ML17276A954
| ML17276A954 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/31/1981 |
| From: | Sharp L WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17276A953 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM WNP-2-01, WNP-2-01-R02, WNP-2-1, WNP-2-1-R2, NUDOCS 8201260009 | |
| Download: ML17276A954 (399) | |
Text
{{#Wiki_filter:Rev. 2 Report No. NNP 52-01 NASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 INTERIM SHIELDING EVALUATION RADIATION REPORT REVISION 2 Report Date: December 1981 ~'iocaYOOO9 05000397 BVOiiS
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Rev. 2 REPORT APPROVAL COVER SHEET Project: Washington Public Power Suppl System WNP-2 Title: Shieldin Evaluation Reports WNP02-01 Revisions 2 Prepared By: Reviewed By: Approval By:
Rev. 1 Report No. NNP N2-Ol WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 INTERIM SHIELDING EVALUATION RADIATION REPORT REVISION 1 Report Date: December 1981
Rev. 1 REPORT APPROVAL COVER SHEET Project: Washington Public Power Suppl System WNP-2 Title: Shielding Evaluation Reports WNPN2-01 Revis ion~ 1 Prepared By: Rev iewed By: Approval By:
Report No. WNP g2-01 WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECT NO. 2 INTERIM SHIELDING EVALUATION RADIATION REPORT REVISION 0 Report Date: November 1981
REPORT APPROVAL COVER SHEET Project: Washington Public Power Suppl S stem WNP-2 Ti tie: Shielding Evaluation Reports WNPN2-Ol Rev is iong 0 Prepared By: Reviewed By: Approval By: ii/~i
CONTENTS OF THIS REPORT Burns and Roe Incorporated (BRI) performed the analysis of radiation levels occurring inside Primary Containment'; assembled, edited, reviewed, and approved this technical report for the Washington Public Power Supply System. EDS Nuclear. Incorporated performed the analysis of radiation levels occurring in the Reactor Building Secondary Containment under subcontract to BRI. The Washington Public Power Supply System performed the analysis of radiation 1'evels occurring in areas outside the Reactor Building Secondary Containment.
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tv'ASHIibGTON PUBLIC POWER PLANT SHIELDIiVG NASHIilGTO'8 NUCLEAR SUPPLY SYSTEM AilALYSIS PROJECT 82 Page i TABLE OF CONTENTS REPORT APPROVAf COVER SHEET ABSTRACT
SUMMARY
TABLE OF CONTENTS INTRODUCTION 2 ' REQlJIREMENTS
- 2. 1 SHI Ef.DING EVAf UATION REGULATORY REQUIREt1FNTS 2,1.1 ACCXDFNT ANALYSIS REQUIREMENTS 2.1.2 SOURCE TERM ASSUMPTIONS 2.1,3 VITAL AREA ACCESS REQUIREMENTS 2.1.4 SYSTl",MS CONTAINING THE SOURCES 2 '..5 SAFETY-RELATED EQUIPMENT 2 ~ 2 'SHIELDIHG EVAfUATION TASK DESCRIPTION 2 ' SHIEl'ING EVALUATION ITEM DFLETFD FROM SHIELDIHG ANALYSIS CONSIDERATION 2.4 ADDITIONAI REVIEW XTEMS REQ(JIRED FOR FINAL Pf ANT SHXEf DING REPORT 3 ~ 0 ANALYTICAL MFTFtODOLOGY F 1 ACCIDENT SCENARIO
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR
'SUPPLY SYSTEM ANALYSIS ii PROJECT //2 Page TABLE OF CONTENTS (continued) 3.2 CONTAMINATED SYSTEMS 3 ~ 2 ~ 1 SYSTEMS INCLUDED FOR PRIMARY CONTAINMENT ANALYSIS 3.2.2 SYSTEMS INCLUDED FOR SECONDARY CONTAINMENT ANALYSIS 3.2.3 SYSTEMS EXCLUDED 3.3 SOURCE TERM ASSUMPTIONS 3.4 TIME PERIOD CONSIDERED FOR STUDY 4 ~ 0 ACCESS AND OCCUPANCY OF VITAL AREAS 4.1 DOSE RATES OUTSIDE THE REACTOR BUILDING 4.2 VITAL AREAS It'll AND ACCESS ROUTES OUTSIDE THE REACTOR BUILDING 5 ~ 0 METHODS 5.1 THE USE OF COMPUTER CODES'OURCE
- 5. 2 TERM DEVELOPMENT FOR PRIMARY CONTAINMENT 5.3 SOURCE TERM DEVE LOPM ENT FOR SECONDARY CONTA Et JT
- 5. '3. l SOURCE TERM DEVELOPMENT FOR 1E/1M EQUIPMENT OUTSIDE THE REACTOR BUILDING 5.4 PARAMETRIC STUDIES FOR DIRECT PIPING DOSE IN SECONDARY CONTAINMENT
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR ANALYSIS PROJECT 82 SUPPLY SYSTEM P~ge iii TABLE OF"CONTENTS (continued) 5 ' DOSE RATE AND CUMULATIVE DOSE CALCULATION PROCEDURE 5.5.1 CALCULATION OF AIRBORNE GAMMA DOSES INSIDE SECONDARY CONTAINMENT 5 ~ 5 ~ 2 METHODOLOGY OF BETA DOSE ANALYSIS 5.5.3 PROCEDURE FOR THE CALCULATION OF RADIATION ZONE DOSE IN SECONDARY CONTAXNMENT 5 ~ 5 ~ 4 CALCULATION OF RADIATION DOSES DUE TO SPECIAL SYSTEMS AND COMPONENTS INSIDE'ECONDARY CONTAINMENT 5.5 4.1
~ SOURCE TERM ASSUMPTXONS II'l SECONDARY CONTAINiMENT 5.F 4.2 SECONDARY CONTAINMENT ANALYSIS METHOD 5.5.4.3 CALCULATION OF RADIATION DOSES INSIDE SECONDARY CONTAINMENT OF GENERIC MECHANICAL EQUIPMENT 6.0 RESULTS
, 6 ~ 1 RIMARY CONTAXN IENT RADIATXON RESUL TS 6 ~ 2 SECONDARY CONTAINMENT RADIATION RESUI'S RADIATION RESULTS OUTSXDE THE REACTOR 13UILDING 7 .0 REFERENCES
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT /I2 page iv LIST OF TABLES Section 3.0 Anal tical Methodolo Table 3.1 Distribution of Fission products in the >lorst post-LOCA Situation For Areas Inside Containment. (Depressurized Reactor Coolant System) I Table 3.2 Distribution of Fission products in the Worst Post-LOCA Situation for Areas Inside Containment. (Pressurized Reactor Coolant System) Table 3.3 Distribution of Fission Products in the Worst post-LOCA Situation for Areas Outside Containment. - Table 3.4 System Operations and Source Term Assumptions Section 5.0 Methods Table 5.1 Tyoes of Generic Mechanical Equipment II Section 6.0 Results Table 6.1 t&lp 2 IF/IM Primary Containment Equipment List of Total Integrated Dose (40 YR Plus LOCA). Table 6. 2 Individual Zone Sketches of Safety-Related Equipment Location Table 6.3 NNP-2 1E/1M Equipment Vital Area List Outside the Reactor Building for Six Month Total Integrated Dose (LOCA) Table 6.4 WNP-2 Vital Areas and Access Route List of Radiation Exposure to personnel During the Required post-LOCA Operations
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM AVALYSIS PROJECT /2 Page v I LIST OF TABLES (continued) Appendix A Unisolated Leak Path Report Table A-l System Flow Diagrams Employed to Perform the Review Appendix B Source Term Development and Parametric Studies Table B-l Gamma Energy Concentration in Liquid-Containing Systems Table 8-2 Total Gamma Activity of the Airborne Fission Products Table B-3 Comparison of Direct Dose Rate Results Appendix C Procedure for the Calculation of Radiation Zone Doses Table C-l. Diameter Correction Factor (FD) for Targets in Contact with the Source Piping Appendix D Calculation of the Radiation Doses Due to Standb Gas Treatment System Table D-1 Total Gamma Activity of the Released Airborne Halogens Table D-2 Direct Gamma Dose Rate and Integrated Dose Results for Targets in the SGTS Room Appendix F Primar Containment Anal ses Table F-l Time Mesh Spacing Used in Source Calculation Table F-2 ta Average Decay Rate ( Me V/sec), 0-6 Months After C'e LOCA Table F-3 Approximate Dose Rate Reduction Factor vs. Distance from Core Mid-Plane
Kev. I ,NASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS //2 'ROJEC Page vi LIST OF TABLES (continued) TABLES F-4 through F-14 to be issued at a latex date
'WASHINGTON PUBLIC POWER PLANT SHIELDING MASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT f/2 Page v3.3. LIST OF FIGURES Section 5. 0 Methods k Figure 5.1 Six Month Airborne Integrated Dose Figure 5.2 Dose Model Liquid Source Figure 5.3 Six Month Integrated Fluid Dose for Pipes Containing Liquid Source Term (RCIC Liquid System, RHR System, etc) Figure 5.4 Six Month Integrated Fluid Contact Dose for pipes Containing CAC Gaseous Source Term Figure 5.5 Six Month Integrated Fluid Contact Dose for pipes Containing Steam Source Term Diluted LJithin the RCS Steam Space (HS System, RCIC Steam System, and MSlVLC System Upstream of the Header) Section 6.0 Results Figure 6.1 Radiation Zone Map Reactor Building Fl. 422' Figure 6.2 Radiation 'lone Hap Reactor Building Fl. Zone Map Reactor Building El.
'441'adiation Figure 6.3 Figure 6.4 Zone Hap Reactor Building El.
471'adiation Figure 6.5 Zone Hap Reactor Building El. 501'adiation Figure 6.6 Zone Hap Reactor Building El. 522'adiation Figure 6.7 Zone Hap Reactor Building Fl. 572' 548'adiation
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR ANALYSIS PROZECT //2
.SUPPLY SYSTD1 Page viii LIST OF FIGURES (continued)
Figure 6.8 Radiation Zone tlap Reactor Building El. Figure 6.9 Integrated Dose Turbine Generator Bldg. 606'orty-Year (El 441' Figure 6.10 Forty-Year Integrated. Dose Turbine Generator Bldg. (El. 441') Figure 6.11 Forty-Year Integrated Dose Turbine Generator Bldg. ( El. 471' Figure 6.12 Forty-Year Integrated Dose Turbine Generator Bldg. ( El. 471' Figure 6.13 Forty-Year Integrated Dose Turbine Generator Bldg. ( El. 50ls)
'U Figure 6.14 Forty-Year Integrated Dose Turbine Generator Bldg. ( El.
