilities R Rain load T 0 Self-staining load, thermal effects, and loads during normal operating, startup, or shutdown conditions, based on the most critical transient or steady-state condition Ro Pipe reaction s during normal operating, s tartup , or s hutdown conditions , b ase d on th e mo s t critical transient or stea dy-state co ndition 3

  • .**.*.*. ..... ;. NWMI ........ *. * *. * !
  • NOfllTHWEST MlDfCAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-8. Load Symbol Definitions (2 pages) Symbol Definition Severe Environmental Load Cases D; Weight of ice F a Flood load W Load due to wind pressure W a Load based on serv iceability wind speed W; Wind-on-ice E o Where required as part of the design basis , loads generated by the operating basis earthquake, as defined in I 0 CFR 50 ,b Appendix S , "E arthquake Engineering Criteria for Nuclear Power Plants ," or as specified by the authority having jurisdiction. Extreme Environmental Load Cases S, Weight oftbe 48-hour probable maximum winter precipitation superimposed on S W 1 Loads generated by th e specified d es i gn ba s i s tornado, includ ing wind pre ss ur es, pr ess ure differentials , an d torna d o-borne mi ss il es , as d efi n e d in NUREG-0800, c or as s pecifi e d by the a uthority h av in g jurisdictio n E ss Loads generated by the safe shutdown , or de s ign basis earthquake , as defined in IO CFR 50 ,b Appendix S , or as specified by the authority having jurisdiction Abnormal Load Cases P a Maximum differential pressure load generated by the postulated accident R a Pipe an d eq uipm ent reactio n s generated b y the po s tulated acci dent , includ ing R o T a Thermal loads generated by the postulated accident , including T o Y j J e t impingement load generated b y the postulat e d acc ident Y m Missile impact load, s uch as pipe whip generated by or during the postulat ed accident Y, Loads o n th e st ructure ge n erated b y the reaction of t h e broken hi g h-ener gy pip e durin g th e postulated accident a ASCE 7 , Minimum D e si g n loads for Buildin gs and Oth e r Stru c tur e s , American Society of Civi l E n g in eers , R es t o n , Virgi ni a, 2005 (R2010). b I 0 CF R 50, " Dom es ti c Licensing of Pr od u ction an d U tili zatio n Faci lit ies," Cod e of F e d e ral R e gu lati ons , Offic e of th e Fe d eral R eg i ste r , as a mended. c NUREG-0800 , Sta ndard R e vi e w P l a n for th e R e vi e w of Saf e ty A nal y s i s R e port s for Nucl e ar Pow e r P l ants, L WR E dition , U.S. N ucl ear R egu l a t ory Co mmi ss i on, Office ofN u c l ear Material Safety and Safeguards , Was hin g ton , D.C., 1987. 3.2.1 Combinations of Loads Load co mbination s u se d for eva lu a ting s tr e ngth and servicea bility are give n in the following subsections.

Com binations for stre ngth-b ase d acceptance criter ia are give n for both nuclear safety-related SSCs a nd for commercial SSCs. 3-25

.. .. NWMI ...... ..* .... ....... :.* 0 *. ! . NO<<THWEST MEOK:AL &SOTOfl£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components 3.2.1.1 Nuclear Safety-Related Structures, Systems, and Components For nuclear safety-related SSCs , the loading combinations from ACI 349 , Cod e R e quirement s for Nucl e ar Saf ety-R e lat e d Con c r e te Stru c tur e s and Comm e ntary , are used. The load combinations from ACI 349 are essentially identical to the combination from ANSVAISC N690 , Sp ec ification for Saf e ty-Relat e d St ee l Stru c ture s for N ucl e ar Facil i ti e s. Table 3-9 presents nuclear safety-related SSC loads. In addition , the load combination for extreme winter precipitation load (S,) takes DC/COL-ISG-007 , Int e rim Staff Guidan ce on Asses sm e nt of N ormal and E x tr e m e Wint e r Pr ec ipitation Load s on th e Roofs of S e ismi c Cat e gory I Stru c tur e s , guidance into account. Table 3-9. Load Combinations for Strength Based Acceptance Criteria, Nuclear Safety-Related Combination Normal Load Combinations l .4(D + F + R o) + T o I .2 (D + F + T 0 + R 0) + I .6(L + H) + 1.4 Cc r + 0.5(L , or S or R) l .2(D + F + Ro) + 0.8(L + H) + l .4C c r + I .6(Lr or S or R) Severe Environmental Load Combinations 1.2(D + F + Ro) + l.6(L + H + E o) 1.2(D + F +R o)+ l.6(L + H + W) Extreme Environmental and Abnormal Load Combinations D + F + 0.8L + C c r + H + T 0 + Ro + E s s D + F + 0.8L + H +T o+ Ro+ W1 D + F + 0.8 L + C c r + H +T a+ Ra+ l.2P a D + F + 0.8L+ H +Ta+ Ra+ P a+ Y,+ Yj + Ym + E ss D + F + 0.8 L + Cc r + H + T 0 + Ro + s, ea1u* ANSl/AISC N690b (9-1) (NB2-l) (9-2) (NB2-2) (9-3) (NB2-3) (9-4) (NB2-4) (9-5) (NB2-5) (9-6) (NB2-6) (9-7) (NB2-7) (9-8) (NB2-8) (9-9) (NB2-9)

  • AC I 3 4 9, Code R e qui re m en t s for N u cl e ar S af ety-R e lat e d Co n c r e t e St r u ctu r es a nd C omm e n t a ry, Am e ric a n Co n c ret e In s titute , F armin g ton Hill s , Mi c hig a n , 201 3. b ANS UAI S C N 6 9 0 , S p ec ifi c ati o n for Safety-R e l a t e d St ee l S tru c tur es fo r Nu cl e ar Fa c iliti es, Am e ric a n In s titute o f Ste e l C on s tru c ti o n , C hica g o , Illinoi s , Janua ry 31 , 2 01 2. 3.2.1.2 Commercial and Nuclear Non-Safety-Related Structures, Systems, and Components For commercial and nuclear non-safet y-related SSCs , the loading combinations from American Society of Civil Engineers (ASCE) 7 , Chapter 2 are used. When the loading includes earthquake effects , the special seismic load combinations are taken from ASCE 7 , Minimum D es ign Loads for Building s and Oth e r Stru c tur es, Chapter 12. The basic load combinations for the strength design of commercial type and safety-related nuclear SSCs are given in Table 3-10. The combinations li s ted are obtained from the 2012 International Building Code (IBC) and ASCE 7. The crane live load case (C c r) is separated from other live loads in the combinations for de s ign purposes. 3-26

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  • NOllTHWEST MEDICAL ISOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-10. Load Combinations for Strength Base Acceptance Criteria, Commercial Combination Basic Load Combinations 1.4(D +F) I .2(D + F) + I .6(L + Ccr + H) + 0.5(L r or S or R) 1.2(D + F) + 1.6(Lr or Sor R) + l.6H + l/1(L + C cr) or 0.5W] I .2(D + F) +I .OW+ f 1(L + Ccr) +I .6 H + 0.5(Lr or Sor R) l.2(D + F) +I.OE+ f 1(L + Ccr) + l.6H + fiS 0.9D+ I .OW+ I .6H 0.9(D + F) + I.OE+ 1.6H Load Combinations, including Flood Load I .2 D + (0.5W + I .OF.)+ L + 0.5(Lr or Sor R) 0.9D + (0.5W + I .OF a) Load Combinations, including Atmospheric Ice Where: I .2D + I .6L +(0.2Di + 0.5S) l.2D + L +(Di+ Wi + 0.5S) 0.9D +(Di+ Wi) fl = 0.5 for ot h er li ve l oads. 1sca (16-I) (16-2) (16-3) (16-4) (16-5) (16-6) (16-7) §1605.2.1 §1605.2.l §I605.2.l

§1605.2.1

§ 1605.2.I f2 = 0.7 for flat roof co nfi g urations, which do not s h e d snow , a nd 0.2 for other roof config uration s a IB C 2012 , Int e rnational Building Code, Int e rn a tional Code Co uncil , Inc., W as hington D.C. ASCE 7b I 2 3 4 5 6 7 §2.3.3.2 §2.3.3.2 §2.3.4.1 §2.3.4.2 §2.3.4.3 b ASC E 7 , Minimum D es i g n Loads for Buildin gs a nd Oth er St ru c tur es , Americ a n Soc i ety of C i v il E n gi n eers , R es ton , V ir gi nia , 2010. 3.2.2 Combinations for Serviceability Based Acceptance Criteria Based on ASCE 7 , Appendix C Commentary , the load combinations given in Table 3-11 are u se d when evaluating serviceability based acce ptance criteria.

3.2.3 Normal

Loads The RPF is required to resist loads due to: * *

  • Operating conditions of the systems and components within the RPF Normal and severe natural phenomena hazards , remaining operational to maintain life-safety and safety-related SSCs Extreme natural phenomena hazards , maintaining life-safety and related SSCs Table 3-11. Load Combinations for Serviceability Based Acceptance Criteria Combination Short-Term Effects D+L D + 0.5S ASCE7 (CC-la) (CC-lb) Creep, Settlement and Similar Long-Term of Permanent Effects D + 0.5L Drift of Walls and Frames D+0.5L+W a Seismic Drift Per ASCE 7, Section I2.8.6 (CC-2) (CC-3) a Appendix C, Comme nt a ry , of ASC E 7 , Minimum D es i g n Loads for Buildings and Oth e r Structures, American Society of Civi l Engineers, Re s ton , Virgin i a, 2013. 3-27

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  • 0 NOITifMST MEDICAL ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Structural loads are due to the following: * * *
  • Self-weight of building materials and SSC s Occupancy and normal use of the RPF Off-normal conditions and accidents Natural phenomena hazards Section 3.1 de s cribes the s tructural discipline s ource requirements for these crit e ria. Structural load c riteria are summar i zed below. Site-specific natural phenomena hazard criteria a re based on the phy s ical location of the RPF given in Chapter 2.0 , Sections 2.3 and 2.5. 3.2.3.1.1 Dead Loads Dead loads consist of the weight of all materials of construction comprising the building , including walls , floors , roofs , ceilings , confinement doors , stairways , built-in partitions , wall and floor finishes , and cladding. Dead loads also consist of the weight of fixed equipment , including the weight of cranes. The density of all interconnections (e.g., heating , ventilation , and air conditioning

[HV AC] ductwork , conduits , cable trays , and piping) between equipment will be conservatively estimated and included in the final design for dead load for fixtures attached to ceiling s or anchored to floors in the RPF. 3.2.3.1.2 Lateral Earth and Ground Water Pressure Loads Lateral earth and ground w ater pressure loads are lateral pressures due to the weight of adjacent soil and groundwater , respecti v ely. The design lateral earth load is a function of the composition of the soil. The Discovery Ridge Phase I En v ironmental Assessment (Terracon , 2011 a) indicates that the soils present are clayey gravels consistent with the Unified Soil Classification "GC." In addition , the assessment indicates that expansi v e soils are pre s ent. Chapter 2.0 , Section 2.5.3 present s additional on-s ite soil information. The de s ign lateral earth pre ss ure load for the RPF i s based on ASCE 7 , Table 3.2.1 , and ha s been augmented to account for the expansi v e s oils (e.g., s urcharge load is applied to account for the weight of the facility above the soils a djacent to the tank hot cell). The design groundwater depth is estimated to be approximately

5.5 meters

(m) (18 feet [ft]) ground surface and will be v erified pending final g eotechnical investigation.

Additional information i s presented in Chapter 2.0 , Section 2.4.2. The lateral earth pressure loads for the RPF are presented in Table 3-12. Table 3-12. Lateral Ea rth Pressure Loads Element Value B as e design lateral soil load 45 lb/ft2 per ft Design lateral load (expansive increase) 60 lb/ft 2 per ft R efe r e n ce: T a ble 3.2-1 o f A SCE 7 , M in i mum D esig n Loa d s for Buildin gs and Oth e r St ru ctu r es, A m e ri ca n Soc i ety of C i v il E n g in ee r s , R esto n , Vir g ini a , 20 1 3. 3-28

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  • NORTHWEn MEDJCAL lSOTOPH NWMl-2 0 13-02 1 , R ev. 1 C h a p ter 3.0 -D esign of Structures , Systems and C o mp on ents 3.2.3.1.3 Live Loads Floor Live Load Live l oads are produced by the use and occupancy of the RPF , and as such, different live load magnitudes are appropriate for different areas of the facility.

Design floor loads provided in Table 3-13 are based on ASCE 7 , Sections 4.3 and 4.4 , and Section C4.3 Commentary.

During the structural analysis, unknown loads (e.g., hot cell roof in Table 3-13) will have a conservative value assumed and marked with "(HOLD)." As the design matures, the actual values will be inserted in the analysis and the HOLDs removed. Final design media cannot be issued if there are HOLDs identified.

The facility live loads will be established during the completion of the final facility design and provided as part of the Operating License App l ication. Roof Live Load The minimum roof live load (Lr) prescribed by the Tab l e 3-13. F l oor Live Loads Description Uniform Concentrated Production area 250 lb/ft 2 3,000 l b Hot cell roof TBD TBD Cover block la y down TBD TBD Mechanical rooms 200 lb/ft 2 2 , 000 lb Laboratory 100 lb/ft 2 2,000 lb Office 50 lb/ft 2 2 , 000 lb Office partitions 20 lb/ft 2 Corridors 100 lb/ft 2 Truck b ay Per AASHTO Based on Sections 4.3 , 4.4 , an d C4.3 Comme nt ary of ASCE 7, Minimum D es ign loads for Buildings and Other Stru c tur e s , American Socie t y of Civil E n g in eers , Reston , V ir gi nia , 2013. AASHTO TBD American Association of State Highway a nd Transportation Officials.

to be d etermined.

City of Columbia is 20 pounds (lb )/square foot (ft 2), non-reducible (Ordnance No. 21804, Section 6-17). Snow load s (e.g., normal and extreme rain-on-s now) are discussed separately in Section 3.2.5.2. Crane Loads The design ba s i s crane load criteria are given in Table 3-14 and are based on a preliminary quote provided in NWMI-2015-SDD

-001 , RPF Facility SDD. The crane design i s to run a top-running bridge crane with a remotely operated, powered bridge and hoist. The crane design basis will be refined in the fina l de s ign and Operating Licen se Application to account for the following:

  • ASCE 7 , Chapter 3 -Include weights of crane and runway beams in dead load s Tab l e 3-14. Cra n e Loa d C ri teria Element Crane capacity Crane weight (with hoists) Bridge weight Hoist and trolley weight Wheel load (stat ic) Value 75 ton (150 kip) 69,990 !bf 62,330 !bf 7,660 !bf 54.3 kip
  • ASCE 7 , Chapter 4 -Increase wheel load by 25 percent to account for vertical impact *
  • ASCE 7, Chapter 4 -Determine l ateral force by multiplying sum of hoist and trolley weight and rated capacity of crane by 20 percent ASCE 7 , Chapter 4 -Determine longitudinal force by multiplying the wheel load by 10 percent 3-29

..... NWMI ..*... .. *.. ........ *.* * *. * ' . NOITNWUT MUHCAl lSOTWU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components 3.2.4 Wind Loading 3.2.4.1 Wind Load PerNUREG-1537, Section 2.3.1 , "General and Local Climate, wind loads will be based on the 100-year return period wind speed. In addition, based on NUREG-0800 , Standard Revi e w Plan for the R e view of Saf e ty Anal ys is R e port s for Nuclear Pow e r Plants , Section 3.3.1 , the wind speed will be transformed to equivalent pressure per ASCE 7-05. For RPF SSCs per current applicable 2012 IBC guidance , ASCE 7-10 is used for this transformation of wind speed to equivalent pressure.

From Table 1.5-1 of ASCE 7-10 and based on u s e and occupancy of the RPF , a Risk Category IV is assigned to RPF SSCs. Figure 26.5-I B for a Risk Category IV building of ASCE 7-10 is used to obtain the basic wind speed for the RPF site. The mean recurrence interval (MRI) of the basic wind speed for Risk Category IV buildings is 1 , 700 years. Since the MRI stipulated in ASCE 7-10 is more stringent than NUREG-1537 100-year wind speeds , wind loads will be determined in accordance with ASCE 7-10 , Chapters 26 through 30 , as applicable , for a Risk Category IV building.

The surface roughness surrounding RFP SSCs is currently Surface Category C , which in turn indicates Exposure Category C for the RFP per ASCE 7-10. The RPF main building is an enclosed building.

The wind loading criteria are provided in Table 3-15. The basic wind speed given in Table 3-15 is a 3-second (sec) gust wind speed at 10 m (33 ft) aboveground for Exposure Category C and Risk Category IV. The wind loading criteria will be updated in the Operating License Application.

3.2.4.2 Tornado Loading Table 3-15. Wind Loading Criteria Element B as ic wind speed , V Exposure category E nclo s ur e cl ass ific a tion Risk category Value 193.1 km/hr (120 mi/hr) c Enclosed IV S o urc e: AS CE 7-10 , Minimum D es i gn L o ad s f o r Buildin gs and O t h er S tru c t u r es, Am e rican Soci e ty of C i v il E n g in ee r s, R es ton , Vir g inia , 2 010. Tornado loads are a combination of tornado wind effects , atmospheric pressure change , and generated missile impact effects and are discussed separately in the following sections.

NUREG-1520 , Standard R e view Plan for th e R e vi e w of a Li ce ns e Application for a Fu e l C y cl e Fa c ility , Part 3 , Appendix D , states that an annual exceedance probability of 10-5 may need to be considered. The maximum tornado wind speed from NRC Regulatory Guide 1.76 , D es ign-Ba s i s Tornado and Tornado Mi s sil es for N ucl e ar Po we r Plant s , for Region I , has an annual exceedance probability of 10-7 that is significantly lower than the target probability stated in NUREG-1520. For the RPF preliminary safety analysis report , the maximum tornado wind s peed from NRC Regulatory Guide 1.76 for Region I will be used. The tornado load criteria will be updated b y using tornado lo ading in accordance with 10-5 annual probability of exceedance in the Operating License Application.

3-30

..... NWMI ..**.. ..* ... ..... .. .. .. 0 *. NORTHWUT MEDICAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components 3.2.4.2.1 Maximum Tornado Wind Speed Tornado wind field characteristics used to calculate tornado wind pressures on the RPF are provided in Table 3-16 per NRC Regulatory Guide 1.76. The maximum tornado wind speed has two components: translational and rotational.

The maximum total tornado wind speed is the sum of these two components and is applied to the RPF building from each direction separately. Based on NUREG-0800 , Section 3.3.2 , ASCE 7-05 may be used to transform maximum tornado wind speed to equivalent pressure. Table 3-16. Design-Basis Tornado Field Characteristics Description Tornado region Maximum wind speed Translational speed Radius of maximum rotational speed Pre s sure drop , Af> Value Region I 370.1 km/hr (230 mi/hr) 74.0 km/hr (46 mi/hr) 45.7 m (150 ft) (1.2 lb/i n.2) Sourc e: NRC Regulatory Guid e I. 76 , D es i g n-Basi s T o rnado and Tornado Mi ss il es for Nucl e ar Po we r Plant s , R e v. I , U.S. Nuclear R e gulatory C ommission , Washington , D.C., March 2007. For RPF SSCs per current applicable 2012 IBC guidance , Chapters 26 and 27 of ASCE 7-10 is used for this transformation of tornado wind speed to equivalent pressure.

From Table 1.5-1 of ASCE 7-10 and based on use and occupancy of the RPF , a Risk Category IV is assigned to RPF SSCs. Per NUREG-800 , Section 3.3.2 , tornado wind speed is assumed not to vary with the height aboveground.

Additional information is provided in Chapter 2.0 , Section 2.3.1.5 , and Chapter 13.0 , Section 13.2.6. l. 3.2.4.2.2 Atmospheric Pressure Change NRC Regulatory Guide 1. 76 provides guidance for determining the pressure drop and the rate of pressure drop caused by the passing of a tornado. Depending on the final design of the RPF building and whether it is enclosed (unvented) or partially enclosed (vented structure), the procedures outlined in NUREG-800 Section 3.3.2 will be used to account for atmospheric pressure change effects. At the preliminary stage of the design, the RPF building is known not to be open. The value for atmospheric pressure drop, corresponding to the design-basis tornado is given in Table 3-16. 3.2.4.2.3 High Straight-Line Winds Similar to the tornado , high straight-line winds can also damage the facility structure, which in tum can lead to damage to SSCs relied on for safety. This evaluation demonstrates how the facility design addressed straight-line winds with a return interval of 100 years or more, as required by building codes. The RPF is designed as a Risk Category IV structure , a s tandard industrial facility with equivalent chemical hazards , in accordance with ASCE 7. The return frequency of the basic (design) wind speed for Risk Category IV structures is 5.88 x J0-4/year (MRI= 1 , 700 year). The provisions of ASCE 7 , when used with companion standards such as American Concrete Institute (ACI) 318 , Building Code Requir e ment s for Structural Concrete , and American Institute of Steel Construction (AISC) 360 , Sp ec ification for Stru c tural Ste e l Buildings , are written to achieve the target maximum annual probabilities of established in ASCE 7. The highest maximum probability of failure targeted for Risk Category IV structures is 5.0 x l0-6. 3.2.4.2.4 Tornado-Generated Missile Impact Effects Tornado-generated missile impact effects are based on the standard design missile spectrum from NRC Regulatory Guide 1. 7 6 and are presented in Table 3-17. These requirements are considered more severe than the characteristics from DOE-STD-1020 , Natural Ph e nomena Ha z ards D e sign and Evaluation Crit e ria for D e partment of En e rgy Faciliti es, that are cited in NUREG-1520 , Section 3. The recommended RPF roof and wall system design criteria are also taken from DOE-STD-1020, Table 3-4. 3-31

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  • NORTHWEST MEDtCAl JSOTOftl:S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Description Automobile Pipe Steel Sphere Table 3-17. Design-Basis Tornado Missile Spectrum *uwn* Dimensions 4,000 lb I 6.4 ft x 6.6 ft x 4.3 ft 287 lb 6.625 in. diameter x 15 ft long 0.147 lb 1.0 in. di ameter Horizontal velocity 92 mi/hr 92 mi/hr 18 mi/hr Vertical velocity 62 mi/hr 62 mi/hr 12 mi/hr Source: NRC Regulatory Guide 1.76 , D esig n-Ba s i s Tornado and Tornado Missil e s for Nucl e ar Pow e r Plants , U.S. u c l ear R egulatory Commission , Washington , D.C., March 2007. The impact-type missile , an automobile is limited to a hei g ht of no more than 9.1 m (30 ft) above-grade.

Structural wall openings are subjected to the impact of a 0.25 centimeters (cm) (I-inch [in.]) diameter steel s phere. The vertical velocities are taken as 0.6 7 of the horizontal ve locity. For an automobile and pipe missile , a normal impact is assumed. The tornado load criteria will be updated by using tornado loading in accordance with 10-5 annual probability of exceedance in the Operating License Application and accordingly, the design-basis tornado missile spectrum will also be updated. 3.2.4.2.5 Combined Tornado Load Effects After tornado-generated wind pres s ure effects, atmospheric pres sure change effects and mi ssile impact effects are determined

the combination thereof will be established in accordance with procedure s outlined n NUREG-800, Section 3.3.2. The effect of atmospheric pressure drop by itself will be considered, and the total effects of wind pressure and mis s ile impact effects with one-half of the atmospheric pressure drop effects will be considered jointly. 3.2.4.3 Effect of Failure of Structures, Systems, or Components Not Designed for Tornado Loads SSCs, in which failure during a tornado event could affect the safety-re lated portions of the RPF , are either de s igned to
*
  • Resist the tornado loading or the effect on the safety-related s tructures from the failure of these SS Cs Be bounded by the tornado missile or aircraft impact evaluations The effects and mitigations of failure of SSCs not de signe d for tornado loads will be developed during final design and the Operating License Application.

3.2.5 Rain, Snow, and Ice Loading 3.2.5.1 Rain Loads From the National Weather Service (NWS)/National Oceanic and Atmospheric Administration (NOAA) Hydrometeorological Report No. 51 , Probable Maximum Pr ec ipitation Estimates , United States East of the 105th Meridian, the probable maximum precipitation (PMP) is defined as "t heoretical greatest depth of precipitation for a given duration that is physicall y possible over a particular drainage area at a certain time of yea r." Per NUREG-1537, Section 2.3.1 , " General and Local Climate," rain loads will be based on the estimate of the weight of the 48-hour (hr) probable maximum precipitation , as specified by the U.S. Geological Survey. This rain load estimate is compared with the local building code rain load (i.e., ASCE 7-10), and the greater value is used in design of the RPF roof. The roof of the RPF is de signe d to prevent rainwater from accumulating on the roof. In accordance with 2012 IBC and ASCE 7-10 , the roof structure is designed to safely support the weight of rainwater accumulation with the primary drainage system blocked and the secondary drainage system at its design flow rate when subjected to a rainfall intensity based on the 48-hr probable maximum precipitation.

3-32

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  • NOmfWEU MEDtCAl tsOTOPl:S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-18. Rain Load Criteria Element Value Rain loads are determined by the amount of water that can accumulate on the undeflected building roof if the primary drainage system becomes blocked (static head), plus a uniform depth of water above the inlet of the secondary drainage system at its design flow (hydraulic head). The rain load criteria are determined per Static head Hydraulic head Rainfall int e nsity 5 cm (2-in) TBD 3 .14 in./hr" ASCE 7-10 , Chapter 8 , and are provided in Table 3-1 8.
  • N O AA Atl as 14 , Pr ec ipita t i o n-Fr eq u e n cy A tlas o f th e U nit e d Stat es , V o lume 8 , Ver s ion 2.0: Midw es t e rn Sta te s , N a ti o n a l O cea nic a nd Atmo s ph e ri c Admini s trati o n , Silver Sprin g, M a ryland , 201 3. TBD = to be d e t e rm i n e d. The hydraulic head is dependent on the roof drain size , roof area drained , and the rainfall intensity. The rainfall intensity used to determine the hydraulic head is taken from NOAA Atlas 14 , Pre c Frequenc y Atlas of the Unit e d State s, web tool for the 100-year storm , 1-hr duration. The rain load criteria will be updated in the Operating License Application. 3.2.5.2 Snow Load Per the guidance in DC/COL-ISG-007, two types of snow load are considered
normal snow load and the extreme winter precipitation load. The normal snow load will be included in normal load combinations given below. Per the guidance in the DC/COL-ISG-007 , the extreme winter precipitation load i s included in the extreme environmental load combinations.

The snow load criteria will be updated in the Operating License Application. 3.2.5.2.1 Normal Snow Load Per NUREG-1537 , Section 2.3.1 and DC/COL-ISG-007 , the normal snow load is the 100-year ground snow , modified using the procedures of ASCE 7 to determine the roof snow load , including snow drifting. The 100-year ground snow load is calculated by factoring the ground snow load stipulated in the City of Columbia Code of Ordinances amendments (City of Columbia , 2014) and IBC 20 I 2 and is equivalent to the mapped ground snow load from Figure 7-1 of ASCE 7. Thi s information is determined using the conversion factor provided in ASCE 7 , Table C7-3. The exposure factor provided in ASCE 7 , Table 7-2 , for partially exposed roof in terrain category C is similar with the exposure used for determining wind loads. Since the RPF does not fall into any of the special cases indicated in ASCE 7 , Table 7-3 , the thermal factor is assumed to be 1.0. The importance factor is taken to be unity from ASCE 7-10 , Table 1.5-2 , for the RPF , which is designated Risk Category N. Snow load criteria are summarized in Table 3-19. 3.2.5.2.2 Extreme Winter Precipitation Load Table 3-19. Snow Load Criteria Element Mapp e d ground snow load (50-y ear) Conversion factor , I 00-year to 50-year De s ign ground s now load , p g (100-year)

Exposure factor (C e) Th e rm a l factor (Ci) Importance factor Value *20 lb/ft 2 b0.82 24.4 lb/ft 2

  • C ity of Columbi a , "C ity of C olumbia C ode of Ordinance s ," www.goc olumbiamo.com/C ouncil/C od e _of_ Ordinances

_PDF/, a cc esse d Septemb e r 8 , 2 014. b ASCE 7 , Minimum D es i g n L o ad s fo r Buildin g s and Oth e r Stru c tu r es, Am e rican So c i ety of C ivil E ngin ee rs , R es ton , Virgini a, 2 013. Per NUREG-1537 , Section 2.3.1 and DC/COL-ISG-007 , the extreme winter precipitation load is the normal snow load as presented in Section 3.2.5.2.1 , plus the liquid weight of the 48-hr probable maximum winter precipitation (PMWP). 3-33

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  • NORTHWEST MEDtcAllSOTOrES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components The 48-hr PMWP is determined from the NOAA/NWS Hydrometeorological Report (HR) 53 , Seasonal Variation of 10-Squar eMile Probable Maximum Precipitation Est imat es, United States Ea s t of the J05 th Meridian, for a 10-mi 2 area. HR 53 gives midmonth PMP estimates for six 24-and 72-hr durations. Using the NOAA we b tool for Co lumbia (NOAA , 2017), a two-day ( 48-hr) average 100-year rain i s 8. 73 in. precipitation.

To determine the PMWP , the months of December , January , February , and March are co nsidered. Using HR 53 , Figures 26 through 45 , the PMWP was determined to occur in the month of March. The PMWP criteria are g i ve n in Table 3-20. Table 3-20. Extreme Winter Precipitation Load Criteria Element 2 4-hr , I O-mi 2 PMWP 72-hr, I O-mi 2 PMWP 48-hr , I O-mi 2 PMWP (i nterpol ated) Weight of 48-hr PMWP Value 4 6.7 cm (18.2 in.)" 56.9 cm (22.5 in.)" 22.2 cm (8.73 in.) 106 lb/ft 2

  • N WS/NOAA HR 53, Seasona l Variat i on of J O-Square-Mile Probabl e Maximum Pr ec ipitation Estimates, U ni ted States East of t h e 1 05th Meridian , Nat i o n a l Oceanic and Atmospheric Adminis tr at i o n , S il ver Spring , M ary l and, 1 980. PMWP prob ab l e m ax imum winter pre c ipit ation. Winter weather events si nc e 1996 in Boone County, Missouri, are pro v ided in Chapter 2.0 , Table 2-36. 3.2.5.3 Atmospheric Ice Load For SSCs to be considered se nsitive to ice , the ice thickness and concurrent wind load s are determined using the procedur es in ASCE 7 , C hapter 10. Consistent wi th the req uirem en t s for s now and wind load s, the mapp e d va lues a re converted to 100-year values usin g the MRI multiplier s given in ASCE 7 , Tabl e C 10-1. Criteria for ice loading are give n in Table 3-21. Table 3-21. Atmospheric Ice Load Criteria Element Ice thickness (50-year)

Concurrent wind speed Ice thickness MRI multiplier Wind speed MRI multiplier Importance factor Value 3 2.54 cm (I in.) 64.4 km/hr (40 mi/hr) 1.25 1.00 1.00

  • ASCE 7 , Minimum D es i gn Load s for Buildin gs and Oth e r Structures, Am e rican Soc i ety of Civi l Engineers, R es t on , Virgin i a, 2013. MRI = m ea n r ec u rre nc e interval.

3.2.6 Operating

Thermal/Self-Straining Loads The operating thermal/self-straining load s will be evaluated in the Op e rating License Application.

These loads will be consistent with the requirement s of ACI 349 or ANSI/ AISC N690 , as applicable to the material of construction.

3.2. 7 Operating Pipe Reaction Loads The operating pipe reaction loads will be evaluated in the Operating License Application.

These loads will be con sis tent with the requirement s of applicable American Society of Mechanical Engineers (AS ME) B3 l , Standards of Pr ess ure Piping , codes. 3.2.8 External Hazards External ha za rds include aircraft impact , external explosions , and external fire. The RPF is a production facility, as opposed to a nuclear power reactor , as such 10 CFR 50.150(a)(3) is interpreted to mean that the requirement for the aircraft impact assessment is not applicable to this facility. Sources of accidental external explosions have been considered and were found to not be an accident of concern. The RPF is c onstructed of robust, noncombustible materials, and adequate setbacks from transportation routes and landscaping consisting of fire fuels are provided such that externals fires are not an accident of concern. 3-34

3.3 WATER

DAMAGE NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components This section identifies the requirements and guidance for the water damage de s ign of the RPF SSCs. NUREG-1520 and ASCE 7 , Chapter 5 , provide guidance on flood protection of nuclear safety-related SSCs. Updates and development of technical specifications associated with the water damage design of the RPF SSCs will be provided in Chapter 14.0 as part of the Operating License Application. 3.3.1 Flood Protection This subsection discusses the flood protection measures that are applicable to safety-related SSCs for both external flooding and postulated flooding from failures of facility components containing liquid. A compliance review will be conducted of the as-built design against the assumptions and requirements that are the basi s of the flood evaluation presented below. Additional information is presented in Chapter 2.0 , Section 2.4.3 and Chapter 13.0 , Section 13.2.6.4. This as-built evaluation will be documented in a flood analysis report and be part of the Operating License Application.

3.3.1.1 Flood Protection Measures for Structures, Systems, and Components 3.3.1.1.1 Flooding from Precipitation Events Regional flooding from large precipitation events raising the water le v els of local streams and rivers to above the 500-year flood level can have an adverse impact on the structure and SSCs within. These impacts include the structural damage from water and the damage to power supplies and instrument control systems for SSCs relied on for safety. The infiltration of flood water into the facility could cause the failure of moderation control requirements and lead to an accidental nuclear criticality.

Direct damage or impairment of SSCs could also be caused by flooding in the facility. The site will be graded to direct the stormwater from localized downpours with a rainfall inten s ity for the 100-year storm for a I-hr duration around and away from the RPF. Thus , no flooding from local downpours i s expected based on standard industrial de s ign. Rainwater that falls on the waste management truck ramp and accumulates in the trench drain has low to no consequence for radiolo g ical , chemical , and criticality ha z ards. Situated on a ridge , the RPF will be located above the 500-year flood plain according to the flood insurance rate map for Boone County , Mi ss ouri , Panel 295 (FEMA , 2011). The s ite is above the elevation of th e nearest bodie s of water (two small ponds and a lake), and no dams are located upstream on the local streams. This data conservatively provides a 2 x 10-3 year return frequency flood , which can be considered an unlikely e vent according to performance criteria.

However , the site is located at an elevation of 248.4 m (815 ft), and the 500-year flood plain starts at an elevation of 231.6 m (760 ft), or 16.8 m (55 ft) below the site. Since the s ite , located only 6.1 m (20 ft) below the nearest high point on a ridge (relative to the local topography), is well above the beginning of the 500-year flood plain , and is considered a dry s ite , the probable maximum flood from regional flooding is considered highly unlikely , without further e v aluation.

1 1 The re c ommended s tandard for determining the probably m a ximum flo o d , ANS 2.8, D e t e rminin g D es i g n Ba s i s Flo o din g at P o w e r R e a cto r Sit es, has been withdrawn. 3-35

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  • HOITHWEST MEDtCAL tsOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Per NUREG-1520, Section 3.2.3.4(1)(c), and ASCE 7 , Chapter 5, flood loads will be based on the water level of the 100-year flood (one percent probability of exceedance per year). The facility has been determined to be above both the 100-year and the 500-year flood plain. Chapter 2 , Section 2.4.3, provides additional detail for flood protection measures.

Postulated flooding from component failures in the building compartments will be prevented from adversely affecting plant safety or posing any hazard to the public. Exterior or access openings and penetrations into the RPF will be above the maximum postulated flooding level. Therefore , flood loads are considered highly unlikely and are not considered design loads. 3.3.1.1.2 Flooding from Inadvertent Discharge of Fire Protection System Water Design of fire suppression systems using water (e.g., automatic sprinklers, hose stations) includes elements such as the grading and channeling of floors , raising of equipment mounts above floors , shelving and floor drains , and other passive means. These features will ensure sufficient capacity for gravity-dri ven collection and drainage of the maximum water discharge rate and duration to avoid localized flooding and resulting water damage to equipment within the area. In addition, particularly sensitive systems and components, whether electrical, optical , mechanical and/or chemical, are typically protected within enclosures designed for the anticipated adverse environmental conditions resulting from these types of water discharges. If critical for safety, these water-sensitive systems and components will be installed within the appropriate severe environment-rated enclosures in accordance with the relevant industry standard(s) (e.g., Nationa l Electrical Manufacturers Association

[NEMA] enclosure standards).

Selection of specific fire suppression systems for facility locations will be guided by recommendations in relevant industry standards (e.g., NFPA 801, Standard for Fire Protection for Faciliti e s Handling Radioactive Materials) and will depend on the level of fire hazards at those lo cations , as determined from the final facility and process systems designs. These final detailed designs will include any facility design elements and sensitive equipment protection measures deemed necessary for addressing the maximum inadvertent rate and duration of water discharges from the fire protection systems. The final comprehensive facility design , along with commitments to design codes , standards , and other referenced documents (including any exceptions or exemptions to the identified requirements), will be identified and provided as part of the Operating License Application.

3.3.1.2 Flood Protection from External Sources Safety-related components located below-grade will be protected using the hardened protection approach. The safety-related systems and components will be protected from external water damage by being enclosed in a reinforced concrete safety-related structure.

The RPF will have the following characteristics:

  • * *
  • Exterior safety-r elated walls below-grade will be 0.6 I m (2-ft) thick minimum Water stops will be provided in all construction joints below-grade Waterproof coating will be applied to external surfaces below-grade and as required above-grade Roofs will be designed to prevent pooling of large amounts of water in accordance with Regulatory Guide 1.102 , Flood Protection for Nuclear Power Plants Waterproofing of foundations and walls of safety-related structures below-grade will be accomplished primarily by the use of water stops at expansion and construction joints. In addition to water stops , waterproofing of the RPF will be provided to protect the external surfaces from exposure to water. The level above the RPF first level where waterproofing is to be used will be determined in the Operating License Application. 3-36

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  • NOllTIIWEST MEDICAi. ISOTOHS NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components The flood protection measures that are described above will also guard against flooding from the rupture of the on-site fire protection water storage tank (if future design development determines that a fire protection storage tank is necessary).

Any flash flooding that may result from tank rupture will drain away from the RPF and thereby cause no damage to facility equipment.

3.3.1.3 Compartment Flooding from Fire Protection Discharge The total discharge from the failure of fire protection piping consists of the combined volume from any sprinkler and hose systems. The sprinkler system , if used , is capable of delivering a water density of 20 gallons per minute (gal/min) (76 liters per minute [L/min]) over a 139 m 2 (1 , 500 ft 2) design area; therefore, the sprinkler system is calculated to have a flow rate of 1, 136 L/min (300 gal/min).

The hose stream will be a manually operated fire hose capable of delivering up to 946 Umin (250 gal/min). In accordance with NFP A 801 , Section 5 .10 , the credible volume of discharge is sized for the suppression system operating for a duration of 30 min. The design of water-sensitive , safety-related equipment will ensure that potential flooding from sprinkler discharge will not adversely affect the safety features.

For example , equipment may be raised from the floor sufficiently such that the potential flooding due to sprinkler discharge will not impact the criticality analyses.

Outside of the radiologically controlled area (RCA), as defined in Chapter 11.0 , " Radiation Protection and Waste Management,'

' there is limited water discharge from fire protection systems. Any sensitive, safety-related equipment will be installed above the floor slab at-grade to ensure that the equipment remains above the flooded floor during sprinkler discharge. 3.3.1.4 Compartment Flooding from Postulated Component Failures Piping, vessels , and tanks with flooding potential in the safety-related portions of the RPF will be seismically qualified.

Water-sensitive, safety-related equipment will be raised above the floor. The depth of water on the floor will be minimized by using available floor space to spread the flood water and limiting the water volumes. Analyses of the worst flooding due to pipe and tank failures and their consequences will be developed in the Operating License Application.

3.3.1.4.1 Potential Failure of Fire Protection Piping The total discharge from the operation of the fire protection system bounds the potential water collection due to the potential failure of the fire protection piping. 3.3.1.5 Permanent Dewatering System There is no permanent dewatering system provided for the flood design. 3.3.1.6 Structural Design for Flooding Since the design PMP elevation is at the finished plant-grade and the probable maximum flood (PMF) elevation is approximately 6.1 m (20 ft) below-grade , there is no dynamic force due to precipitation or flooding. The lateral surcharge pressure on the structures due to the design PMP water level is calculated and does not govern the design of the below-grade walls. The load from buildup of water due to discharge of the fire protection system in the RCA is supported by slabs-on-grade, with the exception of the mezzanine floor. Drainage is provided for the second level in the RCA to ensure that the second level slab is not significantly loaded. The second level slab is designed to a live load of 610 kilograms (kg)/m 2 (125 lb/ft 2); therefore , the slab is capable of withstanding any temporary water collection that may occur while water is draining from that floor. 3-37

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  • NOmfWUT MEotc:Al ISOTOHS 3.4 SEISMIC DAMAGE NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Seismic analysis criteria used for the RPF will conform to IAEA-TECDOC-134 7 , Con s id e ration of E x t e rnal Ev e nt s in the D es ign of Nucl e ar Faciliti e s Oth e r Than Nucl e ar Pow e r Plants , with Emphasis on Earthquak es. This report provides requirements and guidance for the seismic design of nuclear facilities other than nuclear power plants. NUREG-0800 and other NRC Regulatory Guides provide additional detailed guidance for the seismic analysis and design of the RPF. Additional information is provided in Chapter 2.0 , Section 2.5.4 , and Chapter 13.0 , Section 13.2.6.5.

Updates and development of technical s pecifications associated with the seismic damage design of the RPF SSCs will be provided in Chapter 14.0 as part of the Operating License Application.

3.4.1 Seismic

Input 3.4.1.1 Design Response Spectra Safe-Shutdown Earthquake The NRC has recommended using Regulatory Guide 1.60 , D e sign R es ponse Sp ec tra for Sei s mi c Design o f N uclear Po we r Plant s, for radioisotopes production facilities (e.g., IO CFR 50). NWMI will use a spectrum anchored to 0.20 g peak ground acceleration for the RPF design basis. Regulatory Guide 1.60 is not indexed to any specific soil type , with its frequency content sufficiently broad to cover all soil types. Therefore , soil type for the RPF will not be a parameter used to determine the RPF's design response spectra. The composition of soil in which the RPF is embedded will be included in the interaction analysis as part of the building response analysis.

This information will be provided in the final safety analysis report (FSAR) as part of Operating License Application.

This peak ground acceleration matches that of the University of Missouri Research Reactor and the Calloway Nuclear Generating Station , which both are within 80.5 km (50 mi) of the RPF , as suggested by the NRC staff during the November I 0 , 2016 Public Meeting. The analysis procedure develops ground motion acceleration time histories that match or exceed the Regulatory Guide 1.60 spectrum as i nput to the building finite element model. Structural damping will follow the recommendations of Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Pow e r Plants , which range from about 3 to 7 percent. Response spectra corresponding to the recommended damping values of Regulatory Guide 1.61 will be used to derive seismic loads. Damping varies depending on the type of SSC. Structural damping will follow Regulatory Guide 1.61 guidance (ranging from about 3 to 7 percent).

Plotting response spectra at 5 percent damping for purposes of illustration is a convention within the nuclear industry, but for analysis loads , damping will vary depending on the earthquake level (operating basis earthquake or safe-shutdown earthquake) and the type of SSC. Soil-Structure Interaction and Dynamic Soil Pressures The structure is supported on a shallow foundation system on stiff competent soils. The Phase 1 Assessment (Terracon, 201 la/b) stated the site is classified as Site Class C. Prescribed in ASCE 7 , Table 20.3-1 , the typical shear wave velocities for the soils present at the site are 1 , 200 to 2 , 500 ft/sec. Typical practice is to define competent soil as having a shear wave velocity greater than 1 , 000 ft/sec. The analysis of the RPF building structure to the safe shutdown earthquake will include the effects of a structure interaction. Dynamic so il pressures were determined using ASCE 4 , S e i s mic Anal y si s of Saf eR e lat e d Nucl e ar Structur es and Comm e ntary , Section 3.5.3.2, and applied to the earth retaining walls in the hot cell area.

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3.4.1.2 Method of Analysis The effect of loads other than earthquake-induced (seismic) loads is determined by static analysis methods in accordance with ASCE 7 and the fundamental principles of engineering.

Seismic analysis of SSCs will be performed by either equivalent-static methods or dynamic analysis method s in accordance wi th ASCE 4 and ASCE 43, Seismic D esign Criteria for Structures, Systems, and Compo n e nt s in Nuclear Facilities.

The equivalent-static and dynamic seismic analysis methods are discussed below. 3.4.1.2.1 Equivalent-Static Analysis E quivalent-static seismic analysis of commercial type st ructure will be performed in accordance with ASCE 7 , Section 12.8. Direction of Seismic Loading Design ofIROFS will consider seismic loads in all three directions using a combination of the-sum-of-squared or 10/40/40 methodologies per Regulatory Guide 1.92, Combining Modal R esponses and Spatial Components in Seismic R esponse Analysis.

The I 0/40/40 methodology will be used in the development of the final RPF design and included as part of the Operating License Application. 3.4.1.2.2 Dynamic and Static Analysis Dynamic analyses are only u se d for the evaluation of RPF structural components.

A static analysis will be completed during final design by using a combination of static load computations to ensure the SSCs remain in place and intact , and a combination of existing s hake table te s t data and existing earthquake experience to ensure that the equipment functions following the earthquake.

The analysis of related s tructures may be either completed by the: *

  • Linear-elastic response spectra method performed in accordance with ASCE 4 , Section 3.2.3.1, and ASCE 43 , Section 3.2.2 Linear-elastic time hi s tory method performed in accordance with ASCE 4, Section 3.2.2, and ASCE 43 , Section 3.2.2 Damping -The damping values used for dynamic analysis for the s tructural system considered will be taken from Regulatory Guide 1.6 1. Inelastic energy adsorption factors and damping values used for the analysis of nuclear safety-related structures will be selected from ASCE 43 , Table 5-1. Modeling -Finite element models will only be used for the RPF building structures.

The mesh for plate elements and member nodes will be selected to provide adequate discretization and distribution of the mass. Further , the aspect ratio of plate elements will be limited to no greater than 4: l to ensure accurate analysis results. Direction of seismic loading -Three orthogonal directions of seismic loading are used in the RPF de s ign, two horizontal and one vertical.

The modal components of the dynamic analysis and the spatial components ofresponse analysis are combined as described below. 3-39

......... *.* ..... .. NWMI ..... .... .. * ." NOllTHWUT llEDK:Al ISOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components

  • 3.4.2 Modal combinations -The structure of the RPF is designed to be relatively stiff, and components are combined using the complete quadratic combination method. Spatial component combinations

-Spatial components are calculated separately and combined using the square-root-sum-of-the-squares method to determine the combined earthquake effect and resulting demands. Seismic Qualification of Subsystems and Equipment This subsection discusses the methods by which the RPF systems and components are qualified to ensure functional integrity.

Based on the characteristics and complexities of the subsystem or equipment , seismic qualification will be done by a combination of static load computations to ensure that the SSCs remain in place and intact , and a combination of existing shake table test data and existing earthquake experience to ensure that the equipment functions following the earthquake.

3.4.2.1 Qualification by Analysis NWMI will define specific acceptable qualification methods in the procurement packages to demonstrate seismic qualifications.

Seismic qualification of IROFS will include three options of: (1) calculations and verification that the main structural components of the SSC can withstand the seismic loads derived from the in-structure floor response spectra at the damping value derived from Regulatory Guide 1.61 , (2) reference to available shake table testing that demonstrates the seismic capacity of the SSC or of multiple similar items, and (3) demonstrat i on of the seismic capacity through the performance of the type of SSC in actual earthquakes. 3.4.2.1.1 Equivalent Static Analysis The equivalent static analysis of nuclear safety-related subsystems and equipment is performed in accordance ASCE 43, Section 8.2.1.1. The equivalent static analysis of subsystems and equipment that are not relied on for nuclear safety but are designated as a component of a seismic system per IBC 2012 , Chapter 17 , is performed in accordance with ASCE 7 , Chapter 13. 3.4.2.1.2 Static Analysis The static analysis of non-structural , safety-related subsystems and equipment is performed in accordance ASCE 4 , Section 3.2.3.1 , and ASCE 43 , Section 8.2.1.2. A portion of the seismic qualification process will involve simple static analysis of the main structural elements (anchorage and primary framing) of IROFS components , using seismic loads from in-structure response spectra derived from the RPF building structure dynamic response analysis.

In-structure response spectra are determined using ASCE 4 , Section 3 .4.2, and NRC Regulatory Guide 1.122, Development of Floor Design Response Sp e ctra for Sei s mic Design of Floor-Supported Equipm e nt or Compon e nts. In-structure floor response spectra will be developed through a finite element model of the RPF building using an artificial time history that matches or envelops the Regulatory Guide 1.60 spectrum at a peak ground acceleration

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  • NORTHWEST MEDtCAL ISOTOPES 3.4.2.2 Qualification by Testing NWMI will define specific acceptable qualification methods in the procurement packages to demonstrate se ismic qualifications.

Seismic qualification of IROFS will include three options of: (1) calculations and verification that the main s tructural components of the SSC can withstand the se ismic loads derived from the in-structure floor respon se spectra at the damping value derived from Regulatory Guide 1.61, (2) reference to available s hake table te s ting that demonstrates the seismic capacity of the SSC or of multiple similar items , and (3) demonstration of the seismic capacity through the performance of the type of SSC in actual earthquakes.

Per NRC Regulatory Guide 1.100 , Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active M ec hani ca l Equipment for Nuclea r Power Plants: *

  • Active mechanical equipment relied on for or important to nuclear safety will be required to be seismically qualified in accordance with Regulatory Guide 1.100. Active electrical equipment important to or relied on for nuclear safety will be required to be seismically qualified in accordance with IEEE 344, IEEE Standard for Seismic Qualifi cation of Equipment for Nuclear Pow er Generating Stations.

Subsystems and equipment not relied on for nuclear safety but de s ignated as a component of a seismic syste m per IBC 2012, Chapter 17 , will be required.

Existing databa ses of past shake table tests will be used , such as the Office of Statewide Health Planning and Development database provided by the state of California.

These tests have typically been done based on the ICC-ES AC156, "Acce ptance Criteria for Seismic Certification by Shake-Table Testing of Nonstructural Components," spectrum.

The capacity of the standard s upport design for overhead fixtures mounted above RPF IROFS will be checked to ensure that the supports can withstand the seismic loads derived from the floor spectra (e.g., remain stable during and after postulated earthquake effects) of the attachment floor slab. Thi s information will be provided in the FSAR as part of the Operating License Application.

The RPF seismic design will also include a check to ensure that pounding or sway impact will not occur between adjacent fixtures (e.g., rattle space). Estimates of the maximum displacement of any fixture can be derived from the appropriate floor response spectrum and an estimate of the fixture's lowest response frequency.

This information will be provided as part of the Operating License Application.

3.4.3 Seismic

Instrumentation Seismic recording instrumentation will be triaxial digital systems that record accelerations versus time accurately for periods between 0 and 10 sec. Recorders will have rechargeable batteries such that if there is a loss of power , recording will still occur. All instrumentation will be housed in appropriate weather and creature-proofed enclosures.

As a minimum , one recorder should be located in the free-field mounted on rock or competent ground generally representative of the site. In addition , at sites classified as Seismic Design Category D , E , or Fin accordance with ASCE 7 , Chapter 11 , using Occupancy Category IV , recorders will be located and attached to the foundations and roof s of the RPF and in the control room. The systems will have the capability to produce motion time histories. Response spectra will be computed separately.