501') Figure 6.15 Forty-Year Integrated Dose Radwaste Bldg. (El. 437') Figure 6.16 Forty-Year Integrated Dose Radwaste Bldg. (El. 467') Figure 6.17 Forty-Year Integrated Dose Radwaste.Bldg. (El. 484') Figure 6.18 Forty-Year Integrated Dose Radwaste Bldg. (Kl. 501') Figure 6.19 Vital Areas and Access Routes for Radwaste Bldg. (El. 437') Figure 6.20 Vital Areas and Access Routes for Radwaste Bldg. (El. 467)) Figure 6. 21 Vital Areas and Access Routes for Radwaste Bldg. (El. 484')
( WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT >'!2 Page ix LIST QF FIGURES (continued) Figure 6.22 Vital Areas and Access Routes for Radvraste Bldg. (El. 501') Figure 6.23 Vital Areas and Access Routes for Diesel Generator Bldg. (El. 441') Figure 6.24 Vital Areas and Access Routes for WNP-2 Site
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT 82 Page > LIST OF FIGURES (continued) Figure R-l Model of the Primary and Secondary Containment Figure B-2 'Zime-Dependen t Gamma Dose Rate for a Semi-In f in te i Cloud of Fission Products at Secondary Containment Concentration. Figure 8-3 Illustration of Parameters Used in the Shielding Equation Figure B-4 Standard Gamma Dose Rate Curve for Liquid Containing Systems F.igure B-5 Standard Integrated Gamma Dose Curve for Pipes in Liquid Containing Systems Figure 8-6 Standard Gamma Dose Rate'urve for Pipes in the RCIC Steam System and the NSIVLC Steam Syste,n Before the Header Figure B-7 Standard Integrated Gamma Dose Curve for Pipes in the RCIC Steam System and the HSIVLC Steam System Before the Header Figure R-8 Standard Gamma Dose Rate Curve for Pipes in the HSIVLC Steam System After the Header Figure B-9 Standard Integrated Gamma Dose Curve for Pipes in the HSIVLC Steam System After the Header Figure B-10 Standard Gamma Dose Rate Curve for CAC System Gas Lines Figure B-ll Standard Integrated Gamma Dose Curve for CAC System Gas Lines Figure 8-12 Radial Distance Correction Factor for Liquid Sources )
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WASHINGTON PUBLIC POWER PLANT WASHINGTON NUCLEAR SHIELDING'NALYSIS SUPPLY SYSTEt1 PROJECT r/2 Page xii LIST OF FIGURES (continued) Figure C-8 Gamma Dose Rate at a Target 8 Feet Away From Standard P ipe Figure C-9 Gamma Integrated Dose at a Target 8 Feet Away From Standard Pipe Figure C-10 Pipe Diameter Correction Factor Figure C-ll Radial Distance Correction Factor Figure C-12 Pipe Length Correction Factor Figure C-13 Dose Rate versus Concrete. Shield Thickness for Standard P ipe Figure C-14 Pipe Diameter Correction Factor for Targets Located Axially in Line With Source Piping Figure C-15 Distance Correction Factor for Targets Located Axially in Line With Source Piping Appendix D Calculation of the Radiation Doses due to Standb Gas Treatment S stem Figure D-1 Standby Gas Treatment Filter Figure D-2 Geometry of Prefilters and HEPA Filters Figure D-3 Geometry of Charcoal Filters Appendix E Beta Dose Point Derivation Figure E-l Airborne Beta Dose Rate at T=0.1 Hr. Figure E-2 Airborne Beta Dose Rate at T=1.0 Hr.
0 !v'ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEt1 ANALYSIS PROJECT i'!2 Page xiii LIST'OF FIGURES (continued) Figure E-3 Airborne Beta Dose Rate at T=9.0 Hr. Figure E-4 Airborne Beta Dose Rate at T=72 Hr. Figure E-5 Airborne Beta Dose Rate at T=720 Hr. Figure E-6 Airborne Beta Dose Rate at T=2880 Hr. Figure E-7 Integrated Air Beta Doses for the Reactor Building F igure E-8 Integrated Air Beta Doses Inside Containment F igure E-9 Equipment Beta Dose Versus Volume Appendix F 1 Figure F-1 Node .Point 6 Line Identification RNCU RRC Systems Figure F-2 Dose Rate at Pipe b1id-Plane vs, Distance from Pipe Surtace 2 In. RRC(51)-4 Node 9 ~ 8 4.56 x 106 HeV/cc-sec Figure F-3 Dose Rate at Pipe ibid-Plane vs. Distance from Pipe Surface 3 In. RRC(51)-4 Node 8 ~ 7 3.51 v 106 rleV/cc-sec Figure F-4 Dose Rate at Pipe Hid-Plane vs. Oistance from Pipe Surface 4 In. RRC(4)-4S Node 18 ~?B 3.64 x 106 l1eV/cc-sec Figure F-5 Dose Rate at Pipe Hid-Plane vs. Distance from Pipe, Surface 4 In. RRC(4)-4S Node 1A ~2A 3.50 x 106 HeV/cc-sec
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEt1 ANALYSIS PROJECT //2 Page Xiv LIST OF FIGURES (continued) Figure F-6 Dose Rate at Pipe h1id-Plane vs. Distance from Pipe Surface 4 In. RRC(51)-4S and 4 In. RWCU(4)-4 Node 7~6 8.09 x 105 MeV/cc-sec Figure F-7 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 6 In. RWCU(3)-4 Node 28 ~ 3 2.42 x 106 MeV/cc-sec Figure F-8 Dose Rate at pipe Hid-Plane vs. Distance from pipe Surface 6 In. RWCU{3)-4 Node 2A ~3 2.22 x 106 tleV/cc-sec Figure F-9 Dose Rate at pipe Mid-plane vs. Distance from Pipe Oi Surface 6 In. RWCU(3)-4 Node 3 ~4 2.12 x 106 MeV/cc-sec Figure F-10 Dose Rate't Pipe Mid-Plane vs. Distance from Pipe Surface 6 In. RWCU(3)-4 Node 6 ~ 4 4.31 x 104 MeV/cc-sec Figure F-ll Dose Rate at Pipe Hid-plane's. Distance from Pipe Surface 6.Xn. RWCU(3)-4 Node 4 ~5 8.22 x 105 meV/cc-sec Figure F-12 Dose Rate at pipe Mid-Plane vs. Distance from Pipe Surface 12 In. RRC(l)-4S 5.03 x 106 t4eV/cc-sec Figure F-13 Dose Rate at pipe Mid-Plane vs. Distance from Pipe Surface <16 In. RRC( l)-4S 5.03 x 106 NeV/cc-sec Figure F-l4 Dose 'Rate at Pipe Mid-plane vs. Distance from pipe Surface 24 In. RRC{1)-4S and 24 In. RRC(2)-4S 5..03 x 106 t<eV/cc-sec
Rev. 2 PLANT SHIELDING WASHINGTON NUCLEAR ! WASHINGTON PUBLIC POWER ANALYSIS PROJECT 82 SUPPLY SYSTEM Page xv LIST OF FIGURES (continued) Figure F-15 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 12 In. RHR(1)-4 or 12 In. RHR(l)-4S Figure F-16 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 14 In. RHR(l)-4 or 14 In. RHR(1)-4S Figure F-17 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 20 In. RHR(2)-4 or 20 In. RHR(2)-4S Figure F-18 Dose Rate vs. Distance from Surface of 26 In'. Main Steam Pipe Figure F-19 Containment Cross Section Figure F-20 To be issued at a later date Figure F-21 To be issued at a later date Figure F-22 Wetwell Zone Mode'1 Figure F-23 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 2 In. Pipes Sched 160 RWCU, RRC 6 RHR Systems Figure F-24 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 3 In. Pipes Sched. 160 RWCU, RRC 6 RHR Systems Figure F-25 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 6 In. Pipes Sched. 80 RWCU, RRC 6 RHR Systems Figure F-26 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 16 In. Pipes Sched. 80 RWCU, RRC a RHR Systems Figure F-27 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 20 In. Pipes - Sched. 80 RWCU, RRC a RHR Systems
!WASHINGTON PUBLIC PONER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT iI2 Page xvi LIST OF FIGURES {continued) Figure F-28 Dose Rate at Pipe i'1id-Plane vs. Distance Crom Pipe Surface 24 In. Pipes Sched. 80 RNCU, RRC h RHR Systems Figure F-29 Dose Rate at Pipe i'1id-Plane vs. Distance from pipe Surface LPCI System Schedule 80, D = 12 In. > O.D. = 12.75 In., I.D..= 11.376 In. Figure F-30 Dose Bate at Pipe Mid-Plane vs. Distance from Pipe Surface LPCI System Schedule 80~ D = 24 In.,O.D. 14 In.. I.D. = 12.5 In.
.Figure F-31 Dose Rate vs. Distance from Pipe Surface Pipe Length = 2.5 Ft ~
F igure F-32 Dose Rate vs. Distance f rom Pipe Surface Pipe Length = 5 Ft. Figure F-33 Dose Rate vs. Distance Crom Pipe Surface Pipe Length = 10 Ft. Figure F-34 Dose Rate vs. Distance Crom Pipe Surface Pipe Length = 15 Ft. Figure F-35 Dose Rate vs. Distance Crom Pipe Surface Pipe Length = 25 Ft. Figure F-36 Dose Rate vs. Distance Crom Pipe Surface Pipe Length = 40 Ft. Figure F-37 Dose Rate vs. Pipe Source Length at Various Distances Crom Outer Pipe Surface
0 i PUBLIC POWER '.v'ASHINGTON PLANT SHIELDING . NASHINGTOVi NUCLEAR SUPPLY SYSTEM AViALYSIS PROJECT ,/Z Page xvii LIST OF APPENDICES A~ UNISOLATED LEAK PATH REPORT B ~ SOURCE TERM DEVELOPMENT AND PARAMETRIC STUDIES IN S ECON DARY CONTAINMENT C ~ PROCEDURE FOR THE CALCULATION OF SECONDARY CONTAINMENT RADIATION ZONE DOSES Do CALCULATION OF THE SECONDARY CONTAINMFNT RADIATION DOSES DUE TO STANDBY GAS TREATMENT SYSTEM BETA DOSE POINT DERIVATION F. PRIMARY CONTAINMENT ANALYSES BETA DOSE CONTRIBUTION IN, PRIMARY CONTAINMENT VITAL AREAS AND ACCESS ROUTES ANAIYSED FOR POST-LOCA OPERATIONS
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'.v'ASHINGTON PUBLIC POWER PLANT SHIELDING 'WASHINGTON NUCLEAR
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SUPPLY SYSTEM AVALYSIS PROJECT Page xiii
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SUMMARY
The Three Mile Island (TMI-2) accident has-generated a concern that during an accident in which significant core damage occurs, the post-accident operations requiring the use of systems containing contaminated fluid may induce abnormally high radiation doses to safety-related equipment and components which make difficult to operate the systems. The it NRC initially addressed this concern with Vi UREG-0578 and NUREG-0737 and recommended a design review to evaluate the func-tional capability of safety-related equipment and radiation exposure to per-sonnel during the postulated post-LOCA operations. Radiation levels have been determined for all areas containing safety-related equ iprhen t, v i ta1 areas, and access routes which are required for the postu-lated post-LOCA operation. Radiation levels are currently being determined for safety-related equipment inside primary containment. The safety-related list of equipment located inside primary containment is identified in Table 6.1. Analysis effort in the shadow shielding effect of primary containment hardware and the effect of first order iodine plateout is currently in progress to calculate the radiation levels inside containment. Radiation levels determined for safety-related equipment in Secondary Containment are reported in Figures 6.1 through 6.8. An analysis effort is con-tinuing in Secondary Containment due to , impact of considering iodine 'nticipated
Rev. I !v'ASHINGTON PUBLIC PO'HER PLANT SHIELDING HASHI'GTOtl NUCLEAR
~ SUPPLY SYST&1 AiVALYSIS PROJECT A/2 Page xix plateout and time/pressure history of containment following a LOCA. The radiation source term leaking into Secondary Containment vill be reduced by the loss of halogens to plateout. Also pressure inside Containment will a'educed impact the pressure related time Depen-dent leakage into Secondary Containment.
These impacts are being evaluated to more accurately predict radiation levels inside Secondary Containment. Beta radiation inside Secondary Containment has also been considered and can be determined from Figure 5. l. and Figures E-1 thxough E-9. Radiation levels calculated for safety-re la ted equ ipmen t ou ts ide Second ary Containment axe reported in Table 6.3. Figures 6.19 through 6.23 identify the vital areas outside Secondary Con'tainment which contain safety-related equipment. Safety-related equipment will either be qualif ied for the radiation level it functions in, or it will be relocated to a radiation zone it is qualified for, or i t w i 1 1 be re placed wi th compa rable equipment which is qualifiecl for the par-ticular x'adiation level that has been determined.