The purpose of the instrumentation is to (1) permit a comparison of measured responses of C-1 structures and selected components with predetermined results of analyses that predict when damage might occur, (2) permit facility operators to understand the possible extent of damage within the facility immediately following an earthquake, and (3) be able to determine when an safe-shutdown earthquake event has occurred that would require the emptying of the tank(s) for inspection as specified in NFPA 59A , Standard for the Production , Storage , and H and ling of Liquefied Natural Gas , Section 4.1.3.6( c ). 3-41

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  • NORTHW(n 11(01CA1.ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Seismic instrumentation for the RPF site is not an IROFS; it provides no safety function and is therefore not "safety-related." Although the seismic recorders have no safety function , they must be designed to withstand any credible level of shaking to ensure that the ground motion would be recorded in the highly unlikely event of an earthquake. This capability requires verification of adequate capacity from the manufacturer (e.g., prior shake table tests of their product line), provision of adequate anchorage (e.g., manufacturer-provided anchor specifications to ensure accurate recordings), and a check for seismic interaction hazards such as water spray or falling fixtures.

With these design features , the instrumentation would be treated as if it were safety-related QL-2. Additional information on seismic instruction will be provided as part of the Operating License Application. 3.4.3.1 Location and Description Seismic instrumentation is installed for structural monitoring.

The seismic instrumentation consists of solid-state digital , tri-axial strong motion recorders located in the free-field , at the structure base , and at the roof of the RPF. 3.4.3.2 Operability and Characteristics The seismic instrumentation operates during all modes of RPF operations.

The maintenance and repair procedures provide for keeping the maximum number of instruments in service during RPF operations.

The instrumentation installation design includes provisions for in-service testing. The instruments selected are capable of in-place functional testing and periodic channel checks during normal facility operation. 3-42

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3.5 SYSTEMS

AND COMPONENTS Certain systems and components of the RPF are considered important to safety because they perform safety functions during normal operations or are required to prevent or mitigate the consequences of abnormal operational transients or accidents.

This section summarizes the design basis for design , construction , and operating characteristics of safety-related SS Cs of the RPF. 3.5.1 General Design Basis Information 3.5.1.1 Classification of Systems and Components Important to Safety The RPF systems and components will be classified according to their importance to safety , quality levels , and seismic class. The guidance used in developing these classifications during preliminary design with the support ofregulatory guidance reviews , hazards and operability analysis , accident analysis , integrated s afety analysis , and national consensus code requirements is presented below. The RPF systems identified in Table 3-1 and their a s sociated subsystems and components are discussed in the subsections that follow. 3.5.1.2 Classification Definitions The definitions used in the classification of SSCs include the following. In accordance with 10 CFR 50.2 , "Definitions

," design basis refers to information that identifies the specific functions to be performed by an SSC of a facility and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be: *

  • Restraints derived from generally accepted state-of-the-art practices for achieving functional goals Requirements derived from analy s is (e.g., calculation , experiments) of the effects of a postulated accident for which a SSC must meet its functional goals These reference bounds are to include the bounding conditions under which SSCs must perform design basis functions and may be derived from normal operation or any accident or event s for which SSCs are required to function , including anticipated operational occurrences , design basis accidents , external events , natural phenomena , and other events specifically addressed in the regulations. Safety-related is a classification applied to items relied on to remain functional during or following a design basis event (DBE) to ensure the: * *
  • Integrity of the facility infrastructure Capability to shut down the facility and maintain it in a safe-shutdown condition Capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to the applicable guideline exposures set forth in 10 CFR 70.61 , "Performance Requirements

," as applicable Design basis accident is a postulated accident that a nuclear facility must be designed and built to withstand, without loss to the SSCs necessary to ensure public health and safety. Design basis event (DBE) is an event that is a condition of normal operation (including anticipated operational occurrences), a design basis accident , an external event, or natural phenomena for which the facility must be designed so that the safety-related functions are achievable.

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.... .. NWMI ......... *.* ........ *.* ' *. * ! . NOkTHWEST MEOK:Al lSOTOl'lS NW Ml-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Design basis accidents and transients are those DBEs that are accidents and transients and are postulated in the safety analyses. The design basis accidents and transients are used in the design of the facility to establish acceptable performance requirements for SSCs. Single failure is considered a random failure and can include an initiating event (e.g., component failure, natural phenomenon, external man-made hazard) or consequential failures. Mechanical, instrumentation, and electrical systems and components required to perform their intended safety function in the event of a single failure are designed to include sufficient redundancy and independence.

This type of design verifies that a single failure of any active component does not result in a loss of the capability of the system to perform its safety functions.

Mechanical, instrumentation , and electrical systems and components are designed to ensure that a single failure, in conjunction with an initiating event, does not result in the loss of the RPF's ability to perform its intended safety function.

Design techniques such as physical separation, functional diversity, diversity in component design, and principles of operation, will be used to the extent necessary to protect against a single failure. An initiating event is a single occurrence , including its consequential effects , that places the RPF (or some portion) in an abnormal condition.

An initiating event and its resulting consequences are not considered a single failure. Active components are devices characterized by an expected significant change of state or discernible mechanical motion in response to an imposed demand on the system or operation requirements (e.g., switches , circuit breakers , relays , valves , pressure switches , motors, dampers , pumps, and analog meters). An active component failure is a failure of the component to complete its intended safety function(s) on demand. Passive compo n ents are devices characterized by an expected negligible change of state or negligible mechanical motion in response to an imposed design basis load demand on the system. Defense-in-depth is an approach to designing and operating nuclear facilities that prevents and mitigates accidents that release radiation or hazardous material through the creation of multiple independent and redundant layers of defense to compensate for potential human and mechanical failures so that no single layer , no matter how robust, is exclusively relied on. Defense-in-depth includes the use of access controls, physical barriers , redundant and diverse key safety functions , and emergency response measures.

The RPF structure and system designs are based on defense-in-depth practices. The RPF design incorporates:

  • *
  • Preference for engineered controls over administrative controls Independence to avoid common mode failures Other features that enhance safety by reducing challenges to safety-related components and systems Safety-related systems and components identified in this section are described in Chapters 4.0; 5.0, "Coolant Systems;" 6.0; 7.0; 8.0, "Electrical Power Systems;" and 9.0 , "Auxiliary Systems ," as appropriate.

3.5.1.3 Nuclear Safety Classifications for Structures, Systems, and Components SSCs in the RPF are classified as safety-related and non-safety-related.

The safety-related SSCs include IROFS to meet the performance requirement of 10 CFR 70.61 and other safety-related SSCs to meet the requirements of 10 CFR 20. The purpose of this section is to classify SSCs according to the safety function being performed. 3-44

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  • 0 HOllTHWEST MEOtCAl tsOlWH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components In addition, design requirements wil I be placed on SS Cs to ensure the proper performance of their safety function , when required.
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  • Safety-related is a classification applied to items relied on to remain functional during or following a postulated DBE to ensure the: Integrity of the facility infrastructure (e.g., water , sewer , electricity)

Capability to shut down the facility and maintain it in a safe shutdown condition Capability to prevent or mitigate the consequences of postulated accidents identified through accident analyses that could result in potential offsite and worker exposures comparable to the applicable guideline exposures set forth in 10 CFR 70.6l(b), 10 CFR 70.6l(c), and 10 CFR 70.61 (d) Operation of the facility without undue risk to the health and safety of workers , the public, and the environment to meet 10 CFR 20 norma l release or exposure limits for radiation doses and applicable limits for chemical exposures Safety-related IROFS -SSCs identified through accident analyses that are required to meet the performance requirements of 10 CFR 70.6l(b), 10 CFR 70.6l(c), and 10 CFR 70.6l(d) (Table 3-2). Safety-related Non-IROFS

-SSCs that provide reasonable assurance that the facility can be operated without undue risk to the health and safety of workers, the public, and environment , and includes SSCs to meet 10 CFR 20 normal release or exposure limits. Non-safety-related

-SSCs related to the production and delivery of products or services that are not in the above safety classifications 3.5.1.3.1 Quality Group Classifications for Structures, Systems, and Components The assignment of safety-related classification and use of codes and standards conforms to the requirements NWMI's Quality Assurance Program Plan (QAPP) for the development of a Quality Group classification and the use of codes and standards.

The classification system provides a recognizable means of identifying the extent to which SSCs are related to safety-related and seismic requirements , including ANS nuclear safety classifications, NRC quality groups , ASME Code Section III classifications, seismic categories , and other applicable industry standards, as shown in Table 3-7. Quality assurance (QA) requirements are defined in the NWMI QAPP (Chapter 12.0 , "Conduct of Operations," Appendix C). The definitions of QA Levels 1 , 2, and 3 are provided below. QA Level 1 will implement the full measure of the QAPP and will be applied to IROFS. IROFS are QA Level 1 items in which failure or malfunction could directly result in a condition that adversely affects workers, the public, and/or environment, as described in 10 CFR 70.61. The failure of a single QA Level 1 item could result in a high or intermediate consequence. The failure of a QA Level 2 item , in conjunction with the failure of an additional item, could result in a high or intermediate consequence.

All building and structural IROFS associated with credible external events are QA Level 1. QA Level 1 items also include those attributes of items that could interact with IROFS due to a seismic event and result in high or intermediate consequences , as described in 10 CFR 70.61. Examples include: * *

  • Items to prevent nuclear criticality accidents (e.g., preventive controls and measures to ensure that under normal and credible abnormal conditions , all nuclear processes are subcritical)

Items credited to withstand credible design-bases external events (e.g., seismic , wind) Items to prevent degradation of structural integrity (e.g., failure or malfunction of facility) 3-45

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  • NomtWfSTMEOtCAl.tSOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components QA Level 2 will be applied to non-QA Level 1 safety SSCs. The QA program is important to the acceptability and suitability of the item or service to perform as specified.

Acceptance methods shall be specified (including acceptance and other applicable performance criteria), documented , and verified before use of the item or s ervice. Some of the required characteristics may be examined less rigorously than for QA Level 1. Examples of QA Level 2 items include: * * *

  • SSCs to meet 10 CFR 20 normal release or exposure limits Fire protection systems Safeguards and security systems Material control and accountability systems QA Level 3 will include non-safety-related quality activities performed by NWMI that are deemed necessary to ensure the manufacture and delivery of highly reliable products and services to meet or exceed customer expectations and requirements.

QA Level 3 items include those items that are not classified as QA Level 1 or QA Level 2. QA Level 3 items are controlled in accordance with standard commercial practices.

These quality activities are embodied in NWMI's QAPP and will be further specified in the Operating License Application , and when necessary. 3.5.1.3.2 Seismic Classification for Structures, Systems, and Components SSCs identified as IROFS will be designed to satisfy the general seismic criteria to withstand the effects of natural phenomena (e.g., earthquakes , tomados , hurricanes , floods) without loss of capability to perform their safety functions.

ASCE 7 , Chapter 11 , sets forth the criteria to which the plant design bases demonstrate the capability to function during and after vibratory ground-motion associated with the shutdown earthquake conditions.

The seismic classification methodology used for the RPF complies with the preceding criteria , and with the recommendations stated in Regulatory Guide 1.29 , S e ismic De s ign Classification.

The methodology classifies SSCs into three categories:

seismic Category I (C-I), seismic Category II (C-11), and seismic (NS). Seismic C-1 applies to both functionality and integrity , while C-11 applies only to integrity.

SSCs located in the proximity of IROFS , the failure of which during a safe-shutdown earthquake could result in loss of function of IROFS , are designated as C-11. Specifically:

  • C-I applies to IROFS. C-I also applies to those SSCs required to support shutdown of the RPF and maintain the facility in a safe shutdown condition C-11 applies to SSCs designed to prevent collapse under the safe-shutdown earthquake. SSCs are classified as C-11 to preclude structural failure during a safe-shutdown earthquake , or where interaction with C-1 items could degrade the functioning of a safety-related SSC to an unacceptable level or could result in an incapacitating injury to occupants of the main control room.
  • NS SSCs are those that are not classified seismic C-1 or C-11. 3-46

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3.5.2 Radioisotope

Production Facility Systems and components within the RPF are presented in Section 3.5.1. The RPF design basis evaluated the general design criteria from 10 CFR 70.64, " Requirements for New Facilities or New Processes at Ex isting Facilities." Thi s evaluation is presented in Table 3-22. These general design criteria provide a rational basis from which to initiate design but are not mandatory.

There are some cases where conformance to a particular c riterion is not directly measurable.

For each of the criteria, a specific assessment of the RPF design is made , and a complete list of references is included to identify where detailed design information pertinent to each criterion is treated. The Chapter 13.0 accident seq uence s for cre dible events define the DBE. The safety-related parameter limit s ensure that the associated de s ign basis is met for the events presented in Chapter 13.0. Table 3-22. Design Criteria Requirements (4 pages) Design criteria and description Application and compliance 10 CFR 70.64 , "Requirements for New Facilities or New Processes at Existing Facilities"" Quality standards and records

  • Develop and implement de s ign in accordance with management measures to ensure that IROFS are available and reliable to perform their function when needed.
  • Maintain appropriate records of these items by or under the control of the licensee throughout the life of the facility. Natura l phenomena hazards Provide for adequate prot ec ti on ag a in s t natural phenomena , with considerat i on of the mo st severe documented historical events for the s it e.
  • SSCs important to safety will be designed , fabricated, erected, tested , operated, and maintained to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used , they will be identified and evaluated to determine their applicability , adequacy, and sufficiency and will be supplemented or modified as necessary to ensure a quality product in keeping with the required safety function.
  • NWMl's QAPP will be established and implemented to provide adequate assurance that SSCs satisfactori l y perform their safety functions.
  • Appropr iat e re cords of design, fabricatio n , erection, a nd testing of SSCs important to safety will be m a in tained by or und er control ofNWMJ for the li fe of RPF.
  • NWMI will use a graduated QAPP that links quality classification and associated documentation to safety classification and to the manufacturing and delivery of highly reliable products and equipment.
  • The WMI QAPP will provide details of the procedures to be applied, including quality and safety level classifications.
  • SSCs important to safety will be designed , fabricated, e r ecte d , tested, operated , a nd maintained to quality sta nd ards co mm e n s ur ate with the import ance of the safety functions to be performed.

Where genera ll y r ecogn i zed codes and standa rd s are u sed , they will b e id e ntifi e d and eva lu ated to determin e their applicabi li ty , adeq u ac y, and sufficiency a nd w ill b e supp l e ment e d or modified as necessary to ensure a quality product in keeping w ith the required safety function.

  • The de s ign basi s for these SSCs wi II include: -Approp ri a t e considerat i on of the most severe natural phenomena that have been historically reported for the RPF site and surro undin g area , includin g sufficient margin for limi ted accuracy , quantity , and period of time for which historical data h as b ee n accum ul ate d -Appro pri ate combinat i o n s of n atural phenomena eff e cts durin g n ormal and accide nt operatin g cond ition s -Imp o rt a n ce of the safety function s to be p e rform ed
  • Specific RPF d es i gn c rit er ia a nd NRC ge n eral d es ign criteria are discussed in Sections 3.1 and 3.5 , re s pectively.

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.. .. NWMI ...... ..* .... ........... , * "NORTHWlSTMEDM:AllSOTOf'ES NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-22. Design Criteria Requirements (4 pages) Design criteria and description Application and compliance Fire protection Provide for adequate protection against fire s and explosions Environmental and d y namic effects Pro v id e for a d e qu a t e prot e ction fro m e n v i ro nm e nt a l co nditi o ns a nd d yna mi c e ff ec t s assoc i a t e d w ith n o rm a l o p era tion s, m ai nt e nan ce, t es tin g, a nd pos tul a t e d acci d e nt s th a t co uld l ea d to loss o f sa f e t y fun ct ion s Chemical protection Provide for adequate protection against c hemical risk s produced from licen s ed material, facility conditions that affect the safety of licensed material , and hazardous chemicals produced from licensed material Emergenc y capability P rov id e fo r e m e r ge n cy ca p a bili ty t o maintain c ontr o l o f:

  • L ic e n sed m ate ri a l a nd h aza rd o u s c h e mi ca l s p ro du ce d from li ce n se d mat e ri a l
  • Evac u a ti o n of o n-s it e p e r so nn e l
  • O n-s it e e m e r ge n cy fac iliti es a nd se rvi ces th a t fac ilit a t e th e u se of ava il a bl e off-s i te serv ic es
  • SSCs important to safety will be designed and located throughout the RPF to minimize , consistent with other safety requirement s, the probability and effect of fires and explosions.
  • Noncombustible and heat resi s tant mat e rials will be u s ed wherever practical throughout the RPF , particularly in locations such as confinement and the control room.
  • Fire detection and suppre s sion systems of appropriate capacity and capability will be provided and designed to minimize the adverse effects of fires on SSCs important to safety.
  • Firefighting systems will be designed to ensure that their rupture or inadvertent operation does not significantly impair the safety capability of these SSCs.
  • Where necessary , within zoned areas or where criticality and access are an issue , required system s will be manually initiated by operations after revi e w ofa detection signal.
  • RPF fire protection system will be designed s uch that a failure of any compon e nt will not impair the ability of safety-related SSCs to safely shut down and isolate the RPF or limit the relea s e of radioactivity to provide reasonable as s urance that the public will be protected from radiological risks resulting from RPF operations
  • RPF fire protection system will be designed to provide reasonable assurance that the public will be protected from radiological risks resulting from RPF operations (e.g., failure of any component will not impair the ability of safety-related SSCs to s afely s hutdown and isolate the RPF or limit the release of radioactivity).
  • Chapters 6.0 and 9.0 provide additional information.
  • SSCs imp o rt a nt t o safe t y are d es i gne d to acco mm o d ate effec t s of, a nd t o b e co mp a tibl e with , the e n v ironm e nt a l co nditi o n s assoc i a t e d with n o rm a l o p e r a tion , m ai nt e n a n ce, t es tin g, a nd po s tul a t e d acc id e nt s. Du e to l ow t e mp era ture a nd pr ess ur e RP F proc esses , dynami c effec t s du e to pip e ruptur e and di sc h a r g in g flu i ds a r e n o t a ppli ca bl e t o th e RP F.
  • Chemical protection in the RPF will be provided by confinement isolation systems , liquid retention features , and use of appropriate personal protective equipment.
  • Chapter 6.0 , Section 6.2.1 , provides a dditional information.
  • E m e r ge n cy p roce d ures w ill b e d eve l o p e d a nd m a in tai n e d for th e RP F t o co ntr o l SNM a nd h aza rd o u s c h e mical s produ ce d fr o m th e SN M.
  • A pr e limin ary E m erge n cy Pr e par e dn ess Pl a n i s pro v id e d in C h a p ter 1 2.0 , A p pe ndi x B. 3-48

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  • NORTHWUT MEOtCAL tsOTOPlS NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-22. Design Criteria Requirements (4 pages) Design criteria and description Application and compliance Utility services Provide for continued operation of esse ntial utility services Inspection, testing , and maintenance Provid e for a d eq ua te in s p ec ti o n , tes tin g , a nd maint e n a n ce ofIROFS to e n s ur e ava ilabili ty and r e li a bility to perform their function when n ee d e d Criticality control Provide for criticality control , including adherence to the double-contingency principle Instrumentation and control The d es ign mu s t provide for in c l us i o n of l&C sys t e m s to monitor a nd control th e b e ha v ior of it e m s r e li e d on for safety.
  • The RPF is designed for passive , safe shutdown and to prevent uncontrolled release of radioactive material if normal electric power is interrupted or lost.
  • A stan dby diesel generator will be provided for asset protection of selected RPF systems.
  • Uninterruptable power supplies will automatically provide power to systems that support the safety functions protecting workers and the public.
  • A combination ofuninterruptable power supplies and a standby electrica l power system will provide emergency electrical power to the RPF. A 1 ,000 kW (1,34 l hp) diesel generator will provide facility electric power.
  • Chapter 8.0, Section 8.2 provides additiona l information.
  • Th e RPF i s d es igned to provide access a nd co ntrol s fo r te st in g , maint e n a nc e, and in spec tion of safety-r e l ate d SSCs , as ne e d e d , throu g h ou t th e RPF.
  • C h apte r s 4.0 , 6.0 , 7.0 , a nd 9.0 pro v id e additiona l in forma tion.
  • The RPF design will provide adequate protection against criticality hazards related to th e storage, handling, and processing of SNM , which will b e accomplished by: -Including equipment , facilities , and procedures to protect worker and public health and to minimize danger to life or property Ensuring that the design provides for criticality control , including adherence to the double-contingency principl e Incorporating a criticality monitoring and alarm system into the facility design
  • Compliance with the requirements of criticality control, including adherence to th e double-co ntingency principle , are described in detail in Chapter 6.0, Section 6.3.
  • RPF SN M pro cesses w ill be e ncl osed pr e domin a t e l y by hot ce ll s and g l ove box d es i gns exce pt for th e targe t fabrication a r ea.
  • The FPC syste m w ill provide m on it oring a nd con tr o l of safety-r e lat e d co mpon e nt s and process systems within th e RP F.
  • The BMS (a s ub set of the FPC system) w ill m o nit or t h e RP F venti lati on sys t e m a nd mechanical ut i l ity syste m s.
  • ESF sys tem s will o p erate ind e p e nd e ntly from the FPC sys t e m or BMS. Eac h ESF safety function will u se h a rd-wir e d a nalo g co ntrol s/int e rlock s to prot ect wo rk e r s , th e public , and e nvironm e nt. Th e ESF parameters a nd a l a rm function s wi ll b e int egra t e d in to a nd m o nitor e d by the F PC s y s t e m or BMS.
  • RPF designs a r e b ased on d efe n se-in-d e pth practi ces a nd incorp ora te a preference for engineere d contro l s over a dmini strat i ve co ntr o l s , independence to avoid co mm o n mod e fai lur es , a nd incorp orate other features that e nh a n ce safety by redu c in g ch a ll e n ges to safety-r e l a t ed co mpon e nts a nd s ys t e m s.
  • The F P C sys t e m w ill provide th e capabi lity to monit or and co ntr o l the b e h av i or of safety-r e l a t e d SSCs. These sys t e ms e n s ure a d eq u a t e safe t y of proce ss an d utili ty service operations in co nnection with their safety function.

Co ntrols a r e provid e d to maintain th ese variab l es a nd sys t e m s with in th e presc rib e d o p era tin g ra n ges und er a ll normal co ndi t i o n s.

  • The F P C sys t em i s d es igned to fail to a s afe-s t ate or to ass ume a state dem o n s trat e d to b e accep t a bl e if conditions s u c h as loss of s i g n a l , l oss of e ner gy or m ot i ve pow e r , o r a d ve r se e n vironments a re experienced.
  • Chapter 7.0 provid es a dditional I&C sys t e m inform ation. S afe ty-related SSCs are d escri bed in Section 3.5 a nd Chapters 4.0 , 5.0, 6.0 , 7.0 , a nd 8.0. 3-49

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  • NORTHWEST liEMCAL. tsOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-22. Design Criteria Requirements (4 pages) Design criteria and description Application and compliance Defense-in-depthb Base facility and system design and facility layout on defense-in-depth practices. The design must incorporate , to the extent practicable
  • Preference for the selection of engineered controls over administrative controls to increa s e overall system reliability
  • Features that enhance safety by reducing challenges to IROFS
  • Defense-in-depth is a design philosophy that NWMI has applied from the beginning of the project and will continue through completion of a design that is based on providing successive levels of protection such that health and safety are not wholly dependent on any single element of the design , construction , maintenance, or operation of the RPF.
  • NWMJ's risk insights obtained through performance of the accident analysis will be used to supplement the final design by focusing attention on the prevention and mitigation of the higher risk potential accidents.
  • Chapter 6.0 and 13.0 provide additional information.
  • I 0 C FR 7 0.64 , " Requir e m e nt s for N e w Fa ciliti es or N e w Pro cesses a t Ex i s tin g Fa ciliti es ," Code of F e d e ral R e gu lati o n s , Offi ce of th e F e d e ral R eg i s ter , as a m e nd e d. b A s u se d in I 0 C FR 70.6 4 , r e quir e m e nt s for n ew faciliti es or n e w proc esses a t ex i s tin g fa cili t i es , d efe n se-in-d e pth pr ac ti ces means a d es ign philo s oph y , a ppli e d fr o m th e outs e t a nd throu g h c ompl e tion o f th e d es i gn , th at i s bas e d on pro v idin g s u ccess i ve l eve l s of p ro t ec ti o n s u c h th at h ea lth a nd safety will n o t b e w h o ll y d e p e nd e nt on a n y s in g l e e l e m e nt of th e d es i gn , co n s truction , m a inten a n ce , or op e r a tion of th e fac ility. Th e n e t e ff ec t of incorporatin g de fe n se-in-d e pth practi ces is a co n s e rv a ti ve l y d es i gne d faci li ty a n d sys t e m t h a t wi ll e xhibit grea t er to l e ran ce t o fa ilur es a nd ex t e rn a l c h a ll e n ges. BM S bu i ldin g mana ge m e nt sys t e m. N R C U.S. Nucl ear R eg ul a tory C ommi ss i o n. CF R Co d e of F e d era l R eg ul a ti o n s. NWM I N orth west M e d ical I so t o p es, LL C. ESF e ngin ee r e d s a fe t y fea tur e. QAPP qu a lity a ss uran ce program plan. F P C fac ili ty pro cess co ntr o l. RP F R a dioi s ot o p e P ro du c ti o n Fac ili ty. l&C in s trum e nt a ti o n and c ontrol. S NM s peci a l nucl e ar ma te ri a l. IROFS it e m s r e lied on for sa fet y. SSC s tru c tur es, s y s t e m s, a nd c omp o nent s. The criteria are generic in nature and subject to a variety of interpretations
howe v er , they also establish a proven basis from which to provide for and assess the s a fety of the RPF and de v elop principal design criteria. The g eneral de s ign criteria establish the nece ss ary design , fabrication , construction , testing , and performance requirements for SSCs important to safety (i.e., SSCs that provide reasonable assurance that the facility can be operated without undue ri s k to the health and s afety of worker s , the public , and en v ironment). Safety-related SSCs that are determined to have safety significance for the RPF will be designed , fabricated , erected , and tested as required by the NWMI QAPP , described in Chapter 12.0 , Appendix C. In addition , appropriate records of the design , fabrication , erection , procurement , testing , and operations of SSCs will be maintained throughout the life of the plant. The RPF design addre s se s the following
* * * * * * *
  • Radiological and chemical protection Natural phenomena hazards Fire protection Environmental and dynamic effects Emergency capability (e.g., licensed material , hazardous chemicals , evacuation of on-site personnel , on-site emergency facilities

/off-site emergency facilities)

Utility service s Inspection , te s tin g, and maintenance Criticality safety 3-50

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  • . *
  • NORTHWEST MEotCA.l ISOTOl'U *
  • Instrumentation and controls Defense-in-depth Safety-related systems and components will be qualified using the applicable guidance in the Institute of Electrical and Electronics Engineers (IEEE) Standard IEEE 323, IEEE Standard for Qualifying Class 1 E Equipment for Nuclear Power Generating Stations.

The qualification of each safety-related system or component needs to demonstrate the ability perform the associated safety function: *

  • Under environmental and dynamic service conditions in which they are required to function For the length of time the function is required Additionally, non-safety-related components and systems will be qualified to withstand environmental stress caused by environmental and dynamic service conditions under which their failure could prevent satisfactory accomplishment of the safety-related functions.

The RPF instrumentation and control (l&C) system (also known as the facility process control [FPC] system) will provide monitoring and control of the process systems within the RPF that are significa nt to safety over anticipated ranges for normal operations and abnormal operations.

The FPC system will perform as the overall production process controller.

This sys tem will monitor and control the process instrumented functions within the RPF , including monitoring of process fluid transfers and controlled inter-equipment pump transfers of process fluids. The FPC system will also ensure that process and utility systems operate in accordance with their safety function.

Controls will be provided to maintain variables and systems within the prescribed operating ranges under all normal conditions.

In addition, the FPC system i s designed to fail into a safe state or to assume a state demonstrated to be acceptable if conditions such as lo ss of signal, loss of energy or motive power, or adverse environments are experienced.

The building management system (BMS) (a subset of the FPC system) will monitor the RPF ventilation system and mechanical utility systems. The BMS primary functions will be to monitor the facility ventilation system and monitor and control (turn on and off) the mechanical utility systems. ESF systems will operate independently from the FPC system or BMS. Each ESF safe ty function will use hard-wired analog controls/interlocks to protect workers, the public, and environment.

The ESF parameters and alarm functions will be integrated into and monitored by the FPC system or BMS. The fire protection system will have its own central alarm panel. The fire protection system will report the sta tus of the fire protection equipment to the central alarm station and the RPF control room. This integrated control system will be isolated from safety-related components consistent with IEEE 279, Criteria for Protection Systems for Nuclear Power Generating Stations.

In addition, the RPF is designed to meet IEEE 603, Standard Criteria for Safety Systems for Nuclear Power Generating Stations, for se paration and isolation of safety-related systems and components.

Chapter 7 .0 provides additional details on the integrated control system. 3-51 _J

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  • NOmfWEST MEDICAl 1.SOTOPES 3.5.2.1 System Classification The RPF is classified as a non-reactor nuclear production faci li ty per 10 CFR 50. In addition, a portion of the RPF will fabricate LEU targets , similar to fuel fabrication per 10 CF R 70. Due to the nature of the work performed within faci lit y, a hazardous occupancy app li es. Table 3-2 3 provides the RPF classification for hazards occ up ancy, construct ion , risk , and seismic design categories.

3.5.2.2 Clas s ification of Systems and Components Important to Safety RPF SSCs, including their foundations and NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components Table 3-23. System Classifications Classification description Hazard category Occupancy type Construction type Risk category Seismic d es ign category Classification Intermedi ate hazard Mixed, A-2 , B, F-1, H-3 and H-4 II-B IV c Source NRC me 2012* IBC 2012* ASCE 7b ASCE 7b

  • IBC 20 1 2, " Internati ona l Bui ldin g Code," as amended , Int e rnational Co de Co uncil , Inc., Washington , D.C., February 20 1 2. b ASC E 7, Minimum D esig n L o ads for Buildings and Oth e r Structures, American Society of C ivil Engineers , R esto n , Virginia , 20 13. NRC = U.S. N ucl ear Regulatory Commiss ion. s upports , designed to remain functional in the event of a DBE are designated as C-1. SSCs designated IROFS are also classified as C-1. SSCs co-located with C-1 systems are reviewed and supported in accordance with II over I criteria.

This avoids any unacceptable interactions between SSCs. C-1 structures shou ld b e designed using dynamic ana l ysis procedures , or when justified, equivalent static proce dure s using both horizontal and vertical input ground motions. For dynamic ana l yses , either response spectra or time history analyses approaches may be used. Dynamic analysis sho uld be performed in accordance with the procedures of ASCE 4 , with the exception of the damping limitations presented in Section 3 .4.1. Table 3-24 lists the RPF SSCs and associated safety and seismic classifications and quality l evel group for the top-level systems. Subsystems within these systems may be identified with lower safety classifica ti ons. For example, the day tanks of the chemical supp l y system are IROFS , while the rest of the c h emical supp l y system is classified as safety-related or not-safety-related.

Table 3-24. System Safety and Seismic Classification and Ass ociated Qualit y Level Group (2 pages) System name (code) Facility structure (RPF) Target fabrication (TF) Target receipt and disassembly (TD) Target dissolution (DS) Mo recovery and purification (M R) Uranium recovery and recycle (UR) Waste handling (WH) Criticality accident alarm (CA) Radiation monitoring (RM) Standby electrical power (SEP) Norma l e l ectrical power (NEP) Highest safety classification*

IROFS IROFS IROFS IROFS IROFS IROFS IROFS IROFS IROFS IROFS SR 3-52 Seismic classificationb C-1 C-1 C-I C-I C-1 C-1 C-1 C-1 C-1 C-1 C-1 Quality level group QL-1 QL-1 QL-1 QL-1 QL-1 QL-1 QL-1 QL-1 QL-1 QL-1 QL-1 I ......... *. .; ... ;. NWMI ....***. *.* 0 ffOlfTlfWfSTllEDtCALISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Table 3-24. System Safety and Seismic Classification and Associated Quality Level Group (2 pages) System name (code) Process vessel ventilation (PVV) Facility ve ntil ation (FV)c Fire protection (FP) Plant and i nstrument air (PA) Emergency Purge gas (PG) Gas s uppl y (GS) Process chilled water (PCW) Facility c hilled water (FCW) Facility heated water (HW) Proc ess steam Demineralized water (DW) Chem ic a l supp l y (CS) Biological shield (BS) Facility process co ntrol (FPC) Highest safety classificationa IROFS IROFS SR NSR IROFS NS R IROFS NSR NSR IROFS NSR IROFS IROFS SR Seismic classificationb C-1 C-1/11 C-II C-11 C-II C-II C-1 C-11 C-11 C-1 C-II C-I C-1 C-II Quality level group QL-1 QL-1/2 QL-2 QL-2 QL-1 QL-2 QL-1 QL-2 QL-2 QL-1 QL-2 QL-1 QL-1 QL-2

  • Safety classification acco unt s for hi g h est classification in the syste m. Systems that a re c l assifie d as safety-r elated may include b ot h safety-re l ated a nd n on-safety-r e l ated co mp o n e nt s. Only safe t y-r e l ated co mpon e nt s will be u se d to satisfy t h e safety fun ct i ons of th e sys t em , whereas non-safety-related components can b e u sed to perform n o n-safety fun ctio n s. Fo r examp l e, there are n on-safety-r elated co mp o n ents , s u c h as fa n s , wit hin the safety-related ve ntil atio n syste m s th at p erform n o nsafety-r e l a t ed functions.

b Seismic category may be locally revised to account for II over I d esign criteria and to e l iminate p otent i a l system degradation due to se i s mic int eractions.

c Vent il ation zone c l assifications vary -Venti l ation Zone I and II are co n sidered safety-re l ated , C-1 and QL-1; Ventilation Zone III and IV are cons id ered non-safety-related , C-11 a nd QL-2. IROFS = items relied on for safety. RPF NS R = n o n-safety r e l ated. SR = Radioisotope Production Facility.

= safety-re l ated (not IROFS). SSCs that must maintain struc tural integrity post-DBE , but are not required to remain functional are C-11. All other SSCs that hav e no s pecific NRC-regulated requirements are designed to local juri s dictional requirements for structural int egr ity and are C-III. All C-1 SSCs are analyzed under the loading co ndition s of the DBE and consider mar gi n s of safety appropriate for that earthquake.

The mar g in of sa f e ty provided for safety-class SSCs for the DBE are s ufficient to ens ure that th e ir de s ign function s are not put at ri s k. Table 3-25 presents the likelihood inde x limit guidelines and associated event frequency and risk inde x limit s. Table 3-25. Likelihood Index Limit Guidelines Likely normal faci lit y proc ess co ndition Not unlikely (frequent facility process condition)

U nlik e l y (infr eque nt faci lit y process co ndition) Highly unlikely (limiting facility process condition)

      • 4 3 2 3-53 Event frequency limits Multiple events per yea r > or= 0 More than 10 4 per event, per year >-4 <O B etween I 0 4 a nd l 0-5 per eve nt , -4 to 5 p er yea r Less than 10-5 per event, per year < -5

...... .. NWMI ...*.. ... **.* ........ *.* *. NOmfWEST MEDtCAl tsOTOl'£1 NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components 3.5.2.3 Design Basis Functions, Values, and Criteria The design basis for systems and components required for safe operation and shutdown of the RPF are established in three categories , which are described below. The preliminary design basis functions and values for each major system are provided in the following subsections.

Design Basis Functions

  • License conditions, orders , or technical specifications Functions credited in the safety analysis to ensure safe shutdown of the facility is achieved and maintained, prevent potential accidents, or mitigate the potential consequences of accidents that could result in consequences greater than applicable NRC exposure guidelines Design Basis Values *
  • Values or ranges of values of controlling parameters established as reference bounds for RPF design to meet design basis function requirements Values may be established by an NRC requirement , derived from or confirmed by the safety analysis , or selected by the designer from an applicable code, standard, or guidance document Design Basis Criteria
  • Code-driven requirements established for the RPF fall into seven categories , including fabrication , construction, operations, testing , inspection, performance , and quality
  • Codes include national consensus codes, national standards, and national guidance documents
  • Design of safety-related systems (including protection systems) is consistent with IEEE 3 79 , Standard Application of th e Single-Failure Criterion to Nuclear Power G e nerating Station Safety Systems, and Regulatory Guide 1.53 , Application of the Single-Failure Criterion to Nuclear Power Plant Prot ectio n Systems
  • Protection system is designed to provide two or three channels for each protective systems and functions and two logic train circuits: Redundant channels and trains will be electrically isolated and physically separated in areas outside of the RPF control room Redundant design will not prevent protective action at the system level 3.5.2.4 System Functions/Safety Functions The NWMI RPF will provide protection against natural phenomena hazards for the personnel , SNM, and systems within the facility. The facility will also provide protection against operational and accident hazards to personnel and the public. Table 3-2 lists the IROFS defined by the preliminary hazards analysis.

3.5.2.5 Systems and Components 3.5.2.5.1 Mechanical RPF C-1 mechanical equipment and components (identified in Table 3-24) will be qualified for operation under the design basis earthquake (DBEQ) seismic conditions by prototype testing , operating experience , or appropriate analysis.

The C-1 mechanical equipment is also designed to withstand loadings due to the DBEQ , vibrational loadings transmitted through piping, and operational vibratory loading , such as floor vibration due to other operating equipment , without loss of function or fluid boundary.

This analysis considers the natural frequency of the operating equipment , the floor response spectra at the equipment location, and loadings transmitted to the equipment and the equipment anchorage.

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.-.;. NWMI ..*... ..* ... ........... * "NOmfWHTMfDICALISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components The qualification documents and all supporting analysis and test reports will be maintained as part of the permanent plant record in accordance with the requirements of the NWMI QAPP. The safety-related equipment and components within the RPF will be required to function during normal operations and during and following DBEs. This equipment will be capable of functioning in the RPF environmental conditions associated with normal operations and design basis accidents. Certain systems and components used in the ESF systems will be located in a controlled environment.

This controlled environment is considered an integral part of the ESF systems. 3.5.2.5.2 Instrumentation and Electrica l C-1 instrumentation and electrical equipment (identified in Table 3-24) is designed to resist and withstand the effects of the postulated DBEQ without functional impairment.

The equipment will remain operable during and after a DBEQ. The magnitude and frequency of the DBEQ loadings that each component experiences will be determined by its location within the RPF. In-structure response curves at various building elevations will be developed to support design. The equipment (e.g., batteries and instrument racks , control consoles) has test data , operating experience , and/or calculations to substantiate the ability of the components and systems to not suffer loss of function during or after seismic loadings due to the DBEQ. This information will be completed during final design of the RPF and provided in the Operating License Application.

This certification of compliance with the specified seismic requirements , including compliance with the requirements of IEEE 344 , is maintained as part of the permanent plant record in accordance with the NWMIQAPP.

3.5.2.6 Qualification Methods Environmental qualification of safety-related mechanical , instrumentation , and electrical systems and components is demonstrated by tests , analysis , or reliance on operating experience.

Qualification method testing will be accomplished either by tests on the particular equipment or by type tests performed on similar equipment under environmental conditions at least as severe as the specified conditions.

The equipment will be qualified for normal and accident environments.

Qualification data will be maintained as part of the permanent plant record in accordance with the NWMI QAPP. 3.5.2.7 Radioisotope Production Facility Specific System Design Basis Functions and Values The design basis functions and values for each system identified in Table 3-1 are discussed in the following subsections.

Additional details for each system described below will be updated during the de v elopment of the Operating License Application.

3.5.2.7.1 Target Fabrication System An overview and detailed description of the target fabrication system are provided in Chapter 4.0 , Sections 4.1.3.1and4.4 , respectively. Design Basis Functions

  • * * * *
  • Store fresh LEU , LEU target material , and new LEU targets Produce LEU target material from fresh and recycled LEU material Assemble , load, and fabricate LEU targets Reduce or eliminate the buildup of static electricity Minimize uranium losses through target fabrication Safety-related functions: 3-55

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  • NOml'WEST M£tNCAl lSOTOffl NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Maintain subcriticality conditions within target fabrication system Prevent flammable gas composition within target fabrication system Limit personnel exposure to hazardous chemicals and off gases Design Basis Values *
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBE s 3.5.2.7.2 Target Receipt and Disassembly System An overview and detailed description of the target receipt and disassembly system are provided in Chapter 4.0 , Section 4.1.3.2 , and Sections 4.3.2/4.3.3 , respectively. Design Basis Functions
  • * * *
  • Handle irradiated target shipping cask, including all opening , closing , and lifting operations Retrieve irradiated targets from a shipping cask Disassemble targets and retrieving irradiated target material from targets Reduce or eliminate the buildup of static electricity Safety-related functions:

Provide radiological shielding during receipt and disassembly activitie s Maintain subcriticality conditions within target receipt and disassembly system Prevent radiological materials from being released during target receipt and disassembly operations to limit the exposure of workers , the public, and environment to radioactive material Maintain positive control of radiological materials (LEU target material and radiological waste) Protect personnel and equipment from industrial hazards associated with system equipment (e.g., moving parts) Design Basis Values

  • 30-year design life *
  • Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs Crane designed for anticipated load (e.g., hot cell cover block) of approximately 68 metric tons (MT) (75 ton) 3.5.2.7.3 Replace Target Dissolution (DS) An overview and detailed description of the target dissolution system are provided in Chapter 4.0 , Sections 4.1.3.3 and 4.3.4 , respectively. Design Basis Functions
  • * *
  • Fill the dissolver basket with the LEU target material Dissolve the LEU target material within dissolver basket Treat the off gas from the target dissolution system Handle and package solid waste created by normal operational activities 3-56

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  • Safety-related functions: Provide radiological shielding during target dissolution activities Control and prevent flammable gas from reaching lower flammability limit conditions Maintain subcriticality conditions through inherently safe design of target dissolution equipment Maintain positive control of radiological materials (LEU target material and radiological waste) Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) *
  • Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs Prevent radiological materials from being released during target dissolution operations to limit the exposure of workers, the public , and environment to radioactive material per I 0 CFR 20 3.5.2.7.4 Molybdenum Recovery and Purification (MR) An overview and detailed description of the Mo recovery and purification system are provided in Chapter 4.0 , Sections 4.1.3.4 and 4.3.5 , respectively. Design Basis Functions
  • *
  • Recovery of Mo product from a nitric acid solution created from dissolved irradiated uranium targets Purification of the recovered Mo product to reach specified purity requirements , followed by shipment of the Mo product Safety-related functions:

Maintain subcriticality conditions through inherently safe design of components that could handle high-uranium content fluid Prevent radiological materials from being released by containing fluids in appropriate tubing , valves , and other component s Control and prevent flammable gas from reaching lower flammability limit conditions Maintain positive control of radiological materials (9 9 Mo product , intermediate streams , and radiological waste) Provide appropriate containers and handling systems to protect personnel from industrial hazards such as chemical exposure (e.g., nitric acid , caustic , etc.) Design Basis Values * *

  • Maintain primary fission product boundary during and after normal operations , shutdown conditions, and DBEs 30-year design life with the exception of common replaceable parts (e.g., pumps) Replace consumables after each batch 3.5.2.7.5 Uranium Recovery and Recycle (UR) An overview and detailed description of the uranium recovery and recycle system are provided in Chapter 4.0, Sections 4. l.3.5 and 4.3.6, respectively.

3-57

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  • NORTHWESTMEDICAi.

ISOTOPf S NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Design Basis Functions

  • Receive and decay impure LEU solution
  • Recover and purify impure LEU solution
  • Decay and recycle LEU solution
  • Transfer process waste
  • Safety-related functions: Provide radiological shielding during uranium recovery and recycle system activities Prevent radiological release during uranium recovery and recycle system activities Maintain subcriticality conditions through inherently safe design of the uranium recovery and recycle equipment Control and preventing flammable gas from reaching lower flammability limit conditions Maintain positive control of radiological materials Protect personnel and equipment from industrial hazards associated with the system equipment , such as moving parts , high temperatures , and electric shock Design Basis Values *
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs 3.5.2.7.6 Waste Handling An overview and detailed description of the waste handling sy s tem are provided in Chapter 4.0 , Section 4.1.3.6 and Chapter 9.0 , Section 9.7.2, respectively.

Design Basis Functions

  • Receive liquid waste that is divided into high-dose source terms and low-dose source terms to lag storage * * * * *
  • Transfer remotely loaded drums with high-activity solid waste via a solid waste drum transit system to a waste encapsulation cell Encapsulate solid waste drums Load drums with solidification agent and low-dose liquid waste Load high-integrity containers with solidification agent and high-dose liquid waste Handle and load a waste shipping cask with radiological waste drums/containers Safety-related functions: Maintain subcriticality conditions through mass limits Prevent spread of contamination to manned areas of the facility that could result in personnel exposure to radioactive materials or toxic chemicals Provide shielding, distance , or other means to minimize personnel exposure to penetrating radiation Design Basis Values *
  • Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs 30-year design life with the exception of common replaceable parts (e.g., pumps) 3-58
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  • NCMmlWESTMEDK:AllSOTDPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components 3.5.2.7.7 Criticality Accident Alarm System Chapter 6.0 , Section 6.3.3.1 , and Chapter 7.0 , Section 7.3.7 , provide descriptions of the criticality accident alarm system. Design Basis Functions
  • Provide analysis for criticality accident alarm system coverage in all areas where SNM is handled , processed , or stored * *
  • Provide for continuous monitoring , indication , and recording of neutron or gamma radiation levels in areas where personnel may be present and wherever an accidental criticality event could result from operational processes.