~j'ital Areas anal Access Routes were eva-luated for post-LOCA operations and are reported in Table 6.4 and Figures 6.19 through 6.?4. All areas and access routes are in compliance with NUREG-0737 except the security guardhouse and the auxiliary
WASHINGTON PUBLIC POMER PLANT SHiELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS P"OJECT !t2 Page xx security center. Security personnel will he relocated in the Technical Support Center per the ilNP-2 Emergency Preparedness plan if ( guardhouse radiation levels approach, those radiation guidelines presented in NUREG-0737 during the post-LOCA situations.
KQV L '.v'ASHINGTON PUBLIC POWER PLAiNT SHIELDING HASHINGTOVi NUCLEAR SUPPLY SYSTEH AiVALYSIS PROZECT !/2
>>Se xxi 4 ABSTRACT This report presents a radiation shielding design review of the equipment and systems of the-Naehing ton Public Power Supply System Nuclear Project Unit 2 (NNP-2) that may, as a result of an accident, in addition to normal plant radiation levels during its 40 year life contain highly radioactive fluids. This design review recommended by the NRC (NUREG 0570 and
'UREG 0737) evaluates the Functional Capability of safety-related equipment and personnel radiation exposure during the post-accident operations.
This design review evaluates the post-accident radiation conditions for person-nel located in vital areas (areas which require access or occupancy during the post-LOCA scenario) on either a con-tinuous or infrequent basis. The postulated Loss of Coolant Accident ( LOCA) scenarios and the operations of the safety-related systems were reviewed. Radioactive sources contained within each system were developed. Radiation levels were calculated at sa f e ty-re la ted equipment locations, as we as at selected locations outside the Reactor ll Building to which access may be required for pos t-acc iden t opera t ions . 1i
lcev ~ WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR l SUPPLY SYSTEM ANALYSIS PROJECT i/2 Page 1-1 1 ~ 0 INTRODUCTION This report presents a detailed description of and the results from the review of plant shielding and radiation environmental con-ditions for equipment and systems which may be used in post-accident operations for WNP-2. The review was initiated in response to Section 2.1.6.b of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendation" and to Part II.B.2 of NUREG-0660 "NRC Action Plan Developed as a
'Result of the TMI-2 Accident."
The design review determined the post-accident radiation environmental conditions for equip-required for post-accident operations 'ent inside the Primary Containment, inside the Secondary Containment, and outside the Secondary Containment. The six month total post-accident radiation dose rate as a function of time and the integrated dose are curre'ntly being calculated for safety-related equipment locations inside the WNP-2 Primary Containment. The six month total post-accident radiation dose rate as a function of time and the integrated dose were calculated at safety-related equip-ment locations inside the WNP-2 Reactor Building and at selected locations (vital areas) outside the Reactor Building. Section 2. 0 discusses the regulatory requirements upon which this report is based and provides a description of the tasks performed for this Shielding Fvaluation. Section 3.0 provides the systems review and source term assumptions used as input for the definition of the post-accident radiolog ical env ironmen t.
ii'ASHINGTON PUBLIC POWER PLANT SHIELDING 4'ASHI'.!GT6N NUCLEAR SUPPLY SYST121 AVALYSIS PROJECT ~'/2
'I Page l-2 Section 4. O discusses the ivork performed during this project relating to safety-related-equipment located outside of the Reactor Building and the Access and Occupancy of Vital Areas. This consists of the calculation of dose rates outside the Reactor Building.
Section 5.0 discusses the methods of Calculation including the use of computer codes, identifying the parameters that have a significant effect on the radiation dose rates, and the dose rate and cumulative dose 'calculation proce-dure. Section 6.0 presents a summary of the resul ts.
PLANT SHIELDING WASHINGTON NUCLEAR WASHINGTON PUBLIC POWER ANALYSIS PROJECT g2 SUPPLY SYSTEM Page 2 2.0 REQUIREMENTS General Design Criterion 4 requires that systems and components important to safety be designed to accommodate the environmental con-ditions associated with accidents. The Three Mile Island (TMI-2) accident has generated a concern that during an accident in which significant core damage occurs, the post-accident operations requiring the use of systems containing contaminated fluid may induce abnormally high radiation doses to safety-related equipment and components which may make it difficult to operate the systems. The Nuclear Regulatory Commission (NRC) Lessons Learned Task Force initially addressed this concern in Section 2.1.6.b of NUREG-0573 ( Ref. 2.1) and recommended a design review be performed on such systems such that the func-tional capability of safety-related equipment located in close proximity to the resulting high radiation field will not be unduly degraded. Described in this section is a discussion of the current regulatory requirements and guide-lines used.
?.1 Sh ield ing NUREG-0578 Section 2.1.6.b requires that each Evaluation licensee perform a radiation and shielding Regulatory design review of the spaces around systems Requirements that may, as a result nf an accident, contain highly radioactive materials. The scope of the review includes the following:
- l. Identification of the locations of vital areas and safety-related equipment.
- 2. Evaluation of the radiation level at each location.
<ASHINGTON PUBLIC. POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEf ANALYSIS PROJECT f12 Page 2-2
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Provision for adequate access to
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3. vital areas and assurances o'f pos't-accident equipment operation through design changes, increased-permanent or temporary shielding, or post-accident procedural controls. In order to perCorm this review, the NRC has provided guidance in the following documents ("documents of record" ). o NUREG-0578 Section 2.1.6.b Reference 2.1
.o NUREG-0588 Section 1.4 Re ference 2. 2 o NUREG-0660 Section II.R.2 Reference 2.3 o Clarification Letter to NUREG-0578, dated Sept. 5,
.1980, Section II.B.2 Reference 2.4 o NUREG-0737 Section II.B.2 Reference 2.5
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o, IE. Bulletin No. 79-018 Reference 2.6 o IF.'ulletin No. 79-018 Supplement 2, dated Sept. 30, 1980 Reference 2.7 The regulatory requirements in the above men-tioned documents are summarized in the following sections. 2.1.1 Accident The post-accident radiation environment should Ana1ysis be based on the most severe design basis acci-Requiremen ts dent (DBA) during or following which equipment must remain fun'ctional. 'his includes the consideration of the entire spectrum of Loss of Coolant Accident (LOCA) events which can lead to a degraded core condition. These accident conditions include: a., LOCA events which completely depressurize the primary system.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT ~/2 Page 2-3
- b. LOCA" events in which the primary system may not depressurize.
2.1.2 Source Term The radioactive source terms for the postulated Assumptions accident conditions as described in Section 2 1 1 should be equivalent to the source terms
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recommended in Regulatory Guide 1.3 and 1.7 and Standard Review Plan 15.6.5. The source term assumptions consistent with current licensing- requirements used for equipment qualification and access evaluations are summarized as follows:
- 1. The fission product fractions assumed to be released from the fuel rods during a LOCA are:
noble gases 100% halogens 50% remaining fission pr'oducts l These release fractions are not the sum of (double counting) the assumed liquid and gaseous releases. In effect, 254 of the equilibrium halogen activity is assumed to be mixed in the containment atmosphere and 50~ is assumed to be mixed in the Reactor Coolant and recirculated liquids. The post-LOCA .source contri-bution from liquid and gaseous sour-ces are analized separately and the worst dose is tabulated for that eva-luation rather than the sum of both doses. Thus, double counting of the fission product fractions is elimi-nated.
WASHINGTON PUBLIC PO'HER PLANT SHIELDING WASHINGTON NUCLEAR ! SUPPLY SYSTEH ANALYSIS PROJECT f12 Page 2-4
- 2. The above release is assumed to occur and be distributed instantaneously at the start of the accident.
- 3. Until depressurized, liquid in the Reactor.
Coolant System (RCS) and other systems which are not isolated from the core and which contain the reactor coolant at the start of the LOCA contain 100% noble gases, 50% halogens and 1% of the remaining fission products. These radioactive materials are mixed homoge-neously in a volume no greater than the RCS liquid space.
- 4. Liquid in the suppression pool and any system not isolated from the core at the start of the LOCA, and containing only liquid from a depressurized source, is assumed to contain 50% halogens and 1% of the remaining fission products. These radioactive materials are diluted homoge-neously in a volume no greater than the
-combined volumes of the suppression pool and the RCS liquid space.
- 5. The Primary Containment atmosphere and systems which are not isolated from the Primary Containment atmosphere at the start of the LOCA are assumed to contain at least 100e nohle gases and gge halo-gens. These radioactive materials are diluted homogeneously in a volume no greater than the combined volumes of the drywell and suppression pool air spaces.
-6. Until the Reactor Vessel is depressurized, gases in the steam lines and any other
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page 2-5 vapor-containing lines not isolated from the core at the start of LOCA are assumed to contain at least 100% noble gases and 25%, halogens. These are dilutecl uniformly in a volume no greater than the RCS steam space and ad joining, unisolated steam lines. 2.1.3. ~Iital Area As defined in NUREG-0737 (Ref. 2.5), a vital Access area is an area which vill or may require Requirements occupancy to permit an operator to help in the
'itigation of an -accident or perform post-accident operations. The accident scenarios discussed in Section 2.1.1 and the source term assumptions in Section 2,1,2 are used for the evaluation of vital area access and occupancy.
The total radiation exposure to personnel in vital areas should not be in excess of 5 rem whole body, or its equivalent, to any part of the body for the duration of the accident. For areas requiring continuous occupancy (e.g., the Control Room, Onsite Technical Support Center, etc.), the dose rate criteria limits the total radiation exposure to person-nel to less than 15 mrem/hr (averaged over 30 days).
- 2. 1. 4 Systems Systems considered in the shielding review Containing are those systems that could have the he Sources potential of containing a high level of radioactivity post-accident. <or those systems connected directly to tne Reactor Coolant System or to the Primary Containment atmosphere and not isolated at the start of the accident, the radioactivity is assumed to be instantaneously mixed within the uniso-lated parts of the system.
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTV1 ANALYSIS /t2 'ROJECT Page 2-~ 2.1. 5, Safety-Related The safety-related (1E/1N) equipment list con-Equipment tains all equipment necessary to mitigate the consequences of an accident, bring the plant . to a safe shutdown condition, and provide long-term cooling capability. This list includes equipment located inside as well as outside the Primary Containment. 2.2 ., Shieldiqg The Shielding Evaluation tasks which have been Evaluation completed to date are as follows: Task Descript'ion 1. Review all accident scenarios and accident conditions that could result in a limiting radiation environment for all the pieces of safety-related equipment on the
] E/]N (Safetv-Re].ated~ list that are locater1 in the Reactor Building.
2., Identify systems and components that could potentially contain radioactive materials pos t-acc ide'n t.
- 3. Generate source term assumptions based on regulatory requirements discussed in Section 2. l.
- 4. Calculate accident radiation service con-ditions for the safety-related equipment located inside the Reactor Building.
- 5. Calculate gamma dose rates at selected locations outside the Reactor Building due to radioactive sources inside the Reactor Build ing.
6.. Identify vital areas and equipment outside the Reactor Building to evaluate the access to and occupancy of the vital areas
Rev. 1 iv'ASHIi~GTONi'UBLIC POWER PLANT SHIELDING HASHINGTOiit NUCLEAR SUPPLY SYSTEi~i ANALYSIS PROJECT 'I2 Page 2-7 I in accordance with the requirements listed in -Section 2.l.