Provide both local and remote annunciation of a criticality excursion Remain operational during DBEs Design Basis Values

  • 30-year design life
  • Capable of detecting a criticality accident that produces an absorbed dose in soft tissue of 20 absorbed radiation dose (rad) of combined neutron or gamma radiation at an unshielded distance of 2 m from reacting material within one minute 3.5.2.7.8 Continuous Air Monitoring System Chapter 7.0, Section 7.6 , and Chapter 11.0 , Section 11.1.4, provide detailed descriptions of the RPF continuous air monitoring system. Design Basis Functions
  • * * *
  • Provide real-time local and remote annunciation of airborne contamination in excess of preset limit s Provide real-time local and remote annunciation of radiological dose of excess of preset limits Provide environmental monitoring of nuclear radioactive stack releases Provide the capability to collect continuous samples Remain operational during DBEs Design Basis Values * *
  • Activate when airborne radioactivity levels exceed predetermined limit s Activate when radiological dose levels exceed predetermined limits Adjust vo lume of air samp led to ensure adequate sensitivity with minimum sampling time 3.5.2.7.9 Standby Electrical Power Chapter 8.0 , Section 8.2 provides a detailed description of the RPF standby electrical power (SEP) system. Design Basis Functions SEP includes two types of components:

uninterruptible power supplies (UPS) and a standby diesel generator:

  • UPS -Provides power when normal power supplies are absent Standby diesel generator

-Provides power when normal power supp lies are absent to allow continued RPF processing 3-59

] ::.**.*.*.* .. ; ... ... NWMI ........ *.* ,

  • NOmlWESTMEDICAllSOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Design Basis Values * *
  • 30-year design life Maintain power availability for a minimum of 120 min post-accident (UPS) Maintain power availability for 12 hr (diesel generator) 3.5.2.7.10 Nor mal Electrical Power Chapter 8.0 , Section 8.1 pro vi des a detailed description of the RPF normal electrical power (NEP) system. Design Ba sis Functions
  • Provide facility power during normal operations Design Ba sis Values
  • 30-year design life 3.5.2.7.ll Process Vessel Ventilation System Chapter 9.0 , Section 9.1 provides a detailed description of the process vesse l ventilation system. Design Basis Functions
  • Provide primary system functions to protect on-site and off-site personnel from radiological and other industrial related hazards *
  • Collect air in-leakage sweep from each of the numerous vessels and other components in main RPF processes and maintain hydrogen concentration process tanks and piping below lower flammability limit Minimize reliance on administrative or complex active engineering controls to provide a confinement sys tem as simple and fail-safe as reasonably possible Design Basis Values * *
  • Maintain primary fission product boundary during and after normal operations , shutdown conditions, and DBEs 30-year design life Contain and store noble gases generated in the RPF to meet 10 CFR 20 requirements 3.5.2.7.12 Facility Ventilation System Chapter 9.0 , Section 9.1 provides a detailed description of the facility ventilation sys tem. D esign Basis Functions
  • Provide confinement of hazardous chemical fumes and airborne radiological materials and conditioning of RPF environment for facility personnel and equipment
  • * *
  • Prevent release and dispersal of airborne radioactive materials (e.g., maintain pressure gradients to ensure proper flow of air from least potentially contaminated areas to most potentially contaminated areas) to protect health and minimize danger to life or property Maintain dose uptake through ingestion to levels as lo w as reasonably achievable (ALARA) Provide makeup air and condition the RPF environment for process and electrical equipment Process exhaust flow from the process vessel ventilation system 3-60

.; ... ;. NWMI ...... ..* .... ........ *.* * *. * .. NOmfWEST MEOtCAL lSOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components

  • Provide confinement of airborne radioactive materials by providing for the rapid , automatic closure of isolation dampers within confinement zones for various accident conditions Provide conditioned air to ensure suitable environmental conditions for personnel and equipment inRPF Design Basis Values
  • Maintain primary fission product boundary during and after normal operations , shutdown conditions , and DBEs * * *
  • Provide an integrated leak rate for confinement boundaries that meets the requirements of accident analy s es that complies with 10 CFR 10.61 30-year design life Maintain occupied space at 24 degrees Celsius (0 C) (75 degrees Fahrenheit

[°F]) (summer) and 22°C (72°F) (winter), with active ventilation to support workers and equipment Maintain air quality that complies with 10 CFR 20 dose limits for normal operations and shutdown 3.5.2.7.13 Fire Protection System Chapter 9.0 , Section 9.3 provides a detailed description of the RPF fire protection system. Design Basis Functions

  • Provide detection and suppression of fire s *
  • *
  • Generate alarm signal s indicating presence and location of fire Execute commands appropriate for the particular location of the fire (e.g., provide varying levels of notification of a fire event and transmitting notification to RPF central alarm station and RPF control room) Provide fire detection in RPF and initiate fire-rated damper closures Remain functional during DBEs Design Basis Values * *
  • 30-year design life Provide a constant flow of water to an area experiencing a fire for a minimum of 120 min based on the size of the area per International Fire Code (IFC, 2012) Provide sprinkler systems , when necessary , per National Fire Protection As s ociation (NFPA) 13 , Standard for th e In s tallation of Sprinkl er S ys t e m s 3.5.2.7.14 Plant and Instrument Air System Chapter 9.0 , Section 9.7.1 provides a detailed description of the RPF plant and instrument air system. Design Basis Functions
  • Provide small , advective flows of plant air for several RPF activities (e.g., tool operation , pump power , purge gas in tanks , valve actuation , and bubbler tank level measurement)
  • Provide plant air receiver buffer capacity to make up difference between peak demand and compressor capacity 3-61

.; .. ; .. *NWMI ..**.. .. *.. ........... 0 NORTHWEnllBMtAl.ISOTOPfS NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components

  • Provide plant air to instrument air subsystem for bubblers and valve actuation Provide instrument air receiver buffer capacity to make up difference between peak demand and compressor capacity Design Basis Values *
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) Provide instrument air dried in regenerable desiccant beds to a dew point of no greater than -40°C (-40°F) and filtered to a maximum 40 micron (µ) particle size 3.5.2.7.15 Emergency Purge Gas System Chapter 6.0 , Section 6.2.1.7.5 provides a detailed description of the emergency purge gas system. Design Basis Functions
  • *
  • Provide > 12 hr of nitrogen to the emergency purge gas system Emergency purge gas system to provide nitrogen to the required process tanks Remain functional during DBEs Design Basis Values
  • 30-year design life with the exception of common replaceable parts
  • Maintain hydrogen gas (H 2) concentrations less than 25% of the lower flammability limit 3.5.2.7.16 Gas Supply System Chapter 9.0 , Section 9.7.1 provides a detailed description of the gas supply system. Design Basis Functions
  • Provide nitrogen from a tube truck to the chemical supply room where manifold piping will be used to distribute the gas Provide adequate flow to ensure that the accumulation of combustible gases is below hazardous concentrations and reduces radiological hazards due to accumulation of gaseous fission products Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g., pumps)
  • Provide standard gas bottles , with capacity of approximately 8 , 495 L (300 cubic feet [ft 3]) 3.5.2.7.17 Process Chilled Water System Chapter 9.0 , Section 9.7.1 provides a detailed description of the RPF chilled water system. Design Basis Functions
  • Provide process chilled water loop for three secondary loops heat exchangers One large geometry secondary loop in hot cell One criticality-safe geometry secondary loop in hot cell One criticality-safe geometry secondary loop in target fabrication area Provide monitoring of chilled water loops for loss of primary containment 3-62
.**.*.*.* ..... ; .. NWMI ........... ' !*. * . NOmfWEST llEDICAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Components
  • Provide cover gas to prevent flammable conditions in secondary loops Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) *
  • Chilled water to various process equipment at no greater than 10°C (50°F) during normal operations Maintain the hydrogen concentration in the coolant system at less than 25 percent of the lower flammability limit of 5 percent H 2 3.5.2.7.18 Facility Chilled Water System Chapter 9.0 , Section 9.7.1.2.2 provides a detailed description of the RPF facility chilled water system. Design Basis Functions
  • Provide cooling media to heating , ventilation , and air conditioning (HVAC) system Supply HVAC system with cooling water that is circulated through the chilled water coils in handling units Design Basis Values *
  • Provide cooling water at a temperature of9°C (48°F) to the HV AC air-handling unit cooling coil s 30-yeardesign life with the exception of common replaceable parts (e.g., pumps) 3.5.2.7.19 Facility Heated Water System Chapter 9.0 , Section 9.7.1.2.2 provides a detailed description of the RPF heated water system. Design Basis Functions
  • Provide heated media to HV AC system Supply the HV AC system with heated water that is circulated through the heated water coils in the air-handling unit s Design Basis Values
  • Provide heated water at a temperature of 82°C (180°F) to HV AC air-handling unit heating coils and reheat coil
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) 3.5.2.7.20 Process Steam System -Boiler Chapter 9.0 , Section 9. 7.1 provides a detailed description of the RPF process steam sy s tem for the boiler. Design Basis Functions
  • Generate low-and medium-pressure steam using a natural gas-fired package boiler * *
  • Provide a closed loop steam system for the hot cell secondary loops that meets criticality control requirements Provide monitoring of steam condensate for loss of primary containment Limit sludge or dissolved solids content with automatic and makeup water streams in the boiler 3-63

.; ... .. NWMI ...... ... ... ........ *.* *

  • 0 MOkTNWEST llEDtCAL ISGTOfO NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Design Basis Values *
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) Provide saturated steam at 1.7 kg/square centimeters (cm 2) (25 lb/square inch [in.2]) and 4.2 kg/cm 2 (60 lb/in.2) gauge to various process equipment 3.5.2.7.21 Process Steam System -Hot Cell Secondary Loops Chapter 9.0 , Section 9.7. l provides a detailed description of the RPF process steam system for the hot cell secondary loops. Design Basis Functions
  • *
  • Provide a closed loop steam system for the hot cell secondary loops Generate low-pressure steam using a vertical shell-and-tube heat exchanger Provide monitoring of steam condensate for loss of primary containment Design Basis Values
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) 3.5.2.7.22 Demineralized Water System Chapter 9.0 , Section 9.7.1 provides a detailed description of the RPF demineralized water system. Design Basis Functions
  • Provide demineralized water to RPF except for administration and truck bay areas * *
  • Remove mineral ions from municipal water through an ion exchange (IX) process and accumulate in a storage tank Provide regenerable IX media using a strong acid and a strong base Feed acids and ba s es from local chemical drums by toe pump s Design Basis Values *
  • 30-year design life with the exception of common replaceable parts (e.g., pumps) Provide the water at 4.2 kg/cm 2 (60 lb/in.2) gauge 3.5.2.7.23 Supply Air System Chapter 9.0 , Section 9.1.2 provides a detailed description of the supply air system. The design basis functions and values are identified in Section 3.5.2.7.12. 3.5.2.7.24 Chemical Supply System Chapter 9.0 , Section 9.7.4 provides a detailed description of the chemical supply s ystem. Design Basis Functions
  • Provide storage capability for nitric acid , sodium hydroxide, reductant , and nitrogen oxide absorber solutions , hydrogen peroxide, and fresh uranium IX resin
  • Segregate incompatible chemicals (e.g., acids from bases)
  • Provide transfer capability for chemical solutions mixed to required concentrations and used in target fabrication , target dissolution , Mo recovery and purification , and waste management systems 3-64

.. NWMI ...... ..* .... ........ *.* 0 ! *. * ! . NOITHWlST llfDICAl tsOTOPH NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Design Basis Values

  • 30-year design life with the exception of common replaceable parts (e.g., pumps) 3.5.2. 7.25 Biological Shielding System Chapter 4.0 , Section 4.2 , provides a detailed description of the RPF biological shielding. Design Basis Functions
  • Provide biological shielding from radiation sources in the hot cells for workers in occupied areas of the RPF
  • Limit physical acce s s to hot cells
  • Remain functional through DB Es without loss of structural integrity Design Basis Values
  • 30-year design life
  • Provide dose rates consistent with ALARA goals for normall y occupied areas 3.5.2.7.26 Facility Process Control System Chapter 7.0 , Section 7.2.3 provides a description of the FPC system. Design Basis Functions
  • Perform as overall production process controller
  • * * * *
  • Monitor and control process instrumented functions within the RPF (e.g., process fluid transfers , controlled inter-equipment pump transfers of process fluids) Provide monitoring of safety-related components while BMS (a subset of the FPC system) monitors v entilation system and mechanical utility systems Ensure ESF systems operate independently from FPC sy s tem or BMS Use hard-wired analog controls/interlock s for each ESF safety function to protect workers , public , and en v ironment Integrate into and monitor ESF parameters and alarm functions by FPC sy s tem or BMS Initiate actuation of i s olation dampers for hot cell area or analytical area on receipt of signals from fire protection system Design Basis Values
  • 30-yearde s ign life with the exception of common replaceable parts (e.g., controllers) 3-65

..... ;. NWMI ..**.. ..* *.. ........... * *: .° NOlllTNWHT MEDM:Al ISOTOPES

3.6 REFERENCES

NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components 10 CFR 20, " Standards for Protection Against Radiation ," Cod e of F e d e ral R e gulation s, Office of the Federal Register, as amended. I 0 CFR 30, " Rules of General Applicability to Domestic Licensing of Byproduct Material," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities

," Code of F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CFR 50.2 , " Definitions

," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. I 0 CFR 50.31 , "Combining Applications

," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CFR 50.32 , " Elimination of Repetition

," Code of F e d e ral R e gulation s, Office of the Federal Register , as amended. I 0 CFR 70 , " Domestic Licensing of Special Nuclear Material ," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CFR 70.61 , " Performance Requirements

," Cod e of F e d e ral R e gulations , Office of the Federal Register , as amended. I 0 CFR 70.64 , "Requirements for New Facilities or New Processes at Existing Facilities

," Code o f F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CFR 71 , " Energy: Packaging and Transportation of Radioactive Material ," Cod e of Fed e ral R e gulations , Office of the Federal Register , as amended. 10 CFR 73 , " Physical Protection of Plants and Materials ," Cod e of F e d e ral R e gulations , Office of the Federal Register , as amended. 10 CFR 74 , " Material Control and Accounting of Special Nuclear Material ," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CFR 851 , " Worker Safety and Health Program ," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. 10 CSR 10-6.01 , " Ambient Air Quality Standards," Mi ss ouri Cod e of Stat e R e gulations , as amended. 20 CSR 2030 , " Missouri Board for Architects , Professional Engineers , Professional Land Surveyors , and Land s cape Architects

," Cod e of Stat e R eg ulation s, Jeffer s on City , Missouri , as amended. 21 CFR 210 , " Current Good Manufacturing Practice in Manufacturing, Processing, Packaging , or Holding of Drugs," Cod e of F e d e ral Regulation s, Office of the Federal Register , as amended. 21CFR211 , " Current Good Manufacturing Practice for Finished Pharmaceuticals

," Cod e of Fed e ral R e gulations , Office of the Federal Register , as amended. 29 CFR 1910 , "Occupational Safety and Health Standards ," Cod e of F e d e ral R e gulations , Office of the Federal Register , as amended. 40 CFR 61 , "National Emissions Standards for Hazardous Air Pollutants

," Cod e of F e d e ral R e gulation s, Office of the Federal Register , as amended. 40 CFR 63 , "NESHAP for Source Categories," Code of Federal Regulations , Office of the Federal Register , as amended. 3-66

.... ; .. NWMI ........... ........ *.* * !*. * !

  • NORTHWEST MmlCAL tsOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures , Systems and Comp o nents 40 CFR 141 , "National Primary Drinking Water Regulations

," Code of Federal Regulations , Office of the Federal Register , as amended. ACGIH 2097, Industrial V e ntilation:

A Manual of R eco mm e nd ed Practice for Design, 28 1 h Edition, American Conference of Governmental Industrial Hygienists , Cincinnati, Ohio , 2013. ACI 318, Building Cod e R e quirements for Structural Concret e Commentary, American Concrete Institute , Farmington Hills , Michigan , 2014. ACI 349 , Code Requir e ments for Nuclear Safety-R e lat e d Concrete Structures and Commentary, American Concrete Institute , Farmington Hills , Michigan , 2013. AISC 360 , Specification for Structural Steel Building s, American Institute of Steel Construction, Chicago , Illinois , 2010. AMCA Publication 201 , Fans and Systems , Air Movement and Control Association International , Inc., Arlington Heights , Illinois , 2002 (R2011 ). AMCA Publication 203 , Field Performanc e Measurem e nt of Fan S ys tems, Air Movement and Control Association International, Inc., Arlington Heights, Illinois , 1990 (R2011 ). AMCA Publication 211, C e rtifi ed Rating s Program -Product Rating Manual for Fan Air P erfor man ce, Air Movement an d Control Association International , Inc., Arlington Heights, Illinois , 2013. AMCA Publication 311, C e rtifi e d Rating s Program -Product Rating Manual for Fan Sound Performance , Air Movement and Control Association International, Inc., Arlington Heights , II lino is , 2006 (R2010). ANS 2.8, D eter mining D esig n Basis Flooding at Po wer R eacto r Sites , American Nuclear Society , La Grange Park , Illinois, 1992 (W2002). ANSI C84.l, American Nationa l Standard for Electri c Power Systems and Equipment

-Voltage Rating s (60 Hertz), American National Standards Institute, Inc., Washington, D.C., 2011. ANSI Nl 3.1, Sampling and Monitoring R e l eases of Airborne Radioa ctive Substances from the Stacks and Ducts of Nuclear Facilities , American Nuclear Society , La Grange Park , Illinois, 2011. ANSI N42. l 7B, A m e rican Nationa l Standard Performanc e Sp ec ifications for H ea lth Physics Instrum entation -Occupational A irborn e Radioa ctiv i ty Monitoring In strumentat ion , American National Standards Institute , Inc., Washington , D.C., 1989. ANSI N42. l 8 , Specification and P erfo rman ce of On-Sit e In s trum e ntation for Continuously Monitoring Radioa c tivity in Effluents, American National Standards Institute , Inc., Washington, D.C., 2004. ANSI N323D, American Nationa l Standard for In sta ll e d Radiation Protection In stru m e ntation , American National Standards Institute , Inc., Washington , D.C., 2002. ANSI/AHR!

Standard 365 , P e rformance Rating of Commercial and Industrial Unitary Air-Conditioning Condensing Units , Air-Conditioning , Heating , and Refrigeration Institute , Arlington, Virginia , 2009. ANSI/ AHRI Standard 390 , Performance Rating of Single Packag e Vertical Air-Conditioners and Heat Pumps, Air-Conditioning , Heating , and Refrigeration Institute , Arlington, Virginia , 2003. ANSI/ AHRI Standard 410 , Forced-Circulation Air-Cooling and Air-Heating Coils , Air-Conditioning , Heating , and Refrigeration Institute , Arlington , Virginia, 2001. ANSI/AHR!

Standard 430 , Performance Rating of Central Station Air-Handling Units, Air-Conditioning , Heating , and Refrigeration Institute , Arlington , Virginia, 2009. 3-67

... :*:; ..... NWM I .*.******** ' !*. * ! ." NO<<TKWEST MEDtcAl '5GTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components ANSl/AHRI Standard 850 , P e rforman ce Rating of Comm e r c ial and Indu s trial Air Filter Equipm e nt , Conditioning, Heating , and Refrigeration Institute , Arlington , Virginia , 2013. ANSl/AIHA/ASSE Z9.5, Laboratory Ventilation, American Society of Safety Engineers , De s Plaines , Illinois , 20 I 2. ANSl/AISC N690 , Spe c ifi c ation for Safe ty-R e lated St ee l Structures for Nuclear Fa c iliti es, American Institute of Steel Construction , Chicago , Illinois , January 31 , 2012. ANSl/AMCA 204 , Balanc e Quality and Vibration Lev e ls for Fan s, Air Movement and Control Association International , Inc., Arlington Heights , Illinois , 2005 (R2012). ANSl/AMCA 210, Laborato ry M e thods for T es ting Fan s for Rating s, Air Movement and Control Association International , Inc., and American Society of Heating , Refrigerating and Air Cond i tioning Engineers , Inc., Arlington Heights , Illinois , 1999. ANSI/ ANS-2.26 , Categori z ation of Nuclear Facility Stru c tures, System s , and Components for Seismi c D es ign , American Nuclear Society , La Grange Park , Illinoi s, 2004 (R2010). ANSl/ANS-2

.2 7 , Criteria for Inv es tigation s of Nuclear Fa c ility Sit es for S eis mi c Ha za rd A ssess m e nt s, American Nuclear Society , La Grange Park , Illinois , 2008. ANSl/ANS-2.29 , Probabili st i c Seismic Ha z ard Anal ys i s, American Nuclear Society , La Grange Park , Illinois , 2008. ANSl/ANS-6.4 , N uclear Analysis and D es ign of Con cre t e Radiation Shi e lding for Nuclear Po we r Plant s, American Nuclear Society , La Grange Park , Illinois , 2006. ANSI/ANS-6.4.2 , Specification for Radiation Shielding Mat e rials, American Nuclear Society, La Grange Park , Illinois , 2006. ANSI/ ANS-8.1 , Nuclear Criticality Saf ety in Op e rations with Fissionabl e Mat e rial s Outsid e R e a c tors , American Nuclear Society , La Grange Park, Illinois , 1998 (R2007) (W2014). ANSI/ ANS-8.3, Critically Accident Alarm S ys t e m , American Nuclear Society , La Grange Park , Illinois , 1997 (R2003, R2012). ANSI/ ANS-8. 7 , N ucl e ar Criticality Saf ety in th e Stora ge of Fissile Materials, American Nuclear Society , La Grange Park , Illinois , 1998 (R2007). ANSI/ ANS-8.10, Criteria for Nuclear Criti c ality Control in Op e rations with Shielding and Confinement, American Nuclear Society , La Grange Park , Illinois , 1983 (Rl988 , R1999 , R2005). ANSI/ ANS-8.19 , Admini st rati ve Practi ces for Nuclear Criticality Saf ety, American National Standards Institute/American Nuclear Society, La Grange Park , Illinoi s, 1996 (R2014). ANSI/ ANS-8.20 , Nuclear Criti c ality Saf ety Training , American National Standards Institute/ American Nuclear Society , La Grange Park , Illinois , 1991 (R2005). ANSI/ ANS-8.21 , Use of Fix e d Neutron Absorbers in N ucl e ar Faciliti es Outsid e R e a c tor s, American Nuclear Society, La Grange Park , Illinois , 1995 (R2011). ANSl/ANS-8.24 , Validation of Neutron Transport M e thods for Nuclear Criticali ty Safety Cal c ulations , American National Standards Institute/ American Nuclear Society , La Grange Park, Illinois, 2007 (R2012). ANSl/ANS-10.4 , V e rifi c ation and Validation of Non-Saf ety-Relat ed S c i e ntific and Engine e ring Comput e r Programs for the Nuclear Indu stry , American Nuclear Society , La Grange Park , Illinois , 2008. 3-68

.; ... ; .. NWMI ...... ..* .... ...........

  • NCMITifWEST MEOtcAl tsOTOP(S NWMl-201 3-0 2 1 , Rev. 1 Chapter 3.0 -Design of Structures , Systems and Component s ANSVANS-10

.5 , A c commodating Us e r N ee ds in Comput e r Program Dev e lopm e nt , American Nuclear Society , La Grange Park , Illinois , 2006 (R201 l). ANSVANS-15

.1 7, Fir e Prote c tion Pro g ram Crit e ria for R e s e arch R e a c tor s, American Nuclear Society , La Grange Park , Illinois , 1981 (R2000) (W2014). ANSVANS-40

.37, Mobil e Low-L eve l Radioa c tiv e Wa s t e Pro cess ing S ys t e ms , American Nuclear Society , La Grange Park , Illinois , 2009. ANSVANS-55.1 , Solid Radioa c tiv e Wa s t e Pro c e ss ing S ys t e m for Light Wat e r Cool ed R e a c tor Plant s, American Nuclear Society , La Grange Park , Illinois , 1992 (R2000 , R2009). ANSVANS-55.4 , Gas e ou s Radioa c tiv e Wa s t e Pr ocess ing S ys t e ms f or Li g ht Wat er R e actor Plant s, American Nuclear Society , La Grange Park , Illi n ois, 1993 (R l 999 , R2007). ANSVANS-55

.6 , Liquid Radioactiv e Wast e Proc ess in g S ys tem for Light Wat er R e a c tor Plant s , American Nuclear Society , La Grange Park , Illinois , 1 993 (R l 999 , R2007). ANSVANS-58.3 , Ph ys i c al Prot e ction for N ucl e ar Saf ety-R e l at ed S ys t e ms and Component s, Ame r ica n Nuclear Society , La Gra n ge Park , Ill i n ois , 1992 (Rl 998 , R2008). ANSl/ANS-58.8 , Tim e R es pon se D es i g n C rit e ria f or S a f ety-R e lat e d Op e rator A c tion s , Ameri c an N u clear Society , La Grange Park , Ill inoi s, 1994 (R2001 , R2008). ANSVANS-59.3 , N ucl e ar Saf ety Crit e ria f o r Control A ir S ys t e m s, American Nuclear Society , La Gra n ge Park , Illinois , 1992 (R2002) (W2012). ANSVASHRAE 51-07 , Laborato ry M e thod s of T es tin g Fan s for Ce rtifi ed A e rod y namic P e rf o rmanc e Rating , American Society of Heating , R efrigerating , and Air-Conditionin g Engineers , At l anta , Georgia , 2007. ANSI/ASHRAE 1 10 , M e th o d of T es tin g P e rforman ce of Labo r ato ry Fum e Hood s, Amer i can Soc i ety of Heating , Refrigerating , and Air-Con d itioning Engineer s, Atlanta , Georgi a, 1995. ANSI/ ASHRAE 111 , M e a s ur e ment , T es tin g, Adju s tin g and Balan c ing of Buildin g H e ating , V e ntilation , Air-Co nditionin g and R e fri ge ration S ys t e ms , American Society of Heating , Refrigerating , and Air-Conditioning Engineer s, Atlanta , Georgia , 2008. ANSVASHRAE S t an d ard 15 , Saf ety Standard for R e fri ge ration S ys t e m s, American Society of Heating , Refrigerating , and Air-Conditioning E n gineer s, Atlanta , Geor g ia , 2013. ANSVASHRAE Standard 52.2 , M e tho d of T e sting G e n e ral V e ntilati o n Air-Cl e aning Devi ces for Removal Effi c i e n cy by Particl e Si ze, American Society of Hea t ing , Refrigera t i n g , and Air-Conditioning Engineers , At l anta , Georg i a , 2007. ANS V ASHRA E Stan d ard 55 , Th e rmal En v ironm e ntal C onditions f o r Human O cc upan cy, American Society of Heating , Refr i gerating , and Air-Conditioning Engineers , Atlanta , Georgia , 2013. ANSVASHRAE Stan d ard 62.1 , V e ntilation/or Acce p t abl e Indoor A ir Quali ty, American Society of Heating , R efrigerating , and Air-Con d itioning Engi n eers , Atlanta , Georgia , 20 1 0. ANSVASHRAE

/IES 90.1 , En e r gy Standard for Building s Exc e pt L ow-Ri se R es id e ntia l Building s, American Society of Heat i ng, Refrigerating , and Air-Co n ditioning Engineers , At l an t a , Georgia , 2010. ANSI/HI 3.1-3.5 , Rotary Pump s, Hydraulic Inst itu te , Parsippany , New Jersey , 200 8. ANSVIEEE C2, 2012 National El ec tri c al Safety Cod e (N ESC), In s titute of E l ectrica l a nd E l ectronics Engineers , Pi scataway , New Jersey , 20 1 2. 3-69

..... ... NWMI ..*... ..* **: ........ *.* NOmfWEST MEDtCAllSOTOPfS NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components ANSI/IEEE N320 , American National Standard P e rformance Spe c ifi c ations for R e a c tor Em e rgen cy Radiological Monitoring Instrum e ntation , Institute of Electrical and Electronics Engineers , Piscataway, New Jersey, 1979. ANSI/JES RP-I -I 2, American National Standard Practice for Office Lighting , Illuminating Engineering Society , New York , New York , 2012. ANSl/ISA-5

.06.01-2007 , Fun c tional R e quir e m e nts Do c um e ntation for Control Software Appli c ations, The International Society of Automation , Research Triangle Park , North Carolina , 2007. ANSl/ISA-5.

1-2009 , In s trum e ntation S y mbols and Id e ntifi c ation , The International Society of Automation , Research Triangle Park , North Carolina , 2009. ANSl/ISA-7

.0.01-1996 , Quality Standard for In s trum e nt Air , The International Society of Automation, Research Triangle Park , North Carolina , 1996. ANSl/ISA-12.01.01-2013 , D e finition s and Information P e rtaining to El e ctrical Equipment in Ha z ardous (Cla ss ified) Location s, The International Society of Automation , Research Triangle Park , North Carolina , 2013. ANSl/ISA-67.04

.01-2006 , S e tpointsfor N uclear Saf e ty-R e lat e d In s trum e ntation, The International Society of Automation , Research Triangle Park , North Carolina , 2006 (R201 l). ANSl/ISA-75.05.01-2000 , Control Val ve T e rminology , The International Society of Automation , Research Triangle Park , North Carolina , 2000 (R2005). ANSl/ISA-82.03-1988 , Saf ety Standard for El e ctri cal and El ec troni c Te s t , M e a s urin g, Controlling , and Related Equipment , The International Society of Automation , Research Triangle Park , North Carolina , 1988. ANSl/ISA-TR99.00.0l-2007 , S e curity Te c hnologi es for Industrial Automation and Control S ys t e m s , The International Society of Automation , Research Triangle Park , North Carolina , 2007. ANSl/ITSDF B56. l , Safety Standard for Low Lift and High Lift Trucks , Industrial Truck Standards Development Foundation , Washington , D.C., February 2013. ANSl/NEMA Z535. l, Safe ty Color s, American National Standards Institute , Inc., Washington , D.C., 2006 (R2011). ANSl/NEMA Z535.2 , Environmental and Fa c ility Saf ety Sign s, American National Standards Institute , Inc., Washington , D.C., 2011. ANSl/NEMA Z535.3, Crit e ria for Saf ety Symbols , American National Standards Institute , Inc., Washington , D.C., 2011. ANSl/NEMA Z535.4 , Produ c t Saf e ty Sign s and Lab e l s, American National Standards Institute , Inc., Washington , D.C., 2011. ANSI/NET A ATS-2013 , Standard for Acce ptanc e Te s ting Sp ec ifi c ations for Ele c trical Power Distribution Equipm e nt and S ys t e m s, InterNational Electrical Testing Association , Portage, Michigan, 2013. ANSI/NET A ETT-2010 , Standard for Certification of El ec tri c al T es ting T e chni c ian s, Inter National Electrical Testing Association , Portage , Michigan , 2010. ANSI/NETA MTS-2011 , Maintenanc e T es ting Sp ec ifi c ations for El ec trical Pow er Di s tribution Equipm e nt and S ys t e m s, InterNational Electrical Testing Association , Portage, Michigan , 2011. 3-70

.. NWMI ...*.. .. **: ....... !.* . ! *.* . NOmlWlST .OICAl tsOTOPfl NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components ANSI/SMACNA 001-2008 , S e ismi c R e straint Manual: Guid e lin es for Me c hani c al S y st e m s, Sheet Metal and Air Conditioning Contractors

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NECA 409 , Standard/or In s talling and Maintaining Dry-T y p e Tran s f or m e r s (A N SI), National E lectrical Contr a ctors Association , Bethesda , Maryland , 2009. NECA 410 , Standard/or In s talling and Maintaining Liquid-Fill e d Tran s former s (AN SI), National Electrical Contractors Association , Bethesda , Maryland , 2013. NECA 411 , Standard for In st alling and Maintaining U nint e rruptibl e Pow e r Suppli es (UPS) (A N SI), Nation a l Electrical Contractor s A s sociation , Bethesda , Maryland, 2006. N E CA 420 , Standard/or Fu se A ppli c at i on s (AN SI), National Electrical Contractors Association , Bethe s da , Maryland , 2014. NECA 430 , Standard/or In s talling M e dium-Voltag e M e tal-Clad S w it c h ge ar (A NS I), National Electrical Contractors Association, Bethe s da , Maryland , 2006. NECA/AA 104 , Standard/or In s tallin g Aluminum Buildin g Wir e and Cable (A N SI), National Electrical Contractors Association , Bethesda , Maryland , 2012. NEC A/BICSI 568 , Standard/o r In s tallin g Buildin g T e l ec ommuni c a t i o n s Cablin g (AN SI), National Electrical Contractor s A ss ociation , Bethesda , Maryland , 2006. NEC A/EGSA 404 , Standard/or In s tallin g G e n e rat o r Se t s (AN SI), National Electrical Contractor s As s ociation , Bethe s da , Maryland , 2014. NEC A/FOA 301 , Standa r d/o r In s tallin g and T es tin g Fib e r Opti cs, National Electrical Contractors Association , Bethesda , Maryland , 2009. NEC A/IESNA 500 , R eco mm e nd e d Pra c ti ce/or In s tall i n g Indo o r Li g h t in g S ys t e m s (AN SI), National Electrical Contractors Association , Bethesda , Maryland , 2006. NEC A/IESNA 50 l , R ec omm e nd e d Pra c ti ce for In s tallin g E x t e rior Li g htin g S ys t e ms (ANSI), National Electrical Contractor s Association , Bethesda , Maryland , 2006. NEC A/IESNA 502, R ec omm e nd e d Pra c ti ce/or In s tallin g Indu s trial Li g htin g S yste m s (A N SI), National Electrical Contractor s Association , Bethesda , Maryland , 2006. NECA/NCSCB 600 , R ec omm e nd ed P r a ct i ce for In s talling and Ma i ntaining M e dium-Volta ge C abl e (AN SI), National Electrical Contractors Association , Bethe s da , Maryland , 2014. NECA/NEMA 105 , Stand a rd/or In s tallin g M e tal C abl e Tra y S ys t e m s (AN SI), N a tional Electrical Contractor s As s ociation , Bethe s da , Maryland , 2007. N ECA/NEMA 605 , In s tallin g U nd e rgr o und N onm e talli c U tili ty D uc t (AN SI), National Electri c al Contractors Association , Bethe s da , Maryland , 2005. NEMA MG-1 , Motors and G e n e rat o r s, National Electrical Manufa c turers Association , Ro ss lyn , Virginia , 2009. NFPA 1 , Fir e Cod e, National Fire Protection Association , Quincy , Massachusett s, 2015. NFPA 2 , H y d r o gen T ec hn o lo g i es Cod e, National Fire Protection A s sociation , Quincy , Massachusetts , 2011. NFPA 4 , Standard/or Int eg r a t e d Fir e Pr o t ec tion and Lif e Saf ety S ys t em T es tin g, National Fire Protection Association , Quincy , Massachusetts , 2015. 3-78

.; ... ; .. NWMI ..*... ..* **: .-.* .. *:.* * ! ." NORTHWEST MEOtcAI. ISOTOPES NWMl-201 3-0 2 1 , Rev. 1 Chap t e r 3.0 -Des i gn o f Struc t ures , Sys t ems and Compo n en ts NFPA 10 , St a ndard for P o rtabl e Fir e E x tinguish e r s, National Fire Protection Association, Quincy , Massachusetts , 2013. NFPA 13 , Standard for th e Installation of Sprink l er S ys t e m s, National Fire Protection Association , Quincy, Massachusetts , 2013. NFPA 14 , Standard for th e Installation of Standpip e and Hose S ys t e m s, National Fire Protec t ion Association , Quincy , Massachusetts , 2013. NFPA 20 , Standard for th e In s tallation of Stationary Pump s for Fir e Prot ec tion , Nationa l Fire Protection Association , Quin c y , Massachusetts , 2013. NFPA 22 , Standard for W a t e r Tanks fo r Privat e Fir e Prot ec tion , Nationa l Fire Protection Association , Quincy , Massachusetts , 2013. NFP A 24 , Standard for th e In s tallation of Privat e Fir e S e rvi ce Main s and Th e ir A ppurt e nan ces, National Fire Protection Association , Quincy , Massachusetts , 2013. NFPA 25 , St a ndard for th e In s p e ction , T es ting , and Maint e nan ce of Wat e r-Ba se d Fir e Prot ect i o n S ys t e m s, National Fire Protection As s ociation , Quincy , Ma s sachusetts , 2014. NFPA 30 , Flammabl e and C o mbu s tibl e Liquids Cod e, Nationa l Fir e Protection Association , Quincy , Mas s achusetts , 2015. NFPA 37 , Standard for th e In s tallation and Us e of Stationary Combu s tion Engin es and Ga s Turbin es, National Fire Protection Association , Quincy , Massachusetts , 20 1 5. NFPA 45 , Standard on Fir e Prot e ction f or Laborator ies Using Ch e mi c al s, Nationa l Fire Protection Association , Quincy , Massachusetts , 2015. NFPA 55 , Compr esse d Ga ses and Cry oge ni c Fluid s C od e, National Fire Protection Association , Q u incy , Massachusetts , 2013. NFPA 59A , Standard for th e Produ c tion , Storag e, and Handling of Liqu e fi e d Na t ural Gas , National Fire Protection A s sociation , Quinc y, Ma ss achusett s, 2013. NFPA 68 , Standard on E x plo s ion Prot e ction by D e jla g ration V e ntin g, National Fire Protection Association , Quincy , Ma s sachusetts , 2013. NFPA 69 , Standard on E x plo s ion Pr e v e ntion S ys t e m s, Nationa l Fire Protect i on A ss ociation , Quincy , Massachusetts , 2014. NFPA 70 , N ational El ec t r i c al Cod e (NE C), National Fire Pro t ection Association , Quincy , Ma ss ac hu setts , 2014. NFPA 7 0B , R ec omm e nded Pra c tic e for El ec tri c al Equipm e nt Maint e nan ce, National Fire Protection Association, Quincy , Massachu s etts , 2013. NFPA 70E , Standard for El ec tri c a l Saf e ty in th e Workpla c e , National Fire Protection Association , Quincy , Massachusetts , 2015. NFPA 72 , Nationa l Fir e Ala r m and Si g naling Cod e, Nationa l Fire Protection Association , Q u incy , Massachusetts , 2013. NFPA 75 , Standard for th e Fir e Prot ec ti o n of Information T e chnolo gy Equipm e nt , Nat i ona l Fire Protection Association , Quincy , Massachusett s, 2013. NFPA 7 9 , El ec tri c al Standard for I ndu s trial Ma c hine ry, National Fire Protection Association , Q u incy , Massachusetts , 2015. 3-7 9

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  • NOllT1fWUT MEDtCAl ISOTOPH NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components NFPA 80, Standard for Fire Doors and Oth er Op ening Protectives, National Fire Protection Association, Quincy , Massachusetts, 2013. NFPA 80A, R eco mm e nd ed Pra ctice for Protection of Buildings from Exterior Fire Exposures, National Fire Protection Association, Quincy , Massachusetts , 2012. NFPA 86, Standard for Ov e n s and Furnaces, National Fire Protection Association, Quincy , Massachusetts , 2015. NFPA 86C, Standard for Indu stria l Furnaces Using a Special Pro cessi n g Atmosphere, National Fire Protection Association, Quincy , Massachusetts, 1999. NFPA 90A , Standard for the In sta llation of Air-Conditioning and Ventilating System, National Fire Protection Association, Quincy , Massachusetts, 2015. NFPA 90B , Standard for the In sta llation of Warm Air H eat ing and Air-Conditioning Systems, National Fire Protection Association, Quincy , Massachusetts , 2015. NFP A 91 , Standard for Exhaust Systems for Air Conveying of Vapors, Gases, Mists, and Noncombustible Particulate Solids, National Fire Protection Association, Quincy , Massachusetts , 2015. NFPA 92 , Standard for Smoke Control Systems, National Fire Protection Association, Quincy , Massachusetts , 2012. NFPA 92A, Standard for Smoke-Control Systems Uti li zing Barri ers and Pressure Differences, National Fire Protection Association, Quincy , Massachusetts , 2009. NFPA 92B , Standard for Smoke Management Systems in Malls, Atria, and Lar ge Spaces, National Fire Protection Association , Quincy , Massachusetts , 2009. NFPA 101 , lif e Safety Code, National Fire Protection Association, Quincy , Massachusetts , 2015. NFPA 10 lB , Code for Means of Egress for Buildin gs and Structures, National Fire Protection Association, Quincy , Massachusetts, 2002 (W-Next Edition).

NFPA 105 , Standard for the In sta llation of Smoke Door Assemblies and Oth er Opening Protectives, National Fire Protection Association, Quincy , Massachusetts, 2013. NFPA 110 , Standard for Emerge n cy and Standby Power Systems, National Fire Protection Association, Quincy , Massachu se tts , 2013. NFPA 111 , Standard on Stored Electrical Energy Emergency and Standby Power Systems, National Fire Protection Association , Quincy , Massachusetts, 2013. NFPA 170 , Standard for Fire Safety and Eme rg ency Symbols, National Fire Protection Association, Quincy , Massachusetts, 2012. NFPA 204, Standard for Smoke and H eat Venting, National Fire Protection Association, Quincy , Massachusetts , 2012. NFPA 220, Standard on Types of Buildin g Construction, National Fire Protection Association, Quincy , Massachusetts , 2015. NFPA 221, Standard for High Challenge Fire Walls, Fire Walls, and Fire Barri er Walls, National Fire Protection Association, Quincy , Massachusetts, 2015. NFPA 262, Standard Method of T est for Flame Tra vel and Smoke of Wires and Cables for Use in Handling Spaces, National Fire Protection Association, Quincy, Ma ssac hu se tts , 2015. 3-80

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  • NOfllltWEST 11£111CAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components NFPA 297, Guide on Principles and Practices for Communications S ys tems , National Fire Protection Association, Quincy , Massachusetts , 1995. NFPA 329, R ec ommend e d Practi ce for Handling Rel e as es of Flammable and Combustibl e Liquids and Gas es, National Fire Protection Association, Quincy , Massachusetts , 2015. NFPA 400 , Hazardous Mat e rials Code , National Fire Protection Association , Quincy , Massachusetts, 2013. NFPA 496 , Standard for Purged and Pr ess urized Enclosures for Electrical Equipment , National Fire Protection Association , Quincy , Massachusetts , 2013. NFPA 497 , R ec ommend ed Practice for the Classification of F lammabl e Liquids , Gases, or Vapors and of Ha z ardou s (Cla ss ifi e d) Lo c ation s for El ec trical Installation s in Ch e mi ca l Pro cess Areas, National Fire Protection Association, Quincy, Massachusetts, 2012. NFPA 704 , Standard Sy ste m for th e Id en tifi c ation of the Hazard s of Materials for Emergency R es pon se, National Fire Protection Association , Quincy , Massachusetts , 2012. NFPA 730 , Guide for Pr e mi ses Securi ty , National Fire Protection Association, Quincy , Massachusetts , 2014. NFP A 731, Standard for the In stallation of E l ec troni c Pr e mis es Security Systems , National Fire Protection Association, Quincy , Massachusetts , 2015. NFPA 780 , Standard for the Installation of Lightning Protection Systems, National Fire Protection Association , Quincy , Massachusetts , 2 014. NFPA 791 , R ec ommend ed Practic e and Proc e dur es for Un l abeled Electrical Equipment Evaluation , National Fire Protection Association , Quincy , Massachusetts , 20 14. NFPA 801, Standard for Fire Prot ec tion for Facilities Handling Radioactiv e Materia l s, National Fire Protection Association, Quincy , Massachusetts , 2014. NIOSH 2003-136, Guidance for Filtration and Air-Cleaning Systems to Prot ect Building Environments from Ai rborne Chemical, Biol ogic al , and Radiological Attacks, National Institute for Occupational Safety an d Health , Cincinnati , Ohio , 2003. NOAA, 20 17 , " NOAA Atlas 14 Point Precipitation Frequency Estimates: Mo," https://hdsc.nws.noaa.gov/hd sc/pfds/pfds _map_ cont.html?bkmrk=mo, Nationa l Oce anic and Atmospheric Administration, Silver Spring , Maryland , accessed 2017. NOAA Atlas 14 , Pre c ipitation-Fr e quen cy Atlas of th e United States, Volume 8, Ve r sion 2.0: Midwestern States , National Oceanic and Atmosp heri c Adminis trat ion , Silver Spring , Maryland , 2013. NRC, 2012 , Fina l Interim Staff Guidanc e Augmenting NUREG-1 53 7 , " Guid e lin es for Preparin g and R eview ing Applications for the Licensing of Non-Po we r R eactors," Part s 1 and 2,for Licensing Radioi so top e Produ cti on Faciliti es and Aqueous Homog e n eous Rea c tor s, Docket Number: NRC-2011-0135 , U.S. Nuclear Regulatory Commission, Washington , D.C., October 30 , 20 12. NUREG-0700 , Human-S yste m Int e rfa ce D es ign R ev i ew Guid e lin es, Rev. 2 , U.S. Nuclear Regul atory Commissio n , Office of Nuclear Regulatory Research , Washington , D.C., 2002. NUREG-0800, Standard R evie w Plan for th e Revi ew of Saf ety Anal ys i s Reports for Nuclear Pow e r Plan ts, L WR Edition, U.S. Nuclear Regulatory Commission , Office of Nuclear Material Safety and Safeguards, Washington , D.C., 1987. NUREG-1513 , Int eg rat ed Safety Analysis Guidance Do c um e nt, U.S. Nuclear Regulatory Commission , Office of Nuclear Material Safety and Safeguards , Washington , D.C., May 2001. 3-81

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  • DttAL ts011)f(S NWMl-2013-021 , Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components NUREG-1520 , Standard R ev i ew Plan f o r t h e R ev i ew of a Li ce n se A ppli c ati o n for a Fu el C y cl e Fa c ili ty, Rev. 1 , U.S. Nuclear Regulatory Commission , Office of Nuclear Material Safety and Safeguards , Washington , D.C., Ma y 2010. NUREG-1537 , Guid e lin es for Pr e paring and R e vi ew in g Appli c ation s for th e Li ce n s in g of Non-Po we r R e actor s -Format and Cont e nt , Part 1 , U.S. Nuclear Regulatory Commis s ion , Office of Nuclear Reactor Regulation , Wa s hington , D.C., February 1996. NUREG/CR-4604/PNL-5849 , S t ati st i c al M et hod s f o r N ucl e ar Mat e r i al Mana ge m e n t, Pacific Northwest Laboratory , Richland , Washington , December , 1988. NUREG/CR-6410 , N ucl e ar Fu e l C y cl e Fa c ili ty Acc id e nt Anal ys i s Handbook , U.S. Nuclear Regulatory Commi s sion , Wa s hington , D.C., 1998. NUREG/CR-6463 , R e vi e w Guid e lin es on Softwar e Lan g uag es for Use in Nucl e ar Po we r Plant Saf e ty S ys t e m s -Final R e p o rt , U.S. Nuclear Regulatory Commis s ion , Office of Nuclear Regulatory Research , Washington , D.C., 1996. NUREG/CR-6698 , Guid e for Validation o f N ucl e ar Criti c ali ty Saf ety Cal c ulational M e thodolo gy , U.S. Nuclear Regulatory Commission , Office of Nuclear Material Safety a nd Safeguard s, Washington , D.C., January 2001. NWMI-2013-043 , N WMI Radioi s otop e Produ c tion Fa c ili ty Stru c tu r al D es i g n B as i s, Rev. B , Northwest Medical Isotopes , Corvallis , Ore g on , 2015. NWMI-2015-LIST-003 , N WMI Radioi soto p e Produ c ti o n Fa c ili ty Ma s t e r Equipm e nt Li s t , Rev. A , Northwest Medical I s otopes, Corvallis , Oregon , 2015. NWMI-2015-SAFETY-011 , Ev aluation o f Na tural Ph e nom e non and Man-Mad e Eve nt s on Saf ety F e atur es and It e m s R e li e d o n for Saf ety, Rev. A , Northwest Medical Isotopes , Corvallis , Oregon , 2015. NWMI-2015-SDD-001 , RPF Fa c ili ty SDD , Rev. A , Northwest Medical Isotope s, C orvallis , Oregon , 2015. NWMI-DRD-2013-030 , N WMI Radioi so top e Produ c ti o n Facili ty D es i g n R e quir e m e nt s Do c um e nt , Rev. B , Northwe s t Medical Isotopes , Corvalli s , Oregon , 2015. Open-File Report 2008-1128 , Do c umen ta tion for th e 200 8 Updat e of th e Unit e d S ta t es National S e ismi c Ha z ard Maps , U.S. Geological Survey , Washington , D.C., 2008. Regulatory Guide 1.29 , S e i s mi c D es ign C la ss ifi c ation , Rev. 3 , U.S. Nuclear Regulatory Commission , Washington , D.C., September 1978. Regulatory Guide 1.53 , A ppli c ation of th e Sin g l e-Failur e Crit e rion to Saf ety S ys t e m s, Rev. 2 , U.S. Nuclear Regulatory C ommi s sion , Wa s hington , D.C., November 2003 (R2011). Regulatory Guide 1.60 , D es i g n R es pon se Sp ec tra for S e i s mi c D es i g n of N ucl e ar P owe r Plant s, Re v. 2 , U.S. Nuclear Regulatory Commission , Washington , D.C., July 2014. Regulatory Guide 1.61 , Damping Valu es for S e i s mi c D es ign of Nucl e ar Po we r Pl a nt s , Rev. 1 , U.S. Nuclear Regulatory Commission , Washington , D.C., March 2007 (R2015). Regulatory Guide 1.76 , D es ign-Ba s i s T o rnado and Torn a do Mis s il e s for Nucl e ar P o w e r Plant s, Rev. 1 , U.S. Nuclear Regulatory Commission , Wa s hington , D.C., March 2007. Regulatory Guide 1.92 , Co mbinin g Mod a l R es pon ses and Spatial C o mpon e nt s in S e i s mi c R es pon se Anal ys i s, Rev. 2 , U.S. Nuclear Regulatory Commission , Washington , D.C., July 2006. 3-82

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  • NOllTltWUT MEOICAl. ISOTOPES NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants , Rev. 4 , U.S. Nuclear Regulatory Commission , Washington, D.C., June 2006 (R2013). Regulatory Guide 1.100 , S e ismic Qualification of El ec trical and Active Mechanical Equipment and Fun c tional Qualification of Activ e Mechanical Equipment for Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission , Washington, D.C., September 2009. Regulatory Guide 1.102 , Flood Protection/or Nuclear Power Plants, Rev. 1 , U.S. Nuclear Regulatory Commission , Office of Standards Development , Washington , D.C., September 1976. Regulatory Guide 1.122 , D e v e lopm e nt of Floor De s ign R esponse Spectra for Seismi c Design of Supported Equipment or Components, U.S. Nuclear Regulatory Commission , Office of Standards Development, Washington , D.C., February 1978. Regulatory Guide 1.152 , Criteria for Use of Computers in Saf ety S yste ms of Nuclear Power Plant s, Re v. 3 , U.S. Nuclear Regulatory Commission , Washington , D.C., July 2011. Regulatory Guide 1.166 , Pr e-Earthquak e Planning and Immediat e Nuclear Pow e r Plant Op e rator Post Earthquak e Actions , U.S. Nuclear Regulatory Commission , Washington , D.C., March 1997. Regulatory Guide 1.167 , Re s tart of a Nuclear Power Plant Shut down by a Seismi c Event , U.S. Nuclear Regulatory Commission, Washington , D.C., March 1997. Regulatory Guide 1.208 , P e rformance Based Approa c h to Defin e the Site-Specific Earthquake Ground Motion , U.S. Nuclear Regulatory Commission, Washington , D.C., March 2007. Regulatory Guide 3.3 , Quality Assuranc e Program R e quirem e nts for Fuel R e pro cessi ng Plant s and for Plutonium Proc ess ing and Fu e l Fabrication Plants , Re v. 1 , U.S. Nuclear Regulatory Commission , Washington, D.C., March 1974 (R2013). Regulatory Guide 3.6 , Content of Technical Specification for Fu e l Reprocessing Plants, U.S. Nuclear Regulatory Commission , Washington , D.C., April 1973 (R2013). Regulatory Guide 3.10 , Liquid Waste Tr e atment System Design Guid e for Plutonium Processing and Fuel Fabrication Plants , U.S. Nuclear Regulatory Commission, Washington , D.C., June 1973 (R2013). Regulatory Guide 3 .18 , Confinement Barri e rs and S ys tems for Fuel R e proces s ing Plant s, U. S Nuclear Regulatory Commission, Washington , D.C., February 1974 (R2013). Regulatory Guide 3.20, Pro cess Offga s S yste ms for Fu el R e proc ess ing Plants , U.S. Nuclear Regulatory Commission , Washington , D.C., February 1974 (R2 013). Regulatory Guide 3. 71 , Nuclear Criticali ty Safety Standards for Fu e l s and Mat e rials Faciliti es, Rev. 2 , U.S. Nuclear Regul atory Commission , Washington , D.C., December 2010. Regulatory Guide 5.7, Entry/Exit Control for Protected Areas, Vital Areas, and Material Acc ess Areas , Rev. 1 , U.S. Nuclear Regulatory Commission , Washington , D.C., May 1980 (R2010). Regulatory Guide 5 .12 , G e n e ral Use of Locks in the Prot ec tion and Contro l of Facilities and Special Nuclear Material s, U.S. Nuclear Regulatory Commission, Washington , D.C., November 1973 (R2010). Regulatory Guide 5.27 , Spe c ial Nuclear Material Doorway Monitor s, U.S. Nuclear Regulatory Commission , Washington , D.C., June 1974. Regulatory Guide 5.44 , Perimeter Intru sion Alarm Systems, Rev. 3, U.S. Nuclear Regulatory Commission , Washington , D.C., October 1997 (R2010). 3-83

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NWMl-2013-021, Rev. 1 Chapter 3.0 -Design of Structures, Systems and Components Regulatory Guide 5.57, Shipping and Receiving Control of Strategic Special Nuclear Material , U.S. Nuclear Regulatory Commission, Washington , D.C., June 1980. Regulatory Guide 5.65, Vital Area Access Control, Protection of Physi c al Security Equipment, and Key and Lock Controls , U.S. Nuclear Regulatory Commission, Washington , D.C., September 1986 (R2010). Regulatory Guide 5.71, Cyber Security Programs for Nuclear Facilitie s, U.S. Nuclear Regulatory Commission , Washington, D.C., 2010. SMACNA 1143 , HVAC Air Duct Leakage Test , Sheet Metal and Air Conditioning Contractors' National Association , Chantilly , Virginia , 1985. SMACNA 1520, Round Indu s trial Duct Construction Standard, Sheet Metal and Air Conditioning Contractors

' National Association, Chantilly, Virginia, 1999. SMACNA 1922 , Rectangular Industrial Duct Constru c tion Standard , Sheet Metal and Air Conditioning Contractors' National Association, Chantilly, Virginia , 2004. SMACNA 1966, HVA C Duct Construction Standard -M e tal and Flexible, Sheet Metal and Air Conditioning Contractors' National Association, Chantilly , Virginia , 2006. SMACNA-2006 , HVAC S yste ms Duct Design , Sheet Metal and Air Conditioning Contractors' National Association, Chantilly , Virginia, 2006. SNT-TC-1 A, Re co mmend ed Practic e No. SNT-TC-1 A: P e rsonnel Qualifi c ation and Certification in Nondestructive Te st ing , American Society for Nondestructive Testing , Columbus , Ohio , 20 11. Technical Paper No. 40 , Rainfall Frequency At la s of the United State s for Durations from 30 Minutes to 24 Hours and Return P e riods from I to JOO Y e ar s, Weather Bureau , U.S. Department of Commerce, Washington , D.C. 1963. Terracon, 2011 a , Phas e I Environmental Sit e Assessm e nt Discovery Ridg e Lots 2, 5 , 6 , 7, 8, 9 , I 0 , I!, I 2, 13 , I 4 , 15 , I 6 , I 7, and I 8, Terracon Consultants , Inc., prepared for University of Missouri and Trabue , Hansen & Hinshaw , Inc., Terracon Project No. 09117701 , March 23, 2011. Terracon , 2011 b , Pr e limina ry G e ot ec hnical Engineering Report Di scovery Ridge-Ce rtified Site Program Lots 2, 5 , 6, 7, 8, 9 , I 0 , !!, 12 , 13 , 14 , I 5 , 16 , I 7, and 1 8, Terracon Consultants , Inc., prepared for University of Missouri and Trabue, Hansen & Hinshaw, Inc., Terracon Project No. 09105094.1 , February 11 , 2011. UL 1 8 1 , Standard for Facto ry-Mad e Air Duct s and Connectors, Underwriters Laboratories , Washington , D.C., 2013. UL 499 , Standard for Electric H e ating Appliances, Underwriters Laboratories , Washington , D.C., 2014. UL 555 , Standard for Fir e Dampers, Underwriters Laboratories , Washington , D.C., 2006. UL 586 , Standard for High Effi c ienc y, Parti c ulate , Air Filter Units , Underwriters Laboratories , Washington, D.C., 2009. UL 900, Standard for Air Filter Units, Underwriters Laboratories, Washington , D.C., 2004. UL 1995 , Heating and Cooling Equipment , Underwriters Laboratories , Washington, D.C., 2011. USGS, "2008 U.S. Geological Survey National Seismic Hazard Maps ," U.S. Geological Survey, Rolla , Missouri , 2008. 3-84

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  • Chapter 6.0 -Engineered Safety Features Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021, Rev. 1 June 2017 Northwest Medical Isotopes , LLC 815 NW gth Ave , Suite 256 Corvallis, OR 97330 This page intentionally left blank.

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  • NORTHWEST MEDtCAl ISOTOHS NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Chapter 6.0 -Engineered Safety Features Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 6.0 -Engineered Safety Features Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

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.; ... ; .. NWMI ..*... ..* .... ........ *.* " "NDlrTHWHTMBHCAllSOTOPH Rev Date 0 6/29/2015 1 6/26/2017 NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features REVISIO N HISTORY Reason for Revision Revised By Initial Application Not required Incorporate changes based on responses to C. Haass NRC Requests for Additional I nformation

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  • NOITHWBT IWMCAl lSOTOPU NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features CONTENTS 6.0 ENGINEERED SAFETY FEATURES ......................

................................................................... 6-1 6.1 Summary Description

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............... 6-1 6.2 Detailed Descriptions

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6-5 6.2.1 Confinement

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6-5 6.2.1.1 Confinement System ............

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.................................... 6-7 6.2. I .2 Accidents Mitigated

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.... 6-1 I 6.2. I .3 Functional Requirements

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.................. 6-1 I 6.2.1.4 Confinement Components

................................................................. 6-1 I 6.2.1.5 Test Requirements

............................................................................. 6-12 6.2.1.6 Design Basis ......................................................................................

6-13 6.2.1. 7 Derived Confinement Items R elied on for Safety ............................. 6-13 6.2. I .8 Dis solver Off gas Systems ................................................................. 6-23 6.2.1.9 Exhaust System ...................................................................

.............. 6-26 6.2.1.10 Effluent Monitoring System ....................................

.......................... 6-26 6.2.1.11 Radioactive Release Monitoring

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..................... 6-26 6.2.1.12 Confinement Sy s tem Mitigation Effects .................................

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6-26 6.2.2 Containment

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6-27 6.2.3 Emergency Cooling System ............................................................................... 6-27 6.3 Nuclear Critica lity Safety in the Radioisotope Production Facility .................................

6-28 6.3.1 Criticality Safety Controls .................................................................................. 6-36 6.3.1.1 Preliminary Criticality Safety Evaluat ion s ..............

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6-36 6.3.1.2 Derived Nuclear Criticality Safety Items Relied on for Safety ......... 6-59 6.3.2 Surveillance R equirements

................................................................................. 6-71 6.3.3 Technical Specifications

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................................................................. 6-71 6.4 Reference s ............................................................

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...................................... 6-72 6-i

.; ... ; ... NWMI ...... ..* .... ........ *.* * .. ." . NOUHWf.ST M£DfCAI. ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Figure 6-1. Figure 6-2. Figure 6-3. Figure 6-4. Figure 6-5. Figure 6-6. Figure 6-7. Table 6-1. Table 6-2. Table 6-3. Table 6-4. Table 6-5. Table 6-6. Table 6-7. Table 6-8. Table 6-9. Table 6-10. Table 6-11. Table 6-12. Table 6-13. FIGURES Simplified Zone I Ventilation Schematic

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........................................ 6-6 Ground Le v el Confinement Boundary ..............................

............................................... 6-8 Mechanical Level Confinement Boundary ...........

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6-9 Lower Level Confinement Boundary ........................................................

.................... 6-10 Di s sol v er Off gas System Engineered Safety Features ............................................

....... 6-14 Dissolver Off gas Hot Cell Equipment Location ...................................

......................... 6-15 Proposed Location of Double-Wall Piping (Example)

......................

............................ 6-21 TABLES Summary of Confinement Engineered Safety Feature s (2 pages) ...................................

6-2 Summary of Criticality E ngineered Safety Feature s (2 pages) ........................................ 6-3 Confinement System Safety Function s .................................................

.....................

...... 6-7 Area of Applicability Summary ..................................................................................... 6-37 Controlled Nuclear Criticality Safety Parameters

.......................................................... 6-38 [Proprietary Information]

Double-Contingency Controls .............................................. 6-39 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-40 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-41 [Proprietary Information]

Double-Contingency Controls (8 page s) .............................. 6-43 [Proprietary Information]

Double-Contingency Controls (2 page s) .............................. 6-5 I [Proprietary Information]

Double-Contingency Controls (3 page s) .............................. 6-53 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-56 [Proprietary Information]

Double-Contingency Controls (2 pages) .............................. 6-57 6-ii

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  • NCMn"tfWHT Mf.DtCAl ISOTOPU TERMS NWMl-2013-021 , Rev. 1 C h apter 6.0 -Engineered Safety Feature s A cron y ms a nd A bbr ev i at i o n s 99 Mo 23s u ADUN AEC ANECF ANS ANSI CAAS CFR CSE DBE HEGA HEPA HVAC IEU IX IROFS Kr LEU MCNP Mo N0 2 NO x NRC NWMI PEC PHA RPF SSC SPL UN [Proprie t ary Information]

USL Xe mo l ybdenum-99 uranium-235 acid-deficient uranium nitrate active engineered control average neutron energy causing fission American Nuclear Society American National Standards Institute critica l ity accident a l arm system Code of Federal Regu l ations criticality safety eva l uation design basis eart h q u ake high-efficiency gas adsorber high-efficiency particulate air heat i ng , ventilation , and air conditioning intermediate

-enric h ed uranium ion exchange item re l ied on for safety kryp t on low-enriched uranium Monte-Car l o N-Particle mo l ybdenum nitrogen dioxide nitrogen oxide U.S. N u clear Regu l atory Commission Northwest Medica l Isotopes , LLC passive engineered control preliminary hazards analysis radioisotope pro du ction facility structures , systems , and components sing l e parameter l imit urani u m nitride [Proprietary Information]

upper subcritica l limits xenon 6-iii

..... ... NWMI ...... ..* **: .*.* .. *.*.* . ' NCMIT H W'EST MEDtCAl ISOTOfl(S Units o c o p atm cm cm 3 ft ft 2 ft 3 g hr m. L m m 2 mm mL mol rad wt% yr degrees Celsius degrees Fahrenheit atmosphere centimeter cubic centimeter feet square feet cubic feet gram hour inch liter meter square meter minute milliliter mole radiation absorbed dose weight percent year 6-iv NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features

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  • 0 N0<<11IWEST MEDICAL lSOTOPO NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.0 ENGINEERED SAFETY FEATURES 6.1

SUMMARY

DESCRIPTION Engineered safety features are active or passive features designed to mitigate the consequences of accidents and to keep radiological exposures to workers , the public , and environment within acceptable values. The engineered safety features associated with confinement of the process radionuclides and hazardous chemicals for the Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF) are summarized in Table 6-1 , including the accidents mitigated; structures , systems, and components (SSC) used to provide the engineered safety features; and references to subsequent sections providing a more detailed engineered safety feature description. Confinement is a general engineered safety feature that is credited as being in place as part of the preliminary hazards analysis (PHA) described in Chapter 13.0, " Accident Analysis." Additional items relied on for safety (IROFS) associated with the confinement system were derived from the accident analyses in Chapter 13 .0. The derived IROFS are also listed in Table 6-1 , with reference to more detailed descriptions in Section 6.2.1. The current design approach does not anticipate requiring containment or an emergency cooling system as engineered safety features , as discussed in Sections 6.2.2 and 6.2.3. Nuclear criticality safety is discussed in Section 6.3. Criticality safety controls are described in Section 6.3.1. The currently defined criticality safety controls are derived from a combination of preliminary criticality safety evaluations (CSE) and accident analyses , which are described in Chapter 13.0. The criticality safety analyses produce a set offeatures needed to satisfy the contingency requirements for nuclear criticality control. These features are evaluated by major systems within the RPF and listed by major system in Section 6.3.1.1 , Table 6-6 through Table 6-13. The accident analyses in Chapter 13.0 identify IROFS for the prevention of nuclear criticality , which are summarized in Table 6-2 , with reference to more detailed descriptions in Section 6.3.1.2. 6-1

.... ;. NWMI ...*.. ..* ... ........ *.* ' *,* ' NOllTifWHT MEDtCAl. tsOTOPES NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-1. Summary of Confinement Engineered Safety Features (2 pages) Detailed Engineered safety SSCs providing engineered description feature IROFS Accident(s) mitigated safety features section C onfinement . Equipment . Confin e ment enclo s ure s 6.2.1.1 includes: m a lfunction an d/o r includin g p e n e tration s e a l s throu g h Hot cell liquid RS-01 m ai nten a n ce . Zon e I ex h a ust ve nt i l a ti o n 6.2.1.6 . confinem e nt . H aza rdous c h e mi ca l sys t e m , includin g du c tin g , b o undar y s pill s filter s , a nd e xh a u s t s t a ck Hot ce ll RS-0 3 . Zone I inl e t ve ntil a ti o n sy st e m , . sec ondary includin g ductin g, filt e r s , and c onfinement bubble-ti g ht i s olati on d a mper s boundary . Ventil a tion control sys t e m Hot c e ll s hieldin g R S-04 . Second a r y iodine remo va l bed . boundar y . Berm s Confinement IROFS Derived from Accident Analyses and Potential Technical Specifications Prim a ry off g a s reli e f R S-0 9 Di ss ol ve r off gas fa ilur e . Pr ess ur e r e li ef d ev i ce 6.2.1.7.1 sys t e m durin g di ss oluti o n . Pr ess ur e r e li ef t a nk op era t io n Active radiation RS-JO Transfer of high-dose Radiation monitoring and isolation 6.2.1.7.2 monitoring and process liquid outside the system for low-dose liquid isolation of low-hot cell shielding transfers dose waste transfer boundary Cas k lo c al R S-1 3 T a r ge t cl a dding l ea k age Local captur e v e ntil a ti o n s y s t e m 6.2.1.7.3 ve ntil a tion durin g during s hipm e nt ov e r c l os ur e lid durin g lid removal clo s ure lid r e m ov al a nd dockin g prep a ration s Cask docking port RS-15 Cask not engaged in cask Sensor system controlling cask 6.2.1.7.4 enabler docking port prior to docking port door operation opening docking port door Proces s ve ss el FS-0 3 SS C damage du e t o B a ckup bottled nitro ge n g a s 6.2. 1.7.5 e m e r g enc y pur ge hydro g en d e tla gra ti o n or s uppl y sys t e m deton a tion Irradiated target FS-04 Dislodging the target . Cask lifting fixture design that 6.2.1.7.6 cask lifting fixture cask shield plug while prevents cask tipping workers present during . Cask lifting fixture design that target unloading prevents lift from toppling activities during a seismic event 6-2

.: . .-.;* .. NWMI ..**.. ..* .... .. .. . ...*.. * * *. . NORTHWEn MEDICAL ISOTOP£S NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-1. Summary of Confinement Engineered Safety Features (2 pages) Detailed Engineered safety SSCs providing engineered description feature IROFS Accident(s) mitigated safety features section E xhau s t s t a ck h e i g ht F S-05 . E quipm e nt . Zo n e I ex h a ust s t ac k 6.2.1.7.7 m a lfun c tion r es ultin g in liquid s pill o r s p ray . Ca rbon bed fire Double-wall piping CS-09 Solution spill in facility Double-wall piping for selected 6.2.1.7.7 area where spill transfer lines containment berm is neither practical nor desirable for personnel chemical protection purposes B a ckflo w CS-1 8 Hi g h w ork er ex p os u re B ac k flow pr eve nti on d ev ic es 6.2.1.7.9 pr eve nti on d ev i ces fr om b ac kfl o w of h ig h-l ocate d o n pro cess li nes c ro ss in g Safe geom e tr y d ay CS-1 9 do se so lution th e h ot ce ll s hi e ldin g b o undar y t a nk s Dissolver offgas . Potential limiting Dissolver offgas iodine removal 6.2.1.8 iodine removal unit" control for operation units (DS-SB-600A/B

/C) . Primary iodine control system during normal operation Di ss ol ve r offgas . P o t e nti a l limitin g Di sso l ve r offgas p r im ary a d s orb e r 6.2.1.8.2 primary a d so rb er" co ntrol for op e rati o n unit s (D S-S B-62 0 A/B/C) . Pri m ary nobl e g as co ntrol sys t e m durin g n o rm a l op e ration Dissolver offgas . Potential limiting . Dissolver offgas vacuum 6.2.1.8.3 vacuum receiver or control for operation receiver tanks (DS-TK-700A/B) vacuum pump" . Motive force for . Dissolver offgas vacuum pumps dissolver offgas (DS-P-710A/B)

  • Exa mpl es of ca ndid a t e tec hni ca l s p ec ifi cat i o n ra t h e r th a n e n g in ee r e d safety fea tur e. I R OFS ite m re li e d o n fo r sa f e t y. SSC = struc tur es, sys t ems , a nd co mp o n e n ts. Table 6-2. Summary of Criticality Engineered Safety Features (2 pages) Engineered safety feature Int e ra c tion c ontr o l s p a cin g pro v id e d b y p ass iv e l y de s i g n e d fi x tur es and w ork s t a ti o n pl a c e m e nt Pencil tank , vessel , or piping safe geometry confinement using the diameter of tanks , vessels , or piping P e ncil t a nk ge om e tr y control on fix e d int e raction s p a cing o f indi v id ua l ta nk s SSC features providing engineered safety features CS-04 D e fin es s p a cing b e t wee n SSC comp o n e nt s u s in g ge om e t ry t o prevent nucl ea r c ritic a lit y CS-06 Defines dimensions of SSCs using geometry to prevent nuclear criticality CS-07 D e fin es s p a cing b e tw ee n diff e rent SSCs u s in g ge om etry to pr eve nt nucl ea r c ritic a li ty 6-3 -*
  • 6.3.1.2.1 6.3.1.2.2 6.3.1.2.3

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  • _-NORTHWEST llfDICAL ISOTOH:S NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-2. Summary of Criticality Engineered Safety Features (2 pages) Engineered safety feature Floor and sump geometry control on slab depth , and sump diameter or depth for floor dikes D o uble-w a ll pipin g Closed safe-geometry heating or cooling loop with monitoring and alarm S impl e o ve rfl ow t o n o rm a lly e mpt y safe-ge om e t ry t a nk w ith l eve l a l ar m Condensing pot or seal pot in ventilation vent line Si mpl e o ve rfl ow t o n o rm a ll y em pt y safe ge om e tr y fl o or w ith l eve l ala rm in t h e hot ce ll c ont a inm e nt b o u n d a r y Active discharge monitoring and isolation I nd e p e nd e nt ac ti ve di s ch a r ge m o n i toring a nd i so l a tion Backflow prevention device Safe ge om e tr y d ay t a nk s Evaporator or concentrator condensate monitoring Process in g co m po n e nt s a fe vo lum e co nfin e m e nt Closed heating or cooling loop with monitoring and alarm I ROFS i t em r e li e d o n fo r safety. SSC features providing engineered safety features CS-08 Defines sump geometry and dimension s for SSCs using geometry to prevent nuclear criticality CS-09 D e fin es t ra n sfe r lin e l ea k co nfin e m e nt in lo ca ti o ns w h e re s ump s und e r piping a r e n e ith e r feas ibl e n or d es irabl e -* . 6.3.1.2.4 6.3.1.2.5 CS-10 Closed-loop heat tran s fer fluid systems to 6.3.1.2.6 prevent nuclear criticality or transfer ofhigh-dose material across shielding boundary in the event of a leak into the heat transfer fluid CS-11 O ve rfl ow to pr eve nt nucl ea r c ritic a lit y fro m 6.3 .1.2. 7 fi ss il e so luti o n e nt e r i n g n o n-ge om etrica ll y favora bl e ve ntil a ti o n e quipm e nt CS-12 Seal pots to prevent nuclear criticality from 6.3.1.2.8 fissile solution entering non-geometrically favorable ventilation equipment CS-1 3 O ve rfl ow to pr eve nt nu clear cr itic a l i t y fro m 6.3.1.2.9 fi ss il e so lu t i o n e nt er in g n o n-geo m e tr ica ll y fa vo rabl e ve ntil a ti o n e quipm e nt CS-I 4 Information to be provided in the Operat i ng 6.3 .1.2. l 0 License Application CS-1 5 In fo rm a ti o n w ill b e p rov id e d in th e Op era tin g 6.3 .1.2.11 Lic e n se A ppli ca tion CS-18 Backflow prevention to preclude fissile or high 6.3.1.2.12 dose solution from cros s ing shielding boundary to non-geometrically favorable chemical supply tanks and prevent nuclear criticality CS-1 9 A lt e rnat e b ac kflo w pr eve n tion d ev i ce 6.3.1.2. 1 3 CS-20 Prevent nuclear criticality from high-volume 6.3.1.2.14 transfer to non-geometrically favorable vessels in solutions with normally low fissile component concentrations CS-26 D e fin es vo lum e of S SCs to pr eve nt nu clear c ri ti ca li t y CS-27 Closed-loop, high-volume heat transfer fluid systems to prevent nuclear criticality or transfer of high-dose material across shielding boundary in the event of a leak into the heat transfer fluid with normally low fissile component concentrations SSC = s tru c tur es , syste m s, a nd com p o n e nt s. 6-4 6.3.1.2.15 6.3.1.2.16

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  • UEDICAl lSOTOPH NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2 DETAILED DESCRIPTIONS The PHA used to identify accidents in Chapter 13.0 , Section 13.1.3 , assumed the following known and credited safety features , or IROFS, are in place for normal operations:
  • *
  • Hot cell shielding boundary, credited for shielding workers and the public from direct exposure to radiation (a normal hazard of the operation)

Hot cell confinement boundaries , credited for confining the fissile and high-dose solids , liquids , and gases , and controlling gaseous releases to the environment Administrative and passive design features on uranium batch , volume , geometry , and interaction controls on the activities, credited for maintaining normal operations involving the handling of fissile material subcritical (the PHA identified initiators for abnormal operations that require further evaluation for IROFS satisfying the double-contingency principle)

This section provides detailed descriptions of the engineered safety features identified by the accident analyses shown in Chapter 13.0. 6.2.1 Confinement The PHA was based on a definition for confinement , a s follows: Confinement

-An enclosure of the facility (e.g., the hot cell area in the RPF) that is designed to limit the exchange of effluents between the enclosure and its external environment to controlled or defined pathways. A confinement should include the capability to maintain sufficient internal negative pressure to ensure inleakage (i.e., prevent uncontrolled leakage outside the confined area), but need not be capable of supporting positive internal pressure or significantly shielding the external environment from internal sources of direct radiation.

Air movement in a confinement area could be integrated into the heating , ventilation , and air conditioning (HVAC) system s, including exhaust stacks or vents to the externa l environment , filters , blowers , and dampers (ANSl/ANS-15.1 , The D eve lopm e nt of T ec hni c al Sp ec ifi c ation s for R ese ar c h R e a c tor s). Confinement describes the low-leakage boundary surrounding radioactive or hazardous chemical materials released during an accident to facility regions surrounding the physical process equipment containing process materials. The confinement systems localize releases of radioactive or hazardous materials to controlled areas and mitigate the consequences of accidents. The principal design and safety objective of the confinement system is to protect on-site workers , the public , and environment.

Personnel protection control features (e.g., adequate shielding and ventilation control) will minimize hazards normally associated with radioactive or chemical materials. The second design objective is to minimize the reliance on administrative or complex active engineering contro l s and provide a confinement system that is as s imple and fail-safe as reasonably possible. This subsection describes the confinement systems for the RPF. The RPF confinement areas will consist of hot cell and glovebox enclosures housing process operations , tanks , and piping. Confinement will be provided by a combination of the enclosure boundarie s (e.g., walls , floor , and ceiling), enclosure ventilation , and ventilation control system. The enclosure boundaries will restrict bulk quantities of process materials , potentially present in solid or liquid forms , to the confinement and limit in-leakage of gaseous components controlled by the ventilation system. The ventilation and ventilation control systems will restrict the gaseous components (including gas phase components and solid/liquid dispersions) to the confinement.

Figure 6-1 provides a simplified schematic of the confinement ventilation system , which is described in more detail as the Zone I ventilation system in Chapter 9.0 , "Auxiliary Systems." 6-5

..... NWMI ...... ..* *.. ........ *.* * *.

  • 0 NOmtWtn MEDICAL ISOTOPfS NWMl-2013-021 , Rev. 1 Chapter 6.0 -Eng i neered Safety Features [P ro p r i e t ary In fo rm a ti o n] Source: Figure 2-5 ofNWMI-20 I 5-SDD-013 , System D es i gn D es c ription for Ventilation, Rev. A , Northwest Medical I s otopes , LLC, Corva lli s, Orego n , Ma r c h 2015. Figure 6-1. Simplified Zone I V entilation S chematic 6-6

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  • NOmlWEST llEIHCAl tSOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The enclosure bounda ry of the hot cells will also function as biolo gica l s hieldin g for operating personn el. Shielding functions of the hot cells are discussed in C h a pter 4.0 , " Radioisotop e Production Facility D esc ription." Ha za rdous chemical confinement will be provided by berm s located within the RPF to confine spilled material to the vicinity where a spill may originate. 6.2.1.1 Confinement System Co nfinem ent system e n closure s tru c ture s , ve ntilation ducting , i so l ation damper s , and Zone I exha ust filter trains are de signate d as IROFS. Table 6-3 pro v id es a description of the system compo nent safety functions. Figure 6-2 , Figure 6-3 , and Figure 6-4 indi cate the ge neral lo cat ion of confi nement s tructure boundaries to the facility gro und le vel , mechanical le ve l , a nd lo wer le vel l ayo ut s , respectively.

The co nfinement sys tem i s an e ngineered safety feature that perform s the fun ct ion s identified b y IROFS RS-01, RS-03 , and RS-04 i n C hapter 1 3.0. Table 6-3. Confinement System Safety Functions System, structure, component Zo n e I enclosure inlet i so l ation d ampers and ducting l eading from isolation dampers to enc l osu r es Zone I enclosure exhaust ducting leading from enclosures to the exhaust stack, filters, and ex haust stack Process vesse l ve nt ex h a u s t ducting l ea din g from process vesse l s to Zone I exha u st plenum Ventilation control system Secon d ary iodine removal bed Hot cells, tank vaults, and glovebox enclosure structures IROFS = it e m re li ed o n for s afety. Description Provide confi n eme nt isolation at Zone I/Zo n e II e nclosur e boundaries Provides confinement to the confinement exhaust boundary Pro v id es confineme nt to the co nfin ement ex h a u st boundary Provides stack monitoring and interlocks to monitor di sc harge and signal changing on service tilter trains during normal and abnormal operation Mitigate s a re l ease of the iod in e inventor y in the dissolver offgas treatment system Provide solid, liquid , gas confinement 6-7 Classification IROFS IROFS IROFS IROFS IROFS IROFS

...... ;* .. NWMI ..**.. ... .... ......... *.* , "NORTHWEnMBUCAll$0TOPH NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

So ur ce: Figur e 2-1 of NWMJ-2 015-S DD-01 3, System D e s i gn D e s c rip ti on for Ve n ti l a ti on , R ev. A , No rth west M e di cal I so top es, LLC , Corv alli s , O r ego n , March 20 1 5. Figure 6-2. Ground Level Confinement Boundary 6-8

...... ;* .. NWMI ::.**.-.*.* ........ *.* '. *. , NORTHWtn MllHCAl tSOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

Source: Figure 2-2 ofNWMl-2015-SDD-O 13 , System D es i g n D e scr iption for Ventilatio n , Rev. A , Northwest Medical I so t opes , LLC, Corva lli s, Oregon , March 2015. Figure 6-3. Mechanical Level Confinement Boundary 6-9

...... ; .. NWMI ..*... ..* **: ............ . "NomfWESTMEDtCAllSOTDPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

Source: Figure 2-3 ofNWMI-2015-SDD-013, System De s ign D escri ption for Ventilation, Rev. A, Nort hwest Medical Isotope s, LLC , Corva lli s, Oregon , March 2015. Figure 6-4. Lower Level Confinement Boundary During normal operation, passive confinement is provided by the contiguous boundary between the hazardous materials and the surrounding environment and is credited with confining the hazards generated as a result of accident scenarios.

The boundary includes the enclosure st ructures and extension of the structures through the Zone I ventilation components.

The intent of the pa ssive boundary is to confine hazardous materials while also preventing disturbance of the ha z ardou s material inventory by ex ternal energy s ources. This passive confinement boundary extends from the isolation valve downstream of the intake high-efficienc y particulate air (HEPA) filter to the exhaust s tack. An event that results in a release of process material to a confinement enclosure wi ll be confined by the enclosure structural components.

Each process line that connects with vessels located outside of a confinement boundary with vesse ls located inside a confinement boundary will be provided with backflow prevention de v ices to prevent releases of gaseous or liquid material.

The backflo w prevention devices on piping penetrating the confinement boundary are designed as passive devices and will be located as near as practical to the confinement boundary or take a position that provide s greater safety on loss of actuating power. The consequences of an uncontrolled release within a confinement enclosure, and the off-site consequences of releasing fission products through the ventilation system, will be mitigated by use of an active component in the form of bubble-tight isolation dampers as IROFS on the inlet ventilation ducting to each enclosure.

This engineered safety feature reduce s the ducting to the confinement v olume that needs to remain intact to achieve enclosure confinement.

The dampers will close automatically (fail-closed) on lo ss of power , and the ve ntilation system will automatically be placed into the passive ventilation operating mode. Overall performance a ssurance of the active confinement components will be achieved through factory testing and in-place testing. Duct and housing leak tests will be performed in accordance with minimum acceptance criteria, as specified in ASME AG-I , Cod e o n N ucl ear Ai r and Ga s Tr e atment. Specific owner requirements with respect to acceptable leak rates will be based on the safety analysis.

6-10

..... ; .. *NWMI ::.**.*.* . . *.******** * *. * ! . NOmlWEST MEDICAL lSOTOrU NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Berms will employ a passive confinement methodology.

Passive confinement will be achieved through a continuous boundary between the hazardous materials and the surrounding area. In the event of an accidental release, the hazardous liquid will be confined to limit the exposed surface area of the liquid. 6.2.1.2 Accidents Mitigated The hot cell confinement system and shielding boundary are credited as being in place by the accident analysis in Chapter 13.0 , Section 13.1.3. l. Accidents mitigated consist of equipment malfunction events that result in the release of radioactive material or hazardous chemicals to a confinement enclosure.

The confinement system is also credited with mitigating the impact of a non-specific initiating event resulting in release of the iodine inventory in the dissolver off gas treatment system. 6.2.1.3 Functional Requirements Functional requirements of the confinement structural components include: *

  • Capturing and containing liquid or solid releases to prevent the material from exiting the boundary and causing high dose to a worker or member of the public or producing significant environment contamination Preventing spills or sprays of radioactive solution that are acidic or caustic from causing adverse exposure to personnel through direct contact with skin , eyes, and mucus membranes where the combination of chemical exposure and radiologica l contamination would lead to serious injury and long-lasting effects Functional requirements of the confinement ventilation components include: * *
  • Providing negative air pressure in the hot cell (Zone I) relative to lo wer zones outside of the hot cell using exhaust fans equipped with HEPA filters and high-efficiency gas adsorbers (HEGA) to reduce the release of radionuclides (both particulate and gaseous) outside the primary confinement boundary to below Title I 0 , Cod e of F e deral Regulations , Part 20, "Standards for Protection Against Radiation" (10 CFR 20) release limits during normal and abnormal operations.

Mitigating high-dose radionuclide releases to maintain exposure to acceptable levels to workers and the public in a highly reliable and available manner. The hot cell secondary confinement boundary will perform this function using a system of passive and active engineered features to ensure a high level of reliabi li ty and availability.

Removing iodine isotopes present in the process vessel vent under accident conditions to comply with 10 CFR 70.61, " Performance Requirements," for an intermediate consequence release. Berms confining potential hazardous chemical spills are designed to hold the entire contents of the container in the event the container fails. 6.2.1.4 Confinement Components The following components are associated with the confinement barriers of the hot cells , tank vaults, and gloveboxes.

The specific materials, construction , installation, and operating requirements of these components are evaluated based on the safety analysis. 6-11

.; ... NWMI ...*.. ..* ... ........... * * * !'

  • NMTNWEn MElHCAl tSOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Confinement structural components include the following. * * * *
  • Sealed flooring will provide multiple layers of protection from release to the environment.

Diked areas will contain specific releases.

Sumps of appropriate design will be provided with remote operated pumps to mitigate liquid spills by capturing the liquid in appropriate geometry tanks. In the molybdenum-99 (99 Mo) purification clean room, smaller confinement catch basins will be provided under points of credible spill potential in addition to the sealed floor. Entryway doors into a designated liquid confinement area will be sealed against credible liquid leaks to outside the boundary.

Piping penetrations and air ducts will be located to minimize the potential for liquid leaks across the confinement boundary.

Ventilation system components that are credited include the following. * * * * * *

  • Zone I inlet HEPA filters will provide an efficiency of greater than 99.9 percent for removal of radiological particulates from the air that may reverse flow from Zone I to Zone IL Zone I ducting will ensure that negative air pressure can be maintained by conveying exhaust air to the stack. Bubble-tight dampers will be provided to comply with the requirement s of ASME AG-1 , Section DA-5141. Ventilation ductwork and ductwork support materials will meet the requirements of ASME AG-1. Supports will be designed and fabricated in accordance with the requirements of ASME AG-1. Zone I exhaust train HEPA filters will provide an efficiency of greater than 99.95 percent for removal of radiological particulates from the air that flows to the stack. Zone I exhaust train HEGA filters will provide an efficiency of greater than 90% for removal of iodine. The Zone I exhau s t stack will provide dispersion of radionuclides in normal and abnormal releases at a discharge point of 23 meters (m) (75 feet [ft]) above the building ground level. Stack monitoring and interlocks will monitor discharge and signal changing of service filter trains during normal and abnormal operations. Secondary process offgas treatment iodine removal beds (VV-SB-520) will mitigate an iodine release. 6.2.1.5 Test Requirement s Engineered safety features will be tested to ensure that components maintain operability and can provide adequate confidence that the safety system performs satisfactorily during postulated events. The confinement engineered safety features that initiate the system interlocks are de s igned to permit testing during plant operation.

The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application. 6-12

.. ; ... ; .. NWMI ..*... ..* .... ........ *.* * "NOtlTKWUTMEOK:AltSOTOP£S 6.2.1.6 Design Basis NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Codes and standards are discussed in Chapter 3 .0, " Design of Structures, Systems , and Components." The design bases for Zone I and Zone II venti lation systems are described in Chapter 9.0. The design basis of confinement enclosure structures is described in Chapter 4.0. Chap ter 7.0, " Instrumentation and Contro l Systems," identifies the engineered safety feature-related design basis of the venti lation control system. The following information was developed for the Construction Permit Application to describe the proces s offgas secondary iodine removal bed: * * * * *

  • Sorbent bed of [Propri etary Information]

Iodine removal efficiency greater than [Proprietary Information]

Nominal superficial gas flow velocity of [Propri etary Information]

Nominal sor bent bed operating temperature of l ess than [Proprietary Information]

Nominal sorbent bed depth of [Proprietary Information]

Nominal gas relative humjdity less than [Propri etary Information]

Additiona l detailed information on the process offgas iodine retention bed design basis will be developed for the Operating License App lication.

Potential variables , conditions , or other items that will be probable subjects of a technical specification associate d with the RPF confineme nt systems and components are discussed in Chapter 14.0, " Technical Specifications

." 6.2.1.7 Derived Confinement Items Relied on for Safety The following su bsection s describe additional engineered safety features that are derived from the accide nt analyses described in C hapter 13.0 and are projected technical spec ification s defining limited condit ion s for operation.

6.2.1.7.1 IROFS RS-09, Primary Offgas Relief System IROFS RS-09 , " Primary Off gas Relief System," is identified by the accident analysis in Chapter 13.0. As an active engineered control (AEC), the primary off gas relief system wi ll be a component includ ed in the offgas train for the two irradiated target dissolvers.

The dissolver offgas system is intended to operate at a pressure that is less than the confinement enclosures to maintain gaseous components generated during dissolution within the vessels and route the gaseo us components through the offgas treatment unit operations.

The primary off gas relief system, or pressure relief tank, will be u sed to confine gases to the dissolver and a portion of the dissolver off gas equipment , if the off gas motive force (vac uum pumps) ceases operation during di ssolutio n of a dissolver batch. 6-13

...... ; ... NWMI ..**.. ..* .... ..... .. .. .. . * ! : . NOlmlWEn MEOM:Al tsoTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Figure 6-5 is a diagram of the dissolver off gas system process , which s hows the pressure relief tank position in the off gas treatment equipment train. Figure 6-6 shows the location of the pres s ure relief tank within the RPF hot cell (identified as " pressure relief'). [Proprietary Information]

Figure 6-5. Dissolver Offgas System Engineered Safety Features 6-14

.. NWMI ..*... ..* ... ........ *.* ' ." NORTifWUT MEOtcAl lSOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

Figure 6-6. Dissolver Offgas Hot Cell Equipment Location The pressure relief tank will be evacuated to a specified, subatmospheric pressure prior to initiating dissolution of a target batch and selected valves (indicated as 2 , 3, and 4 on Figure 6-5) closed. Valve I will be open during normal dissolver operation.

An upset during the dissolver operation (e.g., loss of vacuum pump operation) will result in closing Valve 1 and opening Valve 2 to contain dissolver offgas within the dissolver and off gas vessels. Due to the short duration of dissolver operation , dissolution is assumed to go to completion independent of an off gas system upset. The pressure relief tank will contain the offgas as dissolution is completed. Valves 3 , 4 , and 5 are provided for upset recovery.

After correction of the upset cause , gases collected in the pressure relief tank will be routed to the downstream treatment unit operations via Valve 3 or returned to a caustic s crubber via Valve 4. Liquid condensed in the pressure relief tank as a result of activation will be routed to the dissolver offgas liquid waste collection tank via Valve 5 for disposal.

6-15

.**.*.*.* .. ;.-.;* .. NWMI ......... *.* . NOATHWESTMEOICAUSOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Accident Mitigated
  • Irradiated target dissolver off gas system malfunctions, including loss of power during target dissolution operations System Components Pressure relief valves Pressure relief tank (DS-TK-500)

Functional Requirements

  • As an AEC , use relief device to relieve pressure from the system to an on-service receiver tank maintained at vacuum with the capacity to hold the gases generated by the dissolution of one batch of targets in the target dissolver Prevent a failure of the primary confinement system by capturing gaseous effluents in a vacuum receiver tank Design Basis The following information was developed for the Construction Permit Application describing the pressure relief tank. * * * * * *
  • Pressure-relief tank sizing is based on a maximum dissolver batch of [Proprietary Information]

that has just started dissolution when the pressure relief event is initiated.

The non-condensable gas volume to the pressure relief tank is equivalent to all nitrogen oxide (NO x) generated by dissolution , plus the sweep gas flow for flammable hydrogen gas mitigation.

Worst-case reaction stoichiometry of [Proprietary Information]

dissolved is used . No credit is taken for reaction of N0 2 with water to produce nitric acid . Dissolver gas additions , other than the minimum sweep gas flow for hydrogen mitigation , are terminated by the pressure relief event. Gas contained by the pressure relief tank and associated dissolver off gas piping is saturated with water vapor. The pressure change from [Proprietary Information], absolute activates the pressure relief tank . Additional detailed information on the pressure relief tank design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application. 6-16

... ; .. NWMI ...... ..* ... ** ** ...... *

  • NOmlWEST llB>tCAl lSOTOf'H NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2.1.7.2 IROFS RS-10, Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer IROFS RS-10 , "Active Radiation Monitoring and Isolation of Low-Dose Waste Transfer," is identified by the accident analyses described in Chapter 13.0. As an AEC , the recirculating stream and the discharge stream of the low-dose waste tank will be simultaneously monitored in a background shielded trunk outside of the hot cell shielded cavity. The continuous gamma instrument will monitor the transfer lines to provide an open permissive signal to dedicated isolation valves. Accident Mitigated Transfer of high-dose process liquid solution s outside the hot cell shielding boundary System Components Additional detailed information of the radiation monitor and isolation of low-dose waste transfers will be developed for the Operating License Application.

Functional Requirement Maintain worker and public exposure rates within approved limits Design Basis Additional detailed information of the radiation monitor and isolation of low-dose waste transfers will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.3 IROFS RS-13, Cask Local Ventilation During Closure Lid Removal and Docking Preparations IROFS RS-13 , "Cask Local Ventilation During Closure Lid Removal and Docking Preparations

," is identified by the accident analyses described in Chapter 13.0. As an AEC , a local capture ventilation system will be used over the irradiated target cask closure lid to remove any escaped gases from the worker breathing zone during removal of the closure lid , removal of the shielding block bolts , and installation of the lifting lu gs. Accident Mitigated

  • Irradiated target cladding fails during transportation , releasing gaseous radionuclides within the cask containment boundary System Components
  • Use a dedicated evacuation hood over the top of the cask during containment closure lid removal Remove gases to the Zone I secondary confinement system for proces s ing 6-17

... NWMI ::.**.*.-.. ........... . * * ! ." . NOflTHWEn MmtcAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Functional Requirement

  • Prevent exposure to workers by evacuating any high-dose gaseous radionuclides from the worker breathing zone and preventing immersion of the worker in a high-dose environment Design Basis The following information was developed for the Construction Permit Application describing the cask local ventilation system: Use the local capture ventilation system to evacuate and backfill the cask with fresh air (from a protected pressurized source such as a compressed bottle) until the atmospheres are within approved safety limits Additional detailed information on the cask local ventilation system design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.4 IROFS RS-15, Cask Docking Port Enabling Sensor IROFS RS-15 , "Cask Docking Port Enabling Sensor , is identified by the accident analyses described in Chapter 13.0. As an AEC, the cask docking port will be equipped with sensors that detect when a cask is mated with the cask docking port door. Accident Mitigated

  • Cask lift failure occurs after shield plug removal (but before target basket removal) with targets inside the cask System Components Enabling contact signal and positive closure signal when the sensor does not sense a cask mated to the cask docking port, causing the cask docking port door to close Functional Requirement
  • Prevent the cask docking port door from being opened and allowing a streaming radiation path to areas accessible by workers Design Basis Detailed information on the system design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6-18

..... : ... NWMI ...... ..* .... ......... *.*

  • NOATHWUTMEDICA.LtsOTOHI NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2.1.7.5 IROFS FS-03, Process Vessel Emergency Purge System IROFS FS-03 , "Process Vessel Emergency Purge System," is identified by the accident analyses described in Chapter 13.0. Hydrogen gas will be evolved from process solutions through radiolytic decomposition of water in the high radiation fields. An air purge to the vapor space of selected tanks will be provided by the facility air compressors to control the hydrogen concentration from radiolysis in vessel vapor space to below the flammability limit for hydrogen. As an AEC, an emergency backup se t of bottled nitrogen gas will be provided for all tanks that have the potential to evolve significant volumes of hydrogen gas through the radiolytic decomposition of water (in both a short-and long-term storage condition). Accident Mitigated Hydrogen deflagration or detonation in a process vessel System Components Information will be provided in the Operating License Application. Functional Requirement
  • Pre ve nt development of an explosive hydrogen-air mixture in the tank vapor spaces to prevent the deflagration or detonation hazard Design Basis The following information was developed for the Construction Permit Application describing the proces s vesse l emergency purge syste m: * * *
  • Monitor the purge pressure going into the individual tanks and open an isolation valve on low pre ssure (setpoint to be determined) to restore the continuous sweep of the system using nitrogen Provide sweep gas sufficient for the facility to allow repair of a compressed gas system outage Activate by sensing low pressure on the norm a l sweep air syste m , introducing a continuous purge of nitrogen from a reliable emergency backup sta tion of bottled nitrogen into each affected vessel near the bottom (e.g., through a liquid level detection leg) of the vesse l Dilute hydrogen as it rises to the top of the vessel and is vented to the respective vent system Additional det ai led information on the process vessel emergency purge system design basi s will be developed for the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application. 6.2.1.7.6 IROFS FS-04, Irradiated Target Cask Lifting Fixture IROFS FS-04 , "Irradiated Target Cask Lifting Fixture ," is identified by the accident analyses described in Chapter 13.0. As a passive engineered control (PEC), the irradiated target cask lifting fixture will be designed to prevent the cask from tipping within the fixture and the fixture itself from toppling during a seis mic event. 6-19

.. ... ;-.. NWMI ..*... ..* .... ........ *.* ' * ! *, ! ." . NOtmfWfST MEDtCAL ISOTOPU Accident Mitigated NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features

  • Dislodged irradiated target shipping cask shield plug in the presence of workers during target unloading activities System Components Detailed information on the system components will be developed for the Operating License Application. Functional Requirements Detailed information on the system functional requirements will be developed for the Operating License Application.

Design Basis Detailed information on the system design basis will be developed for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.7.7 IROFS FS-05, Exhaust Stack Height IROFS FS-05 , "Exhaust Stack Height ," is identified by the accident analyses described in Chapter 13.0. Accidents Mitigated Process solution spills and sprays Carbon bed fire System Component Zone I exhaust stack Functional Requirement

  • Provide an off gas release height for ventilation gases consistent with the stack height used as input to mitigated dose consequence evaluations. Design Basis The Zone I exhaust stack height is 23 m (75 ft). Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6-20 NWMI ..**.. .. ... **** .. .. .. * *. * ! 0 NOltTifWEST MEDtCAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2.1.7.8 IROFS CS-09, Double Wall Piping IROFS CS-09 , "Double Wall Piping," is identified by the accident analyses in Chapter 13.0. This IROFS has both a confinement and nuclear criticality prevention function. As a PEC, the piping system conveying fissile solution between credited confinement locations will be provided with a double-wall barrier to contain any spills that may occur from the primary confinement

[Proprietary Information]

Figure 6-7. Proposed Location of Double-Wall Piping (Example) piping. This IROFS will be used at those locations that pass through the facility , where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection purposes. Figure 6-7 provides an example location where IROFS CS-09 will be applied (e.g., the transfer line between the recycle uranium decay tanks and the [Proprietary Information]).

Accident Mitigated Leak in piping that passes between confinement enclosures System Components The following double-wall piping segments are identified at this time: * *

  • Transfer piping containing fissile solutions traversing between hot cell walls Transfer piping connecting the uranium product transfer send tank (UR-TK-720) and uranyl nitrate storage tank (TF-TK-200)

Other locations to be identified in final design Functional Requirements

  • Double-wall piping prevents personnel injury from exposure to acidic or caustic licensed material s olutions conveyed in the piping that run s outside a confinement enclosure Double-wall piping routes pipe leaks to a critically-safe leak collection tank or berm as a nuclear criticality control feature Design Basis The double-wall piping arrangement is designed to gravity drain to a safe-geometry set of tanks or to a safe geometry berm. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6-21

... .-.;:.NWMI ..**.. ..* ... ........... * *

  • NOmfWEST MUHCAl lSOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2.1.7.9 IROFS CS-18, Backflow Prevention Devices, and IROFS CS-19, Safe-Geometry Day Tanks IROFS CS-18, "Backflow Prevention Devices," and IROFS CS-19, "Safe-Geometry Day Tanks ," are identified by the accident analyses in Chapter 13.0. As a PEC or AEC , chemical and gas addition ports to fissile process solution systems will enter a confinement enclosure through a backflow prevention device. Backflow prevention devices and safe-geometry day tanks will provide alternatives for preventing process addition backflow across confinement boundaries. The device may be an anti-siphon break, an overloop seal, or other active engineering feature that addresses the conditions of backflow and prevents fissile solution from entering non-safe geometry systems or high-dose solutions from exiting the hot cell shielding boundary in an uncontrolled manner. Therefore , these IROFSs have both a confinement and a nuclear criticality prevention function.

Accident Mitigated Backflow of process material located inside a confinement boundary to vessel located outside confinement via connected piping due to process upset. System Components System component information will be provided in the Operating License Application.

Functional Requirements

  • * * *
  • Prevent fissile solutions and/or high dose solutions from backflowing from the tank into systems outside the confinement boundaries that may lead to accidental nuclear criticality or high exposures to workers Provide each hazardous location with an engineered backflow prevention device that provides high reliability and availability for that location Locate the backflow prevention device features for high-dose product solutions inside the confinement boundaries Support the backflow prevention devices with safe-geometry day tanks located inside the confinement boundary Direct spills from the backflow prevention device to a safe-geometry confinement berm Design Basis Design basis information will be provided in the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application. 6-22

..... ; ... NWMI *********. ........ *.* *

  • NOmfWEST lllEDICAl lSOTOPll 6.2.1.8 Dissolver Offgas Systems 6.2.1.8.1 Dissolver Offgas Iodine Removal Unit NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features A s ignificant fraction of iodine entering the RPF in targets is projected to be released to dissolver off gas during target dissolution. The dissolver off gas iodine removal units will be included in the RPF as the primary SSCs for controlling the relea se of iodine isotopes to the environment or facility areas occupied by workers. Components of the di sso l ver off gas system , beginning with the iodine removal unit , will also be used to treat ve nt gas from the target disa sse mbly syste m. Target disa ssem bl y vent gas is treated by dissolver off gas components for the Construction Application Permit configuration as a measure to mitigate the unverified potential for a release of fi ssion gas radionuclides durin g target transportation.

Figure 6-5 (Section 6.2.1.7.1) shows the iodine removal unit position in the offgas treatment equipment train. The dissolver offga s iodine removal unit location in the facility is shown in Figure 6-6 (identified as " primary fission gas treatment").

Accidents Mitigated Projected limitin g control for operation Required for normal operation and not for accident mitigation System Components Iodine removal unit A (DS-SB-600A)

Iodine removal unit B (DS-SB-600B)

Iodine removal unit C (DS-SB-600C)

Functional Requirement Remove iodine isotopes from the dissolver offgas during normal operations such that the dose to workers complies with 10 CFR 20.120 I , "Occ upational Dose Limits for Adults ," and the dose to the public complies with 10 CFR 20.1301 , " Dose Limits for Individual Members of the Public." Design Basis The following information was developed for the Construction Permit Application describing each individual iodine removal unit: Sorbent bed of [Proprietary Information]

Iodine removal efficiency greater than [Proprie tary Information]

Nominal superficial gas flow ve locity of [Proprietary Information]

Nominal sorbent bed operating temperature of [Proprietary Information]

Nominal sorbent bed depth of [Proprietary Information], pro vi ding iodine removal capacity of greater than l year (yr). Additional detailed information on the iodine removal unit design basis will be developed for the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6-23

..... .. NWMI ..**.. ..* .... .*.* .. *.*.* . ' *.* ." . '"*11fWDT MEDICAL lSOTOPf.S NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.2.1.8.2 Dissolver Offgas Primary Adsorber Noble gases (krypton [Kr] and xenon [Xe]) entering the RPF in targets are projected to be released to dissolver offgas during target dissolution. The dissolver offgas primary adsorber units will be included in the RPF as the primary SSCs for controlling the release of noble gas isotopes to the environment or facility areas occupied by workers. Components of the dissolver off gas system will also be used to treat vent gas from the target disassembly system. Target disassembly vent gas is treated by dissolver offgas components for the Construction Application Permit configuration as a measure to mitigate the unverified potential for a release of fission gas radionuclides during target transportation. Figure 6-5 (Section 6.2.1. 7.1) shows the primary adsorber position in the off gas treatment equipment train. The dissolver offgas primary adsorber location in the facility is shown in Figure 6-6 (identified as "primary fission gas treatment"). Accidents Mitigated Projected limiting control for operation Required for normal operation and not for accident mitigation System Components Primary adsorber A (DS-SB-620A)

Primary adsorber B (DS-SB-620B)

Primary adsorber C (DS-SB-620C)

Functional Requirement Delay the release of noble gas isotopes via the dissolver off gas during normal operations such that the dose to workers complies with 10 CFR 20.1201 and the dose to the public complies with 10 CFR 20.1301. Design Basis The following information was developed for the Construction Permit Application describing each individual primary adsorber unit: * * * * *

  • Sorbent bed of [Proprietary Information]

Nominal sorbent bed operating temperature of [Proprietary Information]

Nominal gas relative humidity less than [Proprietary Information]

Average gas flow rate of [Proprietary Information]

Nominal superficial gas flow velocity of [Proprietary Information]

Delay time for release of Xe isotopes of 10 days and Kr isotopes of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (hr) (additional delay time is provided by the secondary adsorber)

Additional detailed information on the primary adsorber unit design basis will be developed for the Operating License Application.

6-24

..... .. NWMI ...... .. *.. ........ *.* * *

  • NOmfWEIT MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.8.3 Dissolver Offgas Vacuum Receiver/Vacuum Pump The dissolver off gas vacuum pump will provide the motive force for transferring off gas , generated in the dissolvers and disassembly equipment during operation , through the dissolver offgas equipment train while maintaining dissolver vessels at a pressure less than the equipment enclosure pressure.

Vacuum receiver tanks will be provided as part of the motive force system to allow the vacuum pumps to cycle on and off less frequently and accommodate the wide variations in gas flow rate associated with a target dissolution cycle. Figure 6-5 (Section 6.2.1.7.1) shows the vacuum receiver tank and vacuum pump positions in the offgas treatment equipment train. The vacuum receiver tank and vacuum pump location in the facility is shown in Figure 6-3 in the vicinity of equipment identified for the process off gas secondary iodine removal bed. Accidents Mitigated Projected limiting control for operation Required for normal operation and not for accident mitigation System Components Vacuum receiver tank A (DS-TK-700A)

Vacuum receiver tank B (DS-TK-700B)

Vacuum pump A (DS-P-710A)

Vacuum pump B (DS-P-710B)

Functional Requirements

  • Maintain the dissolver vessel gas space at a pressure less than the dissolver vessel enclosure pressure throughout the target dissolution cycle Accommodate pressure drops associated with dissolver off gas unit operations over the range of gas flow rates generated in both dissolvers and the target disassembly equipment vent throughout a target dissolution cycle Design Basis The following information was developed for the Construction Permit Application describing the vacuum receiver tanks and vacuum pump: * * * *
  • Minimum inlet setpoint pressure of [Proprietary Information]

Maximum inlet setpoint pressure of [Proprietary Information]

Outlet pressure of [Proprietary Information]

Maximum sustained gas flow into [Proprietary Information]

Receiver tank provides a [Proprietary Information]

with the vacuum pump off and inlet at the maximum sustained gas flow 6-25

.. .. ... NWMI ...... ..* .... ......... *.* * ! *.

  • NORTNWEn MEDIC.Al lSOTOrEI NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Additional detailed information on the vacuum receiver tank and vacuum pump design basis will be de ve loped for the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6.2.1.9 Exhaust System The ventilation exhaust system is described in Chapter 9.0, Section 9. I .2. Additional detailed information will be developed for the Operating License Application, including:

  • *
  • Describing changes in operating conditions in response to potential accidents and the mitigation of accident radiological consequences Demonstrating how dispersion or distribution of contaminated air to the environment or occupied spaces is controlled Identifying the design bases for location and operating characteristics of the exhaust stacks 6.2.1.10 Effluent Monitoring System Each RPF exhaust stack will include an effluent monitoring system. The monitoring system sample lines are designed to comply with ANSI NB.I , Sampling and Monitoring R eleases of Airborne Radioa ct ive Substances from the Stacks and Ducts of Nuclear Facilities.

Additional detailed information on the effluent monitoring systems will be developed for the Operating License Application. 6.2.1.11 Radioactive Release Monitoring The effluent monitoring system will provide flow rate, temperature , and composition inputs for dispersion modeling of releases from the exhaust stacks. These inputs will pro vi de the capability for calculating potential exposures as a ba sis for actions to ensure that the public is protected during both normal operation and accident conditions.

Additional detailed information on radioactive release monitoring will be developed for the Operating License Application.

6.2.1.12 Confinement System Mitigation Effects Detailed information describing the confinement syste m mitigation effects will be developed for the Operating License Application.

This information will compare the radiological exposures to the facility s taff and the public with and without the confinement sys tem engineered safety feature. The comparison will be based on analyses showing airflow rates, reduction in quantities of airborne radioactive material by filter systems, system isolation , and other parameters that demonstrate the effectiveness of the system. 6-26

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  • NOmlWUT MEDtCAl tSOTOPH NWMl-2013-021 , Rev. 1 Chapter 6.0 -Eng i neered Safety Feature s 6.2.2 Containment Containment for the RPF is defined based on NUREG-153 7 , Guidelines for Pr e paring and R eview in g A ppli catio n s fo r th e Li cens in g of Non-P ower R eactors -Format and Content, Part 1 interim staff guidance.

Containments are required as an e ngin ee r ed safety feature on th e ba s i s of th e radioi so top e production facility d es ign , operating characteristics, accidents sce nario s , and location.

A potential sce nario for s u c h a r e l ease cou ld b e a significant lo ss of integrity of th e radioisotope extraction system or th e irradiat e d fu e l proc ess ing sys tem. Th e containment is d es igned to control the rel ease to th e environment of airborne radioa ctive material that i s r e l eased in the facility even if the accident is accompanied by a pressure s ur ge or stea m release. The NUREG-153 7 Part 1 interim staff guidance has been applied to the RPF target processing systems. The current accident analysis described in C hapter 13 .0 has not identified a need for a containment system as an engineered safety feature. 6.2.3 Emergency Cooling System An emergency cooling system for the RPF is defined by NUREG-1537 Part 1 interim staff guidance.