- 7. Conduct a Primary Containment analysis of LOCA events in which the RCS may not depressurize (or may repressurize) with a degraded core condition.. The. Primary Containment radiation environment was determined with the use of l00% noble gases, 50% halogens and 1% o f the remaining fission products for the period of time during which the activity is iso-la ted to the RCS.
2'3 Shield ing NNP-2 has addressed all the issues needed to Evaluation comply with the t>UREG-660 II.B.2 position I tern Deleted except as follows: NNP-'2 takes exception to From Shielding the portion of the task that specif ies that .~ Analysis review of "sa"ety-related equipment w!sich may Consideration be degraded by radiation during post-accident operation be provided for a Non-LOCA, High-Energy Line Break Source Term". The pipe
.break/missile analysis performed in Sections 3.5 and 3.g of.the FSAR addresses non-mechanistic pipe breaks inside and outside containment. These pipe breaks do not lead mechanistically to a radiation release due to fuel failures beyond those allowed in normal operation. Hence, the source term identified and applied outside containment is entirely hypothetical and would be a new design basis beyond the scope of current regulations.
2.4 Add i t ional The i tems wh ich need to be per Cormed to Review Items achieve Cul l compl iance wi th the regula-Required For tory requirements are summarized as Fi nal Plan t fol lows: Sh ield ing Report
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR ANALYSIS ~'!2 SUPPLY SYST&1
'ROJECT Page 2-8
- 1. The equipment considered for this study "was-limited to the initial lE/1H (safety-related) list. Mditional 1E/1M equipment subsequently identified will he evaluated with integrated radiation results presented 'in the Final Shielding Evaluation Report,
- 2. For this report no attempt was made to verify the completeness of the safety-related safety-related equipment list.
The safety-related equipment list will contain all equipment required to "mitigate" the consequences of an acci-dent, bring the plant to a safe shutdown condition, and provide long-term cooling capability". This effort is currently underway and will be addressed in the Final Shielding Evaluation Report.
- 3. Complete the calculation of radiation levels inside primary containment due to the analysis effort in the shadow shielding effect of Primary Containment hardware and the effect of first order iodine plateout.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT f12 Page 3-1 3.0 ANALYTICAL Xn olde~ to develop the method used in the HETHODOLOCiY calculation of radiation doses, a review of all the postulated accident scenarios and system operations were performed. Source term assumptions were developed baseR on the results of accident analysis and system review, as well as the regulatory guidelines described in Section 2,1. The systems and components inside the Reactor Building that have the potential of becoming contaminated during or following the accident were identif.ied. The following subsections describe these acti-vities in greater detail. Section 3.1 descri-bes the accident scenario chosen for this analysis. Section 3.2 identifies all the con-taminated systems. Section 3.3 describes the source term assumptions generated for each contaminated system. Section 3.4 identif ies the time period consiRereR for this study.
- 3. 1 Accident The accident analyses consistent with FSAR Scenario Chapter 15 for small and large break Loss of Coolant AcciRents (LOCA's) were considered.
The entire spectrum of lOCA conditions that could result in a Regraded core configuration was revieweR and it was concluded that there is no single accident scenario that could result in a limiting radiation environment all the safety-relateR equipment located in for the Reactor RuilRing. Therefore, the accident scenario chosen here is based on a non-mechanistic LOCA in which core damage is experienced at the beginning of the accident, primary Containment isolation is assumed to be achieved prior to radioactivity transport.
PLANT SHIELDING WASHINGTON NUCLEAR WASHINGTON PUBLIC POWER ANALYSIS PROJECT 82 SUPPLY SYSTEM 3-2 Page A review of the post-accident operation of the 1E/1M (safety-related) systems was conducted. The result of this review indicated that the worst case accident for the steam supply system (highest source term) was the pressurized Reactor Coolant System .(RCS). For the liquid systems (the ECCS, the RHR, and. the RCIC system), as well as the Primary Containment Atmosphere and Primary Containment Atmosphere Control System (CACS), the worst case accident is the depressurized Reactor Coolant System with the post-LOCA core release fractions dispersed within the Primary Containment.
,3.2 Contaminated In order to perform the radiation dose calcu-Systems lations, it was necessary to identify the systems which would or could contain highly iadioactive materials during the post-accident period. Systems required to operate during the post-accident oeriod are those:
o Systems necessary to mitigate the con-sequences of a large or small break LOCA. o Portions of systems that are in com-munication with systems containing radioactive liquids or gases. o Def ined by the ilRC as being required; such as Gaseous Radwaste System. (See Section 3.2.3).
NUCLEAR CWASHINGTON PUBLIC POWER SUPPLY SYSTEM ANALYSIS'ASHINGTON PLANT SHIELDING PROJECT /t2
~
Page 3-3 3 ~ 2 ~ 1 Systems The following systems are being considered: Included for Primary o 3ffgh Pressure Core Spray (HPCS) Containment Analysis o Low Pressure Core Spray ( LPCS) o Residual Heat Removal (RHR) o Reactor Core Isolation Cooling (RCIC) o Floor Drains and Equipment Drains ( FDR-EDR) o Reactor Water Cleanup (RWCU) o Hain Steam (NS) o Reactor Recirculation (RRC) o Sample Lines (PSR) o Automatic Depressurization System (ADS) o Low Pressure Coolant Injection (LPCI) Function of the RHR system after depressurization 3 ~ 2~2 Systems The following systems were considered: Included for Secondary o Reactor Core Isolation Cooling (RCIC) Containment Analysis o Residual Heat Removal (RHR) o Low Pressure Coolant Injection (LPCI). o Low Pressure Core Spray (LPCS) o High Pressure Core Spray (HPCS)
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT 82 Page 3-4 o Containment Atmosphere Control (CAC, the Hydrogen Recombiners) o Hain Steam (NS, Up to Second Isolation Valve) o ilain Steam Line Isolation'alve Leakage Control (NSIVLCS) o Primary Containment o Secondary Containment Atmosphere o Standby Gas Treatment (SGT) 3.2.3 Systems All systems required to mitigate the con-
=-Exclud ed sequences of an accident have been included.
Of those systems recommended for consideration in regulatory documents, one system (Gaseous Radwaste) has been excluded. The Gaseous Radwaste is isolated by the Primary Containment and Reactor Vessel 1solation Control System and will not receive contaminated gas unless operation is manually initiated. The WÃP-2 operating and accident procedures do not take credit for nor antici-pate using this system. Since WNP-2 philo-ophy is based on containment of. the core releases within the Primary Containment, this system will not be requir, d and <<as, there-fore, excluded from consideration. 3 ' Source Term Regulatory requirements specify that source Assumptions terms equivalent to those recommended in Regulatory Guides 1. '3 and 1. 7 and Standard Review Plan 15.6.5 be used in the LOCA acci-den't analysis. Additional guidance is given
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT /f2 Page in NQREG-'0588 ( Re f . 2. 2) and NUREG-0737 ( Re f . 2.5) and is documented in Section 2.1, Source term assumptions were generated based on the review of the operation of the safety syst ms. Because a non-mechanistic LOCA scenario was chosen for this analysis, the worst con-taminatecl situation for the fluid contained within each system was conservatively assumed. Tables 3.1, 3.2 and 3.3 list the assumptions involved in the distribution of. fission products used in this analysis. These assump-tions-are. consistent with the regulatory requirements discussed in Section 2.1. A review of the ooeration of each of the systems discussed in Section 3.2 was also con-ducted. This review identified the source of contaminatecl fluid contained within each system post-accident. Using the source term assumptions cliscussecl i.n Tables 3.1, 3.2 and 3.3, together. with the results of this system review, the. limiting,source term (activity dividecl by dilutio'n factor) was determine'd for each system. Table 3.4 is a summary of tne system'perations and source term assumptions developed for each contaminated system iden-tified in Section 3.?. 7 ime Pe r'iod All systems were conservatively assumed to Cons ice re<1 become contaminated at the start of the acci-For S tuel y clent and remain contaminatecl until the i n teg ra ted'acl ia t ion dose reached i t asym tot ic value. It was noted that the integrated dose becomes nearly asymptotic to a constant value beyond about 6 months. Therefore, 6 months is the time period chosen for accident dose qualif ication in this report.
Rev. 2 HASHIiXGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH AiVALYSIS PROJECT i'l2 Page 3-5 Table 3.. 1 Distribution of Fission Products in the Worst Post-LOCA Situation for Areas Inside Containment. Depressurized Reactor Coolant System. Suppress'.on Pool (1) Primary Containment and Reactor Coolant Air and Steam Space S stem Wa te r Volume Fission Fraction Dilution Fraction Dilution Products (2) Volume (3) Plateout (2) Volume (3) Noble Gases 100% Drywell Air 0% 0% Suppression Plus Pool Water Halogens 50% (4) Suppression 0-47.5%(5) 50% and RCS Pool Water Particulates 0% Air 0% 1% Volume If equipment must be qualified in the suppression pool atmosphere, then a uniform distribution between drywell and suppression pool atmosphere will he assumed. (2) Expressed in percentage (%) of total core inventory at End-of-Life conditions (1,000 days at 3481 mwt) . (3) Represents the total volume in which the fraction of. core fission products is assumed to be homogeneously mixed. (4) In calculating the radiation dose at a particular location, it is not necessary to assume that all source distribution assumptions are conservative simultaneously. Instead, a set of mutually com-patible ~assumptions will he used which gives the maximum dose for the location being considered. The post-LOCA source contributors are used to calculate independant doses for each contributor. The worst dose is tabulated for that system rather than the sum of. all contributors (i.e., 50% halogens airborne and 50% halogens i.n the water) . Thus double counting of the f ission pro-duct fractions is eliminated. (5) First order iodine plateout occurs during the first 5-6 hours of the post-LOCA time f rame when the elemental halogen concentration is reduced by a factor. of 200. This methodology is in accordance w i th NUREG/CR-0009.
r Rev. 2 IWASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR I SUPPLY SYSTEM ANALYSIS PROJECT //2 Page 3-7 Table 3.2 Distribution of Fission Products in the Worst Post-LOCA Situation for Areas Inside Containment. Pressurized Reactor Coolant System. (1) Suppresszon- Reactor Coolant Reactor Coolant pool Water System Water System Steam Drywell Air Volume and Volume Space Space (1) Air Space (1) (1) (1) Fraction Fraction Fraction Dilution Fraction Dilution Fission (2) (2) (2) Volume (2) Volume products (3) (3) Noble 05 09 100% 100%
- Normal Gases (4) (5) RCS Halogens 0% 0% 504 RCS 25% Steam Water Space Particu- Volume (6) lates 0% 0% 0%
) The reactor coolant system will remain pressurized for a short period of time (17 hours) and then will be depressurized.
(2) Expressed in percentage (%) of, total core inventory at End-of-Life conditions (1,000 days at 3481 mwt). (3) Represents the total volume in which the fraction of fission prodcuts is assumed to be homogenously mixed. core (4) In calculating the radiation dose at a particular location, it is not necessary to assume that all source distribution assumptions are conservative simultaneously. Instead, a set of mutually com-patible assumptions will be used which gives the maximum dose for
'the location being considered. The post-LOCA source contributors are used to calculate indepen'dant doses for each contributor.
The worst dose is tabulated for that system rather than the sum of all contributors (i.e., 50% halogens airborne and 50% halogens in the water ) . Thus double counting of the f ission pro-duct fractions is eliminated. (5) The dilution volume i. the RCS water volume plus the RWCU lines up to the isolation valves, RHR lines up to the isolation valves, and the RRC lines during the 17 hours of the pressurized RCS scenario. The dilution volume is the normal RCS steam space plus the NS lines up to the isolation valves during the 17 hours of the pressurized RCS scenario. The 100% of noble gases, present during the 17 hours of the oressurized RCS during a LOCA, are homogeneously mixed in the water and steam dilution volumes identified
Table 3.3 Distribution of Fission Products In The llorst Post-LOCA Situation For Areas Outside Containment Primary Containment Suppression Pool Reactor Coolant Reactor Coolant Air Space Water Volume System Steam Space System Water Volume Fission Fraction Dilution Fraction Dilution Fraction Dilution Fraction Dilution Products (1) Volume (2) (1) Volume(2) (1) Volume ( 2) Noble Gases 100% Drywell Suppres- 100% Normal 100% sion RCS Air Plus Pool RCS Halogens 25% (4) 50% (5) Water 50% Water Suppression Plus Steam RCS Volume Particulates 0% Pool Air Water Space (1) Expres-ed in percentage (%) of total core inventory at End-of-Life Conditions (1,000 days at 3481 t4Wt) ~ (2) Represents the total volume in which the fraction of core Fission products is assumed to be homogeneously mixed. (3) Based on pressurized Reactor Coolant System (4) Half nf the 50% of the halogens released from the core are assumed to plate-out instantaneously as allowed by NUREG-0588 Rev. 1. The plate-out dose was considered in the total calculation of radiation dose to equipment inside Primary Containment. (5) In calculating the radiation dose at a particular location, it is not necessary to assume that all source distribution assumptions are conservative simultaneously. Instead, a set of mutually compatible assumptions will be used which gives the maximum dose for the location being considered. The post-LOCA source contributors are used to calculate independant doses for each contributor. The erst dose is tabulated for that system rather than the sum of all contributors (i.e., ?5% halogens and 50% halogens). Thus double counting of the fission product fractions is eliminated.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT g2 Page 3-9 Table 3.4 System Operations and Source Term Assumptions System Operation Post-Accident Contaminated Source Term l Source Assumptions HPCS Suction from Condensate Storage Suppression I Tank and/or Suppression Pool and Pool discharge to the Reactor Vessel. I LPCS Suction from Suppression Pool and Suppression discharge to the Reactor Vessel. Pool LPCI Suction from Suppression Pool and Suppression (1) discharge to the core. Pools i (6)'C IC Steam bled-off from Reactor Steam RCS Steam (2) Steam Space is used to drive the RCIC tur- Space System bine, and evhausts into the Sup-pression Pools
!
RCIC Liquid Suction from Condensate Storage Suppression (1) System Tank or Suppression Pool and dis- Pool charge to the Reactor Vessel. RHR System (1) Shutdown Cooling Mode - suction RCS Liquid Note A from reactor recirculation sy- Space stem suction line and discharge into the reactor recirculation discharge line. (2) Alternate Shutdown Cooling Suppression Mode suction from Suppression Pool Pool and discharge to core re-circulates and cools the water in the Suppression Pool. (3) Containment Spray Cooling i'1ode- Suppression suction from Suppression Pool Pool and discharge into the Drywell and Suppression Pool.
WASHINGTON PUBLIC POWER PLANT SHIELDING !WASHINGTON NUCLEAR SUPPLY-SYSTEM ANALYSIS PROZECT 82 Page 3-10 Table 3.4 cont'd System Operation Post-Accident Contaminated Source Term Source .Assumptions (6) RHR 4) Reactor Steam Condensing Mode RCS Note A System - steam bled off from reactor ves- Liquid (cont'd) sel, condensed through the RHR Space heat exchanger and directed to the RCIC pump suction or Sup-pression Pool. Main Stagnant steam fzom the reactor RCS (2) S'team vessel terminates at the second Steam Supply MSIV ~ Space ~ MSIVLCS Steam bled off from main steam RCS (2) line, diluted and discharged Steam into the SGT filter zoom. Space Note B SGT Process the halogens fr'om pri- ~ Primary Contain- (3) Filters mary contai.nment leakage and ment Atmosphere MSIVLCS ~ CAC Process the Primary Containment Primary Contain- (4) Atmosphere. (Hydrogen Recom- ment Atmosphere bination) Primary Primary ontainment is isolated Primary Contain- (4) Containment . Post-Accident. ment Atmosphere Suppression The primary function of the sup- Suppression Pool Pool pression "pool is to contain and Liquid condense the blowdown from the RCS post-accident. ~ Containment Secondary The primary function of the Primary Contain-Secondary Containment is to ment Atmosphere contain all the leakage from the Primary Containment Poet:- Accident.
lie V ~ A
- v'ASHINGTOh'UBLIC PObER PLANT SHIELDING VASHI GTON %NUCLEAR SUPPLY SYSTEi'1 ANALYSIS PROJECT lI2 Page 3-11 Table 3.4 cont'd System Operation Post-Accident Contaminated Source Term Space Assumptions Sample Actuated to obtain Primary Pritnary Contain- (2)
Lines Containment Atmosphere ment Atmosphere samples per NUREG-0737 Sample Actuated to obtai.n RCS liquid RCS Liquid Space I,ines sample per NUREG-0737 Reactor Reactor water cleanup system Reactor Coolant Hater isolated during post-LOCA. System Liquid Cleanup Liquid up to the second iso-lation valve is considered contaminated. Reactor Suction from reactor recircu- Reactor Recir-Recirculation lation system suction line and culation I,iquid, discharge into the reactor re-circulation discharge line. Floor Drains Liquid from ruptured pipes or RCS L iqu id and Equipmen leaky sea,ls discharged into Drains the suppression pool. Au toma tie Automatic or manual depres- RCS Steam (2) Depressuriza- surization of the reactor tion System vessel by blowdown of. the RCS into the suppress ion pool. Au toma tie Alternate Shutdown cooling Suppression, Pool Depressur iz'a- mode wi th reflood of reactor tion System vessel and d ischarge into su'ppression pool. "
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS f!2 'ROZECT Page 3-12 Table 3.4 cont'd Source Term Assumptions (1) .50% flalogens and 1't Solid Fission Products diluted with Suppression Pool water plus RCS water. (2) 100% Noble gases and 254 Halogens diluted with the RCS steam space. (3) 25% Halogens leaked from the Primary Containment is assumed to be
,
deposited in the SGT Filters at the rate of 0.73% per day. See section 5.5.4.1 for justification. 100% Noble Gases pass through also but are not absorbed. (4) 100% Noble gases and 50% Halogens diluted with the Primary Containment air space. 47.5% Halogen plate-out inside Primary Containment was considered. (5) Assumptions involved in the calculation of source terms for Secondary Containment Atmosphere is discussed in detail in Section 5. 5.1. (6) Based on a pressurized Reactor Coolant System. NOTFS: A According to accident mitigation procedures, this mode of operation is not 'used after a degraded core condition is identified. B For the portion of system after the distribution header, credit is taken for dilution by clean air. See section 5.5.4.1 for justification.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH - ANALYSIS PROJECT I!2 Page 4-j S 4.0 ACCESS AND NUREG-0578 initiated the requirement for a OCCUPANCY OF design review to identify the location of VITAL AREAS vital areas in which personnel occupancy may be unduly limited by the radiation f ields during post-accident operations. I t. requ ired that each licensee provide adequate access to v.ital areas by design changes, increased permanent or temporary shielding, or post-accident procedural controls. NURFG-0737 further makes the point that the purpose of this design review is to determine what actions can be taken over the short-term to reduce radiation levels and increase the capability of operators to control and miti-gate the consequences of an accident. This shielding evaluation includes the calcu-lation of gamma, dose rates at selected loca-tions outside the Reactor Building due to radioactive sources inside. The radioactive source terms obtained from ORIGFN computer calculations coupleR with recommendations from Regulatory Guide 1.109 were the basis for the assumptions used in evaluating vital areas and acces. routes outside the Reactor Build ing. 4.1 Dose Rates An analysis was conducted to determine the Ou ts ide the dose rates at selected 1ocations outside the Reac tor Bu i ld ing Reactor Building for personnel access pur-pose". The radiation level in the various areas outside the Reactor Building is defined by the following three radioactive sources:
- 1. Direct gamma ray dose from radioactive piping located inside the Reactor Building and attenuated through the walls o f the Re ac tor Bu i ld ing.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTE'f ANALYSIS PROJECT f32 Page 4
- 2. - Gamma shine dose from airborne activity inside the Reactor Building.
- 3. Gamma dose from airborne activity outside the Reactor Building.
Radiation levels outside the Reactor Building were determined by the zone dose method as discussed in Sectiongaby 5.5. Representative zones were chosen at selected locations out-side the Reactor Building such as ground level outside the railroad bay, sampling room, etc. The worst point in a zone was chosen to be the point directly outside the Reactor, Building wall, at. a height of six feet above floor elevation, at a lateral point determined inspection to receive the highest dose along that wall. The zones outside the Reactor Building are indicated by the letters Y and 7 in the various elevations indicated in the radiation zone maps (shown in Figures 6.1 through 6.8) . The result of the close calculations are shown in Figures 6.19 through 6.24.
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page 4-3 Appendix H presents the methodology used to calculate the radiation doses for the various vital areas and access routes. 4;2 Vital Areas Radiation calculated for the access routes and Access were based on the assumption that no indivi-Routes Outside dual would be in an access route longer than the Reactor 30 minutes for the first 8 hours after the Building postulated LOCA before reaching the vital area of interest. The assumption was also made that no indivi-dual would occupy an infrequent occupied vital area longer than 30 minutes for the first 8 hours after the postulated LOCA. All integrated radiation doses calculated for the access routes were less than the guideli-nes presented in NUREG-0737. All vital areas evaluated had radiation doses less than the guidelines presented in NUREG-0737 except for the security guardhouse and auxiliary security center. Security per-sonnel will he relocated to the Technical Support Center per the NNP-2 Emergency preparedness plan if radiation levels approach those radiation guidelines presented in NUREG-0737 during the post-LOCA operations.
Rev. 2 4ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT !f2 Page 5-l 5 ' METHODS The radiation dose assessment of safety-related equipment inside containment is being done by calculating the radiation dose to each component from all appli-cable radiation sources,(contaminated liquid, steam, and airborne. sources). The Secondary Containment radiation dose assessment portion of the Shielding Evaluation was initiated by dividing the Reactor Building into radiation zones. Because of the large number of radioactive piping and safety-related equipment in the building, the division of the various regions of the Secondary Containment into radiation zones permits a precise, detailed calculation - of the total integrated dose at the "worst target" location. The methods for performing the calculations are discussed in detail in the following sections. The radiation dose assessment of safety-related equipment outside of the Reactor Building was done by calculating the radiation dose of each vital area where safety-related equipment was located. The assumptions and methodology used to perform these calculations are discussed in detail in the following sections and Appendix H. 5.1 The Use of The two computer codes used in the Primary Computer Codes Containment shielding evaluation are ORIGFN 2 and QAD-CG. Descriptions of the two codes are found in References 5.1, 5.2, 5.3 and 5.10. ORIGFN 2 com-putes the radioactive source terms (inside containment) used by QAD-CG to calculate the radiation doses from piping and various pieces of equipment.
Rev. 2 I I <WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR ~ SUPPLY'YSTBi ANALYSIS PROJECT 82 i 5-2 Page The two computer codes used in the Secondary Containment radiation shielding review were ORIGEN and QAD-P5A. Descriptions of the codes are found in References 5.4 and 5.3. ORIGEN computes the radioactive source terms used by QAD-P5A to compute the radiation from piping and other source configura-tions to pieces of equipment. ORIGEN and ORIGEN 2 are a versatile fis-sion product source term codes which solves the equations of radioactive growth and decay for large numbers of isotopes. ORIGEN 2 is being used to calculate the radioactivity of fission products and fuel 'materials that were assumed to be released from the reactor core during the postulated LOCA to become the Primary Containment source terms for the dose rate calculations.