In the eve nt of the loss of any r e quir e d prima ry or normal coo lin g syste m, an e m e rg e n cy coo ling syste m ma y b e r e quir ed to remove decay h ea t from the fuel to prevent th e failure or degradation of th e gas manag e m e nt sys t e m , the i so top e extrac tion s y stem, or the irradiat ed fu e l processing system. An evaluation of RPF cooling requirements provided in Chapter 5.0 , "Coo lant S ys tems ," indicates that an emergency cooling system wi ll not be required to avoid rupture of the primary process vessels. In addition , the current accident analysis described in Chapter 13.0 has not identified a need for an emergency cooling system as an engineered safety feature. 6-27

..... .. NWMI ..*... ..* *.. ........ *. * *

  • 0 NOITHWEST MmtCAl lSOTOPH NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.3 NUCLEAR CRITICALITY SAFETY IN THE RADIOISOTOPE PRODUCTION FACILITY The RPF design will provide adequate protection against criticality hazards related to the storage, handling , and processing of SNM outside a reactor. This is accomplished by: * *
  • Including equipment, facilities, and procedures to protect health and minimize danger to life or property Ensuring that the design provides for criticality control, including adherence to the contingency principle Incorporating a criticality monitoring and alarm system into the facility design For the Construction Permit Application, the design has assumed that a nuclear criticality accident is a high-consequence event independent of whether shielding or other isolation is available between the source of radiation and facility personnel.

While not considered likely at this time , justification for considering criticality events as other than a high-consequence event will be provided in the Operating License Application, if this assumption is changed for s pecific locations by future design activities.

The nuclear criticality safety program defines the programmatic elements that work in concert to maintain criticality controls throughout the operating life of the RPF. The nuclear criticality safety program and facility design are developed based on the following American National Standards Institute/ American Nuclear Society (ANSI/ ANS) standards, with exceptions described in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 3. 71 , Nuclear Criticality Safety Standards for Fuels and Material Facilities.

  • * * * * * * *
  • ANSI/ ANS-8.1, Nuclear Criticality Safety in Op e ration s with Fissionable Materials Outsid e Reactors ANSI/ ANS-8.3, Criticality Accident Alarm System ANSI/ ANS-8. 7 , Nuclear Criticality Safety in the Storage of Fissile Materials ANSI/ ANS-8. l 0 , Crite ria for Nuclear Criticality Safety Controls in Op erations With Shielding and Confinement ANSI/ ANS-8.19, Administrative Practices for Nuclea r Criticality Safety ANS I/ ANS-8.20 , Nuclea r Criticality Safety Training ANSl/ANS-8.22, Nuclear Criticality Safety Ba sed on Limiting and Controlling Moderators ANSI/ ANS-8.23 , Nuclear Criticality Accident Emergency Planning and R espo n se ANSl/ANS-8.24, Validation of Neutron Transport Methods for Nuclear Critica lity Safety Calculations ANSl/ANS-8.26, Criticality Safety Engineer Training and Qualification Program For the Construction Permit Application, no deviations from standards or requirements have been identified that would require development of equivalent requirements for the RPF. NWMI commits to the following standards and guides during design and construction:
  • ANSI/ ANS-8. l -Nuclear criticality safety practices , including administrative practices , technical practices , and validation of a calculational method 6-28

.......... ..... ; .. NWMI ........ *.* 0 ! *. * ! . NOmlWUT MEJHCAL ISOTOP'(S NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features * * * * * *

  • ANSVANS-8.3

-Criticality accident alarm system (CAAS) placement analysis and procedure development

the standard is used as modified by NRC Regulatory Guide 3.71 ANSVANS-8.19 -NWMI nuclear criticality safety program development as it applies to organization, administration , roles, and responsibilities ANSVANS-8.20 -Nuclear criticality safety s taff and contractor qualification and training procedure development ANS V ANS-8.24 -Validation of a calculational method NUREG-1520 , St a ndard R e vi ew Plan for th e R ev i e w of a Li ce n se A ppli c ation for a Fuel C y cl e Facili ty-Guidance for meeting 10 CFR 70.61 NUREG/CR-4604 , Stati s ti c al M e thod s for N ucl e ar Mat e r i al Manag e m e nt-Guidance for normality testing of the data from critical experiment calculations NUREG/CR-6698 , Guid e for Validation of N ucl e ar Criticality Safety Cal c ulational M e thodologyGuidance for validation of a calculational method The nuclear criticality safety program includes the following elements
  • Responsibilities Criticality safety evaluations Criticality safety control implementation Nuclear criticality safety training Criticality safety assessments Criticality prevention specifications Operating procedures and maintenance work Criticality safety postings Fissile material container labeling , storage , and transport Criticality safety nonconformance response Criticality safety configuration control Criticality detector and alarm system Criticality safety guidelines for firefighting Emergency preparedness plan and procedure s Components of the nuclear criticality safety program specifically implemented during the design and construction phases of the RPF will include: Nuclear criticality s afety program policy Nuclear criticality safety program procedur e Nuclear criticality safety evaluation procedure Nuclear criticality safety technical/peer review procedure
  • Nuclear criticality safety engineer training and qualification procedure Nuclear criticality safety validation procedure Preliminary descriptions of the nuclear criticality safety program elements developed for the Construction Permit Application are summarized below. Modifications to the nuclear criticality safety program elements are anticipated as the design matures and will be included in the Operating License Application. Responsibilities This element describes the re s ponsibilities of management and staff in implementing the nuclear criticality safety program. 6-29

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  • General facility management will ensure that the nuclear safety function is as independent as practical from the facility operating functions. A Nuclear Criticality Safety Manager will be assigned and responsible for overall coordination , maintenance , and management of the nuclear criticality safety program. A Criticality Safety Representative will be assigned who is qualified to interpret criticality safety requirements and serve as a liaison between custodians of fissionable material and other operations , advising operating personnel and supervisors on questions concerning conformance to criticality safety requirements.

Qualified Criticality Safety Engineers will responsible for performing criticality analyse s and evaluations of systems , maintaining current verified and validated criticality computer codes, advising staff on technical aspects of criticality controls, and supporting/participating in inspections and management assessments.

Operations management will be responsible for establishing the responsibility for criticality safety throughout the operations organization , communicating criticality safety responsibilities for each individual involved in operations , ensuring that controls identified by CSEs are implemented , ensuring each worker has necessary training and qualifications , and ensuring that procedures that include control s significant to criticality safety are prepared before operations commence.

Supervisors and workers will be responsible for completing training before performing fissile material operations , understanding and ensuring compliance with all applicable criticality safety controls, and reporting any proposed change in fissile material operations to the Criticality Safety Representative for evaluation and approval before the operation commences. Criticality Safety Evaluations This element describes the process for preparing CSEs that demonstrate fissile material operation will be subcritical under both normal and credible abnormal conditions.

  • * * *
  • CSEs will determine, identify, and document the controlled parameters and associated limits on which criticality safety depends. CSEs will be required to evaluate normal operations , and contingent and upset conditions . Preliminary CSEs prepared for the Construction Permit Application, including verification and validation of supporting computer codes , are described in Section 6.3.1.1 and provide examples of the CSEs. Design changes impacting criticality will be reviewed by the Criticality Safety Representative . CSEs will be independently reviewed to confirm the technical adequacy of the evaluation prior to commencing new or modified fissile material operations.

Nuc l ear criticality safety limits established for controlled parameters in the NWMI facility processes will ensure that all nuclear processes are subcritical , including an adequate margin of subcriticality for safety in accordance with the Interim Staff Guidance augmenting NUREG-15 3 7 , Guideline s for Preparing and R e viewing Appli c ations for the Lic e nsing of Non-Power R e actors: Standard R e vi ew Plan and Acceptan ce Criteria , Part 2 , Section 6.b.3 (NRC , 2012). Monte-Carlo N Particle (MCNP) calculation results used to set limits on parameters are compared to the upper subcritical limit (USL) established in the NWMI MCNP code validation report ([Proprietary Information]), after applying a 2cr calculation uncertainty. 6-30

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  • NOfllllfW(n MEDICAl tsOTOPf:I NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The USL includes the method bias and uncertainty established in [Proprietary Information]

and a 0.05 margin of subcriticality. In addition , the area of applicability , also estab l ished in [Proprietary Information], is checked to ensure that the NWMI RPF process model physics and materials are within the bands of applicability.

If either the physics or materials are outside the bands of applicability , an additional margin of subcriticality will be app l ied. Criticality Safety Control Implementation This element describes the process for implementing criticality safety controls defined by the CSEs. * *

  • Implementation includes confirming that: All required engineered criticality safety controls are maintained by a configuration management system. Equipment dimensions, volumes , or other features relied on for controls are with limits documented in the CSEs. Administrative critica l ity safety controls from CSEs are implemented in written operating and maintenance procedures.

Fissile material inventories will be monitored and incorporated into implementation of criticality safety controls.

Access to fissionable material will be controlled . N u clear Cr i tica li ty Safety Tra in i n g This element describes the training program for nuclear criticality safety based on the worker's duties and responsibilities. * * *

  • This training program is developed and implemented with input from the nuclear criticality safety staff , training staff , and management , with a focus on: Knowledge of the physics associated with nuclear criticality safety Analysis of jobs and tasks to determine the knowledge a worker must have to perform tasks efficiently Design and development of learning objectives based on the analysis of jobs and tasks that reflect the knowledge, skills , and abi l ities needed by the worker Implementation of revised or temporary operating procedures Testing methods to demonstrate competence in training materials dependent on an individual's responsibility Training records maintenance General training on criticality hazards and alarm responses will be provided to a ll RPF personnel and visitors.

Operators responsible for some aspect of nuclear criticality safety will: Satisfy defined minimum initial qualifications Complete an initial criticality safety training course designed for operators Perform periodic requa l ification training Management, operations supervisor , and technical staff responsible for some aspect of nuclear criticality safety will: Satisfy defined minimum initial qualifications Comp l ete an initial criticality safety training course designed for managers and engineers 6-31

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  • Perform periodic requalification training The Criticality Safety Representative will: Satisfy defined minimum initial qualifications Complete an initial criticality safety program designed for the Criticality Safety Representative Demonstrate competence in understanding facility nuclear criticality controls and procedures Perform periodic requalification training Criticality Safety Engineers will be trained and qualified in accordance with ANSI/ ANS-8.26 . Nuclear criticality safety staff members and contract support will meet the qualification and training requirements specified in the NWMI nuclear criticality safety qualification and training program. The NWMI nuclear criticality safety qualification and training program is compliant with ANSI/ ANS 8.26. Criticality Safety Assessments This element describes the periodic criticality safety inspections and assessments conducted to ensure that the criticality safety program is maintained at an adequate level for the RPF. * * * *
  • Annual criticality safety inspections will be conducted to satisfy the requirement of ANSI/ ANS-8.1 and 8.19 for operational review s to be conducted at least annually.

Procedures will be developed for performing periodic criticality safety inspections.

The facility Criticality Safety Representative and inspection team will comprise individuals (typically from Engineering) who are knowledgeable of criticality safety , and who , to the extent practicable, are not immediately responsible for the operation being inspected. Facility inspections are conducted to verify that the facility configuration and activities comply with the nuclear criticality safety program. Facility inspections generally consist of observation of task preparation and verification of field procedures and training. Management assessments will be conducted of the nuclear criticality safety program. These assessments will be led by the Nuclear Criticality Safety Manager , with assistance from other members of the criticality safety staff. The criticality safety staff is independent of the operating organization and not directly responsible for the operations. Records generated during performance of criticality safety inspections and assessments will be included in a criticality safety inspection report or specialty assessment report. An audit to assess the overall effectiveness of the nuclear criticality safety program will be performed at least once every three years. The audit will be led by a qualified senior criticality safety engineer from outside the NWMI organization.

The senior nuclear criticality safety engineer conducting the audit will be independent of the NWMI program and will not have participated in any nuclear criticality safety evaluation that will be a subject of the audit. In addition to the triennial audit from an outside organization, NWMI senior management will perform periodic audits of the NWMI nuclear criticality safety program. The senior manager will be chosen from an NWMI organization other than the nuclear criticality safety group. The NWMI Quality Assurance Manager will select and assign auditors who are independent of the NWMI nuclear criticality safety program. Criticality Prevention Specifications This element describes the requirements for the criticality prevention specifications used to implement limits and controls established in the CS Es for safe handling of fissionable material and implement the ANSI/ANS-8 series requirement for clear communication of criticality safety limits and controls.

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  • Each criticality prevention specification will: Be based on an approved CSE and refer to the CSE used as a specification source Be prepared by either the Criticality Safety Representative of a qualified Criticality Safety Engineer Emphasize limits controllable by the operator Have clear and unambiguous meaning and be written , to the extent practical , using operations terminology with common units of measure Operating Procedures and Maintenance Work This element describes the requirements for implementing nuclear criticality controls in written procedures for operations and maintenance work. * * * *
  • Procedures will meet the intent of ANSVANS-8.19 . Procedures for operations and maintenance work will be prepared according to approved procedure control programs, developed and maintained to reflect changes in operations, and written so that no single inadvertent failure to follow a procedure can cause a criticality accident.

Operating procedures will include: Controls and limits significant to nuclear criticality safety of the operation Periodic revisions , as necessary Periodic review of active procedures by supervisors Operating procedures will be supplemented by criticality safety postings on equipment or incorporated in operating checklists. Maintenance work procedures associated with SSCs affecting nuclear criticality safety will be reviewed by the Criticality Safety Representative or a Criticality Safety Engineer for compliance with nuclear criticality safety limits based on current RPF conditions present prior to initiating each maintenance evolution.

Criticality Safety Postings

  • Criticality safety postings will be developed for the Operating License Application . Fissile Material Container Labeling, Storage, and Transport
  • Fissile material container labeling , storage , and transport will be developed for the Operating License Application. Criticality Safety Nonconformance Response This element describes the response to deviations from defined nuclear criticality safety controls. *
  • Deviations from procedures and unforeseen alterations in process conditions that affect criticality safety will be immediately reported to management and the Criticality Safety Representative or a Criticality Safety Engineer.

NWMI management will provide the required notifications of the deviation to the U.S. Nuclear Regulatory Commission Operations Center. 6-33

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  • The Criticality Safety Representative or a Criticality Safety Engineer will support an investigative team comprising , at a minimum , the Operations Manager and operations personnel familiar with the operation in question during the development of a recovery plan for safely returning to compliance with the procedures.

The deviation will be corrected per the recovery plan and the incident documented . Action is to be taken to ensure that a similar situation does not exist in another part of the facility and to prevent recurrence of the nonconformance. Criticality Safety Configuration Control This element describes the criticality safety configuration controls. * * * * * *

  • The primary criticality safety control , performed at the start of a proposed activity or equipment change , is for the Criticality Safety Representative to confirm if an existing active CSE is applicable. All dimensions , nuclear properties , and other features on which reliance is placed will be documented and verified prior to beginning operations , and control will be exercised to maintain them. The nuclear criticality safety staff will provide technical guidance for the design of equipment and processes and for the development of operating procedures. All proposed criticality safety-related changes to design or process configuration will be reviewed by a Criticality Safety Representative or Criticality Safety Engineer to ensure that the change can be performed under an approved CSE. All operational changes that impact criticality safety will be documented and include proper approval designation.

The project manager will request a CSE applicability review at the earliest practical stage of a project to determine if there could be criticality safety impacts. If the potential exists for the physical configuration or operating parameters for new or revised equipment to affect criticality safety , the drawings and process control plans will be reviewed and approved by a Criticality Safety Representative or Criticality Safety Engineer , in compliance with standard engineering practices and procedures. Facility and process change control will include the following . The change management process will be in accordance with ANSVANS-8.19. All dimensions , nuclear properties , and other features on which reliance is placed will be documented and verified prior to beginning operations , and control will be exercised to maintain them. Changes that involve or could affect nuclear criticality controls will be evaluated under 10 CFR 50.59 , "Changes , Tests , and Experiments." Changes include new designs , operation , or modification to existing SSCs, computer programs, processes , operating procedures , or management measures.

Changes that involve or could affect nuclear criticality controls will be reviewed and approved by the Criticality Safety Representative. Prior to implementing the change , the process will be determined to be subcritical (with an approved margin for safety) under both normal and credible accident scenarios. 6-34

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  • NOITIIW£ST MBMCAl. tsOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Testing and Calibration of Active Engineered Controls
  • Testing and calibration of AECs will be developed for the Operating License Application . Criticality Safety Guidelines for Firefighting
  • Criticality safety guidelines for firefighting will be developed for the Operating License Application.

Emergency Preparedness Plan and Procedures This element describes the response to criticality accidents.

  • * * * * * * * * * * * *
  • The CAAS will be used as described in Section 6.3.1.1 and provides for detection and annunciation of criticality accidents.

Emergency procedures will be prepared and approved by management.

Facility and off-site organizations expected to respond to emergencies will be informed of conditions that might be encountered.

Procedures will: Designate evacuation routes that are clearly identified and follow the quickest , most direct routes practical Include assessment of exposure to individuals Designate personnel assembly stations outside the areas to be evacuated.

A method to account for personnel will be established and arrangements made in advance for the care and treatment of injured and exposed personnel.

The possibility of personnel contamination by radioactive material will be considered . Personnel will be trained in evaluation methods , informed of routes and assembly stations , and drills performed at least annually. Instrumentation and procedures will be provided for determining radiation in an evacuated area following a criticality accident and information collected in a central location.

Emergency procedures will be maintained for each area in which special nuclear material is handled , used , or stored to ensure that all personnel withdraw to an area of safety on sounding the alarm. Emergency procedures will include conducting drills to familiarize personnel with the evacuation plan , designation ofresponsible individuals to determine the cause of the alarm , and placement of radiation survey instruments in accessible locations for use in such an emergency.

The current emergency procedures for each area will be retained as a record for as long as licensed special nuclear material is handled , used , or stored in the area. Superseded sections of emergency procedures will be retained for three years after the section is superseded. Fixed and personnel accident dosimeters will be provided in areas that require a CAAS . Dosimeters will be readily available to personnel responding to an emergency and a method provided for prompt on-site dosimeter readouts. 6-35

..... ; .. NWMI ...*.. ..* ... ....... ' *. * ! ." NCMllTIIWEST MfOICAL tsOTOPU NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features 6.3.1 Criticality Safety Controls The following sections describe criticality safety controls based on information developed for the Construction Permit Application. Section 6.3.1.1 summarizes the results of preliminary CSEs that define PECs and AECs credited to satisfy the double-contingency control principle.

Section 6.3.1.2 summarizes IROFS related to preventing a nuclear criticality identified by the accident analyses described in Chapter I 3.0. 6.3.1.1 Preliminary Criticality Safety Evaluat ion s A series of calculations were performed to support the Construction Permit Application investigating parameters associated with prevention of nuclear criticality in the current equipment configuration of major process systems. The calculations are described in the following documents: * * * *

  • NWMI-2015-CRITCALC-001 , Singl e Param e t e r Sub c riti c al Limit s for 20 wt% 235 U-U ranium M e tal , Uranium Oxid e, and Homog e nou s Wat e r Mixtures NWMI-2015-CRITCALC-002 , I r radiat e d Tar ge t Low-Enri c h e d Uranium Mat e rial Di ss olution NWMI-2015-CRITCALC-003 , 55-Gallon Drum Arra ys NWMI-2015-CRITCALC-005 , Targ e t Fabri ca tion Tanks , W e t Pro cesses, and Storag e NWMI-2015-CRITCALC-006 , Tank Hot Ce ll Calculations were performed using the MCNP 6.1 code (LA-CP-13-00634 , MC N P6 Use r Manual). Validation of the MCNP 6.1 code used in the calculations i s described in [Proprietary Information].

The validation report documents the methodology and results for the bias and bias uncertainty values calculated for homogeneous and heterogeneous uranium systems for the MCNP 6.1 code system. The bias is expressed as USLs calculated using a facility-specific

[Proprietary Information].

The primary focus of the validation was to determine the bias and bias uncertainty for intermediate-enriched uranium (IEU) systems. However , sufficient experiments for low-enriched uranium (LEU) and high-enriched uranium were included to demon s trate that there is no v ariation in the USL with varying enrichment.

Similarly , the primary focus of the validation was on thermal neutron energy systems. Sufficient experiments for intermediate and fast energy experiments were also included to demonstrate that there is no variation in the USL with increasing neutron energy. The purpose of the computer code validation is to determine values of k et rthat are demonstrated to be subcritical (at or below the USL) for areas of applicability similar to systems or operations being analyzed.

The USL is defined by Equation 6-1. USL = 1.0 -Bias -Bias Uncertainty

-Margin of Subcriticality Equation 6-1 [Proprietary Information]

rearranges Equation 6-1 to produce a criterion for model cases that are considered acceptable as s ubcritical , as shown by Equation 6-2 , and incorporate s the margin of s ubcriticality in the USL as required by ANSI/ ANS-8.1. k e ff + (2 X O" c alc) USL where ke ff is the MCNP calculated k-effective and G caJc is the MCNP calcu l ation uncertainty. [Proprietary Information]

6-36 Equation 6-2

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indicates the validation is appropriate for homogeneous and heterogeneous IEU systems. A summary of the area of applicability is provided in Table 6-4. For syste ms outside the validation area of applicability, an increased margin of subcriticality value may be warranted, depending on the specific problem being analyzed.

The analyst must document any extrapolation beyond the validation area of applicability , and justification must be documented for no adjustments to the margin of subcriticality when extrapolating.

Table 6-4. Area of Applicability Summary Parameter Fissile materi a l Fissile material form H/235 U ratio Average neutron energy causing fission E nrichment Moderating materials R eflect in g materials Absorber materials Geometry

  • Source: [Proprietary In format i o n]. ANECF = average neutr on energy causing fission. Area of Applicability

[Proprietary Inform ation] [Proprietary Information]

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The RPF was divided into 13 activity groups for de ve lopment of preliminary CS Es of the activities and associated equipment.

Controlled nuclear criticality safety parameter s vary with the activity group and are summarized in Table 6-5. A minimum of two nuclear criticality safety parameter s are controlled to sat isfy the double-contin ge nc y principle.

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Nuclear parameter

Mass y y y y y y y N y y y b y y Geometry y y y y y y e y e y N y y y y Moderation y N N N N N N N N N N N N Interaction y y y y y y y y N y y y y Volume y y y y y y y N N N y N y Concentration/

N yd yd yd yd N N N y e y e ye N N density R e flection N N N N N N N N N N N N N Absorbers N N N N N N N N N N N N N E nrichment f N N N N N N N N N N N N N

  • D e ri ve d fr o m the indi ca t e d C S E r efe r e nc e d oc um e nt. b Limit ed b y n a ture of p rocess in the a ir filtr a ti o n. c Limit ed b y ta r get d es i gn. d Co ntroll e d thr o u g h input fiss il e m ass. e Limit ed b y t o tal uranium m ass a llow e d in th e sys t e m. r Fa c ili ty li c en se limit e d to ::::2 0 wt% m u. m u uranium-235. NWMI N orthw est M e di ca l I so top es , LL C. CSE = critic a lity sa fety e v a lu a tion. y yes. N = no. The preliminary CSEs define a series of PECs , AECs , and administrative control s that are credited to s ati s fy the double-contingency control principle for pre v ention of nuclear criticality events such that at least two changes in proce s s condition s must occur before criticality i s po s sible. P E C s, AEC s, and administrative controls are described for the 13 activity groups in the followin g referenced tables: * * * * * * * * * * *
  • NWMI-2015-CSE-01 , Irradiat e d Targ e t Handlin g and Di s a sse mbly (Table 6-6) NWMI-2015-CSE-02 , Irradiat e d Low-En r i c h e d U ranium Targ e t Mat e ri a l Di ss olution (Table 6-7) NWMI-2015-CSE-03 , Mol y bd e num-99 R ecovery (Table 6-8) NWMI-2015-CSE-04 , Low-Enri c h e d U ranium Tar ge t Mat e rial Produ ct i o n (Table 6-9) NWMI-2015-CS E-05 , Tar ge t Fabri c ation Ura nium S o luti o n Pro cess e s (Table 6-9) NWMI-20 l 5-CS E-06 , Tar ge t Fini s hin g (T a ble 6-9) NWMI-2015-CSE-07 , Targ et and Can Stora ge and Car ts (Table 6-9) NWMI-20 l 5-CSE-08 , Hot C e ll Uranium Purifi c ation (Table 6-10) NWMI-2015-CSE-09 , Wast e Liquid Pro cess ing (Table 6-11) NWMI-2015-CSE-10 , Solid Wa s t e Coll ec ti o n , En c ap s ulation , and Sta gi n g (Table 6-11) NWMI-2015-CSE-l l , Off g a s and V e ntilati o n (Table 6-12) NWMI-20 l 5-CSE-12 , Targ e t Tran s port Ca s k o r Drum Handling -The s hipping packages dictate design feature s that must be properly implemented for legal over-the-road transport. This CSE does not impose or credit additional passi v e control s other than those already incorporated in the respective shipping packages. 6-38

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  • NWMI-2015-CSE-13 , Analytical Laborato ry (Tab le 6-13) The CSEs will be updated for final design and the Operating License Application. Criticality controls are selected based on the following order of preference:

Passive engineered controls Active engineered contro ls Enhanced administrative controls Administrative controls Note that a number of features li sted in the preliminary CSEs are duplicated in multiple activity groups (e.g., the floor of cells is verified to be flat , with no collect ion points deeper than 3.5 centimeters

[cm]). Duplications are included in the current listings to c learly identify minor dimension variations that may exist in the defined features for different activity groups. Table 6-6. [Proprietary Information]

Double-Contingency Controls Identifier" CSE-0I-PDF1 [Proprietary Information]

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). HEPA = high-efficiency particu l ate a ir. Feature description and basis S PL s in g le parameter limit. 6-39

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6-43

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CSE-05-AC 1 b [Proprietary Information]

CSE-05-AC2b

[Proprietary Information]

CSE-05-AC3b

[Proprietary Information]

CSE-0 6-PDFI c [Proprietary Information]

CSE-06-PDF2c

[Proprietary Information]

CSE-06-ACl c [Proprietary Information]

CSE-06-AC2c

[Proprietary Information]

CSE-0 6-A C3 c [Proprietary Information]

CSE-06-AC4c

[Proprietary Information]

CSE-06-ACSC

[Proprietary Information]

CSE-06-AC6c

[Proprietary Information]

CSE-07-PDFI d [Proprietary Information]

CSE-07-PDF2d

[Proprietary Information]

CSE-07-PDF3d

[Proprietary Information]

CSE-07-PDF4d

[Proprietary Information]

CSE-07-ACl d [Proprietary Information]

CSE-07-AC2d

[Proprietary Information]

CSE-07-A C3 d [Proprietary Information]

CSE-07-AC4d

[Proprietary Information]

CSE-07-AC5d [Proprietary Information]

CSE-07-AC6d

[Proprietary Information]

CSE-07-AC7 d [Proprietary Information]

  • [Propriet a ry Information]

b [Proprietary Information]

c [Propriet a ry Information]

d [Proprietary Information]

ADUN DBE u acid-deficient uranium nitrate. design basi s earthquake. uranium. Feature description and basis UN = urani um ni tride. [Proprietary Inform a tion] [Proprietary Information]

6-44

.; ... ; .. NWMI ...... ..* *.. ....... !.* * *. *

  • NOmfWUT MEDICAi. lS01WEI NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-45

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..*... ..* **: .......... *.* . ' ! ! ." . NOllTHW'EST MlOtCAL lSOl'OP£1 NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-46

...... ;*. NWMI ..*... ..* .... ............ , * ! *. NOlllfWHT MlDtCAl tsOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-47

.: .

.*:.**.*.* . ..............

  • 0 ! ." . NGmfWEST MEDICAl. ISOl'Of'H NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-48

.; ... ; ... NWMI ......... *.* ........ *. ' !*.

  • NORTNWEn lrffOK:Al lSOTOPlS NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-49

.. NWMI ...*.. ..* *... ........ *.* * *.*

  • NOUHWEST MEDICAL ISOTDPD NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [P ro p rietary In format i o n] 6-50

.; ... .. NWMI ...... ..* ... ........ *. * !*.

  • NOllTNWUT llEDtCAl tsOTOPH NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-10. [Proprietary Information]

Double-Contingency Controls (2 pages) ldentifiera Feature description and basis C S E-0 8-PDFl [Propri et a ry Inform a tion] CSE-08-PDF2

[Proprietary Information]

CSE-0 8-PD F3 [Propri etary In fo rm a tion] CSE-08-PDF4 [Proprietary Information]

CSE-0 8-PD F 5 [P ro pri etary In fo rm a ti on] CS E-08-PDF6 [Proprietary Information]

C S E-0 8-PD F 7 [Propri etary In fo rm a tion] CS E-08-PDF8 [Proprietary Information]

C S E-0 8-PD F 9 [Propri e t ary Inform a tion] CSE-08-[Proprietary Information]

PDFlO CSE-0 8-[Propri e t ary I nfo rm a ti o n] PDF!! CS E-08-[Proprietary Information]

PDF12 CSE-0 8-AE F I [Propri e t ary Inform a tion] CSE-08-ACI

[Proprietary Information]

CSE-0 8-AC2 [Propri e t ary Inform a tion] * [P ro pri etary I nfo rm a t io n] DBE = des i gn b as i s earthq u ake. IX io n exc h a n ge. 6-51

..... .. NWMI ...*.. ..* **: ..... .. .. .. ' *. * !

  • NOATIIWUT ll£OICAl tSOTOPD NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-52

.;.-.;:. NWMI ...... .. *.. ........ *.* * *

  • MOmlWEST MEDtCAl. ISOTOPU NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-11. [Proprietary Information]

Double-Contingency Controls (3 pages) Identifier Feature description and basis CSE-09-AEF l" [Propriet ary Information]

CSE-09-AC 1 * [Proprietary Information]

CSE-09-A C2" [Proprietary Information]

CSE-09-AC3" [Proprietary Information]

CSE-10-[Proprietary Information]

PDF l b CSE-10-AEflb [Proprietary Information]

CSE-10-A C 1 b [Propri etary Information]

CSE-10-AC2b

[Proprietary Information]

CSE-10-A C3b [Proprietary Information]

CSE-10-AC4b

[Proprietary Information]

CSE-1 O-AC5 b [Proprietary Information]

CSE-10-AC6b

[Proprietary Information]

CSE-I 0-AC7 b [Proprietary Information]

CSE-10-AC&b

[Proprietary Information]

CSE-1 O-AC9 b [Proprietary Information]

m u HT C RPF * [Proprietary In formation]

b [Proprietary Inform at ion] urani um-235. high-int egr i ty co nt a iner. R a dioi so top e Product i on Fac i lity. S PL u 6-53 sing le parameter limit. uranium.

..

.*:.**.*.* . ..............

  • 0 *. ."
  • NORTHWEST MEDtcAl tsOTOP£S NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-54

.. ; ... .. NWMI ..*... ... **.* ......... *.* . *. * ." NOmfWEST MEDICAi. ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-55

...... NWMI ..*... ..* .... ......... *.* * ' "* NOmfWEST MEOICAl tS01VU NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-12. [Proprietary Information]

Double-Contingency Controls (2 pages) ldentifiera CSE-11-PD F l [Propri e t a ry In fo rm a tion] CSE-l l-PDF2 [Proprietary Information]

C S E-l l-PDF 3 [Propriet a ry Inform a tion] CSE-l l-PDF4 [Proprietary Information]

CSE-11-PDFS (Propriet ary In fo rmation] CSE-I l-PDF6 (Proprietary Information]

C S E-l l-PDF7 [Propri e t ary In fo rm a tion] CSE-l l-PDF8 [Proprietary Information]

CSE-I 1-AEF l (Propri e t a ry Information]

CSE-11-ACl

[Proprietary Information]

  • [Pr o pri etary In fo rm a ti o n] D B E HE P A des i gn b as i s ea rth q u a k e. = hi g h-e ffi cie n cy p a rti c ul ate a i r. Feature description and basis Mo NO x 6-56 m o l y bd e num. nitroge n ox id e.

.; ... ; .. NWMI ...... ..* ... ........... * ! *. * !

  • NOmlWEST MEDICAL ISOTOPU [Proprietary Information]

NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Table 6-13. [Proprietary Information]

Double-Contingency Controls (2 pages) ldentifiera CSE-13-PDF I [Proprietary Information]

CSE-l 3-PDF2 [Proprietary Information]

CSE-J 3-PDF3 [Proprietary Information]

CSE-13-ACI

[Proprietary Information]

CSE-1 3-AC2 [Proprietary Information]

CSE-13-AC3

[Proprietary Information]

CSE-13-A C4 [Proprietary Information]

CSE-13-AC5

[Proprietary Information]

CSE-1 3-AC6 [Proprietary Information]

a [Proprietary Inform ation] R&D RPF r esea rch and d evelop m e nt. = Radioi soto pe Produ ct ion Facility.

Feature description and basis SPL u 6-57 si ngle param e t e r limit. uranium.

..... ; .. NWMI ..*... ..* .... ..... .. .. . . * * !°

  • NOITHWEST llEDtCAl ISOTOPO NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features [Proprietary Information]

6-58

...... ;:. NWMI ...... ... .... ........ *.* ' * . NORTHWEST MED.CAL tSOTOP£S NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Each of the preliminary CSEs indicates that the process areas evaluated will be within the detector and alarm coverage of the CAAS. Evaluation of the CAAS coverage will be performed after final design is complete and prior to facility startup. To ensure the CAAS coverage is adequate for the facility, NWMI will conduct a coverage analysis using the minimum accident of concern that produces a detector response when the dose rate at the detector is equivalent to 20 rad/min at 2 m (6.6 ft) from the reacting material.

Using the source from the minimum accident of concern , NWMI will conduct one-dimensional deterministic computations , when practical , to evaluate CAAS coverage.

For areas of the facility where the use of one-dimensional deterministic computations is not practical , NWMI will use 3D Monte Carlo analysis to determine adequate CAAS coverage.

The CAAS will be designed to meet the following.

  • *
  • 6.3.1.2 The facility CAAS: Will be capable of detecting a criticality that produces an absorbed dose in soft tissue of 20 radiation dose absorbed (rad) of combined neutron and gamma radiation at an unshielded distance of 2 m from the reacting material within 1 minute; two detectors will cover each area needing CAAS coverage Will use gamma and neutron sensitive radiation detectors that energize clearly audible alarm signals if an accidental criticality occurs Will comply with ANSl/ANS-8

.3 , as modified by NRC Regulatory Guide 3.71 Will be appropriate for the type of radiation detected, the intervening shielding , and the magnitude of the minimum accident of concern Will be designed to remain operational during design basis accidents Will be clearly audible in areas that must be evacuated or there will be alternative notification methods that are documented to be effective in notifying personnel that evaluation is necessary Operations will be rendered safe , by shutdown and quarantine , if necessary , in any area where CAAS coverage has been lost and not restored within a specified number of hours. The number of hours will be determined on a process-by-process basis , because shutting down certain processes, even to make them safe , may carry a larger risk than being without a CAAS for a short time. Compensatory measures (e.g., limiting access , halting SNM movement, or restoring CAAS coverage with an alternate instrument) when the CAAS is not functional will be determined for inclusion in the Operating License Application.

Emergency power will be provided to the CAAS by the uninterruptable power supply system . Derived N u clear Criticality Safety Items Relied o n for Safety The following subsections describe engineered safety features that are derived from the accident scenarios that could result in a nuclear criticality, as described in Chapter 13.0. 6.3.1.2.1 IROFS CS-04, Interaction Contro l Spacing Provided by Passively Designe d Fixtures and Wo r kstation Placement IROFS CS-04 , " Interaction Control Spacing Provided by Passively Designed Fixtures and Workstation Placement," is identified by the accident analyses in Chapter 13.0. During handling of uranium solids and solutions outside of processing systems under normal conditions , the material will be handled in safe masses controlled by either physical measurement or batch limits on well characterized devices. 6-59 -----------------------

..... ;* .. *NWMI ...... ...* ... ........ *.* . ' *: ! ." NomtWfST MEDICAL ISOTOH:S NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features Solid uranium will be handled outside of processing systems during: * * * *

  • Receipt and processing of fresh uranium (and presumably shipment of spent uranium back to the supplier)

[Proprietary Information]

Fabrication of targets using [Proprietary Information]

LEU target material (including movement of LEU target material to and from the fabrication workstation and handling of the completed targets) Disassembly of targets following irradiation Laboratory sampling and analysis activities (albeit in smaller quantities) . Each activity is assigned a mass or batch limit for safe handling.

Accident Mitigated The accident occurs when a safe mass or batch limit is exceeded beyond some bounding extent based on the management measures on the control. Note that this accident involves normal condition criticality controlled limits for safe handling, and the upset represents failure of an associated administrative control. The most limiting activity would involve processing the LEU target material from [Proprietary Information].

If the IROFS fails , accidental nuclear criticality is possible without additional control. System Components As a PEC , fixed interaction control fixtures or workstations will be provided for holding or processing approved containers containing approved quantities of uranium metal , [Proprietary Information], batches of targets , and batches of samples. Functional Requirements The fixtures are designed to hold only the approved container or batch and are fixed with 2-ft edge spacing from all other fissile material containers , workstations , or fissile solution tanks, vessels, and ion exchange (IX) columns. Where LEU target material is handled in open containers , the design will prevent spills from readily spreading to an adjacent workstation or storage location. Design Basis Final workstation and fixture spacing will be determined in final design when all process upsets are evaluated.

Workstations with interaction controls include the following (not an all-inclusive listing):

  • * * [Proprietary Information]

[Proprietary Information]

Target basket fixture that provides safe spacing of a batch of targets from one another in the target receipt cell Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6-60

.; ... NWMI ...... ..* ... ..... .... .. * *. * ! . NOtffHWfn MEOtCAl ISOTOPO NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.3.1.2.2 IROFS CS-06, Pencil Tank, Vessel, or Piping Safe Geometry Confinement Using the Diameter of Tanks, Vessels, or Piping IROFS CS-06 , "Pencil Tank , Vessel, or Piping Safe Geometry Confinement using the Diameter of Tanks , Vessels , or Piping," is identified by the accident analyses in Chapter 13.0. The PHA in Chapter 13.0 identified a number of individual potential initiating events that could lead to a spill of fissile solution from the geometrically safe confinement tanks , vessels , or piping that provide the primary safety functions of the processes.

Four processing systems will handle fissile solutions:

Target fabrication (from the [Proprietary Information])

Target dissolution system First stage of molybdenum recovery and purification Entire uranium recovery and recycle sy s tem Three of these systems will be at least partially located within the hot cell wall boundary due to the dose of the fission products.

Initiating events include the general categories of tank , vessel , or piping failure due to operator error (valves out of position), valves leaking , equipment leaking (pumps , piping , vessels , etc.), high pressure events from various causes including high temperature solutions (locked in boundary valves), hydrogen detonation , and exothermic reactions with the wrong resins or reagents used in the respective systems. Some of the initiators result in small leaks that are identified and mitigated (e.g., pump seal and small valve leaks). Over the life of the facility , these types of leaks are to be expected , but do not challenge the overall safety of RPF operations.

Accident Mitigated The accident of concern involves fissile process solution in quantities necessary to sustain accidental nuclear criticality.

Larger catastrophic leaks or ruptures of equipment must occur for enough material to be released. Such leaks would represent a failure of the safe-geometry confinement IROFS for the respective equipment.

Thus , scenarios leading to this accident sequence involve the failure of these IROFS. Due to the nature of the process , the worst-case accident involves the tanks with the largest capacity and the highest normal case concentrations.

System Components As a PEC , pencil tanks and other standalone vessels are designed and will be fabricated with a geometry diameter for safe storage and processing of fissile solutions.

The safe diameters of various tanks , vessels , or components will be provided in the Operating License Application.

Functional Requirements The safety function of safe diameter vessels is also one of confinement of the contained solution.

The safe-geometry confinement of fissile solutions will prevent accidental nuclear criticality, a high consequence event. The safe-geometry confinement diameter will conservatively include the outside diameter of the tank wall or out to the outside diameter of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the vessels. Where insulation is used on the outside wall of a vessel , the insulation will be closed foam or encapsulated type (so as not to soak up solution during a leak) and will be compatible with the chemical nature of the contained solution. Design Basis The safe-geometry diameter of tanks , vessels , and piping will be determined in final design after finalizing the reference CSEs. Note that preliminary vessel sizes for activity groups are listed in the double-contingency parameters described in Section 6.3.1.1. 6-61

.; ... .. NWMI ......... *.* ..... .. .. ..

  • 0 * ! : . NatlTHWEST MEDtw. lSOTDPU Test Requirements NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.3 IROFS CS-07, Pencil Tank Geometry Control on Fixed Interaction Spacing of Individual Tanks IROFS CS-07 , "Pencil Tank Geometry Control on Fixed Interaction Spacing of Individual Tanks ," is identified by the accident analyses in Chapter I 3.0 (see description in Section 6.3.1.2.2). Accident Mitigated See description in Section 6.3.1.2.2. System Components As a PEC , pencil tanks and other standalone vessels (controlled with s afe geometry or volume constraints) are designed and will be fabricated with a fixed interaction spacing for safe storage and processing of the fissile solutions.

Tanks, vessels, and components requiring fixed interaction control spacing of the barrels within each set of pencil tanks and between various tanks , vessels , or components will be provided in the Operating License Application.

Functional Requirements The safety function of fixed interaction spacing of individual tanks in pencil tank s and between other single processing vessels or components is designed to minimize interaction of neutrons between vessels such that under normal and credible abnormal process upsets , the systems remain subcritical.

The fixed interaction control of tanks , vessels , or components containing fissile solutions will prevent accidental nuclear criticality , a high consequence event. The fixed interaction spacing will be measured from the outside of the respective tanks, vessels, or component or from the outside of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the vessels or component.

The fixed interaction control distance from the safe slab depth spill containment berm will also be specified where applicable.

Design Basis Actual interaction control parameters will be defined during final design. In addition , the following generic interaction control parameters apply during design. *

  • Connecting piping between fissile material components will not exceed a cross-sectional density to be determined during final evaluation of systems. Edge-to-edge spacing between fissile material-bearing vessels and components and the concrete reflector presented by the hot cell shielding walls will be fixed at a distance to be determined during final evaluation of all components. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application. 6-62

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  • NOmlWEST 11EDtCA1 ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.3.1.2.4 IROFS CS-08, Floor and Sump Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Dikes IROFS CS-08 , "Floor and Sump Geometry Control on Slab Depth, Sump Diameter or Depth for Floor Dikes," is identified by the accident analyses described in Chapter 13.0 (see description in Section 6.3.1.2.2). Accident Mitigated See description in Section 6.3.1.2.2. System Components As a PEC, the floor under designated tanks , vessels , and workstations will be constructed with a spill containment berm using a safe-geometry slab depth , and one or more collection sumps with diameters or depths, to be determined in final design. Functional Requirements The safety function of a spill containment berm is to contain spilled fissile solution from systems overhead and prevent an accidental nuclear criticality if one of the tanks or related piping leaks , ruptures , or overflows (if so equipped with overflows to the floor). Each spill containment berm will be sized for the largest single credible leak associated with overhead systems. The sump will have a monitoring system to alert the operator that the IROFS has been used and may not be available for a follow-on event. A spill containment berm is operable if it contains reserve volume for the largest single credible spill. Spill containment berm sizes and locations will be determined during final design. Design Basis The safe-geometry slab depth under designated tanks , vessels , and workstations will be determined during final design after finalizing the reference CSEs. Note that the preliminary slab depth for the activity groups are listed in the double-contingency parameters described in Section 6.3.1. l. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.5 IROFS CS-09, Double-Wall Piping IROFS CS-09 , "Double Wall Piping," is identified by the accident analyses described in Chapter 13.0. As a PEC , a piping system for conveying fissile solution between confinement structures will be provided with a double-wall barrier to contain any spills that may occur from the primary piping. Accident Mitigated

  • Leak in piping that passes between confinement enclosures 6-63

.; ... NWMI ...... ..* **: ..... .. .. . . * *: 0 NORTHW'EIT MEDICAL ISOTOf'll System Components NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features IROFS CS-09 is used at the locations listed below that pass through the facility where creating a spill containment berm under the piping is neither practical nor desirable for personnel chemical protection purposes. The following double-wall piping segments are identified for criticality safety: * *

  • Transfer piping containing fissile solutions traversing between hot cell walls Transfer piping connecting the uranium product transfer send tank (UR-TK-720) and the uranyl nitrate storage tank (TF-TK-200)

Any other locations in final design where fissile solution piping exits a safe-slab spill containment berm and enters another Functional Requirements The safety function of this PEC is to safely contain spilled fissile solution from system piping and prevent an accidental nuclear criticality if the primary confinement piping leaks or ruptures.

The double-wall piping arrangement will maintain the safe-geometry diameter of the solution.

The double-wall piping will also function as a barrier to prevent fissile solution from soaking into the concrete from lines passing through concrete walls where required by the criticality safety analysis (e.g., see PDF2 of Table 6-9). The secondary safety function of double-wall piping is to prevent personnel injury from exposure to acidic or caustic licensed material solutions that are conveyed in the piping. Design Basis The double-wall piping arrangement is designed to gravity-drain to a safe-geometry set of tanks or a geometry containment berm. Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.6 IROFS CS-10, Closed Safe Geometry Heating/Cooling Loop with Monitoring and Alarm IROFS CS-10, "Closed Safe Geometry Heating or Cooling Loop with Monitoring and Alarm ," is identified by the accident analyses in Chapter 13.0. As a PEC , a closed-loop , safe-geometry heating or cooling loop with monitoring for uranium process solution or high-dose process solution will be provided to safely contain fissile process solution that leaks across the heat transfer fluid boundary if the primary boundary fails. Accidents Mitigated The dual-purpose safety function of the closed-loop system is to prevent (I) fissile process solution from causing accidental nuclear criticality , and (2) high-dose process solution from exiting the hot cell containment , confinement , or shielded boundary (or to prevent low-dose solution from exiting the facility, for systems located outside of the hot cell containment , confinement , or shielded boundary), and causing excessive dose to workers and the public , and/or causing a release to the environment.

System Components The closed loop steam and cooling water loop design is described in Chapter 9.0. 6-64

.; ... ;. NWMI ...... .. ... ........ *.* * ! *. * ! : NOmfWEIT MEDICAL ISOTOPH Functional Requirements NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The heat exchanger materials will be compatible with the har s h chemical environment of the tank or vessel process (this may vary from application to application).

Sampling of the heating or cooling media (e.g., steam condensate conductivity , cooling water radiological activity , or uranium concentration) will be conducted to alert the operator that a breach ha s occurred , and that additional corrective actions are required to identify and i so late the failed component and restore the closed loop integrity.

Dis c harged solutions from thi s system will be handled as potentially fissile and sa mpled prior to discharge to a safe geometry.

Design Basis The closed loop s team and coo ling water loop de s ign is de scri bed in Chapter 9.0. Test Requirements The above analysis is ba se d on information developed for the Construction Permit Application. Additional detailed information on test requirements wi ll be developed for the Operating License Application. 6.3.1.2.7 IROFS CS-11, Simple Overflow to Normally Empty Safe Geometry Tank with Level Alarm IROFS CS-I I , "Simple O ver flow to Normally Empty Safe Geometry Tank with Level Alarm," is identified by the accident analyses described in Chapter 13.0. As a PEC , a simple overflow line will be installed below the level of the process vessel ventilation port and any chemical addition port s (where an anti-siphon safety feature will be installed) for each vente d tank containing fissile or potentially fissile process solution for which this IROFS is assigned. Accident Mitigated The overflow drain will prevent the proce ss solution from entering the respective non-geometrically favorable sections of the process ventilation system and any chemical addition ports (where chemical addition port s enter throu gh a nti-siphon de v ices). System Components Locations of the overflow and overflow collection tanks will be pro vi ded with the final design. Functional Requirements The safety function of thi s feature is to pre vent accidental nuclear criticality in non-geometricall y favorable sections of the process ventilation sys tem. The overflow will be directed to a safe-geometry storage tank. The overflow storage tank will normally be maintained empty. The overflow storage tank will be equipped with a le ve l alarm to inform the operator when use of the IROFS has been initiated , so that actions can be taken to restore operability of the safety feature by emptying the tank. Design Basis De s ign basis information will be provided in the Operating License Application.

6-65

.. NWMI ...*.. ** :.*.< ' *. ! ." . NORTHWUT MEOtCAl ISOTOKS Test Requirements NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.8 IROFS CS-12, Condensing Pot or Seal Pot in Ventilation Vent Line IROFS CS-12, "C ondensing Pot or Seal Pot in Ventilation Vent Line," is identified by the accident analyses described in Chapter 13.0. As a PEC , a safe-geometry condensing pot or seal pot will be installed downstream of e ach t a nk for which this IROFS is assigned to capture and redirect liquid s to a safe-geometry tank or flooring area with safe-geometry sumps. One s uch condensing or seal pot may service several related tanks within the sa fe-geometry boundary of the ventilation system. The condensing or seal pot will prevent fissile solution from flowing into the respective geometrically favorable process ventilation system by directing the solution to a safe-geometry tank or flooring area with safe-geometry sumps. Accident Mitigated Where independent seal or condensing pot s are credited, the drain s of the seal or condensing pot s must be d i rected to independent locations to prevent a common clog or over-capacity condition from defeating both. System Components Locations of the condensing pots or seal pots and associated drain points will be provided with the final design. Functional Requirements The safety function of the condensing or seal pots is to prevent accidental nuclear criticality in geometrically favorable sections of the process ventilation system. The safe-geometry tank or sumps will be equipped with a level alarm to inform the operator when use of the IROFS has been initiated. Each individual tank or vessel operation must be evaluated for required overflow capacity to ensure that a su itable overflow volume is available. A monitoring and alarm circuit will be pro vi ded so that common overflow tanks or safe slab flooring or sumps can be used for multiple tanks or vessels , and limiting conditions of operation will be defined to ensure that the IROFS is made available in a timely manner or operations are suspended following an overflow event of a single tank. Design Basis Design basis information will be provided in the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6-66

... NWMI ...*.. ..* .... ........ *.* * ' *.

  • NORTHWEST MEDtcAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features 6.3.1.2.9 IROFS CS-13, Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary IROFS CS-13 , " Simple Overflow to Normally Empty Safe Geometry Floor with Level Alarm in the Hot Cell Containment Boundary ," is identified by the accident analyses described in Chapter 13.0. As a PEC , a simple overflow line will be installed above the high alarm setpoint for each vented tank containing fissile or potentially fis s ile process solution for which this IROFS is assigned.

The overflow will be directed to one or more safe-geometry flooring configurations with safe-geometry sumps. Accident Mitigated This IROFS prevents accidental criticality by ensuring that overflowing fi ssi le sol utions are captured in a safe-geometry slab configuration with safe-geometry s umps. System Components System component information will be provided in the Operating License Application.

Functional Requirements The floor areas (separated as needed to s upport operations in different hot cell areas) will normally be maintained empty. The floor area(s) will be equipped with a s ump level alarm to inform the operator when use of the IROFS has been initiated. Design Basis Design ba sis information will be provided in the Operating License Application.

Test Requirements The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirement s will be developed for the Operating Licen se Application.

6.3.1.2.10 IROFS CS-14, Active Discharge Monitoring and Isolation IROFS CS-14, "Ac tive Di sc harge Monitoring and Isolation ," is identified by the accident analyses described in Chapter 13.0. Additional detailed information describing active di sc harge monitoring and isolation will be developed for the Operating License Application.

System Components System component information will be provided in the Operating License Application.

Functional Requirements Functional requirements information will be provided in the Operating License Application. Design Basis Design basi s information will be provided in the Operating License Application.

6-67

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ISOTOPlS Test Requirements NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The above analysis is based on information developed for the Construction Permit Application.

Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.11 IROFS CS-15, Independent Active Discharge Monitoring and Isolation IROFS CS-15 , " Independent Active Discharge Monitoring and Isolation ," is identified by the accident analyses described in Chapter 13.0. Additional detailed information describing independent active discharge monitoring and isolation will be developed for the Operating License Application. System Components System component information will be provided in the Operating License Application. Functional Requirements Functional requirements information will be provided in the Operating License Application.