- 5. 2 Source Term The radiation level at any given loca-Dev'elopment tion inside the Primary Containment of For Pr imary NNP-2 following the postulated LOCA such Containment as that clescribed in Section 3.1 is determined from the following major source contributors:
Gamma ray dose f rom airborne radioac-tive sources suspended in the drywell and wetwell inside Primary Containment (Airborne Gamma Dose).
- 2. Gamma ray dose from piping and/or
.equipment containing contaminated fluids which are recirculated inside Primary Containment (Direct Gamma Dose).
- 3. Beta ray dose emitted by airborne radioactive sources suspended in the
Rev. 2 IWASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTF21 ANALYSIS fI2 .'ROJECT t Page 5-3 drywell and wetwell inside Primary Containment (Airborne Beta Dose) . 4." Beta ray dose emitted by airborne radioactive sources inside Secondary Containment (Airborne Beta Dose). The initial phase of this analysis was concerned with the determination of radioactive source terms for the liquids and gases inside Primary Containment. The ORIGEN 2 Computer Code was used for this calculation. The fission products at the end of fuel life (maximum burnup at power level of 3481 tlWt for 1,000 Rays) was assumed to be available for release immediately following the acci-dent. The concentrations of noble gases, halogens, and other fission pro-ducts released to the gaseous and liquid sources were computed.
- 5. 3 Source Term The radiation level at a given location Development inside the Secondary Containment of WNP-2 For Secondary following an accident such as that described Containment in Section 3.1 is defined, by the following major source contributors:
- 1. Gamma ray dose from airborne radioactive source inside Secondary Containment (Airborne Gamma Dose) .
- 2. Gamma ray dose from radioactive sources suspended in the drywell and the wetwell inside Primary Containment (Containment Shine Dose) .
- 3. Gamma ray dose from piping and/or equip-ment containing contaminated fluids which are recirculated outside Primary Containment (Direct Gamma Dose).
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT /2 Page 5-4
- 4. 'eta ray dose emitted by airborne radioactive sources inside Secondary Containment (Airborne Beta Dose) .
The initial phase of this analysis was con-cerned with the definition of radioactive source terms for the liquid and gas con-taining systems. 'The OREGEN Computer Code was used for this calculation. The fission products at the end of fuel life (maximum burnup at power level of 3481 NNt for 1000 days) was assumed to be available for release immediately following the accident. The con-centrations of noble gases, halogens, and other fission products released to the gaseous and liquid sources were computed. fission product depletion and 'ubsequent daughter product generation were then calcu-lated for twenty time periods, covering a total, period of one year. A detailed description of the method of analysis, including the assumptions used, as well as results of the source terms, is found in Appendix B and Reference 5.5. 5.3 ~ 1 Source Term The radiation level at any given location Deve lopmen t for outside the Reactor Building of WiJP-2 1F.'/lN Equipmen t following the postulated LOCA as described in Ou tside the Section 3.1 is determined from the Following Reac tor Bu ild ing major source contributors:
- 1. Direct gamma dose from radioactive piping located inside the Reactor Building and attenuated through the walls of the Reac tor Build ing.
- 2. Gamma shine dose from airborne activity inside the Reactor Building.
IP WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT g2 Page 5-5
- 3. Gamma ray dose from airborne activity outside the Reactor Building.
A detailed description of the method of ana-lysis, including the assumptions used, as well as results of the source terms is found in Appendix H. 5.4 Parametric'tudies The purpose of the parametric study was to for Direct Piping identify the parameters which have a signifi-Dose in Secondary cant effect on the radiation dose rates Containment inside Secondary Containment. The computer code QAD-PSA was used to develop a correla-tion scheme Cor the significant parameters such that a simplified procedure for calcu-lating radiation dose rates for complex source and receptor geometries can he deve-loped. The dose rate at a target distance of 8 ft radially outwards Crom the centerline oC an 8-inch schedule 40 pipe, infinitely long (standard pipe) was first calculated and def ined as the standard dose rate. The results of this parametric study were then correlated as a set oC correction factors to the standard dose rate. A simplified proce-dure was developed to calculate the dose rates and cumulate doses Cor complicated source-target configurations by using these correction factors. The development oC these correction -factors and the result of the parametric study inside Secondary Containment is discussed in detail in Appendix B.
WASHINGTON PUBLIC POWER PLANTSHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 3'I2 page 5-6 5.5 Dose Rate and The. res ul ts o f the source te rm calculations Cumulative Dose and those of the parametric study were used Calculation to generate and cumulate doses for compli-Procedure cated source target conf igurations inside Secondary Containment. The following steps were taken to define the radiation service conditions for the pieces of safety-related equipment:
- l. Based on the accident scenarios, con-taminated systems, and assumptions defined in Section 3.0, the radioactive source terms for liquid-containing and gas-containing systems were developed.
?'adiation zones vere selected and the radiation zone boundaries were carefully defined based on shield wall locations, contaminated piping locations, and loca-tions of safety-related (1F/1N) equipment.
- 3. The radiation environment in each Secondary Containment zone { zone 'dose) was calc,ulated {see Appendiv B for detailed procedure) . A zone dose is the radiation dose (gamma) that bounds the magnitude of dose received by all the pieces of safety-related (1E/lM) equip-men t loca teel with in tha t zone.
- 4. The zone dose as calculated in used, as a first cut,"to qualifystepall-3 the was pieces of safety-relateR (lE/lM) equip-ment located within that zone.
- 5. For the pieces of safety-related (1F/1M) equipment that could not be qualified for
<ASHINGTON PUBLIC PONER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYST&f ANALYSIS PROJECT I'32 Page 5-7
~ the. conservative radiation environment calculated in step 3, the integrated dose f or that piece o f equipment was,. rede f ined based on a more realistic and refined approach.
5.5.1- Calculation of The time-dependent post-LOCA activity levels Airborne Gamma as calculated by the ORIGEN computer code Doses Inside were used as input for the calculation of the Secondary Con- airborne gamma dose rates and integrated tainment doses inside the cubicles in the Secondary Containment. The assumptions used in this analysis are as follows:
- 1. Activity that leaks into the Secondary Containment is homogenously mixed with the Secondary Containment atmosphere prior to its removal from the atmosphere through the Standby Gas Treatment System (SGTS).
- 2. The minimum SGTS flowrate of 1100 SCFN was assumed to be the flowrate'f the effluent air.
- 3. Air that leaks out of the Primary Containment flows directly and totally into the Secondary Containment. Bypass leakage was not considered.
4 . Geometric factors were used to convert the semi-infinite cloud gamma dose to a finite gamma dose.
- 5. Primary Containment leakage rate of 0.5%
volume/day was considered. Justifications of the above assumptions are stated in Appendix R. The equations that
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT f/2 Page 5-8 were used for 'the gamma dose calculations are described in Appendix B. 5.5.? Methodology of The source volume used for the beta dose ana-Beta Dose lysis in Secondary Containment is a sphere Analysis surrounded by a shell of sufficient thickness to stop all outside beta particles from entering the source volume. This spherical source volume is conservative for any genera-lized source volume shape (the dose at the center 'of the sphere is higher than the dose at any point of any generalized source shape of equal total volume) . The'ssumptions used in this analysis are as follows:
- l. Atmosphere inside the equipment casing is identical to the atmosphere in the Reactor Building.
- 2. Doses will be calculatecl using an air dose as suggested by NUREG-0588, Revision l.
- 3. The beta source term used was 100% of core noble gases and 25% of core halogens.
- 4. Daughter oroducts of the airborne noble gases and halogens are included in the calculation of the airborne dose.
- 5. The primary to secondary leak rate is
- 0. 5% of primary containment volume per day.
- 5. The SGT system operates at the minimum flow of 1100 scfm.
Kev ~ !v'ASHENGTON PUBLIC POWER PLANT SHIELDING MASHINGTON NUCLEAR SUPPLY SYSTEi~1 .ANALYSIS PROJECT g2 Page 5-9 !
- 7. Primary'o secondary leakage is homoge-neously mixed in the secondary contain-ment atmosphere.
- 8. No halogen plateout after release is assumed.
- 9. A spherical volume and equipment casing will be used.
The equations used for the beta dose calcula-tions are described in Appendix E. The beta dose to equipment is dependant on the internal volume size of the piece of equipment. The beta dose is determined, through the use o f an energy dependant geometry factor and a ratio of the internal equipment volume to an inf inite cloud. The results of these factors are shown in Figure 5.1, Thus, the beta dose contribut'o.; to equipment can be determined from Figure 5.1 once the internal air volume of a piece of equipment is known. 5 ~ 5 "3 Procedure For As discussed previously, the gamma radiation the Calculation ~ level at a given location inside the of Radiation Secondary Containment of Nllp-2 following a Zone Dose in Loss of Coolant Accident is determined for Secondary four types of radioactive source Containment distributions: o Fission products suspended in the atmosphere of: the Secondary Containment ( Airborne t,amma Dose) o Gamma irradiation f rom the Primary. Con tainmen t ( Sh ine Dose)
tcev. !WASHINGTON PUBLIC POWER PL'ANT SHIELDING 4/ASHINGTON NUCLEAR SUPPLY SYSTEM A lALYSIS PROJECT 82 5-10'age o Direct gamma irradiation from the radioactive fluid contained inside recir-
'ulating pipes (Direct Dose) o Airborne radioactive sources inside Secondary Containment (Airborne Beta Dose)
The dose contributed by each of these sources is determined by the location of the equip-ment, the time-dependent distribution of the source and the effects of shielding. A step-by-step procedure for calculating radioactive zone doses is shown in Appendix C. The methods presented in that procedure make it possible to calculate the worst case gamma dose from the above-mentioned source con t r ibu tors 'inside radiation zones o f the Secondary Containment. In general, this procedure for determining zone doses consists of a correction factor method of calculating direct dose rates. 4s d iscussed in Append ix B 'the correc t ion factor method for calculating dose rates pro-vides a convenient and fairly precise way of determining direct dose rates due to generic pipe segments. For radioactive fluid con-tained within components of geometry other than generic pipe segments, such as Residual Heat Removal (RHR) Heat Fxchangers, Standby Gas Treatment System (SGTS) "ilters, Hydrogen Recombiners, etc. special QAD-P5A computer modelling was performed to calculate the gamma dose contribution due to those systems. A brief description of the guidelines used'n
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~
SUPPLY SYSTEH ANALYSIS PROJECT I/2 I ~
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Page 5-ll modelling special components is found in Appendix B. 5 ' ' Calculation of As discussed in Appendices 9 and C, the Radiation Doses correction factor method for calculating Due to Special gamma dose rates and integrated doses is Systems and involved with the application of the dose Components In- correction factors (pipe diameter, pipe side Secondary length and radial distance correction Containment factors) to a standard dose rate curve. A standard dose is defined as the gamma radiation measured at a target distance of 8 feet and emitted by radioactive sources con-tained within the suppression pool liquid and recirculated within infinitely long 8-inch schedule 40 piping. The systems that contain such radioactive fluids are'he Reactor - Coolant system, High Pressure Core Spray, Low Pressure Core Spray, and Residual Heat Removal systems. Other systems which contain fluids of different source terms and dilu-tions are considered special sources. The systems that need to be considered for spe-cial sources are: o SGTS filters o CAC system o Main Steam System o Main Steam Isolation Valve Leakage Control System
- 5. 5. 4. l Source Term The assumptions for the calculations of source terms inside seconSary Containment for Assumptions In Secondary special source systems are listed as follows:
Containment
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page 5-12 SGTS .Filters
- 1. The SGTS Filters will be loaded by halo-gens .at the rate of 0.73% Primary Containment free volume per day. This consists of 0.5% per day of Primary Containment leakage and 0.23% per day of leakage due to the ifSIVLC system. No holdup of this activity in the Secondary Containment is assumed.
2 ~ 25% of the total core halogen inventory is assumed to be released in the drywell free volume. This halogen fraction is assumed to be composed of. 91% elemental, 4% organic and 5% particulate halogens.
- 3. The particulate halogens are assumed to be homogeneously distributed within the prefilters and the particulate filters, while the elemental and organic halogens are assumed to be homogeneously distri-buted within the charcoal "ilters.
Assumption 1 is consistent with the assump-tions used in the Accident Analysis (Ref. 5.6, and Section 3.1). Assumption 2 is the RC recommended assump-
~$
tion for the distribution of halogen inven-tory (Ref. 5.7) . Assumption 3 is necessary because the time dependent distribution of activity within a filter is unknown. The homogeneous assump-tion, therefore, is considered appropriate and conservative for zone dose assessment~
!WASHINGTON PUBLIC POWER PLANT SHIELDING HASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT fI2 Page 5-13 CAC System N ~ The function of the CAC system is to process the Primary Containment atmosphere to remove hydrogen af ter a LOCA accident. Therefore this system is assumed to be filled with gaseous source containing 25% halogens and 100% noble gases diluted with the Drywell free volume. Hain Steam System
.The main steam lines are located inside and outside the Priinary Containment; they include the Hain Steam lines in the steam tunnel and the RCIC turbine supply and exhaust lines.
The radioactive source term for this system is assumed to be composed of 100% noble gase: and 25% halogens, distributed throughout the Reactor Coolant System (RCS) steam space. Hain Steam Isolation Valve Leakage Contxol System The 41SIVLC system of NNP-2 is a vacuum-type syste)n which collects leakage between and downstream of the closed isolation valves and releases it to the atmosphere through the SGT system. Leakage through the valve stems
'maxiinum lea1;age of 11.5 SCFH as described in Ref. 5.8) is directed to a distribution header or low pressure manifold where clean
.air is brought in,to ciilute the contaminated steam before exhausting to the SGTS filter unit at a rated flow rate of 50 SCF'1. Thus the . ource term in the portion of piping system before the 'distribution header is con-servatively assumed to be the same as that of the Hain Steam system. For the portion of
WASHINGTON PUBLIC PLANT SHIELDING WASHI'NGTON NUCLEAR SYST&f POtJER'UPPLY ANALYSIS PROJECT g2 PaSe 5-14 the-piping af ter the header, credit is taken for the dilution by the clean air. Th'is assumption is consistent with that recom-mendeR in Reference 5.9. 5.5.4.? Secondary The corre'ction factor method is used for the Containment calculation of the direct dose contribution Analysis due to the piping systems describeR in Hethod Section 5.5.4.1, with the exception of the SGTS filter system. A description of the analysis of the SGTS f ilter is documenteR in Appendix D. Generic piping dose rate and integrateR dose (dose at a target distance of 8 Eeet away from the centerline of an infini-tely long 8 inch schedule 40 pipe) for each system were developed using .the source term assumptions discussed in section 5.5.4.1 and are shown in Appendix B. parametric studies were also performed to investigate the
~
variation of dose rates due to pipe diameter, pipe length and t'arget distance for pipe segments containing gaseous source terms. The gaseous source term correction factors derived as a result of this parametric study (described in Appendix B), together with the generic Rose rate curves generated Eor each system were useR to calculate the Rirect gamma dose contribution on a target. 5; 5. 4. 3 Calculation Table 5.1 is a sample list of generic mecha-o f Rad ia t ion nical equipment that are on the safety-Doses Inside re la teR equipment l. is t. For conservatism, Secondary the direct dose on the pieces of generic Containment mechanical equipment is assumed to be the on Generic fluid contact Rose. Figure 5.2 is an mechanical illustration of the point where the Rirect Equipment dose is calculateR on a piping segment.
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT iI2 Page5 15 The Secondary Containment source term assump-tions developed in section 5.5.4.1 are used Cor the calculation of radioactive source terms for different systems, and the fluid contact dose was calculated using QAD-P5A by following the guidelines set forth in, Appendix C. Figures 5.3 through 5.5 are 6 month integrated fluid contact doses versus pipe diameter. These curves are intended to give conser-vative, upper-bound direct gamma dose estima-tes for the qualification of the pieces of generic mechanical equipment and components in the various systems. In order to use these curves to calculate the direct doses on generic mechanical equipment, the following steps should be taken:
- 1. Identify the system on which the equip-ment or component is located.
- 2. Identify the diameter of the contaminated pipe on which the equipment is located.
- 3. The six-month integrated dose for that piece of equipment or component can be determined by reading the appropriate curve.
WASHINGTON PUBLIC PONER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 Page 5-16 Table 5.1 GENERIC MECHANICAL EQUIPMENT'alve packing Lubricants Seals Expansion, Joints Pressure RelieC Valve Flow Element Rupture Disc Gasket Material Conductivity Element Valve Strainers Steam Traps Filters (piping) Temperature Elements Tanks Moisture Separators Evaporator Heat Exchanger Air Washer (Scrubber) Pumps
EQUIPMENT DOSE REDUCTlQN- 6 MO. I.O 0.8 LU O 0.5'l a REACTO SVIL DI N 0.2 I-CO T A I N I4 E. N b 0 I
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2.2xlO 8 Figure 5.5 Six Month Integrated Fluid Contact Dose For pipes Containing Steam Source Term Diluted Within The RCS Steam Space (MS System, RCIC Steam System, And NSIVLC System Upstream Of The Header) I I ~ ~ 1.8x10 8 a zH 8 1.4x10 1.0x10 7I 6.0xlO 2.0x10 7t 12 '6 PIPE DIAMETER (IIV) 20
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR I SUPPL'Y SYSTEM ANALYSIS PROJECT f/2 Page 6-] 6.0 REStJLTS All LOCA scenarios and accident conditions that "c'ould result in a limiting radiation environment for all the safety-related equip-ment in the initial lE/1M list were reviewed and analyzed accordingly. Additional safety-related equipment subsequently identified will be evaluated with results presented in the Final Shielding Evaluation Report. For this report, no attempt was made to verify the completeness of the safety-related equipment list. The safety-related equipment list will contain all equipment required to "mitigate the consequences of an accident, bring the plant to a safe shutdown condition, and provide long-term cooling capability." This effort is currently underway and will be addressed in the Final Shielding Evaluation Report. Systems that could potentially contain radioactive material during and following the accident have been identified as listed in Section 3. 2. 1 and 3. 2. 2. The accident radiation doses indicated in Table 6.3 and Figures 6.1 through 6.8, generated as a result of this analysis, are intended solely for the purpose of the quali-f ication of safety-related equipment'. 5.1 Primary Con- Table 6.1 lists the 1E/1M (safety-related) tainment equipment identified inside Primary Cont-Radiation Results ainment. The integrated direct gamma dose (40 yrs and 6 month LOCA -direct gamma and airborne gamma) is currently being evaluated for safety-related components
- WASHINGTON PUBLIC PO'HER PLANT SHIELDING WASHINGTON NUCLEAR
, SUPPLY SYSTEi~f ANALYSIS PROJECT ~2 Page 6-2 insicle Primary Containment. The 40 year integrated gamma Roses due to normal- operation are taken from Reference 6.1. The direct gamma dose contribution insiRe Primary Containment is currently being evaluated.
Airborne beta dose inside Primary Containment was determineR and is Rescribecl in detail in Appendix G. The total beta dose contribu-tion, clue to Primary Containment atmosphere was calculated to he 3.4 x 10- Rads. The beta dose contribution insiRe Primary Containment due to plate-out of 25% halogens corresponds to a total dose of -2 3 x ~ 10~ Rads. Thus, the total beta dose contri-butions reported above should be used in con-, junction with total integrateR gamma radiation results to be presented in Table 6.1 for equipmen" qualification purposes. The radiation levels for Table 6.1 will be determined when the shadow shielRing analy-sis and consideration of drywell plateout activities insiRe Primary Containment have been completed. These additional analyses e f forts will be reportecl in the Final
~
Sh ield ing Evaluat ion Report.