Design Basis Design basis information will be provided in the Operating License Application. Test Requirements The above analysis is based on information developed for the Construction Permit Application. Additional detailed information on test requirement s will be developed for the Operating License Application. 6.3.1.2.12 IROFS CS-18, Backflow Prevention Device IROFS CS-18 , " Back.flow Preventions Device," is identified by the accident analyses described in Chapter 13.0. See description in Section 6.2.1.7.9. Accident Mitigated See description in Section 6.2.1.7.9. System Components See description in Section 6.2.1. 7.9. Functional Requirements See description in Section 6.2.1.7.9.

Design Basis See description in Section 6.2.1. 7.9. 6-68

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  • NotmfWEST umtCAl lSOTO'U Test Requirements See description in Section 6.2.1.7.9. 6.3.1.2.13 IROFS CS-19, Safe-Geometry Day Tanks NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features IROFS CS-19 , "Safe Geometry D ay Tanks ," is identified by the accident analyses described in Chapter 13.0. See de scription in Section 6.2.1.7.9. Accident Mitigated See description in Section 6.2.1.7.9. System Components See description in Section 6.2.1.7.9. Functional Requirement s See description in Section 6.2.1.7.9.

Design Basis See description in Section 6.2.1.7.9. Test Requirements See description in Section 6.2.1.7.9. 6.3.1.2.14 IROFS CS-20 , Evaporator/Concentrator Condensate Monitoring IROFS CS-20 , " Evaporator

/Conce ntrator Condensate Monitoring

," is identified by the accident a n a l yses described in Chapter 13.0. As an AEC , the condensate tanks will use a continuous active uranium detection system to detect high carryover of uranium that s hut s down the evaporator feeding the tank. The purpose of this system is to (I) detect an anomaly in the evaporator or concentrator indicating high uranium content in the condenser (due to flooding or excessive foaming), and (2) prevent high concentration uranium solution from being available in the condensate tank for discharged to a favora ble geometry system or in the condenser for l eaking to the non-safe geometry coo ling loop. Accident Mitigated The s afety function of this IROFS is to prevent an accidenta l nuclear critica lit y because of excessive uranium in the condensate carryover to a non-geometrically favorable waste collection tank. System Components System components consist of: Condensate samp l e tank 1 A Condensate dela y tank 1 Condensate sample tank 1 B Condensate samp l e tank 2A

  • Condensate del ay tank 2 Condensate sample tank 2B Condensate sampling systems Condensate monitors (U R-TK-340) (UR-TK-360) (UR-TK-370) (UR-TK-540) (UR-TK-560) (UR-TK-570) 6-69

..... NWMI ..*... ..* *.. **** 0 :.!! *. 0 NORTHWEn MEOK:Al tSOTOPEI Functional Requirements NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The detection system works by continuously monitoring condensate uranium content and detecting high uranium concentration, and then shutting down the evaporator to isolate the condensate from the condenser and condensate tank. At a limiting setpoint, the uranium monitor detecting device will close an isolation valve in the inlet to the evaporator (or otherwi se secures the evaporator) to stop the discharge of high uranium content solution into the condenser and condensate collection tank. The uranium monitor is designed to produce a valve-open permissive signal that fails to an open state, closing the valve on loss of electrical power. The isolation valve is designed to fail-closed on loss of instrument air, and the solenoid is designed to fail-closed on loss of signa l. Locations where these IROFS are u sed will be determined during final design. Design Basis Design basis information will be provided in the Operating License Application. Test Requirements The above analysis i s based on information developed for the Construction Permit Application. Additional detailed information on test requirements will be developed for the Operating License Application.

6.3.1.2.15 IROFS CS-26, Processing Component Safe Volume Confinement IROFS CS-26 , " Processing Component Safe Volume Confinement," is identified by the accident analyses described in Chapter 13.0 (see description in Section 6.3.1.2.2).

Accident Mitigated See description in Section 6.3.1.2.2. System Components As a PEC , some processing components (e.g., pump s, filter housings , and IX columns) will be controlled to a safe volume for safe storage and proce ssi ng of the fissile solutions.

Components that may be controlled to a safe volume will be described in the Operating License Application. Functional Requirements The safety function of a safe-volume component is also one of confinement of the contained solution.

The safe-volume confinement of fissile solutions will prevent accidental nuclear criticality , a consequence event. The safe-volume confinement will conservatively include the outside diameter of any heating or cooling jackets (or any other void spaces that may inadvertently capture fissile solution) on the component.

Where insulation is used on the outside wall of the component , the insulation will be foam or encapsulated type (so as not to soak up solution during a leak) and will be compatible with the chemical nature of the contained solution. Design Basis The safe-volume confinement components will be determined in final design after finalizing the referenced CSEs. 6-70

.. .-.;* .. NWMI ..**.. .. **.* ........ *.* ' * ' NORTMWUT 110HCAL lSOTOPU Test Requirements NWMl-2013-021, Rev. 1 Chapter 6.0 -Engineered Safety Features The above analysis is based on information developed for the Cons truction Permit Application.

Additional detailed information on test requirements wi ll b e developed for the Operatin g License Application.

6.3.1.2.16 IROFS CS-27 , Closed Heating or Cooling Loop with Monitoring and Alarm IROFS CS-27 , " Closed Heating or Cooling Loop with Monitoring and Alarm ," is identified b y the accident analyses in Chapter 13.0. As a PEC , closed cooling water loops with monitoring for breakthrough of proce ss solution will be provided on the evaporator or concentrator condensers to contain process solution that leak s across this boundary , if the boundary fails. This IROFS will be applied to those high-heat capacity cooling jackets (requiring very large loop heat exchangers) servicing condensers where the leakage is always from the cooling loop to the condenser.

The inherent characteristics of the leak path will reduce back-leakage into the closed loop syste m , and the risk of product solutions entering the condenser will be very low by evaporator and concentrator de sign. System Components The purpose of this safety function i s to monitor the he a lth of the condenser cooling jacket to ensure that in the unlikel y event that a co ndenser overflow occurs , fissile and/or high-dose process solution will not flow into thi s non-safe-geo metry cooling loop and cause nuclear criticality.

The closed loop will also isolate any high-dose fis s ile product solids , from the same event, from penetratin g the hot cell s hielding boundary , and any high-do se fission gases from penetratin g the hot cell shielding boundary during normal operations.

Functional Requirements The heat exchanger materials will be compatible with the harsh chemical environment of the tank or vesse l proce ss (t hi s may vary from application to application).

Sampling of the cooling media (e.g., coo lin g water radiological activ ity , or uranium concentration) will be conducted to alert the operator that a breach ha s occurred, and that a dditional co rr ect i ve actions are required to identify and i so late the failed component and re s tore the c lo se d-loop integrity.

Closed-loop pres sure will also b e monitored to identify a leak from the closed loop to the proc ess sys tem. Di sc har ge d so lutions from this sys tem will be handled as potentiall y fissile and sa mpled prior to discharge to a non-safe geometry.

Design Basis De sign basi s information will be provided in the Operatin g Licen se Application.

Test Requirements The above ana l ysis i s ba sed on information develop e d for the Construction Permit Application.

Additional detailed information on test requirements wi ll be developed for the Operating License Application.

6.3.2 Surveillance

Requirements A review of surveillance requirements to ensure the ava ilability and reliability of safe ty controls when required to perform safe ty functions will be included in the Operating License Application.

6.3.3 Technical

Specifications The technical s pecifications w ill be pro v ided in the Operating Licen se Application.

6-71

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  • NOUKWHT MEDM:Al ISOTOftlS

6.4 REFERENCES

NWMl-2013-021 , Rev. 1 Chapter 6.0 -Engineered Safety Features I 0 CFR 20, "Standards for Protection Against Radiation," Code of Federal Regulations, Office of the Federal Register, as amended. I 0 CFR 20.1201, "Occ upational Dose Limits for Adults," Code of Federal Regulations, Office of the Federal Register, as amended. IO CFR 20.1301, "Dose Limits for Individual Members of the Public," Code of Federal Regulations, Office of the Federal Register, as amended. 10 CFR 50.59, "Changes, Tests , and Experiments," Code of Federal Regulations , Office of the Federal Register, as amended. IO CFR 70.61, "Performance Requirements," Code of Federal Regulations, Office of the Federal Register, as amended. ANSI/ ANS-8.1, Nuclear Criticality Safety in Operations with Fissionable Material Outside of Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2014. ANSI/ ANS-8.3, Criticality Accident Alarm System, American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois , I 997 (Reaffirmed in 2012). ANSI/ ANS-8. 7, Nuclear Criticality Safety in the Storage of Fissile Materia l s, American National Standards Institute/American Nuclear Society, La Grange Park, Illinoi s, 1998 (Reaffirmed in 2007). ANSI/ ANS-8.10, Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement, American National Standards Institute/American Nuclear Society, La Grange Park, Illinoi s, 2015. ANSI/ANS-8.19, Administrative Practices for Nuclear Criticality Safety, American National Standards Institute/American Nuclear Society, La Grange Park , Illinois, 2014. ANSI/ ANS-8.20, Nuclear Criticality Safety Training, American National Standards Institute/ American Nuclear Society, La Grange Park , Illinois, 1991 (Reaffirmed in 2005). ANSI/ANS-8.22 , Nuclear Criticality Saf e ty Based on Limiting and Controlling Moderators , American National Standards Institute/ American Nuclear Society, La Grange Park, Illinois, 1997 (Reaffirmed in 2011 ). ANSI/ ANS-8.23, Nuclear Criticality Accident Emergency Planning and Response , American National Standards Institute/American Nuclear Society, La Grange Park , Illinoi s, 2007 (Reaffirmed in 2012). ANSI/ANS-8.24, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, American National Standards Institute/ American Nuclear Society , La Grange Park , Illinois, 2007 (Reaffirmed in 2012). ANSI/ANS-8.26, Criticality Safety Engineer Training and Qualification Program , American National Standards Institute/ American Nuclear Society , La Grange Park, Illinoi s, 200 7 (Reaffirmed in 2012). ANSI/ ANS-15 .1 , The Development of Technical Specifications for Research Reactors, American National Standards Institute/American Nuclear Society, LaGrange Park, Illinois, 2013. ANS I Nl 3.1, Sampling and Monitoring Releases of Airborne Radioacti ve Substances from the Stacks and Ducts of Nuclear Facilities, American Nuclear Society, La Grange Park, Illinois, 2011. 6-72

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  • NOITifWEn MEIUCAL. 1$0TOP£S NWMl-2 0 13-02 1, R e v. 1 Ch apter 6.0 -E n gineered Safety Featu r e s ASME AG-1, Code on Nuclear Air and Gas Treatment, American Society of Mec h anica l Engineers, New York , New York, 2003. LA-CP-13-00634, MCNP6 User Manual , Rev. 0, Los A l amos Nationa l Laboratory, Los A l amos , New Mexico, May 2013. NRC , 2012, Fina l Interim Staff Guidance Augmenting NUREG-1537, "Guidelines for Preparing and R ev iewing Applications for the Licensing of Non-Power Reactors , " Parts 1 and 2, for Licensing Radioisotope Production Facilities and Aqueous Homogeneous Reactors, Docket Number: NRC-201 1-0135, U.S. Nuc l ear Regu l atory Commission, Washington , D.C., Octo b er 30, 20 1 2. NUREG-1520 , Standard R e view Plan for the Review of a Licens e Application for a Fuel Cycle Facility , Rev. 1, U.S. Nuclear Regulatory Commission , Office of Nuc l ear Material Safety and Safeguards , Washington , D.C., May 2010. NUREG-153 7 , Guidelines for Preparing and Revi ewi ng Applications for the Lic e nsing of Non-Power Reactors -Format and Content , Part 1 , U.S. Nuc l ear Regulatory Commission, Office of Nuclear Reactor Regulation , Was h ington , D.C., February 1996. NUREG/CR-4604 I PNL-5849, Statistical Methods for Nuclear Material Manag e ment , Pacific Northwest Laboratory, Richland, Wash i ngto n , December , 1988. NUREG/CR-6698, Guide for Validation of Nuclear Criticality Saf ety Calculational Methodology , U.S. Nuclear Regulatory Commission, Office of Nuclear Materia l Safety and Safeguards, Washington , D.C., January 2001. [Proprietary Information]

[Proprietary Information]

NWMI-2015-SD D-013, S yste m Design D esc ription for V e ntilation, Rev. A, Northwest Medical Isoto p es , LLC, Corva ll is , Oregon , 2015. NWMI-2015-CRITCALC-001, Sing l e Param e ter Sub c riti c al Limits for 20 wt% 235 U-Uranium Metal , Uranium Oxide , and Homogenous Water Mixtures , Rev. A , Northwest Medica l Isotopes , LLC , Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-002, Irradiated Target Low-Enriched Uranium Material Dissolution , Rev. A Northwest Medical Isotopes , LLC , Corva ll is , Oregon , 2015. NWMI-2015-CRITCALC-003, 55-Gallon Drum Arrays, Rev. A Northwest Medical Isotopes, LLC, Corvallis , Oregon , 2015. NWMI-2015-CRITCALC-005 , Target Fabrication Tanks , W e t Pro cesse s , and Storage , Rev. A , Northwest Medical Isotopes , LLC , Corvallis, Oregon , 2015. NWMI-2015-CRITCALC-006, Tank Hot Ce ll , Rev. A , Northwest Medica l Isotopes , LLC , Corva ll is , Oregon, 2015. NWM I-2015-CSE-001, NWMI Preliminary Criticality Safety Evaluation:

Irradiated Target Handling and Disassemb l y, Rev. A, Northwes t Medical Isotopes , LLC , Corva ll is, Oregon, 2015. NWM I-2015-CSE-002, NWMI Preliminary Criticality Safety Evaluation:

Irradiated Low-Enriched Uranium Targ e t Materia l Dissolution, Rev. A, Nort h west Medical I sotopes, LLC, Corva ll is , Oregon , 20 1 5. NWMI-2015-CSE-003, NWMI Preliminary Criticality Safety Evaluation:

Mo l y bdenum-99 R ec overy, Rev. A , Northwest Me d ica l Isotopes , LLC , Corva ll is, Oregon , 20 1 5. 6-73

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MEOICAl. ISOTOPH NWMl-2013-02 1, Rev. 1 Chapte r 6.0 -Engineered Sa f ety Features NWMI-2015-CSE-004 , N WMI Preliminary Criti c ality Saf e ty Evaluati o n: Low-Enri c h e d Uranium Target Material Produ c tion , Rev. A , Northwest Medica l I s otopes , LLC , Corva ll i s, Oregon , 2015. NWMI-2015-CSE-005 , NWMI Pr e l imina ry Criti c ali ty Saf ety Evaluation

Targ e t Fabrication Uranium So l ution Pro ce ss es, Rev. A , Northwest Medical Isotopes , LLC , Corva ll is , Oregon , 2015. NWMI-2015-CSE-006 , NWMI Prelimina ry Criti c ality Saf e ty Evaluation:

Targ e t Finishing , R ev. A , Northwest Med i ca l I s otopes , LLC , Corva ll is , Oregon, 2015. NWMI-2015-CSE

-00 7 , N WMI Pre l iminary Criti c ality Saf ety Eva l uati o n: Targ e t and Can Stora ge and C art s, Re v. A , Northwest Medical I s otopes , LLC , Corva ll is , O r egon , 2015. NWMI-20 1 5-CSE-008, N WMI Pr e limin ary C riti c ali ty S a f ety Evaluation:

Hot C e ll Uranium Purifi c ation , Rev. A , Nort h west Medical I so t opes , LLC , Corva ll is , Oregon , 2015. NWMI-20 1 5-CSE-009 , N W M I Pr e limin ary Criticali ty S a f ety E v a l uati o n: Liquid Wa s t e Proc ess in g, Rev. A , Nort h west Medica l Isotope s, LLC , Cor v a lli s , Oregon , 2015. NWMI-2015-CSE-010 , NW MI Pr e limina ry C riti c ali ty Saf ety Eva l uati o n: Solid Wa s t e Co ll ec ti o n , En c ap s ulation, and S ta g ing , Rev. A , Nort h west Medica l Isotopes , LLC , Corva ll is , O r egon , 2015. NWMI-2015-CSE-Ol I , NWMI Pr e limina ry Criti c a l i ty S a f ety Evaluati o n: Off g a s and V e nti l ati o n , Rev. A , Northwest Medica l Isotope s, LLC , Corv a ll i s , Oregon , 20 1 5. NWMI-20 1 5-CSE-O 12 , NWMI Pr e limina ry Criti c ality Saf ety E v aluati o n: Targ e t Tran s port Ca s k or Drum Handling , Re v. A , Northwe s t Medical I s otopes , LLC , Corva ll is , Oregon , 20 1 5. NWMI-20 1 5-CSE-O l 3 , N WMI Pr e l imina ry C riti c ali ty Saf ety E v aluati o n: Anal y t ic al Laborat ory, Rev. A , Northwest Medical I s otope s, LLC , Corva ll is , Orego n , 20 1 5. R e gu l atory Guide 3.71 , N u cle ar Cr i ti c al ity Saf ety St a ndard s for Fu e l s and Mat e ri a l Fa c iliti es, Rev. 2 , U.S. Nuclear Regul a tory Commi s sion , Wa s hington , D.C., Decem b er 2010. 6-7 4

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  • _-. NORTHWEST MEDICAL ISOTOPES *
  • Chapter 7.0 -Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility Prepared by: NWMl-2013-021, Rev. 1 June 2017 Northwest Medical Isotopes, LLC 815 NW gth Ave, Suite 256 Corvallis, OR 97330 This page intentionally left blank.

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  • NOITHWlST MEmtAl ISOTOP£S NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control System s Chapter 7.0 -Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility NWMl-2013-021, Rev. 1 Date Published:

June 26, 2017 Document Number: NWMl-2013-021 I Revision Number. 1 Title: Chapter 7.0 -Instrumentation and Control Systems Construction Permit Application for Radioisotope Production Facility Approved by: Carolyn Haass Signature:

.. NWMI ...... ..* *.. ........ *.* ' * ." NOITHWHT MlDICAL ISOTIWI NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems This page intentionally left blank.

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  • NOR11fWEST MEDtcAl lSOTOPH Rev Date 0 6/29/2015 1 6/26/2017 NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems REVISION HISTORY Reason for Revision Revised By Initial Application Not required Incorporate changes based on responses to C. Haass NRC Requests for Additional Information

.; ... NWMI ........... ..... *. NOmfWlSTllEDfCAltsOTOf'(S NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems This page intentionally left blank.

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  • NORTHWESTMHHCAllSOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems CONTENTS 7.0 INSTRUMENTATION AND CONTROL SYSTEMS ................................................................. 7-1 7.1 Summary Description

..............

.......................................

...............

..................................... 7-1 7 .2 Design of Instrumentation and Control Systems ..........................

...........................

........... 7-4 7.2.1 Design Criteria ..................................................................................................... 7-4 7.2.2 Design Basis and Safety Requirements

................................................................ 7-4 7.2.3 S y stem De s cription ..........................................

...................

................

............... 7-1 3 7.2.3.1 Facilit y Process Control Sy s t e m ....................................................... 7-14 7.2.3.2 Engineered Safety Feature Actuation Sy s tems ........................

......... 7-14 7.2.3.3 Control Room/Human-Machine Interface De sc ription ..................... 7-14 7.2.3.4 Building Management Sy s tem .......................................................... 7-15 7.2.3.5 Fire Protection System ...........................................

........................... 7-1 5 7.2.3.6 Fa c ility Communication Sy s tem s ............................

..................

........ 7-15 7.2.3.7 Analytical Laboratory System .......................

.................................... 7-16 7.2.4 S ys tem Performance Analysi s ............................................................................ 7-16 7.2.4.1 Facilit y Trip and Al a rm Design Ba s is .......................................

........ 7-16 7.2.4.2 Anal ys is ...............

..............................................................................

7-17 7.2.4.3 Conclusion

.................................

....................................................... 7-1 8 7.3 Proces s Control System s .......................................

........................................................... 7-2 2 7.3.1 Uranium Reco v ery and Recycle Sy s tem ............................................................ 7-2 2 7.3.1.1 De s ign Criteria ............................................................................

...... 7-23 7.3.1.2 De s ign Basi s and S a fety Requirement s ............................................. 7-23 7.3.1.3 System Description

.............

..............................................................

7-23 7.3.1.4 Sy s tem Performance Analy s is and Conclusion

................................. 7-28 7.3.2 Targ e t Fabrication System ...........

...................................................................... 7-2 8 7.3.2.1 De s ign Criteria .......................................................

........................... 7-29 7.3.2.2 De s ign Basi s and S a fety Requir e ments ............................................. 7-29 7.3.2.3 System Description

...........

............................................................

.... 7-29 7.3.2.4 System Performan ce Analysi s and Conclusion

..............

................... 7-3 2 7.3.3 Target Receipt a nd Disassembl y Sy s tem ........................................................... 7-3 2 7.3.3.1 Design Criteria ...............................

................................................... 7-32 7.3.3.2 De s ign Ba s is and S a fety Requir e ments .............................................

7-3 2 7.3.3.3 Sy s tem Description

...........................................................

................ 7-3 2 7.3.3.4 Sy s tem Performanc e Analysis and Conclusion

.................................

7-3 3 7.3.4 Target Dissolution System ....................................................

............................. 7-33 7.3.4.1 Design Criteria ..................................................................................

7-34 7.3.4.2 Design Basis and Safety Requirements

.............................................

7-34 7.3.4.3 System Description

...................

................................

........................ 7-34 7.3.4.4 System Performance Analysis and Conclusion

................................. 7-37 7.3.5 Mol y bdenum Recovery and Purification S ys tem ............................................... 7-37 7.3.5.1 Design Criteria .................................................................................. 7-37 7.3.5.2 Design Basis and S a fet y Requirement s ...........

.................................. 7-37 7.3.5.3 System Description

........................................................................... 7-38 7.3.5.4 System Performance Analysi s and Conclusion

................................. 7-39 7-i

. .. NWMI ...... ..* **.* .*.* .. *.*.* . * "NOmfWEST IWttCAl ISOTOf'fS NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.3.6 Waste Handling System ...........................................

....................

...................... 7-39 7.3.6.1 Design Criteria ..........

...............................................

.....................

.... 7-39 7.3.6.2 Design Basis and Safety Requirements

............

................................. 7-39 7.3.6.3 System Description

.................

.......................................................

... 7-40 7.3.6.4 System Performance Analysis and Conclusion

................................. 7-43 7.3.7 Criticality Accident Alarm System .....................

...................

..............

.............. 7-43 7.3. 7.1 Design Criteria ..............................................................

....................

7-43 7.3.7.2 Design Basis and Safety Requirements

............................................. 7-43 7.3.7.3 System Description

........................................................................... 7-43 7.3.7.4 System Performance Analysis and Conclusion

................................. 7-43 7.4 Engineered Safety Features Actuation Systems ................................................

...............

7-44 7.4.1 System Description

................................................

...................

......................... 7-44 7.4.2 Annunciation and Display ........................

..................................................

........ 7-45 7.4.3 System Performance Analysis ..............................

.............................................. 7-45 7.5 Control Console and Display Instrument s ........................................................................

7-46 7 .5 .1 Design Criteria .....................................

...................................

...........................

7-46 7.5.2 Design Basis and Safety Requirements

..............................................................

7-46 7.5.3 System Description

..............................

.......................................................

....... 7-46 7.5.4 System Performance Analysis and Conclusion

.................................................. 7-46 7.6 Radiation Monitoring Systems ...............................................

....................

......................

7-47 7.6.1 Design Criteria ................................................................

................................... 7-47 7.6.2 Design Basis and Safety Requir ements ............................................

.............

..... 7-47 7.6.3 System Description

...............................................................

............................. 7-47 7.6.3.1 Air Monitoring

.......................

............

..............

................................. 7-48 7.6.3.2 Stack Release Monitoring

................................................................. 7-49 7.6.4 System Performance Analysis and Conclusions

................................................

7-49 7.7 References

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...................

................... 7-50 7-ii

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  • NOllTHWEST MEDICAL ISOTOPH NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Figu r e 7-1. Ta bl e 7-1. Table 7-2. Table 7-3. Tab l e 7-4. Tab l e 7-5. Table 7-6. Table 7-7. Tab l e 7-8. Table 7-9. Table 7-10. Table 7-11. Table 7-12. Table 7-13. FIGURES Radioisotope Production Faci lit y In s trumentation and Control System Configuration

..........

......................................................................................................... 7-2 TABLES Instrumentation and Control System Design Criteria (10 page s) .................................... 7-5 Instrumentation and Control Criteria Crosswalk with Design Ba s i s Applicability and Function Means ( 5 pages) ....................................................................................... 7-18 Uran ium Recovery and Recycle Control and Monitoring Parameters (2 pages) ........... 7-24 Uranium Recycle and Recovery System Interlocks and Permissive Signa l s (4 pages) .....................

........................

..............

...........................................................

... 7-25 Target Fabrication System Control and Monitoring Parameters (2 pages) .................... 7-29 Target Fabrication System Interlocks and Permissiv e Signals (2 pages) ...................... 7-30 Target Dis s olution System Contro l and Monitoring Parameters

................................... 7-35 Target Dis s o lution System Interlocks and Permis s ive Signa l s (2 pages) ...................... 7-36 Molybdenum Recovery and Purification System Contro l and Monitoring Parameters

...................................................................................................................... 7-38 Molybdenum Recovery and Purification Sy s tem Interlocks and Permissive Signa l s ................

.........................................................................................

................... 7-38 Waste Handling System Contro l and Monitoring Parameters

....................................... 7-41 Waste Handling System Interlocks and Permissive Signa l s .......................................... 7-42 Engineered Safety Feature Actuation or Monitoring System s (2 pa g es) ....................... 7-44 7-iii

..... NWMI ...... ... .... ........ *. 0 !*. * ! 0 NOmfWEST ME.DM:Al ISOTOf'lS Acronyms and Abbreviations 99 Mo molybdenum-99 NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems TERMS ADUN acid-deficient uranyl nitrate ALARA as low as reasonably achievable BMS building management system CAAS criticality accident alarm system CAM continuous air monitor CFR Code of Federal Regulations CGD commercial grade dedication COTS commercial off-the-shelf DCS digital control system ESF engineered safety feature FPC facility process control HMI human-machine interface I iodine I&C instrumentation and control IEEE Institute of Electrical and Electronics Engineers IROFS items relied on for safety ISA integrated safety analysis IX ion exchange Kr krypton LEU low-enriched uranium Mo molybdenum NAVLAP National Voluntary Laboratory Accreditation NO x nitrogen oxide NRC U.S. Nuclear Regulatory Commission NWMI Northwest Medical Isotopes , LLC PLC programmable logic controller RAM radiation area monitor RPF Radioisotope Production Facility SDOE secure development and operational environment SIF safety instrumented function.

SIL safety integrity level. SIS s afety instrumented system SNM special nuclear material SSC structures , systems , and components TCE trichloroethylene U.S. United States [Proprietary Information]

UPS V&V Xe Units m mm rad uninterruptible power supply verification and validation xenon meter minute radiation absorbed dose 7-iv

........... .; ... NWMI .......... ' ! *

  • NORTHWEST MUMCAl lSOTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.0 INSTRUMENTATION AND CONTROL SYSTEMS 7.1

SUMMARY

DESCRIPTION The Northwest Medical Isotopes , LLC (NWMI) Radioisotope Production Facility (RPF) preliminary instrumentation and control (I&C) configuration includes the special nuclear material (SNM) preparation and handling processes (e.g., target fabrication , and uranium recovery and recycle), radioisotope extraction and purification processes (e.g., target receipt and disassembly , target dissolution , molybdenum

[Mo] recovery and purification , and waste handling), process utility systems , criticality accident alarm system (CAAS), and sy s tems associated with radiation monitoring.

The SNM processes will be enclosed predominately by hot cells and glovebox designs except for the target fabrication area. The facility process control (FPC) system will provide monitoring and control of the process systems within the RPF. In addition, the FPC system will provide monitoring of related components within the RPF. The process strategy for the RPF involves the use of batch or batch processes with relatively simple control steps. The building management system (BMS) (a subset of the FPC system) will monitor the RPF ventilation system and mechanical utility systems. The BMS primary functions will be to monitor the facility ventilation system and monitor and control (turn on and off) the mechanical utility systems. Engineered safety feature (ESF) systems will operate on actuation of an alarm setpoint reached for a specific monitoring instrument/device. For redundancy , this will be in addition to the FPC system or BMS ability to actuate ESF as needed. Each ESF safety function will use hard-wired analog controls/interlocks to protect workers, the public , and environment.

The ESF parameters and alarm functions will be integrated into and monitored by the FPC system or BMS. The preliminary concept for the RPF l&C system configuration is shown in Figure 7-1. The green circles identify the FPC and the BMS distributed process control or programmable logic controller (PLC) s ystems. The solid lines and dashed lines show how the SNM processes , support systems , utilities , radiation and criticality systems , and building functions relate to the FPC and BMS and to local machine interface (HMI) stations.

Solid lines indicate the control functions , and dashed lines indicate the monitoring functions. The FPC system will perform as the overall production process controller.

This system will monitor and control the process instrumented functions within the RPF , including monitoring of process fluid transfers and controlled inter-equipment pump transfers of process fluids. Process control systems are described further in Section 7.3. The fire protection system will have its own central alarm panel (green circle). The fire protection system will report the status of the fire protection equipment to the central alarm station and the RPF control room. The fire protection s ystem is discussed further in Section 7.2.3.5. 7-1

.. .. NWMI ...... ..* .... ....... !.* . * "NORTHWUTMf.DICAl tsofOfllS NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Process Support Systems Waste Handling System :-HHIH+--e -------!-II.Ill e -e I I I 1--1 I I I I ------; I I I I I ,--------I I I I ------,-------

-I I .. --e Fire Protection System --, I I I --' I I I I _J Process Systems In Hot Cell Area I I I I I I I r-------I I I I I -------; I I Facility Ventllation Sy stem ---

Only +---+ Control a nd M onitoring Process Systems In Target Fabrication Area I Process Utility Systems Figure 7-1. Radioisotope Production Facility Instrumentation and Control System Configuration Special nuclear material preparation and handling processes

-The FPC system will control and/or monitor the SNM preparation and handling processes , the following. *

  • Target fabrication

-Batch processes located in the target fabrication area will be controlled by operators at local HMls , with surveillance monitoring in the control room. Uranium recovery and recycle -Batch processes located inside the hot cell area will be controlled by operators in the control room. Radioisotope extraction and purification processes

-The FPC system will control and/or monitor the radioisotope production processes , including the following. *

  • Target receipt and disassembly

-Hardware/target movement located in irradiated target basket receipt bay area , target cask preparation airlock , target receipt hot cell , and target disassembly hot cell will normally be controlled by operator s at local HMis , with surveillance monitoring in the control room. Target dissolution

-Batch process located inside the dissolution hot cell will occur at local HMis in the operating gallery , and offgas operations in the tank hot cell will be controlled by operators in the control room, with surveillance monitoring at both locations.

7-2 NWMI ...... ..* **: ..... ...... 0 ." NOfOlfWEST MEDtCAl ISOTOPfS NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems *

  • Mo recovery and purification

-Batch processes located inside the Mo hot cells will be controlled by operators at a local HMI in the operating gallery , with surveillance monitoring in the control room. Waste handling-This system includes liquid waste handling , liquid waste solidification , and solid waste handling.

Operators in the control room will control liquid waste handling , while operators at local HMls in the low-dose liquid solidification room (Wl07) will monitor and control liquid wa s te solidification , and solid waste nondestructive examination and solidification. Process utility and support systems -The FPC system will control and monitor the process utility and process support systems. Operators in the control room will control the following subsystems:

  • *
  • Process chilled water hot cell secondary loops Process steam hot cell secondary loops Process vessel ventilation system Operators at local HMis will control the following subsystems , with surveillance monitoring in the control room using the FPC system or BMS. * * * * * *
  • Plant air system Gas supply system Process chilled water chillers Process steam boiler s Demineralized water system Chemical supply system Standby electrical power system Criticality accident alarm system -The CAAS will be provided as an integrated vendor package. The detectors and alarm response are integral to the individual units/locations. The FPC s ystem will monitor the CAAS status in the control room. The CAAS is de s cribed further in Section 7.3. Radiation monitoring system -The FPC system will monitor the various radiation monitoring systems , including continuous air monitors (CAM), air samplers , radiation area monitors (RAM), and exhaust s tack monitors. The CAMs and RAMs will be strategically placed throughout the RPF to alert personnel of any potential radiation hazards. The CAMs and RAMs will alarm in the control room and locally at locations throughout the RPF. The radiation monitoring s ystems are described further in Section 7.6. Facility ventilation system and mechanical utility systems -The control function for most of the RPF ventilation system and mechanical utility systems will be local HMls and hard-wired interlocks for the ESF function s. The BMS will monitor the systems and provide ventilation and mechanical utility s ystem status as an input to the FPC process controls.

The following subsystems will be monitored by the BMS: * * *

  • Facility ventilation Zones I , II , III , and IV Supply air system Facility chilled water system Energy recovery and heating water 7-3

.;.-.;* .. NWMI ...... ... .... ........ *.* * ! *:

  • NOflTHWUT MlDtCAL asGTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Safety-Related Components and Engineering Safety Features The ESF safety functions will operate independently from the FPC systems as hard-wired analog controls or interlocks. The FPC system will be a digital control system (DCS) that monitors safety-related components within the RPF. The ESFs will be integrated into the FPC systems and provide a common point of HMI , monitoring , and alarming at the control room and , as necessary , local HMI workstations.

Control Console and Display Instruments The control room will be the primary interface location for the RPF support systems and provide centralized process controls, monitoring, alarms, and acknowledgement.

Mechanical utility systems with vendor packages and integrated controls will be controlled at associated local HMi s. The BMS will provide primarily on/off control and system monitoring from the control room. The tank hot cell processes will be controlled primarily in the control room, with surveillance monitoring of the FPC subsystems.

The FPC system will have annunciation, alarms, and HMI displays.

From the consoles, operators will view and trend essential measurement values from the HMI display, and evaluate real-time data from the essential measurements used to control and monitor the RPF process. This system is further described in Section 7.5. Process utility and support systems with vendor package and integrated controls will be operated at associated local HMis. These systems are discussed further in Section 7.5. Local HMis are anticipated in the following locations:

  • * * * * *
  • 7.2 Irradiated target basket receipt bay A/B (Rl 02A/B) Cask preparation airlock (R012) Operating gallery (GlOl A/B/C) Target fabrication (Tl 04 A/B) Low-dose liquid waste solidification (Wl 07) Chemical supply room (Ll02) Local to equipment with integrated control systems DESIGN OF INSTRUMENTATION AND CONTROL SYSTEMS The design criteria and the codes and standards for I&C systems are outlined in Chapter 3.0 , " Design of Structures , Systems, and Components," and discussed below. 7.2.1 Design Criteria The applicable design criteria and guidelines that apply to the RPF I&C systems are summarized in column one of Table 7-1. Additional, design criteria for I&C systems are provided in Chapter 3.0. The detailed and specific design criteria for I&C systems will be confirmed in the Operating License Application.

7.2.2 Design

Basis and Safety Requirements The design basis for I&C systems used in the RPF are presented in the second column of Table 7-1. The second column maps the criteria to l&C systems or components and how compliance will be ensured. Note that the FPC system callouts may also apply to the BMS. The design basis requirements for facility and process systems are described in Chapter 4.0 , " Radioisotope Production Facility Description

," and Chapter 9.0 , "Auxiliary Systems." The I&C system will use hard-wired interlocks for actuated engineered safety functions. Section 7.4 summarizes the I&C ESFs. 7-4

.; ... NWMI ...... .. .. .... .. .. .. ' *. *

  • NOll1IWEST MlOtCAl ISOTOPlS NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages) Design criteria descriptiona Design bases as applied to RPF IEEE 379-2014, IEEE Standard Application of the Application:

Singl e-Failure Criterion to Nucle ar Pow e r Generating

  • Design ofFPC system, ESFs , and ot h er Station Safety System s instrumentation SSCs that are identified as IROFS Description
App li cat i on of the single-failure cr it erion Compliance:

to electrical power , in strumentat i on , and contro l portions

  • Ensure FPC system is a DCS designed , rated , and of nuclear power generating safe ty systems. approved for use in safety instrumented systems , as Keywords:

Actuator , cascaded failu r e, common-cause determined b y ANSI/I SA 84.00.01 fai lur e , design basi s event , detectable failure , effects

  • Use a safety PLC , as r ecog nized by IEC 61508 , in the ana l ysis, safety system , single-failure cr it erion , system FPC sys tem with redundant power supplies , actuation , syste m lo gic processor s, and inpu t/o utput channels IEEE 577-2012, IEEE Standard Requirements for Reliability Analysis in the Design and Operation of Safety Systems for Nuclear Facilities

==

Description:==

Sets minimum acceptable requirements for the performance of reliability analyses for safety systems when used to address the reliability considerations discussed in industry standards and gui d e lines. The requirement that a reliability analysis be performed does not originate with this standard.

However, when reliability analysis is used to demonstrate compliance with reliability requirements, this standard de sc ribes an acceptable response to the requirements.

Keywords: Nuclear facilities, reliability analysis, safety systems

  • Eva lu ate contro l s that are classified as IROFS in C hapt ers 6.0 and 13.0 , or NWMI-2 0 l 5-SAFETY-002 , against s ingle-failure criteria Exce ption:
  • NUREG-1537 allows for s haring and combin in g of sys tem s and components with justification
  • The RPF i s not considered a nucl ea r pow er reactor but a production faci li ty. The facility will not have all of the syste ms detailed in this standard and guidance will be app li ed as appropriate. Application:
  • Use for design ofFPC system, ESFs, and other instrumentation SSCs that are identified as IROFS Compliance:
  • Perform a reliability analysis of the proposed design solutio n for IROFS functions, as identified in Chapters 6.0 and 13.0, or NWMI-2015-SAFETY-002.

The analysis can be qualitative or quantitative in nature, as de scribed in the sta nd a rd 7-5

.; .. ; .. NWMI ..*... ..* .... ....***. *.* . NORTHWEST MlDtCAl ISOTOPES NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages) Design criteria description 3 IEEE 603-2009 , IEEE Standard Criteria for Safety Systems for N uclear Pow e r Generating Stations Description

Estab li shes minimum functional and design criteria for the power , instrumentation , and contro l portions of nuclear power generating station safety systems. Criteria are to be app li ed to those systems required to protect public health and safety by functioning to mitigate the consequences of design basis events. The intent is to promote appropriate practices for design and evaluat ion of safety system performance and reliability. T h e standard is limited to safety systems; many of the principles may have applicabi lity to equ ipm ent provided for safe shutdown , post-accident monitoring display in strumentation, preventive interlock features, or any other systems, structures, or equipment related to safety. Keywords: Actuated equipment, associated circuits, Class IE, design, failure, maintenance bypass , operating bypass, safety functio n , sense and command features, sensor IEEE 384-2008, IEEE Standard Criteria for Independence of Class IE Equipment and Circuits

Description:

Describes independence requirements of circuits and equipment comprising or associated with Class IE systems. Identifies criteria for independence that can be achieved by physical separation, and electrical isolation of circuits and equipment that are redundant.

The determination of what is to be considered redundant is not addressed.

Keywords: Associated circuit, barrier, Class IE, independence, isolation, isolation device , raceway, separation Design bases as applied to RPF Application:

  • Use for d esign ofFPC system , ESFs, and other instrumentation SSCs that are identified as IROFS
  • App ly minimum functional and design cr it eria to safety systems Compliance:
  • Ensure design conforms to the practices detailed in the standard for the IROFS functions identified in Chapters 6.0 and 13.0 , orNWMI-2015-SAFETY-002 Exception:
  • The RPF i s not considered a nuc le ar power reactor but a production facility. The fac ilit y will not have all of the systems detailed in this standard and guida n ce will be app li ed as appropriate.

Application:

  • Use for design of FPC system, ESFs, and other instrumentation SSCs that are identified as IROFS
  • Apply minimum criteria for separation and independence of systems in a physical way Compliance:
  • Ensure design conforms to the practices deta i led in the standard for the IROFS functions identified in Chapters 6.0 and 13.0 , orNWMI-2015-SAFETY-002 Exception:
  • The RPF is not considered a nuclear power reactor but a production facility.

The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

7-6

.; ... ; .. NWMI ::.**.*.-.. ..... .. .. ..

  • NORTHWESTMlDtcALISOTDP(S NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages) Design criteria descriptiona IEEE 323-2003, IEEE Standard for Qualifying Cla s s JE Equipment for N ucl e ar Power G e nerating Station s Description
Id e ntifi es re quir e m e nt s fo r qu a li fy in g C l ass I E e quipm e nt a nd i nt e r faces th a t are to b e u se d in nucl e ar p o w e r ge neratin g sta ti o n s. The prin c ipl es, m e thod s, a nd pro ce dur es a r e intended for u se in qu a lifying e quipment , m a in ta inin g and ex t e ndin g qu a lifi ca ti o n, a nd upd a tin g qu a lifi ca ti o n , as r e quir e d , i f the e qu i pm e nt i s modifi e d. Th e qualifi cat i o n r e quir e m e nts o f the s t a nd ard d e mon s tr a te a nd do c um e nt the a bilit y of e quipm e nt to pe r fo rm safe t y func t ion(s) und e r appli ca bl e se r v ic e co nditi o n s, includin g d es i gn b as i s eve nt s, re ducin g th e ri sk o f c omm o n-ca u se e quipm e nt fai lur e. Keywords: Age c onditi o nin g, ag in g , co nditi o n monit o rin g, d es i gn b as i s eve nt, e quipm e nt qu a lific a ti o n , qu a lific a ti on m e thod s, h a r s h env ironm e nt , m a r g in , mild e n v ironm e nt , qu a lifi e d li fe, ra di a ti o n , sa f e t y-r e l a t e d fun ct ion, s i gni fic a nt ag in g m ec hani s m , t es t pl a n , t es t se qu e nc e, t ype t es tin g IEEE 344-2004, IEEE Recommended Practice for Seismic Qualification of Class JE Equipment/or Nuclear Power Generating Stations

Description:

Identifies recommended practices for e s tablishing procedures that will yield data to demonstrate that the Class IE equipment can meet performance requirements during and/or following one safe shutdown earthquake event , preceded by a number of operating basis earthquake events. This recommended practice may be used to establish tests , analyses , or experience-based evaluations that will yield d a ta to demonstrate Class 1 E equipment performance claims or to evaluate and verify performance of devices and assemblies as part of an overall qualification effort. Common methods currently in use for seismic qualification by test are presented. Two approaches to seismic analysis are described:

one based on dynamic analysis, and the other on static coefficient analysis.

Two approaches to experience-based seismic evaluation are described , one based on earthquake experience and the other on te s t experience.

Keywords:

Class 1 E , earthquake, earthquake experience, equipment qualification, inclusion rules, nuclear, operating basis earthquake , prohibited features , qualification methods , required response spectrum , response spectra , safe shutdown earthquake, safety function , seismic , seismic analysis , test response spectrum, test experience Design bases as applied to RPF Application:

  • Use fo r e quipm e n t qu a li ficat ion w h en n ee d e d to qu a li fy e quipm e nt for a ppl ica tion s or e n vi ronm e nt s t o whi c h the e quipm e nt m ay b e ex po se d
  • Use for qu a lifi ca ti o n o f C l ass IE e quipm e nt l oca t e d in h a r s h env i ro nm e nt s and fo r ce rt a in po s t-a cc id e nt monitoring e quipm e nt; m ay a l so b e u se d fo r th e qu a lifi ca ti o n o f e quipm e nt i n mild e n v ironm e nt s Compliance:
  • E n s ur e d esig n c onform s to t h e pra c ti ces d e t a il e d in the s t a nd a r d fo r tho se sys t e m s det e rmin e d to b e C l ass IE a nd l o c a t e d in h ars h e n v ironm e n ts fo r safe t y fun c tion s id e n t ifi e d in C h a p te r s 6.0 a nd 1 3, or NWMI-2 0 J 5-S A FE TY-00 2
  • Appl y t o SSCs w ithin th e h o t ce ll ar e a; n ot all sa fet y co mpon e n ts res id e in th e h ot ce ll a r e a
  • Apply s t a nd a rd u s ing a gra d e d a ppro ac h Exception:
  • Th e RPF i s n o t co n s id e r e d a nucl ea r po wer r eac t o r but a produ c ti o n facilit y. Th e fa cilit y w ill not h ave a ll o f th e sys t e m s d e t a il e d in this s t a nd a rd a nd g uidan ce wi ll be a ppli e d as a ppropri a t e. Application:
  • Apply sei s mic design requirements for equipment used in Class 1 E systems Compliance:
  • Use in design ofFPC system , ESFs, and other instrumentation SSCs that are identified as a Class IE system Exception:
  • The RPF is not considered a nuclear power reactor but a production facility.

The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate. 7-7

.; ... NWMI ...... ..* *.. .*.******* ' ! *. * ' NOITIIWtsT lllOICAl tsOTOfl(S NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control S y stem Design Criteria (10 pages) Design criteria description s IE EE 338-2012 , IEEE Standard for Criteria for th e P e riodi c Surv e illan ce T es tin g of N ucl e ar Pow e r G e n e rating Station Saf e ty S yste m s Description

Provi d es c ri ter i a fo r t h e p e r forma n ce of per i o dic s u rve i llance tes tin g of n u clear p ower ge n era tin g stat i o n safety systems. T h e scope of p eriod i c surve illance testing co n sists of functio n a l tests a n d checks, ca li b r ation verificat i on, a nd t im e response meas ur e m e n ts, as re qui re d , to ver i fy th at t h e safe t y system p e r fo rm s i ts d efi n e d safety fu n c ti o n. m a in te n a nce a nd p os t-m o d ificat i o n test ing a r e n ot cove r ed b y thi s doc um e n t. T hi s s ta nd a rd a m p li fies th e period ic s u rve i lla n ce t es tin g req u ire m e n ts of ot h er nuc l ea r sa f ety-re l a t ed I EEE standa rd s. Ke y words: Funct i o n a l tes t s , IEEE 338 , p eriodic tes ti ng , risk-in fo rm e d testi n g , s ur vei ll a n ce test in g IEEE 497-2010 , IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations

Description:

Establishe s criteria for variable s election , p e rform a nce , d e sign, a nd qualification of acciden t monitoring instrumentation , and include s the requirements for display alternative s for accident monitoring instrumentation , documentation of design b as e s, and u se o f port a ble in st rumentation. Ke y words: Accident monitorin g, display criteria , se lection criteria , type v ariabl es IE EE 7-4.3.2-2010 , IEEE Standard Crit e ria for Di g ital Comput e r s in Saf ety S ys t e m s of N ucl e ar Pow e r G e n e rating Station s A bstract: S p ecifies a ddi tiona l co m p ut e r-spec i fic req uir e m e nt s to s u pp l e m e n t IEEE 603-2 0 09. T h e sta nd a rd d efi n es t h e t e rm comp ut e r as a syste m t h a t incl ud es com pu ter h a rd ware, software, firmware , a n d i n te r faces, an d esta bli s h es m i n im u m fun ctio n a l a nd d es i gn r e quir e m e nt s fo r co m p ut ers u se d as co mp o n e nt s of a safe t y sys t e m. Ke y words: Co mm erc i a l-gra d e ite m , di versity , safety sys t e m s , softwa r e , softwa r e too l s, softwa r e ve rifi ca ti o n a nd va lid a ti o n Design bases as applied to RPF Application:

  • Use for d es i g n ofF P C system , ESFs , a nd o th er in st rum e n ta ti o n SSCs th a t a re i d e ntifi e d as IRO FS
  • Use m e th o ds a n d cr i ter i a to esta bli s h a p eriod i c s urve ill a n ce progr a m C ompliance:
  • Ens ur e d esign confo rm s to the practices detailed in the sta nd ard for t h e IR OFS funct i o n s i d e nt ified in C h ap t e r s 6.0 an d 1 3.0 , or NWMI-2 0 l 5-SAFETY-0 0 2 Ex c e ption:
  • T h e RP F i s n o t co n s id e r e d a n u c l ear p ower reacto r but a produ c t ion fac ilit y. T h e fac ili ty w ill n ot h ave a ll of th e sys t e m s d eta il e d in t hi s sta nd a rd a nd gu id a n ce w ill be a ppli e d as appro pri ate. Application:
  • U se as selection , de s ign , p e rformance , qualification , and displ a y criteria for a ccid e nt monitoring instrumentation
  • Apply guidance on the u s e of portable in s trumentation and for e x ample s of accident monitoring displa y configuration s Compliance:
  • E nsure design conform s to st and a rd for the monitoring functions determined to be required for health and s a fet y of worker s or the public during normal operat i on and de s ign ba s i s accidents Exception:
  • Th e RPF is not con s idered a nuclear power reactor but a production facility.

The facili t y will not have all of the sy s tems detailed in this s tandard and guidance will b e appli ed a s appropriate.

Application:

  • I n co njun ction with I EEE 603-2009 , u se to esta bli s h m i n im um functiona l a n d de s i gn re qu irements fo r com pu ters t h a t a r e components of a safety syste m
  • D es i g n F P C sys t e m a s a DCS , a nd a ppl y th is sta nd ar d to sys t em d eve l o pm e n t , s p ecifica ll y software deve lopm ent
  • A pply s t a nd ard t o CG D a n d impl e m e nt a n a ppro ac h C ompliance:
  • D eve l o p F P C sys t e m software u si n g th is stan d a rd E x c e ption:
  • T h e RP F is n ot co n si d e r e d a nu c l ear p ower r eactor b ut a p ro du ction fac ili ty. The faci li ty wi ll not h ave a ll of t h e systems d e t ai l e d in t h is sta nd a rd a nd gu id a n ce wi ll be a ppl ie d as ap p ro pr ia t e. 7-8

.. ; ... .. NWMI ...... ..* .... .*...*. *.* , 0 NORTHWUTMmtc.Al ISDTOPfS NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control S y stem Design Criteria (10 pages) Design criteria description 3 IEEE 828-2012, IEEE Standard/or Configuration Management in Systems and Software Engineering Description

Establishe s minimum requirements for configuration management in s ystems and s oftware engineering.

This standard applies to any form , class , or type of software or system , and explain s configuration management , including identifying and acquiring configuration items , controlling changes , reporting the status of configuration it e ms, and performing software builds and rele as e engineering.

This standard addre s ses what configuration management activities are to be done , when they are to happen in the life-cycle , and what planning and resources are required. The content area s for a configuration management plan are also identified. The standard supports IEEE STD 12207 and ISO/IEC/IEEE 15288 , and adheres to the terminology in ISO/IEC/IEEE STD 24765 and the information item requirements of IEEE STD 15939. Keywords: Change control , configuration accounting , configuration audit, configuration item , IEEE 828, release engineering, software builds , software configuration management, system configuration management Design bases as applied to RPF Application:

  • Use to establish configuration management processes , define how configuration management is to be accomplished , and identify who is responsible for performing s pecific activities , when the activities are to happen , and what specifi c resources are required
  • Design FPC system a s a DCS , and apply standard during the development of s oftware for systems with IROFS functions Compliance:
  • Develop FPC system software using this standard for safety function implementation IEEE 1028-2008, IEEE Standard/or Software Review s Application:

and Audit s

Description:

Id e ntifi es fi ve types o f so ft ware r ev i ews a nd a udit s , t oget h e r w ith p roce dur es r e qui re d fo r th e exec uti o n of eac h typ e. This s t a ndard i s co n ce rn e d onl y w ith r ev i ews a nd a udi ts; p roce dur es for d ete rminin g th e n ecess it y of a r ev i e w or a ud it a r e not d e fi ne d, a nd th e d is p os iti o n of t h e r es ult s of t h e r ev i ew o r au dit i s not s p ec ifi e d. T ypes included are m a n age m e n t rev i e w s , tec hnic al r ev i ews , in s p ect i ons , wa lk-thr o u g h s, a nd a udit s. Ke y words: Au dit , in s p ect i o n , r ev i ew , wa l k-thr o u g h

  • Use t o id e n tify minimum acce pt a bl e r e qu i r e m e nt s fo r s y s t e mati c so ftw a r e r ev i ews
  • Id e nti fy or g a ni zat ion al m ea n s for condu cti ng a r ev i ew a nd d o cum e n t in g th e findi ngs
  • D es i gn FP C s yste m as a D CS , a nd a pply s t a nd a rd durin g th e d eve l o pm e nt of so ftw a r e fo r syste ms w ith IROF S fun ct i o n s Compliance:
  • D eve l o p FPC sys t em u s ing t his s t a nd a rd ANS 10.4-2008, Verification and Validation of Non-Application:

Safety-Related Scientific and Engineering Computer

  • Perform software V & V to build quality into the Programs for the Nuclear Industry software during the software life-cycle Description
Provides guidelines for V & V of non-* Use to verify and validate software development for safety-related scienti fie and engineering computer non-safety-related systems programs developed for use by the nuclear industry.
  • Use for software development in the RPF that is not Scope is restricted to research and other non-safety-safety significant (e.g., not safety-related or IROFS) related , noncritical applications.

Compliance:

Keywords: Software integrity level, software life-cycle ,

  • Develop non-safety-related software using this validation , verification , V & V standard 7-9

.; ... .. NWMI ..**.. ..* .... ..... .... .. . . ****** *. * * . NCMl:T H WUT MlD"-'l tsOTDPES NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages) Design criteria descriptiona ANSI/ISA 67.04.01-2006 , Setpoint s for N uclear Safety-Related Instrumentation Description

D e fin es r e quir e m e nt s fo r assessi n g , esta bli s hin g , a nd m a int a inin g nucl ea r safe t y-re l a t e d and ot h e r import a n t in s trum e n t set p oi nt s ass oci ate d w ith nucl ea r po w e r pl a nt s or nucl ear r eac tor faci li ties. Ke y words: Se tpoint , dri ft , ana l og c h a nn e l , re li a bilit y a n a l ys i s ANSI/ISA 84.00.01-2004, Functional Safety: Safety Instrumented Systems for the Process Industry Sector Part 1: Framework, Definitions, System, Hardware and Software Requirements" Part 2: "Guidelines for the Application of ANSI/ISA-84.00.01-2004 Part 1 (IEC 61511-1 Informative" Part 3: "Guidance for the Determination of the Required Safety Integrity Levels -Informative"

Description:

Provides requirem e nts for the specification , design , installation , operation , and maintenance of a safety instrumented system, so the s ystem can be confidently entrusted to place and/or maintain the process in a safe state. This standard has been developed as a proces s sector implementation of IEC 61508. Keywords:

Safety instrumented system (SIS), safety integrated level (SIL), safety in s trumented function (SIF) Design bases as applied to RPF Application:

  • Use m e thods a nd c rit e ri a t o esta bli s h se tp o in ts fo r sa f e t y sys t e ms a nd t o m ai nt ai n th e do c um entat i o n
  • A ppl y to th e d es i g n of t h e F P C syste m a nd ot h e r in st rum e n tat i o n SSCs t h a t a r e id e ntifi e d as IROF S fo r t h e RP F Compliance:
  • E n s ur e d es i g n co nform s to th e prac tic es d eta il e d in t he s t a nd a rd fo r IRO FS fun ctio ns w ith inh e r e n t se tp o int s id e nt ifie d in C hapt ers 6.0 a nd 1 3 .0 , o r NWM I-20 l 5-SAFE T Y-00 2 Application:
  • Apply to the design of safety systems (standard specifically designed for indu s trial processe s)
  • Standard is made up of three parts: -Use Part I to lay the groundwork for the s afety system life-cycle , overall structure of safety systems , definitions used , and to implement safety system design engineering

-Use Part 2 guidance for the specification , design , installation , operation , and maintenance of safety in s trumented functions and related safety in s trumented systems , as defin e d in Part 1 -Use Part 3 to develop underlying concepts of risk in relation to safety integrity, identify tolerable risk , and determine the safety integrity levels of the safety functions

  • Design physical hardware of the FPC system based on this standard and IEC 61508
  • Evaluate the IROFS function s required to be impl e mented by the FPC system using Parts 1 , 2 , and 3 of this standard
  • Use to demonstrate reliab i lity and risk reduction of the FPC sy s tem , while having similar or higher documented and tested ability to reduce risk as fulfillment through other channels Compliance:
  • Use for the design and implementation for IROFS functions that are required of the FPC system 7-10

.... ;. NWMI ...... ..* ... ........... * *. *

  • NOmfWlST MEDttAl lSOTO,H NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages) Design criteria description 3 NUREG-0700, Human-Sy ste m Interface Design Review Guidelines Description
Provides guidance to the NRC on the eva luation of human factors engineering aspects of nuclear power plants in accordance wit h NUREG-0800.

Detailed design review procedures are provided in NUREG-0711.

As part of the review process , the interfaces between plant personnel a nd the plant systems and components are evaluated for conformance with human factors engineering guidelines.

Ke ywo rds: Display , HMI , human-interface system, human-system interface NUREG/CR-6463, Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems

Description:

Provides guidance to the NRC on auditing programs for safety systems written in the following six high-level languages:

Ada, C and C++, PLC Ladder Logic, Sequential Function Charts, Pascal, and PL/M. The guidance could also be used by those developing safety significant software as a basis for project-specific programming guidelines.

Keywords:

Pascal , C, Ladder Logic, PL/M , Ada, C++, PLC, programming , sequential function charts NUREG/C R-6090, The Programmable Logic Controller and It s Application in N ucl ea r Reactor Systems Abstract:

Outlines recommendations for review of the app li cation of PLCs to the contro l , monitoring, and protection of nuclear reactor s. Keywords:

PLC, programming , protection systems Design bases as applied to RPF Application:

  • Use comprehensive design review guidance to develop information displayed in human-interface systems
  • Develop informative and effective designs that will assist operators in the performance of their duties Compliance:
  • Design FPC system to provide information to operators in a display format
  • Display development used in connection wit h the FPC system wi ll be provided in the Operating License App li cation Application:
  • Use guidance to review high-integrity software in a nuclear facility
  • Develop FPC system as a DCS, with associated programming d eve lopment needs for the RPF
  • Use guideline as a means to review FPC system programming code Compliance:
  • Develop FPC system software programs using this guidance Exception:
  • The RPF is not considered a nuclear power reactor but a production facility.

The facility will not have all of the systems detailed in this standard and guidance will be applied as appropriate.

Application:

  • Use gu id ance to implement PLC s for nuclear app li cation and as a forum for what constitutes good practices of previously installed systems
  • Use guidance during se l ection process for hardware , failure analysis , and product life-cycle within the faci lit y Compliance:
  • Design FPC system to use a PLC-type DCS
  • Select design and implement PLC s based on this guide , as applicable Ex ception:
  • The RPF is not co n s id ered a nuclear power r eactor but a production facility. The fac ilit y will not h ave a ll of the systems detailed in this sta nd ard and guidance will be app li ed as appropriate.

7-11

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  • NOITHWEST MEDttAl tSOTOPU NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages) Design criteria descriptiona EPRI TR-106439, Guideline on Evaluation and Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Applications

==

Description:==

Provides a consistent, comprehensive approach for the evaluation and acceptance of Design bases as applied to RPF Application:

  • Use to identify appropriate critical characteristics with subsequent verification through testing , analysis , vendor assessments , and careful review of operating expenence commercial digital equipment for nuclear safety systems.
  • Keywords:

Commercial off-the-shelf(COTS), programming, software , commercial grade dedication Use guidance for digital upgrades to safety-related systems and for non-safety-related applications that require high reliability or are compatible with specific change processes, including graded approaches for quality assurance Regulatory Guide 1.152, Criteria/or Use of Computers in Safety Systems of N uclear Power Plant s Description

D es crib es a m e th o d th a t the N RC s t a ff d ee ms a c ce p ta bl e fo r c ompl y ing w ith N R C reg ul a tion s fo r prom o tin g hi g h function a l re l ia bili ty, d es i gn quality , a nd a sec ur e d eve lopm e nt a nd o p e rati o n a l e n viro nment fo r th e u se of di g i ta l c omput ers in th e safety sys t e ms o f nu clear p owe r p l a nt s. Keywords: S ec ur e d eve lopm e nt a nd op e rati o n a l e n v ironm e nt (S DO E), comput e r s Regulatory Guide 1.53, Application of the F ailure Criterion to Safety Systems Description
Provides methods acceptable to the NRC staff for satisfying NRC regulations with respect to the application of the single-failure criterion to the electrical power and I&C portions of nuclear power plant safety systems. Keywords:

IEEE 379-2014 , single-failure criterion Compliance:

  • Ensure that digital systems components that require CGD apply the guidance of this standard , as applicable Application:
  • Use fo r l&C system d es i gns w ith c ompu te r s in s a fe t yr e lat e d sys t e m s th at m a k e exte n s i ve u se of adva nc e d t ec hn o lo gy
  • Use fo r RP F d es i g n s (th a t a r e ex p ec t ed t o b e s i g nific a ntly a nd fun c tionall y diff ere nt fro m c urr e nt d ay proc ess d es i g n s) with m ic ropro c e ss or s, di g it a l sys t e ms a nd di s pl ays , fiber o pti cs , multipl ex in g, a nd diff e r e nt i so l a ti o n te c hniqu es to ac hi eve s uffi c i e nt ind e p e nd e n ce a nd r e dund a n cy Compliance
  • D eve lop FP C syste m a nd assoc i a t e d HMI u s in g thi s g uid a nc e Exception:
  • Th e RPF i s n o t co n s id e r e d a n ucl ea r pow er r e a c to r but a p ro du c ti o n fa cilit y. T h e fac ili ty will n ot h ave a ll o f th e sys t e m s d e t a il e d i n this s t a ndard a nd gu id a nce w ill be a ppli e d as ap propriat e. Application:
  • Apply single-failure criterion to safety-related I&C systems
  • Apply to end-devices used by the FPC system that are identified as IROFS Compliance:
  • Evaluate FPC system, ESFs , and IROFS end-devices using this guidance 7-12

... ; .. NWMI ...... ..* ... ........ *. ' * ! . NOllTNWEST MlDICAl ISOTOPH NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-1. Instrumentation and Control System Design Criteria (10 pages) Design criteria description 3 Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants Description

Provides a method that the NRC staff co n s ider s acceptable for u se in complying with NRC regulations with respect to sa ti sfy ing criteria for accident monitoring instrumentation in nucle ar po we r pl a nt s. Keywords:

IEEE 497-20 I 0 , acc ident monitoring Regulatory Guide 5. 71, Cyber Security Programs for Nuclear Facilities

==

Description:==

Provides an approach that the NRC staff deems acceptable for complying with NRC regulations regarding the protection of digital computers , communications systems , and networks from a cyberattack, as defined by I 0 CFR 73. l. Keywords: Cybersecurity, 10 CFR 73.54(a)(2), design basis threat a F ull refer e nces provid ed in Sec tion 7.7. CAAS CAM CF R CG D COTS DCS ESF F P C HM! I&C IEEE c r it icality acc id ent a larm sys t e m. co ntinuou s air monitor. Co d e of Federal R egu l at i o n s. co mm e rcial grade dedication.

co mm e rci a l off-th e-s h e l f. digital co ntr o l sys t e m. e n g in ee r e d safety feature. fac ilit y proc ess co ntrol. human-machine interface. instrumentation and co ntrol. In s titut e of E l ectr i ca l and E l ectronics E n gi n eers. Design bases as applied to RPF Application:

  • Use this guidance for developm e nt of accident monitorin g for the RPF Compliance:
  • Desi g n FPC system, CAAS, CAMs, and RAMs u s in g this guidance Exce ption:
  • The RPF i s not considered a nuclear power reactor but a production facility.

The facility will not hav e a ll of the systems detailed in this s tandard and guidance will be applied as appropriate.

Application:

  • Use this guidance for development of cybersecurity protections Compliance:
  • Design the FPC system and associated HMI based on this guidance IROF S N R C PLC RAM RPF S DO E S I F S IL S I S SSC Y&Y items reli e d on fo r safety. U.S. Nuc l ear R egu lat ory Commission. progr a mm ab l e l og i c co ntroll er. radiation alarm monitor. Radioisotope P roduct i o n Facility. sec ur e d eve l opme nt a nd operational env i ro nm e n t. safety in st rum e nt e d fun c tion. safety int egr it y l eve l. safe ty in strumented system. s tructure s, sys t e m s, and components.

verifica tion a nd va lidation. Specific requirements will be developed during the next stages of design for the Operating License Application. The I&C design wi ll be expande d and analyzed to document fulfillment of the design criteria and design basi s requirements for the Operating License Application.

7.2.3 System

Description As described in Section 7 .1 , the RPF I&C system basic compo n e nt s include the FPC system, ESF actuat ion systems, control conso le and HMI display instruments , and BMS. These systems provide an interface for the operator to monitor and control those systems. The FPC system will be a DCS that functions independently and electrica ll y isolated from power systems. The item s relied on for safe ty (IROFS)/ESF safety functions will be activate d via hardwire interlocks. 7-13

...... ; .. NWMI ...... ..* .... ........ *. * *. . NORTHWEST MEDUl ISOTOPH NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.2.3.1 Facility Process Control System The FPC system controls and monitors the target fabrication system, hot cell area (e.g., Mo recovery and purification, uranium recovery and recycle system), process utility and support systems , and waste handling activities.

The FPC system functions also include radiation monitoring, CAAS , HMis , safe shutdown control and initiation , supervisory information , and alarms. The BMS is a subsystem to the FPC system and monitors the facility ventilation system. The primary control location of the FPC system is in the control room. The control room FPC system operates with a synchronized hot standby redundant system structure. The hot standby workstations provide redundant hardware with identical PLC software systems as automatic backup control systems. The primary and backup PLC systems monitor each other. On loss of synchronizing signal from one system, the other system continues with control and monitoring.

This automatic backup control system minimizes the likelihood of downtime during Mo production processing. 7.2.3.2 Engineered Safety Feature Actuation Systems The operator will have direct visualization of critical values and the ability to observe status of the features described in Table 7-13 (Section 7.4.1). The engineered safety feature actuation system dedicated displays will perform the following functions: * *

  • Static display -This display will show critical measurement values and perform the function of an annunciator panel. This fixed display panel will not provide any interactive control functionality.

Alarm/event annunciator display panel -This panel will display any event or alarm that is defined for the process. The display will enable the operator to acknowledge current events and alarms , and will provide a historical record of events. Dynamic interface display panel or HMI -This panel will enable the operator to perform tasks , change modes, enable/disable overrides , and other tasks that require operator input to allow , perform , or modify a task or event. The set of displays will be arranged in a workstation. This workstation will also include a keyboard and mouse that will be used to interface with the system. 7.2.3.3 Control Room/Human-Machine Interface Description The operator will have direct visualization of critical values and the ability to input control functions into the FPC system. The FPC system dedicated displays will perform the following functions: * *

  • Static display -This display will show critical measurement values and perform the function of an annunciator panel. This fixed display panel will not provide any interactive control functionality.

Alarm/event annunciator display panel -This panel will display any event or alarm that is defined for the process. The display will enable the operator to acknowledge current events and alarms , and will provide a historical record of events. Dynamic interface display panel or HMI -This panel will enable the operator to perform tasks, change modes, enable/disable overrides, and other tasks that require operator input to allow , perform , or modify a task or event. The set of displays will be arranged in a workstation.

This workstation will also include a keyboard and mouse that will be used to interface with the system. 7-14

.-.;. NWMI ...... ... ... ........... * !*. * . NOITHWEST MEIMCAL &SOTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.2.3.4 Building Management System The BMS will control the facility ventilation system and receive indications from the fire protection , FPC , and process vessel ventilation systems. The primary purpose of the BMS is to control the air balance of the facility ventilation system and to shut down the facility ventilation system in the event of receiving an alarm from the fire protection system or off-normal conditions indicated by the FPC or engineered safety features.

The operator will have direct visualization of critical va lues and the ability to input control functions into the BMS. The BMS dedicated displays will perform the following functions in the control room: * *

  • Static display -This display will show critical measurement values and perform the function of an annunciator panel. This fixed display panel will not provide any interactive control functionality.

Alarm/event annunciator display panel -This panel will display any event or alarm that is defined for the process. The display will enable the operator to acknowledge current events and alarms, and will provide a historical record of events. Dynamic interface display panel or HMI -This panel will enable the operator to perform task s , change modes , enable/disable overrides, and other ta s ks that require operator input to allow , perform , or modify a task or event. The set of displays will be arranged in a workstation.

This workstation will also include a keyboard and mouse that will be used to interface with the system. 7.2.3.5 Fire Protection System The fire protection system will report the status of the fire protection equipment to the central alarm s tation and the RPF control room with sufficient information to identify the general location and progres s of a fire within the protected area boundaries.

Initiating devices for the fire detection and alarm s ubsystem , including monitoring devices for the fire s uppression subsystem, will indicate the presence of a fire within the facility.

Once an initiating device activates, signals will be sent to the fire alarm control panel. The fire alarm co ntrol panel will transmit signa ls to the central alarm station and perform any ancillary functions. As an example , signals from the fire control panel may initiate actions such as shutdown of the ventilation equipment or actuating the deluge valves. The fire protection system is described in Chapter 9.0 , Section 9.3. 7.2.3.6 Facility Communication Systems The RPF communication sys tems will relay information within the facility durin g normal and emergency conditions.

The sys tems are designed to enable the RPF operator on duty to be in communication with the supervisor on duty , health physics staff, and other personnel required by the technical specifications, and to enable the operator , or other staff, to announce the existence of an emergency in all areas of the RPF complex. Two-way communication will be provided between all operational areas and the control room. Facility communications system is described in Chapter 9.0 , Section 9.4. 7-15

.; .. NWMI ::.**.*.*.* ..... .. .. ..

  • NDllTHWESTMEOtW.ISOTOPfS NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.2.3.7 Analytical Laboratory System The analytical laboratory will support the production of the Mo product and recycle of uranium. Samples from the process will be collected , transported to the laboratory , and prepared in the laboratory gloveboxes and hoods, depending on the analysis to be performed. The analytical laboratory equipment will be provided as vendor package units. Control room monitoring of the analytical laboratory will be limited to the facility systems , including ventilation and radiation monitoring systems. Analytical laboratory system is described in Chapter 9.0 , Section 9.7.3. 7.2.4 System Performance Analysis The RPF I&C system will monitor the processes and ESFs when required.

The IROFS will be managed by the FPC system. The FPC system will provide the central decision-making processor that evaluates monitored parameters from the various plant instrumentation and from the radiation monitoring systems of the CAMs , CAAS , and RAMs. The analysis herein discusses safety as it relates to the IROFS design criteria and design basis. Potential variables , conditions , or other items that will be probable subjects of technical specifications associated with the RPF I&C systems are provided in Chapter 14.0 , "Technical Specifications

." 7.2.4.1 Facility Trip and Alarm Design Basis The design basis information for the FPC system trip functions is based on the following two requirements from Title 10, Cod e of F e d e ral R e gulation s, Part 70 (10 CFR 70), " Domestic Licensing of Special Nuclear Material." *

  • Double-contingency principle

-Process designs should incorporate sufficient factors of safety to require at least two unlikely, independent , and concurrent changes in process conditions before a criticality accident is possible (baseline design criteria of 10 CFR 70.64 , "Requirements for New Facilities or New Processes at Existing Facilities

," paragraph

[9]). The safety program will ensure that each IROFS will be available and reliable to perform its intended function when needed and in the context of the performance requirements of this section (10 CFR 70.61, " Performance Requirements

," paragraph

[e]). The FPC system trip and alarm annunciation are protective functions and will be part of the overall protection and safety monitoring systems for the RPF. The specific equipment design basis for the instrumentation and equipment used for the FPC system trip and alarming functions is discussed in Section 7.2.2. The following discussion relates to the design basis used for monitoring specific signal values for RPF trips and alarms , requirements for performance , requirements for specific modes of operation of the RPF and the FPC system , and the general design criteria noted in Table 7-1. 7.2.4.1.1 Safety Functions Corresponding Protective or Mitigative Actions for Design Basis Events IEEE 603-2009, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations (Sections 4a and 4b). The results of the integrated safety analysis (ISA) for the RPF structures , systems, and components (SSC) are discussed in Chapter 13.0 , " Accident Analysis." Conditions that require monitoring and the subsequent action to be taken are described in Chapter 13.0. 7-16

.;.-.;* .. NWMI ..*... ..* .... ..... .... .. ' !*. * ! .*

  • NCMITHWtST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.2.4.1.2 Variable Monitored to Control Protective or Mitigative Action IEEE 603-2009 (Section 4d). The list of variables to be monitored in the RPF to eliminate or reduce the exposure for the operator will be provided in the Operating License Application.

7.2.4.1.3 Functional Degradation of Safety System Performance IEEE 603-2009 (Section 4h). These design requirements will be factored in and will be evaluated in the Operating Licensing Application.

7 .2.4.2 Analysis 7.2.4.2.1 Facility Process Control System Trip Function Conformance to Applicable Criteria The FPC system will perform a trip as a protective function as part of the RPF safety analysis.

The associated design criteria are discussed in Sections 7.2.1 and 7.2.2. The following discussions relate to conformance to the criteria for the FPC system trip function. 7.2.4.2.2 General Functional Requirement Conformance IEEE 603-2009 (Section 5). The FPC system will initiate and control ESF activation and isolation when the system detects an off-normal event appropriate for activation. The FPC system trips are discussed in Section 7 .2.4.1. These monitored values and subsequent trips are a re s ult of the preliminary accident analysis in Chapter 13.0 and provide a means to mitigate or reduce the consequences from the design basis accident to acceptable levels. 7.2.4.2.3 Requirements on Bypassing Trip Functions Conformance IEEE 603-2009 (Sections 5.8, 5.9, 6.6, and 6.7). Trip override or bypass is recognized as a design requirement.

Channel bypass will be allowed based on the nature of the signal. No channel bypass will be allowed without a visual indication on the FPC system display and recording the bypass event in the historical log. 7.2.4.2.4 Requirements on Setpoint Determination and Multiple Setpoint Conformance IEEE 603-2009 (Section 6.8). Table 7-1 discusses the criteria to be used for setpoint derivation. Setpoints will be calculated in accordance with ISA-RP-67.04.02 , M e thodologi es f o r th e D e t e rmination of S e tpoints for N ucl e ar Saf ety-R e lat e d Instrum e ntation. 7.2.4.2.5 Requirements for Completion of Trip Conformance IEEE 603-2009 (Section 5.2). The ESF and the interaction of a mitigative action going to completion will be provided in the design. The FPC system will monitor for a complete trip of the ESF. This information will be available on the operator display for the FPC system and at the local HMI terminals near the hot cell. An alarm/event annunciation will be displayed to the operator.

Section 7.4.1 describes the activation of the ESF , alarm/event strategy , and operator requirements to manually reset the system after a facility trip. 7.2.4.2.6 Requirements for Manual Control of Trip Conformance IEEE 603-2009 (Section 6.2). The FPC system will have the ability to perform a manual activation of the ESF. Section 7.4.1 describes the activation of the ESF , alarm/event strategy , and operator requirements to manually reset the system after a facility trip. 7-17 "NWMI ...... *.t: ** :.::: .. *****. * *. *

  • NORTHWEST MEDtCAl ISOTDPll NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.2.4.3 Conclusion Th e I&C systems fo r th e RP F w ill m eet t he s t a ted d es i gn c r i t e ri a and d esign ba s i s r e quir e ments o utlin e d i n NU RE G-1 53 7, Guide li nes for P repa r ing a n d R ev i ewing Ap p lications for th e L icensi n g of Non-P owe r R eac t ors -Format a n d Co n tent. A c ro sswa lk of th e I&C s ub sys t e m s, a l o n g with a cross-r efe r e n ce to s p ec ifi c de s i gn c rit e ri a, i s pr ese nt e d in T a b le 7-2. Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Mean s (5 pages) Criteriaa Design basis applicability Functional means IEEE 3 79
  • FPC sys t e m Si n g l e fa ilur e cr i te ri o n
  • FPC sys t e m di s pl ay IEEE 577 Reli a bility a nal y sis criterion
  • FPC sys t e m I R OFS e nd d ev i ces
  • ESFs m a nu al i so l atio n
  • F PC s ystem
  • FPC system di s pla y
  • FPC sys t e m Sta nd a rd criteria safety
  • FPC sys t em d isplay syste m
  • FPC sys t e m IR OFS e nd d ev i ces IEEE 384 Independence of C lass 1 E equipment a nd circuits I EEE 323 Q u a lif y ing C l ass I E E quipm e n t IEEE 344 Recommended practice for s e ismic qualification
  • ESFs m a nu al i solat i o n
  • FPC s y stem display
  • ESFs manual isol a tion
  • FPC sys t e m di s pl ay
  • FPC sys t e m IR OFS e nd d ev i ces
  • ESFs m a nu a l iso l atio n
  • FPC system display
  • ESFs manual isolation 7-18
  • Safety D CS pr eapprove d pl a t fo rm
  • R e d un d a nt ind e p ende n t i so l a t ion compo n e nt s
  • R e d un d a nt o p erator i n te r face workstat i o n s
  • R edu nd a n t se n so r s
  • A lt erna ti ve m a nu al m ea n s fo r ESF in itiatio n
  • Safet y DCS pre-appro v ed pl a tform for a n SIS Redund a nt independent i s ol a tion components
  • Redundant operator interface workst a tions
  • R e dundant s ensor s
  • Alt e rnati v e man ua l means for ESF initiation See Sec t ion 7 .3 for detai l s.
  • IE E E 603 and IE EE 379 wer e used during development o f the Con s truction Permit Application.
  • Additional detail s w i ll be dev e lop e d for the Operating Lic e ns e Application.
  • Standa rd s upp orts se l ec t io n a nd q u a li ficat i o n of equi pm e nt to be C l ass 1 E use q u a lifi e d.
  • T h is sta nd a rd w i ll b e r eeva lu ate d i n t h e Operat ing L i cense A ppli ca ti on for app li ca bilit y.
  • Standard supports s election and qu a lification of equ i pment to be Class IE us e qualified.
  • Standard will be reevaluated in the Operating License Application for applicability.

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  • NOmfWUT llEDtCAl &SOTOPl.I NWM l-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Contro l Systems Table 7-2. Instrumentation and Contro l Criteria Crosswalk with Design Basis Applicability and Function Means (5 pages) Criteriaa Design basis applicability Functional means IEEE 338
  • F P C sys t e m C ri ter i a for t h e p e ri o di c
  • F P C sys t em d is pl ay s ur vei ll a n ce tes tin g of
  • F P C syste m IRO FS e nd d ev i ces safety sys t e m s
  • FPC system Criteria for accident
  • FPC system display monitoring in s truments
  • FPC system JROFS end device s IEEE 7-4.3.2 Cr it e ri a fo r di g i ta l co mput e r s in safe t y sys t e m s
  • F P C sys t e m
  • F P C sys t em d is pl ay
  • FPC s ystem C onfiguration
  • FPC s ystem display management i n sys tems
  • HMI display s and software en g ine e ring I EEE 829 Software a nd system test d oc um e n tation IEEE 1012 Criteria for software verification and validation
  • F P C sys t e m
  • FPC syste m di sp l ay
  • H M I di s pl ays
  • F PC system display
  • HMI displays 7-19
  • S t a nd a rd s upp o rt s se l ec ti o n of e quipm e nt; w h ich r es ult e d in t h e u se o f ge n eral d es i g n crite ri a (pr ese n te d in C h a pt e r 3.0) durin g d eve l o pm e nt of t h e Co n st ru c t io n P e rmi t App l icat i o n.
  • Sta nd a rd w ill b e reeva lu a t e d in the O p era tin g Lice n se A ppli cat i o n for app l ica b i li ty.
  • Standard supports s election of accident monitoring equipment (e.g., radiation monitoring , annunciation), which resulted in the use of general design criteria (presented in Chapter 3.0) during development of the Con s truction Permit Application.
  • Standard will be reevaluated in the Operating Licen s e Application for applicability.
  • P rogra mmin g software mu s t comp ly w ith t h is cr i te ria a nd w i t h th e NW M I Softwa r e Q ua l ity Ass u ra n ce Pl a n (pr e p are d durin g d eve l o pm e n t o f the O p era tin g Lice n se App li cat ion), w h ic h w ill b e d eve l o p ed p e r t h e des ign c r i t e ria o u t lin e d in C h ap t e r 3.0 a nd this s t a nd ard.
  • Softwa re a nd h a r dwa r e u se d for t h e displays fo r t h e F P C syste m a nd H M I mu st a l so fo ll ow g ui de lin es set fo rth in t h is s t anda rd.
  • S t anda r d wi ll b e r eeva lu a t e d i n t h e Op era ting L i ce n se Ap pli ca ti o n for app li ca bilit y.
  • Complie s with IEE E 7-4.3.2 and the NWMI Software Quality Assurance Plan Standard will be re ev aluat e d in the Operating Licens e Application for applicability.
  • Co m p li es wit h IEEE 7-4.3.2 a nd th e NWM I Softwa r e Qu a li ty Ass u ra n ce Pl a n
  • S t a nd a rd w ill b e reeva lu a t e d in t h e O p era ting L i ce n se A pplic a tion fo r a p p li ca bil ity.
  • Complies with IEEE 7-4.3.2 and the NWMI Software Quality A s surance Plan
  • Standard will be reevaluated in the Operating License Application for applicability.

.. ; ... .. NWMI ...... ..* **: ........... . NOllTNWESTMEDtW.JSOTOl"fS NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (5 pages) Criteriaa Design basis applicability Functional means IEEE 1028 Software reviews and audits ANS-10.4 Verification and validation for safety software

  • FPC system display
  • HMI displays
  • FPC system display
  • HMI displays
  • FPC system ANSI/ISA 67.04.01 Setpo int s for nuclear safety-related instruments
  • FPC system IROFS end devices ANSI/ISA 84.00.01, Parts l , 2, and 3 Functional safety: safety instrumented systems for the process industry sector
  • FPC system display
  • FPC system Human-system
  • FPC system display interface design review
  • FPC system Review guidelines on software languages for use in nuclear power plant safety systems NUREG/CR-6090
  • FPC system PLC an d applications in nuclear reactor systems 7-20
  • Complies wit h IEEE 7-4.3.2 and the NWMI Software Quality Assurance Plan
  • Standard wi ll be reevaluated in the Operating License App li cat ion for applicability.
  • Complies with IEEE 7-4.3.2 and the NWMI Software Quality Assurance Plan
  • Standard will b e reevaluated in the Operating License Application for applicability.
  • Incorporated into overa ll design and the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicability.
  • Standard supports the design and development of non-safety-related systems that rely on safety, reliability , and functionality and was used during development of the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicability.
  • Standard supports the design and development of non-safety-related systems that pertain to control room arrangement, scree n developments, and operator interface , and was used during development of the Construction Permit Applicat i on.
  • Standard wi ll be reevaluated in the Operating License Application for applicability.
  • Standard supports the design , development , and re v iew of safety-related software and was used during development of the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicability.
  • Standard s upp orts the d es i gn, development, and r eview of safety-r e lat ed and safety-re l ated software and was used during development of the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicab il ity.

...... .. .... .. .. . ..*... * *. * !

  • NOmfWEST MED1CAl ISOTDPH NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-2. Instrumentation and Control Criteria Crosswalk with Design Basis Applicability and Function Means (5 pages) Criteriaa Design basis applicability Functional means EPRI TR-106439 Guideline on e v aluatio n/acceptanc e of commercial grade digital equipment for nucl e ar safety applications Re g ulatory Guide 1.15 2 C rit e ria for u s e o f comput e r s in safe t y sys t e m s
  • FPC system display
  • HMI displays
  • F PC s y s tem
  • F PC s y s tem di s pl ay
  • HMI di s pl ays
  • FPC system IROFS end devices R eg ulatory Guide 5.71 Cy ber s ecurit y progr a m s for nu c le a r fac iliti es
  • ESFs manual isolat i on
  • F PC s y s tem
  • F P C sys t e m di s pl ay
  • HMI di s pl ay a F ull r efe r ences a r e p rov id ed in Sec ti o n 7. 7. C AAS critica li ty acc i d e nt a l ar m system. C AM co ntinu o us a ir m o n itor. DCS di g i ta l co ntr o l syste m. ESF e n g in ee r e d safe t y fea tur e. F P C fac ili ty proc ess co ntr o l. H M! hum a n-m a chin e in te r face. IR OFS NWM I P LC RAM S I S 7-21
  • Standard supports the design, development, and review of safety-related systems that pertain to obtaining software or hardware for the FPC system , HMI displays , and data acquisition systems, and was used during development of the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicability.
  • St a ndard s upport s th e d e si g n a nd d eve lopm e nt of r e dundant sa f e t y PL C pl a tform s, FP C sys t em r e dund a nt HMI wo rk s t a tion s , a nd o p e r a t o r int e r fa c e wo rk s t a tion s , a nd was u se d durin g d eve lopm e nt o f the C on s truction P e rmit A pplic a tion.
  • S t a nd a rd w ill b e r eeva lu a t e d in th e Op e ratin g Lic e n se Applic a tion for a ppli c abili ty.
  • Standard supports the design and development of high-integrity safety PLCs, redundant channels for ESFs , redundant operator interface workstations , redundant sensors , and alternative manual means for ESF initiation , and was used during development of the Construction Permit Application.
  • Standard will be reevaluated in the Operating License Application for applicability.
  • C r i t e ria r e qu i r e th e d eve lopm e nt o f a d es ign a pp roa ch a nd impl e m e nt a ti o n fo r cy b e r sec uri t y.
  • St an d a rd w ill b e reeva lu a t e d in th e Op era tin g Li ce n se A ppli ca ti o n fo r a pp lica bili ty. i tem s re li ed o n fo r s afety. ort h west M e di cal I so t o p es , L L C. p rogra mm a bl e l o g i c co ntroll e r. ra di a ti o n a l a rm m o nit o r. safety in s trum e nt e d sys t e m.

.; .. ;. NWMI ...... .. *.. .*.******* * ! * * . NORTMWEST MEDICAi. tsOTOPlS NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.3 PROCESS CONTROL SYSTEMS The process control systems for the RPF will include SNM preparation and handling processes and radioisotope production processes.

SNM preparation and handling processes include uranium recovery and recycle , and target fabrication. Radioisotope production processes include target receipt and disassembly , target dissolution , Mo recovery and purification , and waste handling. The RPF process control will be administered by the FPC system and is described in Section 7.2.3. The FPC system will perform the following high-level process functions.

  • *
  • Monitor the remote valve position for routing process fluid for inter-equipment process fluid transfers

-For specific transfers identified by the operator , the FPC system will provide a permissive to allow for the active pump in that circuit to be energized once the operator has manually configured the routing. Monitor and control inter-equipment process fluid transfers in the RPF -For transport requiring a pump , the FPC system will control the ability of the pump to be energized.

For specific transfers , the FPC system will provide controlled fluid flow transfers based on a loop flow control. The operator wil 1 initialize the transfer of fluids. Other process fluid transfers , including: Dissolved low-enriched uranium (LEU) solution to the Mo recovery and purification system Uranium solution to the uranium recovery and recycle system Liquid wastes to the waste handling system The I&C system for process utilities and support systems and for the ventilation systems will be described in more detail in the Operating License Application. The process systems described below provide for reliable control of the SNM preparation and handling process and the radioisotope production processes , and include: * * *

  • Range of operation of the sensor that is sufficient to cover the expected range of variation of the monitored variable during normal and transient process operation Reliable information about the status and magnitude of the process variable necessary for the full operating range of the radioisotope production and SNM recovery and rec y cle processes Reliable operation in the normal range of environmental conditions anticipated within the facility Safe state during loss of electrical power Potential variables , conditions , or other items that will be probable subjects of technical specifications associated with the RPF process control systems are discussed in Chapter 14.0. 7.3.1 Uranium Recovery and Recycle System The uranium recovery and recycle system will process raffinate from the Mo reco v ery and purification system for recycle to the target fabrication system. Two cycles of uranium purification will be included to separate uranium from unwanted fission products using ion exchange. The first ion exchange cycle will separate the bulk of the fission product contaminant mass from the uranium product. Product will exit the ion exchange column as a dilute uranium stream that is concentrated to control the stored volume of process solutions.

Uranium from the first cycle will then be purified by a nearly identical second cycle system to further reduce fission product contaminants to satisfy product criteria.

Each ion exchange system feed tank will include the capability of adding a reductant and modifying the feed chemical composition such that adequate separations are achieved , while minimizing uranium losses. 7-22

.; ... NWMI ...... ..* ... ........... * "* NOmfMST MB>>CAl lSOTOPH NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Due to the variety of process activities performed during uranium recovery and recycle , the system description is divided into the following subsystems

* * * * * * * *
  • 7.3.1.1 Impure uranium collection Primary ion exchange Primary concentration Secondary ion exchange Secondary concentration Uranium recycle Uranium decay and accountability Spent ion exchange resin Waste collection Design Criteria Design criteria for the uranium recovery and recycle I&C systems are described in Section 7.2. 7.3.1.2 De s ign Basi s and Safety Requirement s The de sign basis and safety requirements for the uranium recovery and recyc l e I&C systems are described in Section 7.2. The ESFs for this system are li sted in Chapter 6.0, " Engineered Safety Features." 7.3.1.3 System Description The uranium recovery and recycle I&C system will b e defined in the Operating License Application.

The strategy and associa t ed parameter s for the system are provided below. Preliminary process sequences are provided in Chapter 4.0 to comm un icate the con trol strategy for normal operations , which sets the requirements for the process monitoring and control equipment , and the associated instrumentation. Normal operating functions will b e performed remotely using the FPC system in the contro l room. Table 7-3 lists the anticipated contro l parameters , monitoring parameters , and primary control locations for each sub s ystem. In addition , the implementation of IROFS CS-14 , CS-15 , CS-20 , CS-27 , and RS-10 interlocks for this system are under development.

Details of the control system (e.g., interlocks and permissive signa l s), nuclear and process instrument s, contro l lo gic and elements , indica tion , alarm , and control features wi ll b e developed fo r th e Operating License Applicatio

n. 7-23

.... ;; .. NWMI ..* **: .... .... .. ! *: . NC*TNWEST MEOtcAL tsOTOfl'H NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-3. Uranium Recovery and Recycle Control and Monitoring Parameters (2 pages) Subsystem Control parameters Primary control name (automatic/manual)

Monitoring parameters location Impure . Flowrat e (A) . Density C ontrol room ur a rnum Pump a ctu a tion (M) . Differential pre ss ur e c olle c tion . Pump motor s pe e d (A) . Flowr a t e . T e mp era ture (A) . L e v e l . V a l ve ac tu a tion (NM) . Pr ess ur e . T e mp e ratur e . V a l ve po s ition Primary ion . Flowrate (A) . Analyzer , uranium Control room exchange Pump actuation (NM) . Density . Pump motor speed (A) . Differential pressure . Temperature (A) . Flowrate . Valve actuation (NM) . Flowrate totalizer . Level . Pressure

  • Temperature . Valve position Primary . Den s it y (A) . Anal y z e r , ur a nium C ontrol room co n ce nt ra tion Flowr a t e (A) . D e n s it y . L e v e l (A) . Diff e r e nti a l pr ess ur e . Pump ac tu a tion (NM) . Flowr a t e . Pump motor s pe e d (A) . Level . T e mp era ture (A) . Pr ess ur e . Valv e ac tu a tion (AI M) . T e mp e ratur e . V a l ve po s ition Secondary ion . Flowrate (A) . Analyzer, uranium Control room exchange Pump actuation (AIM) . Density . Pump motor speed (A) . Differential pressure . Temperature (A) . Flowrate . Valve actuation (NM) . Flowrate totalizer . Level . Pressure . Temperature . Valve position S eco nd a ry . Densit y (A) . Anal yz er , uranium C ontrol ro o m co n ce ntr a tion . F lowr a t e (A) . D e n s it y . L ev el (A) . Diff e r e n tia l pr ess ur e . Pump ac tu at i o n (AI M) . Flowr a t e . Pump motor s p ee d (A) . L eve l . Temp era tu re (A) . Pre ss ur e . V a lve a ctu a tion (NM) . Temp e ratur e . Val ve po s ition Uranium . Flowrate (A) . Density Control room recycle . Pump actuation (AIM)
  • Differential pressure . Pump motor speed (A) . Flowrate . Valve actuation (NM) . Level . Pressure . Temperature . Valve position 7-24 "NWMI ..**.. ..* ... ........ *.* . ' !*. * !
  • NOflJTHWEST M£DtCAl ISOTOPU NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-3. Uranium Recovery and Recycle Contro l and Monitoring Parameters (2 pages) Subsystem Control parameters Primary control name (automatic/manual)

Monitoring parameters location Ura nium decay . Flowrate (A) . D ensity Control room a nd . Pump act u at ion (AIM) . Diff erentia l pr ess ur e accountability . Pump motor speed (A) . Flowrate . Temperature (A) . Level . Valve actuation (AIM) . Pr essure . Temp eratu re . Valve position Spent ion . Flowrate (A) . Analyzer, uranium Control room exchange resin . Pump actuation (AIM) . Differential pressure . Pump motor speed (A) . Flowrate . Valve actuation (AIM) . Level . Pressure . Valve position Waste . Flowrate (A) . D ensity Control room co ll ection . Pump actuat ion (AIM) . Differential pr essure . Pump motor spee d (A) . Flowrate . Temperature (A) . Level . Valve actuat ion (AIM) . Pr essure . Valve position Table 7-4 provides a preliminary listing of the interlocks and permissive signals that have been identified. These device s will be further developed and detailed information will be provided in the Operating License Application. Table 7-4. Uranium Recycle an d Recovery System Interlocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock Impure uranium collection tank (U R-TK-lOOA) low-l evel PLC N I A switch (typical of eight tanks) Impure uranium collection tank (UR-TK-lOOA) high-level PLC N I A switch (typical of eight tanks) Impur e uranium collection tank (UR-TK-1 OOA) high-PLC N I A temperature switc h (typical of eig ht tanks) IX feed tank 1 (UR-TK-200) low-level switch PLC N I A IX feed tank 1 (U R-TK-200) hi g h-l eve l switc h PL C N I A IX feed tank 1 (UR-TK-200) high-temperature switch PLC N I A IX column IA (UR-IX-240) high-uranium alarm (AAH-252)

PLC N I A IX column IA U solution filter (UR-F-250) high-differential PLC N I A pressure alarm IX column IA waste filter (UR-F-255) high-diff e renti a l PL C N I A pressure alarm IX column lB (UR-IX-260) high-uranium alarm (AAH-272)

PLC N I A IX column lB U so lution filter (UR-F-270) high-differential PLC N I A pre ss ure alarm 7-25

...... .. NWMI ...... ..* .... ..... .... .. , 0 ! *. *. * . NCNITNWESf MEDK:Al. ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock IX Column lB Waste Filter (UR-F-275) high-differential PLC pressure alarm Co ncentrator 1 feed tank (UR-TK-300) low-level switch PLC Concentrator 1 feed tank (UR-TK-300) high-level switch PLC Co ncentrator 1 (UR-Z-320) low-liquid level a larm PLC Concentrator 1 (UR-Z-320) high-liquid level alarm PLC Co nc e ntrator 1 (UR-Z-320) demi ste r high-differential pre ssure PLC ala rm Concentrator 1 (UR-Z-320) condenser high-differential PLC pressure alarm Co ncentrator 1 (UR-Z-320) condenser high-offga s temperature PL C a larm Condensate sample tank IA (UR-TK-340) high-liquid level PLC alarm Con den sate sample tank 1 A (UR-TK-34 0) high-uranium Hard-wired switc h (AE-356) N I A N I A N I A N I A N I A N I A N I A N I A N I A Reroute condensate transfer to UR-TK-300 (position V-396, close V-397) C lo se IX column eluent addition control valves (V-244 and V-264) Condensate delay tank 1 (UR-TK-370) high-liquid level alarm Condensate sa mple tank lB (UR-TK-340) high-liquid level a larm PLC N I A Condensate sample tank 1B (UR-TK-370) high-uranium switch (AE-386) IX feed tank 2A (U R-TK-400) low-level switc h IX feed tank 2A (UR-TK-400) high-level switch IX feed tank 2A (UR-TK-400) high-temperatur e switch IX feed tank 2B (UR-TK-420) low-level switch IX feed tank 2B (UR-TK-420) high-level switch IX feed tank 2B (UR-TK-420) high-temperature switch IX column 2A (UR-IX-460) high-uranium alarm (AAH-472)

IX column 2A U solution filter (UR-F-470) high-differential pressure alarm IX column 2A waste filter (UR-F-475) high-differential pressure alarm IX column 2B (UR-IX-480) high-uranium alarm (AAH-492) 7-26 PL C N I A Hard-wired Permissive to route condensate to PLC PLC PL C PLC PLC PLC PLC PLC PLC PLC WH-TK-420 (position V-496, open V-397) Permissive to open IX column eluent addition control valves (V-244 and V-264) N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A

.; ... NWMI ..*... ..* ... ........... ' !*. * . NOITifWUT liWMCAl ISOTOf'U NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-4. Uranium Recycle and Recovery System Interlocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock IX co lumn 2B U so luti on filter (UR-F-490) high-differential pressure alarm IX column 2B waste filter (UR-F-495) high-differential pressure alarm Co nc entrator 2 feed tank (UR-TK-500) low-level switch Concentrator 2 feed tank (UR-TK-500) high-level switch Co n centrator 2 (U R-Z-520) low-liquid l eve l a l arm Concentrator 2 (UR-Z-520) high-liquid level alarm Co n centrator 2 (UR-Z-520) demister high-differential pressure a l arm Concentrator 2 (UR-Z-520) condenser high-differential pressure alarm Concentrator 2 (UR-Z-520) condenser high-offgas temperature a l arm Condensate sample tank 2A (UR-TK-540) high-liquid level alarm Condensate sample tank 2A (UR-TK-540) high-uranium sw itch (AE-556) Condensate delay tank 2 (UR-TK-560) high-liquid level alarm Condensate sample tank 2B (UR-TK-570) high-liquid level alarm Condensate sample tank 2B (UR-TK-570) high-uranium switch (AE-586) Concentrate receiver tank (UR-TK-600) high-liquid l evel alarm Concentrate receiver tank (UR-TK-600) high-temperature alarm Product sample tank (UR-TK-620) high-liquid l eve l alarm Product sample tank (UR-TK-620) high-temperature alarm Urani um rework tank (UR-TK-660) high-liquid l evel alarm Uranium rework tank (UR-TK-660) high-temperature alarm Urani um decay tank (UR-TK-700A) high-liquid l eve l a l arm (typica l of 17 tanks) Uranium decay tank (UR-TK-700A) high-temperature alarm (typical of 17 tanks) 7-27 PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC Hard-wired N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A Reroute condensate transfer to UR-TK-500 (position V-596, close V-5 97) Close IX co lumn eluent addition control valves (V-464 and V-4 84) PLC N I A PLC N I A Hard-wired Permissive to route condensate to PLC PLC PLC PLC PLC PLC PLC PLC WH-TK-420 (position V-596, open V-597) Permissive to open IX column eluent addition control valves (V-464 and V-484) N I A N I A N I A NIA N I A N I A N I A N I A NWMI ...*.. .. *.. .*.* .. *.*. 0 !*. * ! 0 NOITH'W(ST MlDICAl tsOlWU NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Tab l e 7-4. Urani um Recycle and Recovery System Interlocks and Permissive Signals (4 pages) Hard-wired or Interlock or permissive input PLC Safety Interlock Uranium accountability tank (UR-TK-720) high-liquid l evel alarm Uranium accountability tank (UR-TK-720) high-temperature alarm Spent resin tank A (UR-TK-820A) high-liquid level alarm Spent resin tank A (UR-TK-820A) high-temperature alarm Spent resin tank B (UR-TK-820B) high-liquid level alarm Spent resin tank B (UR-TK-820B) high-temperature alarm Resin transfer liquid tank (UR-TK-850) high-liquid l eve l alarm IX waste collection I tank (UR-TK-900) high-liquid level alarm IX waste collection I tank (UR-TK-900) high-temperature a l a rm IX waste collection 2 tank (UR-TK-920) high-liquid level alarm IX waste collection 2 tank (UR-TK-920) high-temperature alarm IX ion exc h ange. PL C programmable l o g ic c ontro ll er. 7.3.1.4 System Performance Analysis and Co nclu sion PLC N I A PLC NIA PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A TBD to be determined. The system performance analysis and conclusion for each process system wi ll be provided in the Operating License Application. 7 .3.2 Target Fabrication System The target fabrication system w ill produce LEU targets from fresh LEU material and recycled uranyl nitrate. The syste m will comme nce with the receipt of fresh LEU from the U.S. Department of Ene r gy, and end with packaging new targets for shipment to the university research r eactor faci liti es. Due to th e variety of process activities performed dur ing target fa bri cation , the system description is divided into the following subsystems.

  • * * *
  • * * * *
  • Fresh uranium receipt and dissolution Nitrate extraction Ac id-d eficient uranyl nitrate (ADUN) concentration

[Proprietary Information]

[Propr ietary Information]

[Propri etary Information]

Target fa bric a tion waste Target assem bl y [Proprietary Information]

New target handlin g 7-28

..... ;*. NWMI ...... ..* **: ........... * *.* ! . . NOtmfWHT MEDICAi. tsOTDPH 7.3.2.1 Design Criteria NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Design criteria for the target fabrication I&C systems are described in Section 7.2. 7.3.2.2 Design Basis and Safety Requirements The design basis and safety requirements for the target fabrication I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.3.2.3 System Description The target fabrication I&C system will be defined in the Operating License Application. The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation.

Normal operating functions will be performed remotely using the FPC system HMI in the target fabrication area. Table 7-5 lists the anticipated control parameters , monitoring parameters, and primary control location for each subsystem.

In addition, the implementation ofIROFS CS-14, CS-15, CS-20, CS-27, and RS-10 interlocks for this system are under development.

Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments , control logic and elements , indication, alarm, and control features will be developed for the Operating License Application.

I Table 7-5. Target Fabrication System Control and Monitoring Parameters (2 pages) Subsystem name Fresh uranium receipt and . dissolution . (I 00-series tag numbers) . . . . . . Nitrate extraction . (200-series tag numbers) . . . . . . .

. ADUN concentration . (300-series tag numbers) . . . . . . Control parameters (automatic/manual)

Current (A) Conductivity (A) Flow totalizer (A) Heater actuation (NM) Level (A) Pump actuation (A/M) Temperature (A) Valve actuation (A/M) Analyzer, pH (A) Contactor actuation (M) Flow totalizer (A) Flowrate (A) Level (A) Pump actuation (NM) Pump motor speed (A) Temperature (A) Valve actuation (AIM) Conductivity (A) Density (A) Flowrate (A) Level (A) Pump actuation (A/M) Pump motor speed (A) Valve actuation (A/M) 7-29 Monitoring parameters . Conductivity . Density . Differential pressure . Flowrate . Level . Pressure . Temperature . Analyzer, pH . Density . Differential pressure Flowrate . Level . Pressure . Pump motor speed . Temperature . Conductivity . Density . Flowrate . Level . Pressure . Temperature Primary control location Local Local Local

' ....... ; .. NWMI ...... ..* .... ........ *.*

  • 0 *.*
  • NOAT N WEST MEDICAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-5. Target Fabrication System Control and Monitoring Parameters (2 pages) Subsystem name [Proprietary Information] . (400-series tag numbers) . . . . . [Proprietary Information] . (500-series ta g numbers) . . . . . [Proprietary Information] . (600-series tag numbers) . . . . . . Target fabrication waste . (700-series tag numbers) . . . . Target assembly [Proprietary Information]

New target handling Control parameters (automatic/manual)

Level (A) Pump actuation (NM) Tank agitator actuation (AIM) Tank agitator speed (A) Temperature (A) Valve actuation (AIM) Flowrate (A) Pump actuation (NM) Pump motor s peed (A) Temperatur e (A) Valve actuation (AIM) Vibration di s persion asse mbly actuat ion (M) Analyzer, hydrogen (A) Analyzer , oxygen (A) Flow totalizer (A) Level (A) Tank agitator speed (M) Temperature (A) Valve actuation (AIM) Flowrate (A) Level (A) Pump actuati on (AIM) Pump motor speed (A) Valve act uation (AIM) TBD TBD TBD ADUN ac id-defici e nt uranyl nitrate. TBD LEU = low-enriched uranium. Monitoring parameters . Flowrate . Level . Pressure . Temperature . Den s it y . Differential pres sure . Pre ssure . Level . Temp erat ure . Vibration . Analyzer, hydrogen . Analyzer, oxygen . Flowrate . Level . Pressure

  • Temperature . Densit y . Flowrate . Level . Pre ssure . Temperature TBD TBD TBD to b e d e termin ed. Primary control location Local Local Local Local Local Local Local Table 7-6 provides a listing of the target fabrication l&C system interlocks and permissive signals that have been identified.

These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 pages) Hard-wired or Interlock or permissive input PLC Safety interlock Dissolver column (TF-D-100) high-temperature switch Uranium dissolution heat exchanger (TF-E-120) chilled water return high-conductivity switch Uranium dissolution heat exchanger (TF-E-120) differential pressure alarm Uranyl nitrate storage tank (TF-TK-200) level switch 7-30 PLC N I A Hard-wired Close chilled water return control valve (XV-122) on high conductivity PLC N I A PLC N I A I ::.**.*.* .. .. NWMI ........ *.* . ' ! *. ." . NOmfWEST MEOtcAl ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 pages) Interlock or permissive input ADUN evaporator condenser (TF-E-350) chilled water return high-conductivity switch ADUN product heat exchanger (TF-E-360) differential pressure alarm ADUN product heat exchanger (TF-E-360) chilled water return high-conductivity switch ADUN evaporator reboiler (TF-E-330) steam condensate high-conductivity switch ADUN storage tank (TF-TK-400) low-level switch ADUN storage tank (TF-TK-405) low-level switch ADUN storage tank (TF-TK-410) low-level s witch ADUN storage tank (TF-TK-415) low-level switch ADUN storage tank (TF-TK-400) high-level switch ADUN storage tank (TF-TK-405) high-level switch ADUN storage tank (TF-TK-401) high-level s witch ADUN storage tank (TF-TK-415) high-level switch [Proprietary Information] (TF-TK-480) high-level switch [Proprietary Information] (TF-C-500) high-temperature switch Silicone oil heater (TF-E-550) outlet high-temperature switch [Proprietary Information] (TF-Z-660) high-temperature switch [Proprietary Information] (TF-Z-661) high-temperature switch [Proprietary Information] (TF-Z-662) high-temperature switch [Proprietary Information] (TF-Z-663) high-temperature switch [Proprietary Information] (TF-Z-660) door closed switch [Proprietary Information] (TF-Z-661) door closed switch [Proprietary Information] (TF-Z-662) door closed switch [Proprietary Information] (TF-Z-663) door closed switch Reduction furnace offgas heat exchanger (TF-E-670) outlet high-oxygen concentration Reduction furnace offgas heat exchanger (TF-E-670) outlet high-hydrogen concentration 7-31 Hard-wired or PLC Hard-wired PLC Hard-wired Hard-wired PLC PLC PLC PLC PLC PLC PLC PLC PLC PLC Hard-wired Hard-wired Hard-wired Hard-wired Hard-wired PLC PLC PLC PLC PLC PLC Safety interlock Close chilled water return control valve (HV-352) on high conductivity N I A C l ose chilled water return control valve (HV-361) on high conductivity Close steam condensate control va l ve (XV-333) on high conductivity N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A N I A

...-----------------------------

* -... .. NWMI ::.**.*.*.* ........ *.* NCMTHWESTMEDICAllSOTOPCS NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-6. Target Fabrication System Interlocks and Permissive Signals (2 page s) Interlock or permissive input Hard-wired or PLC Safety interlock Aqueous waste pencil tank (TF-TK-700) high-level alarm Aqueous waste pencil tank (TF-TK-705) high-level alarm TCE tank (TF-TK-760) high-level switch Target fabric a tion overflow tank (TF-T K-770) level switch ADUN PL C acid-defici e nt uranyl nitr a t e. programmable lo g i c controll e r. TBD T CE 7.3.2.4 System Performance Analysis and Conclusion PLC PLC PLC PLC to b e det e rmined. trichloroethylen

e. N I A N I A N I A N I A The system performance analysis and conclusion for each process system will be provided in the Operating License Application.

7.3.3 Target

Receipt and Disassembly System The target receipt and disassembly system will include the delivery and receipt of the irradiated target cask , introduction of the irradiated targets into the hot cell , disassembly of the targets , and retrieval and transfer of the irradiated target material for processing.

This system will feed the target dissolution system by the transfer of recovered irradiated target material through the dissolver I hot cell (DS-EN-100) and dissolver 2 hot cell (DS-EN-200) isolation door interfaces.

Due to the variety of activities performed during target receipt and disassembly , the system description is divided into the following subsystems:

  • *
  • 7.3.3.1 Cask receipt Target receipt Target disassembly Design Criteria Design criteria for the target receipt and disassembly I&C system s are described in Section 7.2. 7.3.3.2 Design Basis and Safety Requirements The design basis and safety requirements for the target receipt and disassembly I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.3.3.3 System Description The target receipt and disassembly I&C system will be defined in the Operating License Application.

The strategy and associated parameters for the l&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations , which sets the requirements for the process monitoring and contro l equipment, and the associated instrumentation.

7-32

.**.*.*.* ..... .. NWMI ........... *. ! ! ." NORTHWEST Mt:DacA.L ISOT01'£S NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Normal operating functions will be performed remotely using the FPC system RMI in the truck bay , cask preparation airlock , and the operating gallery. Redundant control functions will be provided in the control room. In addition, the implementation of IROFS CS-14 , CS-15 , CS-20 , CS-27, and RS-10 interlocks for this system are under development.

Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments, control logic and elements , indication , alarm, and control features will be developed for the Operating License Application.

Prior to the start of disassembly operations , the following process control permissive signals will be required.

  • * *
  • Ventilation inside the hot cell is operable . Fission gas capture hood is on and functional.

Irradiated target material collection container is in position under the target cutting assembly collection bin. Waste drum transfer port is open and there is physical space to receive the waste target hardware after disassembly and irradiated target material recovery.

The control parameters and monitoring parameters will be defined during design development for the Operating License Application.

7.3.3.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application. 7.3.4 Target Dissolution System The target dissolution system process will receive the LEU target material from the target receipt and disassembly system and dissolve the uranium and molybdenum-99 (99 Mo) in the solid irradiated target material in hot nitric acid. The concentrated uranyl nitrate solution will then be transferred to the Mo recovery and purification system for further processing.

The target dissolution process will be operated in a [Proprietary Information]

transferred to a collection container.

The collection container will move through the pass-through to a dissolver basket positioned over a dissolver , the target material will then be dissolved and the resulting solution transferred to the Mo recovery and purification system. Target dissolution of irradiated LEU will result in gaseous fission products (iodine [I], krypton [Kr], and xenon [Xe]) with very high radiation fields. A primary function of the process offgas systems will be to control release of these gases both internal and external to the facility. The dissolver off gas treatment system will include the nitrogen oxide (NO x) treatment and fission gas treatment subsystems. Due to the variety of process activities performed during target dissolution , the system description is divided into the following subsystems:

  • * * * *
  • Target dissolution 1 and target dissolution 2 NO x treatment 1 or NO x treatment 2 Pressure relief Primary fission gas treatment Secondary fission gas treatment Waste collection 7-33

..... .. NWMI ...*.. ..* **. .*.* .. *.*.* . "NOllTHWfST MEDICAL ISOTOfJfS 7.3.4.1 Design Criteria NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Design criteria for the target dissolution I&C systems are described in Section 7.2. 7.3.4.2 Design Basis and Safety Requirements The design basis and safety requirements for the target dissolution I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.3.4.3 System Description The target dissolution I&C system will be defined in the Operating License Application.

The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations, which sets the requirements for the process monitoring and control equipment, and the associated instrumentation. Loading of [Proprietary Information]

into the dissolver will involve mechanical handling of the transfer containers.

Operators using remote in-cell cranes and manipulators will perform these functions.

Other normal operating functions will be performed remotely using the FPC sys tem HMI in the operating ga llery. Redundant control functions will be provided in the control room. Table 7-7 lists the anticipated control parameters , monitoring parameter s, and primary control locations for each subsystem.

Details of the control system (e.g., interlocks and permissive signals), control logic , indication , alarm, and control features will be defined in the Operating License Application.

7-34

.; ... ; ... NWMI ......... *.* ........ *.* .. ****** * .*. *

  • NORTHWEST ll£DICAl ISOTOl'U NWMl-20 1 3-021 , R e v. 1 Chapter 7.0 -Instrumentat i on and Control System s Ta ble 7-7. T ar ge t Di ss olution Sys t e m C ontrol and M onitorin g Param e t e r s Subsystem name T a r ge t di ss o lu ti o n 1 a nd 2 N O x t reatme n t I or 2 Pre ss ur e r e li ef Pri m ary fission gas trea t me n t Sec ond a r y fi ss i o n gas t rea tm e nt Waste co ll e ction . . . . . . . .

. . . . . . . . . . . . . . . Control parameters (automatic/manual)

D i ss ol ve r ag it a tor ac tu a ti o n (AI M) D i ss ol ve r ag it a tor s p ee d (A) F l o wra t e (A) P u mp ac tu a tion (AI M) Pump m o t o r s p ee d (A) T e mp e ratur e (A) Va l ve act u a tion (AI M) F l owrate (A) P um p actuat ion (AIM) Pu m p mot or s p ee d (A) Te mp era tur e (A) Valve act u at i o n (AIM) P u mp a c t u a tion (AI M) P u mp m o tor s p ee d (A) Temp era tur e (A) Va l ve a c t u a t i o n (AIM) Tem p erature (A) Va l ve actuatio n (AIM) Va l ve ac tu a tio n (AIM) P u mp actuation (AIM) P u mp mot o r S p ee d (A) T empe ratur e (A) Va l ve ac t uat ion (AIM) NO x ni t ro ge n ox id e. Monitoring parameters . Di ss ol ve r ag it a tor s p ee d . Fl o wr a t e . F lowr a t e tot a l izer . L eve l . Pr ess ur e . R a d i a tion . T e mp e ratur e . Va l ve po s i ti o n . Diffe r e n tial p ressure . F l owra t e . Flowra t e t o t alizer . Leve l . Pr ess ur e . R adiati on . Te mp erature . Va l ve po sit i o n . Fl owra t e . Leve l . Pr ess ur e . V a l v e po s ition . D iffere n tia l pressure . Flowrate . P ress ur e . R a di atio n . Tempera tur e . Valve posi ti on . D i ff e r e nti a l pr ess ur e . F l o wr a t e . Pr ess ur e . R a d i at ion . Te mp e ratur e . Va l ve po s it i o n . D iffere nt ia l p r essure . F l owra t e . Leve l . T e mp e ratur e . P ress ur e . R a di a t i on . Va l ve po s i t ion 7-3 5 Primary control location Op era tin g ga ll e r y O perating ga ll ery Op e ratin g ga ll ery Op erating gallery Op e ra t in g ga ll e r y Op era t ing ga ll ery

..... ;* .. NWMI ..*...... *. * ........... . 0 * ! ." , NOftTHWUT Mfl>tw ISOnftl NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-8 provides a preliminary listing of the target dissolution I&C system interlocks and permissive signa l s that have been identifi ed. In addition, the implementation ofIROFS CS-14, CS-15, CS-20 , CS-27 , and RS-10 interlocks for this system are und er development.

These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-8. Target Dissolution System Interlocks and Permissive Signals (2 pages) Hard-wired or Interlock or permissive input PLC Safety interlock Dissolver I (DS-D-100) high-liquid level alann PLC N I A Dissolver l (DS-D-100) low-liquid level alarm PLC N I A Dissolver I (DS-D-100) high liquid temperature a l arm PLC N I A Dissolver I Condenser (DS-E-130) high gas temperature alann PLC N I A Dissolver 2 (DS-D-200) high-liquid level alarm PLC N I A Dissolver 2 (DS-D-200) low-liquid level alarm PLC N I A Dissolver 2 (DS-D-200) high liquid temperature a l arm PLC N I A Dissolver 2 condenser (DS-E-230) high gas temperature alarm PLC N I A Primary caustic scrubber I (DS-C-3 10) high-liquid l eve l alarm PLC N I A Caustic scrubber I (DS-C-310) high gas temperature PLC N I A NO x oxidizer I (DS-C-340) high-liquid level alann PLC N I A NOx oxidizer I (DS-C-340) high gas temperature PLC N I A NO x a b sorber I (DS-C-370) high-liquid level alarm PLC N I A NOx absorber I (DS-C-370) high gas temperature PLC N I A Primary caustic scru bber 2 (DS-C-4 10) high-liquid level alann PLC N I A Caustic scrubber 2 (DS-C-410) high gas temperature PLC N I A NO x oxidizer 2 (DS-C-440) hi g h-liquid l evel alann PLC N I A NOx oxidizer 2 (DS-C-440) high gas temperature PLC N I A NO x a b sorber 2 (DS-C-470) high-liquid l evel a l ann PLC N I A NOx absorber 2 (DS-C-470) high gas temperature PLC N I A Pressure relief tank (DS-TK-500) high-pressure a l ar m Hard-wired Opens valve to capture dis s olver gases Pressure relief tank (DS-TK-500) high-liquid level alarm PLC N I A Pressure relief tank (DS-TK-500) low-liquid l eve l alann PLC N I A Dryer A (DS-E-610A) high gas temperature alann PLC N I A Primary ad s orber A (DS-SB-620A) high gas temperature alann PLC N I A Filter A (DS-F-630A) high-pressure differential alarm PLC N I A Dryer B (DS-E-610B) high gas temperature alarm PLC N I A Primary adsorber B (DS-SB-620B) high gas temperature alann PLC N I A Filter B (DS-F-630B) hi gh-pr essure differential a l ann PLC N I A Dryer C (DS-E-610C) high gas temperature alann PLC N I A Primary adsorber C (DS-SB-620C) hi gh gas temperature alann PLC N I A Filter C (DS-F-630C) high-pressure differential alarm PLC N I A 7-36

... .. NWMI ........ *.* * * **

  • NOl1lfMST llEDtCAl tSOTOf'U NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-8. Target Dissolution System Interlocks and Permissive Signals (2 pages) Hard-wired or Interlock or permissive input PLC Safety interlock Secondary adso rb er A (DS-SB-730A) high gas temperatur e a l a rm Secondary adsorber B (DS-SB-730B) high gas temperature a larm Secondary adsorber C (DS-SB-730C) high gas temperature a larm Waste collection and sampling tank 1 (DS-TK-800) high-liquid le ve l alarm W aste collection and sampling tank I (DS-TK-8 00) hi g h-liquid temperature a l ar m Waste collection and sampling tank 2 (DS-TK-820) high-liquid level alarm W aste co ll ection and sa mplin g t a nk 2 (DS-TK-820) hi g h-liquid s temperature a l ar m N I A NO x not app li cab l e. = nitrogen oxide. PLC 7.3.4.4 System Performance Analysis and Conclusion PLC PLC PL C PLC PL C PLC PLC N I A N I A N I A N I A I A N I A N I A programmable l o g ic co ntr o ll er. T h e system performance ana l ysis and conc lu sion for each process system will be provided in the Operating License Application. 7.3.5 Molybdenum Recovery and Purification System The Mo recovery an d purification system will recei ve the impure Mo/uranium solution from the target dissolution system into feed tank lA and fee d tank lB (MR-TK-100 and MR-TK-140) lo cated in the tank hot ce ll. The Mo/uranium so lution will th e n be tran sfe rred to process hot cells and processed through three separate ion exchange unit operations to achieve the desired product criteria.

A co ll ection container holding the s eparated and purified Mo product material wi ll be u se d for final chemica l adjustment and sa mpling for ve rification of batch acceptance. The product wi ll b e samp l ed and weig hed , pl aced in sta inless steel b ottles with lids app li e d an d tightened , loaded into shie ld ed containers , a nd then s hipp ed in an ap pro ved cask. Due to the va riety of activitie s performed during Mo recovery a nd purification , the system description i s divided into the following s ub sys tems: * * *

  • 7.3.5.1 Primary ion exchange Secondary ion exchange Tertiary ion exchange Mo product Design Criteria Design criteria for the Mo recovery and purification J&C sys tems a re described in Section 7.2. 7.3.5.2 Design Basis and Safety Requirements The de sign basis and safety requirements for the Mo recovery and purification I&C systems are described in Section 7.2. The ESFs for thjs system are li sted in C hapter 6.0. 7-37

..... ;. NWMI ...... ..* *.. ........ *.* ' " "NOITtfWEST M£otCAl ISOTOPES 7.3.5.3 System Description NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems The Mo recovery and purification I&C system wi ll be defined in the Operating License Application. The strategy and associated parameters for the l&C system are provided below. Preliminary process seq uence s are provided in Chapter 4.0 to communicate the control strategy for norm a l operations , which sets the requirements for the process monitoring and control equipment , and the associated instrumentation. Operators u sing remote in-cell manipulators wi ll perform the product transfer and packaging functions. All other normal operating functions will be performed remotely using the FPC system HMI in the operating gallery. R edundant contro l functions will be provided in the contro l room. Table 7-9 lists the anticipated control parameters , monitoring parameters , and primary control loc ations for each subsystem.

In addition , the implementation of IROFS CS-14 , CS-15 , CS-20, CS-27 , and RS-10 interlocks for this sys tem are under development.

Details of the control system (e.g., interlocks and permissive signa ls), nuclear and process instrument s, control logic and elements , indicat io n , alarm , and co ntrol features will be developed for the Op erating License Application. Tab l e 7-9. Molybdenum Recovery and Purification System Control and Monitoring Parameters Subsystem name Primary ion exchange . Secondary ion exchange . Tertiary ion exchange . Molybdenum product

  • Control parameters (automatic/manual)

Temperature (A) Valve act u ation (AIM) Pumps (M) Pumps (M) Actuate capping unit (M) Monitoring parameters . Density . Flowrate . Level . Temperature . Pressure . Radiation . Va l ve position . Temperature . Den s ity . Flowrate . Level . Pressure . Temperature . Weight Primary control location Operating gallery Operating gallery Operating gallery Operating gallery Tab l e 7-10 provides a preliminary listing of the Mo recovery and purification system interlock s and permissive signals that have been identified.

These devices will be further developed and detailed information will be pro vided in the Operating License Application.

Table 7-10. Mo l ybdenum Recovery an d Purification System Interlocks and Permissive S i g nal s Interlock or permissive input Feed tank lA (MR-TK-100) high-liquid level alarm Feed tank IA (MR-TK-100) low-liquid l evel a l arm Feed tank lA (MR-TK-100) high-temperature alarm Feed tank lA (MR-TK-100) high-pressure alarm Feed tank lB (MR-TK-140) high-liquid level alarm Feed tank lB (M R-TK-140) low-liquid le vel alarm 7-38 Hard-wired or PLC PLC PLC PLC PLC PLC PLC Safety Interlock N I A N I A N I A N I A N I A N I A

.; ... ; .. NWMI ::.**.*.* .. ........ *.* * . ffOITHWHT MEDICAL ISOTOP'l:S NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-10. Molybdenum Recovery and Purification System Interlocks and Permissive Signals Interlock or permissive input Hard-wired or PLC .. Feed tank lB (MR-TK-140) high-temperature alarm Feed tank lB (MR-TK-140) high-pressure alarm U solution collection tank (MR-TK-180) high-liquid level alarm U solution collection tank (MR-TK-180) low-liquid level alarm U solution collection tank (MR-TK-180) high-pressure alarm Waste collection tank (MR-TK-340) high-liquid level alarm Waste collection tank (MR-TK-340) low-liquid level alarm Waste collection tank (MR-TK-340) high-pressure alarm N I A PL C not a pplicabl e. = programmable lo g ic c ontroller.

u 7.3.5.4 System Performance Analysis and Conclusion uranium. PLC PLC PLC PLC PLC PLC PLC PLC The system performance analysis and conclusion for each process system will be provided in the Operating License Application. 7.3.6 Waste Handling System N I A N I A N I A N I A N I A N I A N I A N I A The waste handling system will consist of storage tanks for accumulating waste liquids and adjusting the waste composition , and the equipment needed for handling and encapsulating solid waste. Liquid waste will be split into high-dose and low-dose streams by concentration.

The high-dose fraction will be further concentrated and adjusted. Liquid waste will then be mixed with an adsorbent material.

The solid waste streams will be placed in a waste drum and encapsulated by adding a cement material to fill voids remaining within the drum. All high-dose waste streams will be held for decay and shipped to a disposal facility. Due to the variety of activities performed during waste handling , the system description is divided into the following subsystems

* * * * * * * *
  • High-dose liquid waste collection Low-dose liquid waste collection Low-dose waste evaporation High-dose liquid waste solidification Low-dose liquid waste solidification Spent resin dewatering Solid waste encapsulation High-dose waste decay High-dose waste handling 7.3.6.1 Design Criteria Design criteria for the waste handling I&C systems are described in Section 7.2. 7.3.6.2 Design Basis and Safety Requirements The design basis and safety requirements for the waste handling I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7-39

..... .. NWMI ..**.. ..* .... ........ *.* .

  • NORTHWESTWDICAllSOTOPES 7.3.6.3 System Description NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems The waste handling I&C system will be defined in the Operating License Application.

The strategy and associated parameters for the I&C system are provided below. Preliminary process sequences are provided in Chapter 4.0 to communicate the control strategy for normal operations , which sets the requirements for the process monitoring and control equipment , and the associated instrumentation.

All normal operating functions for low-dose liquid solidification will be controlled locally using FPC system HMis in the low-dose waste room (Room W 107). A local control room will be provided in this room for most waste handling operations. All normal operating functions for the high-dose liquid waste solidification , high-dose waste decay, spent resin dewatering , and solid waste handling hot cell operations will be controlled from the waste handling control room. Liquid waste collection and low-dose liquid waste evaporation operations will be controlled from the control room. Table 7-11 lists the anticipated control parameters , monitoring parameters , and primary control locations for each subsystem.

In addition , the implementation of IROFS CS-14 , CS-15 , CS-20 , CS-27 , and RS-10 interlocks for this system ar e under development.

Details of the control system (e.g., interlocks and permissive signals), nuclear and process instruments , control logic and elements , indication , alarm , and control features will be developed for the Operating License Application. 7-40

.; ... ;. NWMI ...... ..* .... .-.* .. *:. * * !

  • NOlmfWHT MEDtCAl ISOTGm NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-11. Waste Handling System Control and Monitoring Parameters Subsystem name Control parameters (automatic/manual)

Hi g h-do se liquid was t e

  • V a l ve po s ition c oll e ction High-dose liquid waste
  • Valve position solidification Low-do s e liquid was t e c oll ec ti o n Low-dose liquid waste evaporation Low-do s e liquid wa s t e so lidi fi c a tion
  • F l o wr a te (A)
  • Pump ac tuati o n (A/M)
  • Pump motor s p ee d (A)
  • T e mp e ratur e (A)
  • Va l ve ac tu at i o n (A/M)
  • Flowrate (A)
  • Pump actuation (AIM)
  • Pump motor speed (A)
  • Temperature (A)
  • Valve actuation (A/M)
  • F l owra te (A)
  • Pump ac tu a ti o n (A/M)
  • Pump m o t o r s p ee d (A)
  • Te mp e r a tur e (A)
  • Va l ve a c t u a ti o n (A/M) Monitoring parameters
  • D e n s it y
  • Diff e r e nti a l pr ess ur e
  • F lowr a t e
  • Fl owr a te t ot a l ize r
  • Leve l
  • Te mp era tur e
  • P ress ur e
  • R a di at i o n
  • Va l ve po s iti on
  • Density
  • Differential Pressure
  • Flowrate
  • Flowrate totalizer
  • Level
  • Temperature
  • Pre s sure
  • Radiation
  • Valve Position
  • D e n s it y
  • Di ffe r e nti a l pr ess ur e
  • Fl o wr a t e
  • Flowra t e t o t a li zer
  • Leve l
  • Te mp era tur e
  • Pr ess ur e
  • Va l ve p osi t i o n
  • Differential pressure
  • Flowrate
  • L e vel
  • Temperature
  • Pressure
  • Valve position
  • D e n s it y
  • D i ff ere nti al p ress ur e
  • F l owrate
  • F l owra t e tot a l izer
  • Leve l
  • Te mpe ra tur e
  • Pr ess ur e
  • Va l ve p os iti o n Spent resin dewatering
  • Valve actuation (AIM)
  • Valve position S olid wa s t e
  • Ac tuat e gr out mix e r (M)
  • Pr ess ur e e n c ap s ul a tion High-dose waste decay Hi g h-do se was t e h a ndlin g TB D = t o b e d e t e nnin e d. TBD TBD 7-41 TBD TBD Primary control location C ontrol ro o m Low dose solidification room C ontr o l room Control room L ow d ose s olidific a tion room Low dose solidification room L ow do se s olidific a tion room Low dose solidification room Low do se so lidifi cat i on ro o m
.**.*.*. .; ... NWMI ........ *.* ' ! : . NOlllfWHT MEDtCAl ISOTOPf.S NWM l-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-12 provides a preliminary listing of the waste handling system interlocks and permissive signals that have been identified.

These devices will be further developed and detailed information will be provided in the Operating License Application.

Table 7-12. Waste Handling System Interlocks and Permissive Signals Hard-wired or Safety Interlock or permissive input PLC interlock High-dose waste collection tank (WH-TK-100) bjgh-liquid level alarm High-dose waste collection tank (WH-TK-100) low-liquid level alarm High-dose waste collection tank (WH-TK-100) low-pressure alarm High-dose waste concentrator (WH-Z-200) high-liquid level alarm High-dose waste concentrator (WH-Z-200) low-liquid level alarm High-dose waste concentrator (WH-Z-200) demister high-differential pressure alarm High-dose waste concentrator (WH-Z-200) condenser mgh-differential pressure alarm High-dose waste concentrator (WH-Z-200) condenser offgas high-temperature alarm Low-dose waste collection tank (WH-TK-240) mgh-liquid level alarm Low-dose waste collection tank (WH-TK-240) low-liquid level alarm Low-dose waste collection tank (WH-TK-240) low-pressure alarm High-dose waste container offgas filter (WH-F-330) high-pressure differential alarm Condensate collection tank (WH-TK-400) high-liquid level alarm Condensate collection tank (WH-TK-400) low-liquid level alarm Condensate collection tank (WH-TK-400) low-pressure alarm Low-dose waste collection tank (WH-TK-420) high-liquid level alarm Low-dose waste collection tank (WH-TK-420) low-liquid level alarm Low-dose waste collection tank (WH-TK-420) low-pressure alarm Low-dose waste evaporation tank 1 (WH-TK-500) high-liquid level alarm Low-dose waste evaporation tank 1 (WH-TK-500) low-liquid level alarm Low-dose waste evaporation tank 1 (WH-TK-500) low-pressure alarm Low-dose waste evaporation tank 2 (WH-TK-530) high-liquid level alarm Low-dose waste evaporation tank 2 (WH-TK-530) low-liquid level alarm Low-dose waste evaporation tank 2 (WH-TK-530) low-pressure alarm Low-dose waste container offgas filter (WH-F-630) high-pressure differential alarm PLC = programmable logic controller.

TBD 7-42 PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A PLC N I A to be determined.

..... 'NWMI ..**.. ... ... .*.******* NORTIIWlSTMEDtcAllSOTOrfl NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.3.6.4 System Performance Analysis and Conclusion The system performance analysis and conclusion for each process system will be provided in the Operating License Application.

7.3.7 Criticality

Accident Alarm System The RPF will use a CAAS to monitor for a criticality and provide emergency notifications for evacuation. 7.3.7.1 Design Criteria Design criteria for the CAAS I&C systems are described in Section 7.2. 7.3.7.2 Design Basis and Safety Requirements The design basis and safety requirements for the CAAS I&C systems are described in Section 7 .2. 7.3.7.3 System Description The CAAS will be provided as a vendor package with an integrated control system. The CAAS control HMI will be located in the control room and will provide local alarms at the detector locations and at the CAAS HMI. The FPC sy s tem will provide alarm and status monitoring in the control room. The facility-wide notification system configuration will be provided in the Operating License Application.

The surveillance requirements for the CAAS system are described in Chapter 6.0. 7 .3. 7 .4 System Performance Analysis and Conclusion The system performance analysis for each process system will be provided in the Operating License Application. The overall I&C system performance analysis is discussed in Section 7.2. The CAAS will provide for continuous monitoring , indication , and recording of neutron or gamma radiation levels in areas where personnel may be present and wherever an accidental criticality event could result from operational processes. The CAAS will be capable of detecting a criticality accident that produces an absorbed do s e in soft tissue of 20 radiation absorbed dose (rad) of combined neutron or gamma radiation at an unshielded distance of 2 meters (m) from the reacting material within 1 minute (min), except for events occurring in areas not normally accessed by personnel and where shielding provides protection against radiation generated from an accidental criticality. Two detectors will cover each area needing CAAS co v erage. The control unit electronjcs will actuate local and remote alarms. The locations of the detectors will be provided in the Operating License Application.

The CAAS detectors will provide local annunciation and remote annunciation in the control room to alarm when the radiation levels exceed established setpoints.

Alarming CAAS monitors will communicate the location of the criticality accident alarm to the FPC system. Diagrams of the CAAS and associated systems will be provided in the Operating License Application. The uninterruptible power supply (UPS) will provide emergency power to the CAAS during a loss of off-site power. The CAAS will meet the criteria of 10 CFR 20.1501, " General," and use the guidance provided by ANSI/ANS 8.3 , Criticality Ac c ident Alarm Sy s t e m , and Regulatory Guide 3.71 , Nuclear Criticality Saf ety Standards for Fuels and Material Fa c iliti e s. As a safety-related system , the CAAS will be designed to remain operational during design basis accidents , which are described in Chapter 13.0. 7-43

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...... .. *.. .... .. .. .. ' * ! . NOITNWEST MEOM:Al lSOTOP£1 NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS 7.4.1 System Description The ESF s are ac ti ve or pa ss i ve f e atur es d e signed to miti g at e the con s equence s of acc ident s and to keep rad i olo g ic al e x po s ure s to work e r s , the publi c , a nd e n v ironment within ac c eptable va lu es. Chapter 6.0 pro v ide s a de sc ription of the E SF s, in c ludin g th e ac c id e nt s miti g ated and SS Cs u se d to pro v ide the E SF s. The ESF sy stems w ill operat e independ e n t l y from th e F P C sys t e m s as h a rd-w ir e d c ontrol s. Ho w e v er , the E SF s will inte gr ate into the F P C sys t e m s and provide a c ommon point of HMI , monitorin g, and alarmin g at the control room a nd lo ca l HMI work s t a tions. T a ble 7-13 li s t s the ESF s th a t w ill requir e ac tu a tion b y th e I&C sys t e m. Monitorin g sys tems that ar e credited in the s afety anal ys i s are al s o included in the t a ble. Table 7-13. Engineered Safety Feature Actuation or Monitoring Systems (2 pages) l&C SSCs providing Eng i neered safety feature IROFS Accident(s) mitigated engineered safety feature Pr im ary off gas re li ef sys t e m R S-0 9 Di sso l ve r o ff gas failur e durin g Pr e s s ur e r e li ef d evi c e , pr ess ur e di ss oluti o n o p e ration r e li ef t a nk Active radiation monitoring RS-10 Transfer of high-dose process Radiation monitoring and a nd isolation of low-dose liquid outside the hot cell isolation system for low-dose waste transfer shielding boundary liquid transfers Cas k local v e ntil a tion durin g R S-1 3 Tar ge t c l a ddin g l e aka ge durin g Loc a l ca ptur e ve ntil a ti o n clos ur e lid r e mo va l a nd s hipm e nt sys t e m ove r c lo s ur e l i d durin g d oc kin g pr e p arat ion s lid r e m ova l Cask docking port enabler RS-15 Cask not engaged in the cask Sensor system controlling cask docking port prior to opening the docking port door operation docking port door Proc ess vesse l e m e r ge nc y F S-0 3 H y dro gen d e fl agra tion o r B ackup b o ttl ed n itr o ge n gas p u rge sys t e m deton a ti o n s uppl y Active discharge monitoring CS-14 Accidental criticality To be provided in the Operating and i s olation License Application Ind e p e nd e nt ac ti ve di sc h a r ge C S-15 A cc ident a l c rit i c a lit y To b e pro v id e d in th e Op e ratin g m o nitoring a nd i s ol a tion Li ce n se A pplic a ti o n E vaporator or concentrator CS-20 Prevent nuclear criticality from Conductivity analyzer and condensate mon i toring high-volume transfer to non-control valve geometrically fa v orable vessels in solutions with normally low fissile component concentrations 7-44 NWM I ..... .*.******* ' *

  • NOllTHWEST MEDtCAl ISOTO'U NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Table 7-13. Engineered Safety Feature Actuation or Monitoring Systems (2 pages) l&C SSCs providing Engineered safety feature IROFS Accident(s) mitigated engineered safety feature C lo sed h eating or cooling loop with monitoring and alarm Dissolver offgas vacuum receiver or vacuum pump CS-27 TBD l&C IROFS in s trum e ntation a nd c ontrol. it e m s relied on for sa fety. 7.4.2 Annunciation and Display Accidental critica lit y Potential limiting control for operations; motive force for dissolver offgas C lo s ed-loop , hi gh-vo lum e h eat transfer fluid systems to prevent nuclear cr iti cality or transfer of high-dose material across shielding boundary in the event of a leak into the heat transfer fluid with normally low fissile component concentrations Dissol ver offgas vacuum receiver tanks, dissolver offgas vacuum pumps SS C TBD structure s , syst e m s, and component s. to b e det e rmin e d. The actuation of an ESF will be displayed on the FPC system HMI and locally at the affected system with an audible alarm. The a l arm annunciator display panel and the alarm or event display will show the triggering event. Once actuated, the ESFs will require manual input from the operator to reset the ESF. C learing the triggering event wi ll be required.

7.4.3 System

Performance Analysis Section 7 .2.4 provides additional details on the analysis of system performance. Potential variables, condit ion s , or other items that will b e probable subjects of technical specifications associated with the FPC system are provided in Chapter 14.0. 7-45

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7.5.1 Design

Criteria Design criteria for the control room I&C systems are described in Section 7.2. 7.5.2 Design Basis and Safety Requirements The design basis and safety requirements for the control room I&C systems are described in Section 7.2. 7.5.3 System Description The control room will provide the majority of interfaces for the facility and process control systems, with overall process controls , monitoring , alarms , and acknowledgement.

The control room will consist of a properly sized and shaped control console with two or three operator interface stations or HMis (one being a dedicated engineering interface), a master PLC or distributed controller , and all related and necessary cabinetry and subcomponents (e.g., input/output boards , gateways , Ethernet switches , power supplies , and UPS). This control system will be supported by a data highway of sensing instrument signals in the facility process areas that will be gathered onto the highway throughout the facility by an Ethernet communication-based interface backbone and brought into the control room and onto the console displays.

Dedicated controllers and human-machine monitoring interfaces or stations for other equipment systems will also be in the control room. This equipment includes the facility crane , closed-circuit television system, CAAS , and radiation monitoring system. A control panel for all facility on-site and off-site (if required) communications (e.g., telephone , intercom) will likely also be located there. The control room door into the facility will be equipped with controlled access. The BMS will be primarily controlled and monitored from the control room. Utility systems with vendor packages and integrated controls will provide surveillance monitoring to the control room. The FPC system will operate with a synchronized hot s tandby redundant system structure for all hot cell processes.

Each hot cell process will be an independent subsy s tem having a local HMI with monitoring and control functions from the control room. Workstations for each system within the control room will be hot standby redundant.

The redundant stations will run software on identical PLC systems. The PLC systems will monitor each other. On loss of synchronizing signal from one system, the other system will continue with control and monitoring. Process systems that will be primarily controlled in the control room include uranium recovery and recycle , target dissolution , and liquid waste handling. The target receipt system will be controlled with local HMis in the irradiated target basket receipt bay or target cask preparation airlock. Mo production process hot cell systems , including target disassembly and Mo recovery and purification, will be controlled with local HMis in the hot cell operating gallery. The hot cell processes will have monitoring and redundant control functions from the control room. The FPC subsystem for target fabrication processes will be controlled with local HMis in the target fabrication area , with surveillance monitoring in the control room. Local HMis will be provided in Room W107 , which houses equipment for low-dose waste solidification. Low-dose liquid waste will be piped in from the holding tanks in the utility area above Room W107 , and drums of solidified waste will be transported out by pallet jack. This local HMI will be the primary control location for the high-dose liquid waste solidification , high-dose waste decay, spent resin dewatering , and solid waste handling hot cell operations. 7.5.4 System Performance Analysis and Conclusion The system performance analysis for each process system will be provided in the Operating License Application. The overall I&C system performance analysis and conclusions are provided in Section 7.2. 7-46

..... ; ... NWMI ********** *.-.* .. *.*:. NORTifWESTMEDtcAllSOTOP£S NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems 7.6 RADIATION MONITORING SYSTEMS The radiation monitoring systems will include CAMs , continuous monitoring at the exhaust stacks, process control instruments , and personnel monitoring and dosimetry. Process control instruments used to analyze for uranium concentrations are described in each respective process system in Section 7.3. The objective of the radiation monitoring system is to provide the RPF control room personnel with a continuous record and indication of radiation levels at selected locations where radioactive materials may be present , stored, handled , or inadvertently introduced.

The system is also designed to ensure that there is accurate and reliable information concerning radiation safety as related to personnel safety. The design considerations for the radiation monitoring system include the following: * *

  • Provision of information to RPF operators so that in the event of an accident resulting in a release of radioactive material , decisions on deployment of personnel can be properly made. Indication and recording in the control room of the gamma and airborne radiation levels in selected areas as a function of time , and , if necessary, alarming to indicate any abnormal radiation condition.

These indicators aid in maintaining plant contamination levels as low as reasonably achievable (ALARA) and in minimizing personnel exposure to radiation.

Provision oflocal alarms and/or indicators positioned at key points throughout the RPF where a substantial increase in radiation levels might be of immediate importance to personnel frequenting or working in the area. Radiation Monitoring Locations RAMs will be located in areas where personnel may be present and where radiation levels could become significant based on the following considerations

* *
  • Occupancy status of the area , including time requirements of personnel in the area , the proximity to primary and secondary radioactive sources , and shielding Potential for increase in the background radioactivity level Desirability of surveillance of infrequently v i s ited areas CAMs will be located in work areas where there is a potential for airborne radioactivity. The CAMs will have the capability to detect derived air concentrations within a specified time. 7.6.1 Design Criteria Design criteria for the radiation monitoring I&C sy s tems are described in Section 7.2. 7.6.2 Design Basis and Safety Requirements The design basis and safety requirements for the radiation monitoring I&C systems are described in Section 7.2. The ESFs for this system are listed in Chapter 6.0. 7.6.3 System Description The radiation safety monitoring system will include CAMs , continuous monitors at the exhaust stacks , and personnel monitoring and dosimetry.

7-47

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  • NO<<rHW£STMEOfCALISOTOPES NWMl-2013-021 , Rev. 1 Chapter 7.0 -Instrumentation and Control Systems Three basic types of personnel monitoring equipment will be used at the facility:

count rate meters (friskers), hand/foot monitors , and portal monitors. All personnel whose duties require entry to restricted areas will wear individual external dosimetry devices (e.g., passive dosimeters such as thermoluminescent dosimeters that are sensitive to beta , gamma, and neutron radiation) from a National Voluntary Laboratory Accreditation (NAVLAP)-certified vendor. Personnel monitoring and dosimetry is described in Chapter 11.0, "Radiation Program and Waste Management." 7.6.3.1 Air Monitoring Continuous air monitors -CAM units will consist of a particulate measuring channel with a filter to capture particulate. Air will be drawn through the system by a pump assembly.

The sample will be withdrawn from inside the appropriate area , room , or cell through an isokinetic nozzle with the sampling volume flow at a known fixed rate , so that the accumulation of radioactive particles can be interpreted as a quantitative sample. After passing through the nozzle , the sample will be drawn through tubing and through a fixed or moving filter tape before being discharged to the atmosphere.

The samplers also have a purging system for flushing the volume cell surrounding the gas sample chamber with clean air for purposes of calibration and the removal of crust activity. Replaceable liners will be changed out periodically when contamination becomes excessive. Flow regulating will ensure that flow through the filters remains constant.

Each instrument channel will include a detector , preamplifier , count rate meter , and power supply. The detector may be a scintillation counter or similar device having a gamma sensitive crystal , and a photo multiplier whose output pulses are counted by the rate meter. Each readout module will be equipped with a light that illuminates when the radiation level exceeds preset limits. The setpoint will be adjustable over the entire detection range. Pressing a button will cause the meter to indicate the alarm setpoint.

Visible alarms will be accompanied by a simultaneous local audible a larm with an alarm light in the control room. A normally energized light will deenergize when there is a detector signal failure , circuit failure , power failure , or failure due to a disconnected cable. Power for the monitor s that initiates a safety signal will be provided from the UPS. Loss of power and signal failure will be monitored for each detector.

CAMs will be provided with a check source. This check source will simulate a radiation field and will be used as a convenient operational and gross calibration check of the detectors and readout equipment.

CAM calibration will include , where practical , exposures to the s pecific isotopes that the particular system monitors in the field. Instrument calibrations will be performed at prescribed frequencies. An electronic test signal and/or radioactive check source drift indication may also require CAM recalibration. Radiation area monitors -The RAM detector unit will be housed in an environmentally suitable container that is mounted in a duct , on a wall , or other suitab l e surface. The sensitivity of each detector will be sufficient to have the alarm setpoint an order of magnitude higher than the detection threshold. The detectors are designed to be operational over a wide range of temperatures. The design of the detectors will meet expected normal and abnormal environmental design conditions , as appropriate. Saturation will not be expected to adversely affect operation of the detector within its calibrated range. Sensors will be mounted as close as practical to the most probable radiation sources with no objects , persons, pillars , and piping serving as shielding. The sensors will also be mounted so as to minimize inaccuracies due to any directionality of the detector.

A udible and vis ual alarm devices -When the radiation exceeds predetermined levels , alarms will actuate in the control room and at selected detector locations. 7-48

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  • NORTHWEST MEDICAL ISOTOPES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems The alarms will consist of the following ca pabiliti es: * * * "A lert li g ht" will illuminate when the radiation level exceeds preset limit s with an adjustable se tpoint " High alarm red light" will illuminate when radiation level s exceed a predetermined alarm setpoint "Fai lure alarm" will sou nd when either the power or the channel's electronics fail The visual alarms will be accompanied by a simultaneous audible alarm annunciator at the selected detector locations and in the control room. The annunciator windows for the monitors will be located in the control room. The alarm can be manually re set when the alarm conditions are corrected.

The local a larm horn s and warning lights will remain on until the radiation le ve l is belo w the pre se nt le ve l. Additional CAM requirements and locations are described in Chapter 11.0. 7.6.3.2 Stack Release Monitoring The exhaust stacks will be provided with continuous monitors for noble gases , particulate , and iodine. The stack monitoring system design ba sis is to continuously monitor the radioactive stack relea ses. Additional information will be provided in the Operating License Application. Airborne exposure pathway monitoring is described in Chapter 11.0. 7.6.4 System Performance Analysis and Conclusions The system performance analysis and conclusions for eac h process system will be provided in the Operating License Application.

The overall I&C syste m performance analysis i s provided in Section 7.2. 7-49

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7.7 REFERENCES

NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems I 0 CFR 20.1501, "Ge neral ," Code of Federal R egu lation s, Office of the Federal Register, as amended. I 0 CFR 70 , " Domestic Licensing of Special Nuclear Material ," Code of Federal R egu lations , Office of the Federal Register , as amended. 10 CFR 70.61, "Performance Requirements

," Code of Federal R egu lations , Office of the Federal Register , as amended. I 0 CFR 70.64, " Requirements for New Facilities or New Processes at Existing Facilities," Code of Federal R egu lation s, Office of the Federal Register , as amended. 10 CFR 73.1, " Purpose and Scope," Code of Federal R egu lations , Office of the Federal Register , as amended. I 0 CFR 73.54, "Protection of Digital Computer and Communication Systems and Networks," Code of Federal R egu lation s, Office of the Federal Regi s ter , as amended. ANS 10.4-2008, Verification and Validation of Non-Safety-Related Scientific and Engineering Computer Programs for the Nuclear Indust ry, American National Standards Institute , New York, New York, 2008. ANSI/ ANS 8.3, Criticality Accide nt Alarm System, American National Standards Institute/ American Nuclear Society, La Grange Park , Illinois , 1997 , R2003, R2012. ANSI/ISA 67.04.01-2006, Setpointsfor Nuclea r Safety-Related In strume ntation , American National Standards Institute/International Society of Automation, Research Triangle Park , North Carolina, 2006 (R201 l). ANSI/ISA 84.00.01-2004 Part 1 , Functional Safety: Safety Instrum e nt ed Systems for the Pro cess Indu s try Sector -Part 1: Framework , D efinitio ns , System , Hardwar e and Software R equ ir eme nt s, American National Standards Institute/International Society of Automation, Research Triangle Park , North Carolina, September 2004. ANSI/ISA 84.00.01-2004 Part 2, Functional Safety: Safety Instrument ed Systems for the Process Indu s t ry Sector -Part 2: Guidelines for the Ap plication of ANSI I ISA-84.00.0 1-2004 Part 1 (!EC 61511-1 Mod) -Informativ e, American National Standards Institute/International Society of Automation, Research Triangle Park , North Carolina, September 2004. ANSI/ISA 84.00.01-2004 Part 3, Functional Safety: Safety In st rument ed Systems for the Pro cess Indu s try Sector -Part 3: Guidance for the D ete rmination of the R equ ir e d Safety Int egrity L eve l s -Informative , American National Standards Institute/International Society of Automation, Research Triangle Park , North Carolina, September 2004. EPRI TR-106439 , Guideline on Evaluation and Acceptance of Commercial Grade Digital Equipment for Nuclear Safety Applications, Electric Power Research Institute , Palo Alto, California, November 1996. IEC 61508, Functional Safety of Electrical

/Electronic/Programmable Electronic Safety-Related Systems, Parts 1 -7 , International Electrotechnical Com.rrllssion, Geneva , Switzerland, as amended. IEEE 7-4.3.2-2010 , IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Pow e r Generating Stations, Institute of Electrical and Electronics Engineers, Pi sca taway , New Jersey , 2010. IEEE 323-2003, IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, Piscataway , New Jersey , 2003. 7-50

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  • NOlllfWESTMEDtcAllSOTOitES NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems IEEE 338-2012 , IEEE Standard for Crit e ria for th e P e riodic Surv e illan ce T e sting of Nuclear Power G e nerating Station Saf e ty S y st e m s, Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2012. IEEE 344-2004 , IEEE R ec ommend e d Practice for S e i s mi c Qualifi c ation of Cla ss JE Equipm e nt for Nuclear Pow e r G e nerating Stations , Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2004. IEEE 379-2014 , IEEE Standard Appli c ation of th e Singl e-Failur e Crit e rion to Nuclear Power G e nerating Station Safety S ys t e ms , Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2014. IEEE 384-2008 , IEEE Standard Crit e ria for Ind e p e nd e nc e of Cla ss JE Equipm e nt and Cir c uit s, Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2008. IEEE 497-2010 , IEEE Standard Crit e ria for A cc id e nt Monitorin g Instrum e ntation for Nuclear Power G e n e ratin g Stations , Institute of Electrical and Electronics Engineers , Piscataway, New Jersey , 2010. IEEE 577-2012 , IEEE Standard R e quir e m e nts for R e liability Anal y sis in the D es ign and Op e ration of Saf e ty S y st e m s for Nucl e ar Fa c ilities , Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2012. IEEE 603-2009, IEEE Standard Crit e ria for Saf e ty S ys t e ms for N uclear Pow e r G e nerating Stations , Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2009. IEEE 828-2012 , IEEE Standard for Configuration Manag e m e nt in S y st e ms and Softwar e Engin ee ring , Institute of Electrical and Electronics Engineers , Piscataway, New Jersey , 2012. IEEE 829-2008 , IEEE Standard for Software and S y st e m T e st Do c um e ntation , Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2008. IEEE 1012-2012 , IEEE Standard for S ys t e m and Softwar e Verifi c ation and Validation , Institute of Electrical and Electronics Engineers, Piscataway, New Jersey , 2012. IEEE 1028-2008 , IEEE Standard for Sof tw ar e R ev i ews and Audit s, Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2008. IEEE STD 12207 , ISO I IE C I IEEE Standard for S ys t e m s and Sof tw are Engine e rin g-Softwar e Lif e C y cl e Pro cesses, Institute of Electrical and Electronics Engineers , Piscataway , New Jersey , 2008. IEEE STD 15939 , IEEE Standard Adoption of ISO l lEC 15939: 200 7 S ys t e m s and Softwar e Engin e ering M e a s ur e m e nt Pro ce ss , Institute of Electrical and Electronics Engineers , Piscataway , New Jer s e y, 2008. ISA-RP-67.04.02 , Methodologi es for th e D e t e rmination of S e tpoint s for Nucl e ar Saf ety-R e lat e d In s trumentation, Instrument Society of America , Research Triangle Park , North Carolina, 2010. ISO/IEC/IEEE 15288 , S ys t e ms and Software Engin ee rin g-S ys t e m Lif e Cycl e Pro cesses, International Organization for Standardization , Geneva , Switzerland, 2015. ISO/IEC/IEEE STD 24765 , S y st e m s and Softwar e Engin ee ring-Vo c abulary , International Organization for Standardization , Geneva , Switzerland, 2010. NUREG-0700 , Human-S ys tem Interfac e D e sign R e vi e w Guid e lin es, Rev. 2 , U.S. Nuclear Regulatory Commission , Office of Nuclear Reactor Regulation , Washington , D.C., May 2002. 7-51

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  • NORTHWUTMEDtw.ISOTOf'CS NWMl-2013-021, Rev. 1 Chapter 7.0 -Instrumentation and Control Systems NUREG-0711, Human Factors Engineering Program Review Model, Rev. 3, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington , D.C., November 2012. NUREG-0800, Standard R eview Plan for the R ev i ew of Safety Analysis R eports for Nuclear Pow er Plants, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, Washington, D.C., as amended. NUREG-153 7 , Guidelines for Pr eparing and R eviewi ng Applications for the Li censing of Non-Power R eactors -Format and Content, Part 1 , U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation , Washington , D.C., February 1996. NUREG/CR-6090, The Programmable Logic Controller and It s Applicatio n in Nuclear R eactor Systems , U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation , Washington , D.C., September 1993. NUREG/CR-6463, Revi ew Guidelines on Software Languages for Use in N ucl ear Power Plant Safety Systems, U.S. Nuclear Regulatory Commission, Washington , D.C., June 1996. NWMI-20 l 5-SAFETY-002, Radioisotop e Production Facility Int egrated Safety Ana l ys i s Summary, Rev. 0, Northwest Medical Isotopes , Corvallis, Oregon , 2015. Regulatory Guide 1.53 , Application of the Single-Failure Criterion to Safety Systems, Rev. 2, U.S. Nuclear Regulatory Commission, Washington , D.C., June 2003. Regulatory Guide 1.97 , Criteria/or Accident Monitoring In strumentatio n for N ucl ear P ower Plants , Rev. 4 , U.S. Nuclear Regulatory Commission, Washington, D.C., 2006. Regulatory Guide 1.152 , Criteria for Use of Computers in Safety Systems of Nuclear Power Plan ts, Rev. 3, U.S. Nuclear Regulatory Commission, Washington , D.C., June 2011. Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Material Facilities, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research , Washington , D.C., 2010. Regulatory Guide 5. 71, Cyber Security Program s for Nuclear Facilities, U.S. Nuclear Regulatory Commission, Washington , D.C., 2010. 7-52
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