Rev. 1 I WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT II2 Page ' 3 6.2 Secondary Table 6.2 lists the 1E/lt< (safety-rclatec3) Containment equipment racliation zones and elevations Racliation inside Secondary Containment. Figures 6.1 Resul ts through 6.8 present the radiation "one maps the Secondary Containment,- and the "Horst. Target Integrated Direct Gamma Dose"" asso-ciatecl with all the 1E/1N equipment listed for each zone. The integratecl direct gamma dose ( 40 yrs and 6 month LOCA direct gamma, gamma shine, and airborne gamma) is evaluated Cor the worst target in each zone and is used Cor qualification (in conjunction with air-borne beta doses) oC all the other lF/1H equ i pmen t in that 'zone. The 40 year integrated gamma doses (Figures 6.9 through 6.18) are taken Crom References 6.1 and 6.2. The gamma shine dose contribution outside Primary Containment clue to sources insicle the Primary containment was investigated. No safety-related equipment was located in the clirect shine path through'he penetrations and the shine dose was much lower than the zone Rose (less than 5% in all cases) . There Core, i t was conc lucled tha t the Pr imary Containment shine dose contribution is negli-gible. For the evaluated equipment, this "negl igible shine dose" was less than 10t) rac3s Cor the 6 month LOCA integrated dose. Airborne beta doses outside containment were evaluatecl per the methodology clescribecl in Section 5. 5. 2. The resul tant curve procluced from the beta analysis insicle Secondary Contanment is sho~n in Section 5 (Figure 5.1) This curve will be used to ;letermine the beta dose Cor equipment qualification purposes to a particular piece of equipment insicle Secondary Containment once the internal air volume o the c.qu i pmen t is known. C
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT (/2 Page 6-4 6.3 Radiation Results Figures 6.19*through 6.24 present the Outside the radiation "one maps for 1F/lM Reactor Building (safety-related) equipment located outside the Reactor Building. The doses indicated on each Figure are also the six month LOCA integrated gamma doses to be used fo' lE/1M ( sa f e tv-re la ted) equipment qual i f ica-tion purpose". Table 6.3 also oresents a summary of the six month LOCA integrated gamma doses on all 1F/1M equipment located outside the Reactor Building. Radiation levels of vital areas and access routes were determined at selected location outside the Reac tor Bu ilail ing due to rad ioac-tive sources inside the Reactor Building and release of radiation activity from the Reactor Building elevated vent. The vital areas and access routes analysed are con-sistent with those discussed'n NURFG-0737, I tern II. B. 2. The rad ia t ion levels cle ter~ ines for the vital areas and access routes iden-tif ied in Figures 6.19 through 6.24 are sum-marized in Table 6.4. All of the vital areas and access routes have radiation levels less than the guidelines presented in NUREG-0737 except for the security guardhouse and the auxiliary security center. Security personnel will be relocated to the Technical Support Center per the NNP-2 Emergency Preparedness Plan if guardhouse radiation levels approach tho'se radiation guidelines presented in NtJRFG-0737 during the post-LOCA situation. The analysis completed for vital areas and access routes assumed there would be no
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT /I2 Page 6-5 access to equipment or areas located within the Reactor Building during the post-LOCA scenario.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT f!2 Page 6-6 I a e, WNP-2 lE/1M Primary Containment Equipment List of Total Integrated Dose (40 Yr and LOCA) Radiation Level Equ'ipmen t 40 yr + Direct Gamma Equipment Number Elevation Airborne Gamma (R) MS-V-1 574'-3" MS-V-2 574'-3" MS-V-5 578'-5/8" 505'-llew MS-V-16 502'-2 13/16" MS-V-22A 505'-ll 11/16" MS-V-22B MS-V-22C MS-V-22D, 505 'll 1/2" 1/2" 505'-ll ll/16" RHR-V-9 509'-7" RHR-V-50A 509'-1'/8" RHR-V-50B 509'-1 5/8" RHR-V-123A 509'-1 5/8", RHR-V-123B 509'-1 5/8" to be provided at a later date RRC-P-lA 507'-0" RRC-P-1B 506'-8" RRC-V-19 509'-10" RRC-V-23A 502'-6" RRC-V-23B 502'-6" RRC-V-60A 507'-0" RRC-V-60B 506'-8" RRC-V-67A 507'-0" RRC-V-67B 506'-8" RRC-V-85A Drywe ll RCIC-V-63 551'-4" RCIC-V-76 552'-4" RWCU-V-1 540'-0" RWCU-V-100 500'-3" RWCU-V-101 )14'-0" RWCU-,V-102 500'-3" RWCU-V-106, 500'-3" RCC-V-40 514 '-0" RFW-LMS-10A 512'-0" RFW-LMS-100 513'-0"
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANAL'YSIS PROJECT //2 Page 6-7 I Table 6.1 (co'ntinued) Radiation Level
,Equipment 40 yr + Direct Gamma Equipment Number Elevation Airborne Gamma (R)
MS-RV-1A 546'-5'46'-ll" AS-RV-lB MS-RV-1C 546'-ll'46'-5" NS-RU-lD MS-RU-2A, 546'-5" MS-RV-2B 546'-ll" MS-RU-2C 546'-ll" MS-RV-2D 546'-5" MS-RV-3A 546'-5" MS-RV-38 546'-ll" MS-RV-3C 546'-ll" MS-RV-3D 546'-5" MS-RV-4A = 546'-5" MS-RV-4B 546'-ll" to be provided MS-RV-4C 546'-ll" at a later date MS-RV-4D '546'-5" MS-RV-SB 546'-ll'46'-ll" MS-RV-5C MS-TE-4A 546'-5" MS- Tt";-4B 546'-5" MS-TE-4C 546'-5" MS-TE-4D 546'-6" MS-TE-4E 546'-5" tIS-TE-4F 546'-6" MS-TE-4G 546'-6" MS-TE-4H 546'-6" MS-TE-4J 546'-4" MS-TE-4K 546'-4" AS-TE-4L 546'-5" MS-TE-4M 546'-6" MS-TE-4N 546'-7" MS-TE-4P 546'-6" MS-TE-4R 546'-6" MS-TE-4S 546'-6" MS-TE-4(J 546'-7" MS-TE-4V 546 '-5"
Rev. 2 i
~ ~ &WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT iI2 page 6-8 ~
Table 6.1 (continued) Radiation Level Equipment 40 yr + Direct Gamma Equipment Number Elevation Airborne Gamma (R) MS-RPV-3 Drywell CMS-TE-21 515'-8 1/2" CMS-TE-22 515'-8 1/2" CMS-TE-23 515'-8 l/2" CNS-TE-41 450I-8" CNS-TE-42 492'-8" CNS-TE-43 450'-8" CNS-TE-44 Drywe 1 1 LPCS-ZNS-6 547'-3" SPTM-TE-lA 465'-5" SPTt4-TE-1B 465'-5" SPTN-TE-2A 465i-5 S PTtC-TE-.2B 465'-5" S PTN- TE-3A 465'-5" S PTN-TE-3B 465'-5" to be provided SPTN-~E-4A 465'-5 at a later date SPTN-TE-4B 465I-5 I SPTN-TE-5A 465'-5" SPTM-TE-5B 465 I 5II b PT'4- TF-6A 465 I 5 SPTN-TF.-6B 465'-5" SPTM-TE-7A 465'-5" SPTN-TE-7B 465'-5" SPTM-TE-SA 465'-5" SPTth-TE-SB 465'-5 SPTM-TE-9 447'-10" SPTM-TE-10 447'-10" SPT14-TE-11 447'-10" S PTN-TE-12 447'-10" SPTN-TE-13 447'-10" SPTN-TE-14 447'-10" S PTM-TE-15 447'-10" SPTN-TE-16 447'-10"
- v'ASHIi~GTON PUBLIC POWER PLAilT SHIELDING WASHINGTON .lUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT '!2
?age 6-9 Table 6.1 (continued) ~
Radiation Level Equ ipmen t 40 .yr + Direct Gamma Equipment Number El e va tion Airborne Gamma (R) L D-V- 5A 5081-7'01'-0" LD-V-5AA LD-V-5B 508'-7" f D-V-5BB 500'-10" LD-V-5C 508'-7" LD-V-5CC 546'-7" LD-V-5D S08'-7" LD-V-5DD S79'-5 1/2" LD-V-5E 503'-7" LD-V-SEFi 503'-2" LD-V-5F LD-V-5G SOS 507'-1"
'4 1/2" f D-V-5H 504'-5" LD-V-SL 507'-4"
'-V-SN 507'-5" to he prov ided LD-V-SW 508'-3" at a later date LD-V-SQ 556'-9" LD-V-5.. '61'-ll" f D-V-SS SS6'-9" f D-V-5T 509'-2" f D-~I-SiJ 506'-2" LD-V-5V 546'-8" LD-V-SN 548'-0" LD-V-SX 539'-7" LD-V-5Y "
501'-3" LD-V-SZ '14'-ll" LD-TE-16C1 549'-0" i D- TE-16C3 549'-0" LD- TE-16C4 549'-0" LD-TE-16C5 549'-0" LD-TE-16C6 549'-0" f D-TE-16C7 549'-0"
'- TE-16Dl 549'-0" LD- TE-16 El 849'-0" i D-TE-16E2 54~'-0" LD- TE-16Fl 549 '-0" LD-TE-16G1 549'-0"
Rev. 2
- v'ASHii~GTO'8 PUBLiC POWER PLAiNT SHIELDING WASHINGTON .IUCLEAR SUPPLY SYSTEil PARALYSIS PROJECT :f2 Page 6-10 Table 6.1 (continued)
Radxatzon Level Equipment 40 yr + Direct Gamma Equipment Number Elevation Airborne Gamma (R) RH R- LM S- 1 1 1 A 562'-ll" RHR-LMS-lllB 562'-ll" RHR- LMS-111C 562'-ll" RHR-LMS-112A 509'-2" RHR-LMS-112B 509'-2" RHR-LMS-113 509'-7" TRM-CONN-01 Drywell TRH-CONN-02 Drywell TRM-CONN-03 Drywe 11 TRM-CONN-04 Drywell to - ~ TRM-CONN-05 TRM-CONN-06 TRM-.'CONN- 07 Drywe 11 Drywe 1 1 Drywell at be a provided later date TRM-CONN-08 Drywell RRC-TE-23A Drywell RRC-TE-239 Drywell RRC-TE-,28A RRC-TE-28B Drywe Drywell ll RRC-. TE- 3 5A Drywe ll RRC-TE-35B Drywell
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page 6-11 Table 6.2 Individual Zone Sketches Safety-Related Equipment Location Elevations Zones 422 C DE I JLM. 441 Br CtDi FiGs It J ~ O 471 AgBrDgE/F fHgXgJ.. ~ ~ 480M ~ 501
- B F I F M 0 P...510S.
522 B g C g Dg F gG ) H g J g K g N g Og P ~ 548 B,CgEgFgGgHp JgKgLgMgNgPgQ ~ 572 BfC/DfFfH/IfLgN. 606 Ao
WASHINGTON,PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT /!2 PaS< 6-12 Table 6.3 NNP-2 1E/1N Equipment Vital Area List Outside 'the Reactor Building For Six iilonth Total Integrated Dose (LOCA) Vital Area Radiation Level ~* Description Direct Gamma Shine Gamma
+ Airborne Gamma ( Rad s )
Control Room (EL 503.) 0.58 Technical Support Center 0.58 Sample Area (EI 487) 10.2 Nitrogen Supp'y to ADS Accumulators (EL 437) 7.6 Standby Service Water Pump Valves (cooling ponds) 3.5 Remote Shutdown Room (EL 467) 7.6 Switchgear Room Nl (EL 467) 7.6 Switchgear Room 02 (EL 467) 7.6 Radwaste Control Room (EL 467) 7.6 Battery Racks, DC Battery Chargers and 2HCC's* (EL 467) 7.6 3 blCC's* and 3 s>>itchgears (EL 437) 7.6 DC Battery Charger and Rack ( EL 437) 7.6 Diesel Oil Tanks (EL 437) 7.6 Solid Radwaste Control Panel
'and Decontamination Station Control Panel (EL 437) 7.6
~ l1CC's t1otor Control Centers
*~ Volume Correction Factors Cor a semi-infinite cloud were only applied to the Control Room and TSC due to the minimal exposure to the equipment. If the Volume Correction Factors wire to be appliec to all areas the integrated dose would be re reduced uce b y a minimum of five Cold.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTE1 ANALYSIS PROJECT I2 Page 6-13 Table 6.4 NNP-2 Vital Areas.and Access Route List of Radiation Exposure to Personnel During the Required Post-LOCA Operations Vita1 Area Radiation Exposure Gamma Be ta Description [thole Body Thyroid Skin (Rem) (Rem)(5) (Rads)
*Control Roon (EL 501) 0.58 0.58(>> 3.05
*Technical Support Center 0 '8 0.58(1) 3.05
*Security Center 5.9 26.4(2) 9,4
*Auxiliary Security Center 3 ' 15.3(2)- 5.5
"*Standby Service Hater
- Pump Valves (cooling ponds) 0.6 9(2) 0.9
*Sample Analysis Area (EOC) 0.0013
*"All infrequenty occupied 0.25(4) 3.2(2) 0.95 vital areas inside the Radwaste and Diesel Gene ra tor Bu ild ings
"*All access'routes inside 0. 25 (4) 3.2 (?) 0.95 the Radwaste and Diesel Generator Bu ild ings
**All access routes (3) outside the Radwaste and 2(2)
Diesel Generator Buildings
**Pos t Acc iden t Sample Area ( 487) 0.6 6.4(2) 2-0 Area of. continious occupancy Area occupied 0.5 hours at times after 1 hour 'into the LOCA (1) Assumes self-contained respiratory equipment ~as used by
.personnel during 0 - 3 hours post-IsOCA sit'uation.
(2) No respiratory equipment was assumed. (3) Extremely conservative analysis since the plane of airborne radioactivity cannot simultaneously cover all access routes.
0 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR
~ SUPPLY SYSTEM ANALYSIS PROJECT 82 Page 6-l4 Table 6;4 Continued (4) A Volume Correction Factor for the semi-infinite cloud .was included in the calculation (5) If Self Contained Respirat'ory Equipment (SCBA) is used the thyroid dose will essentially equal the whoe body dose.
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Rev. WASHINGTON PUBLIC POWER 'LANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS 82 .'ROJECT Page 7-1
7.0 REFERENCES
2.1 NUREG-0578, "TMI Lessons Learned Task Force Status Report and Short-Term Recommendations". 2.2 NUREG-0588 Rev. 1, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", 2.3 NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident". 2.4 Clarif ication letter to NUREG-0578, September 5, 1980. 2.5 NUREG-0737, "Clarification of TMI Action Plant Requirements". 2.6 IE Bulletin No. 79-01B, "Environmental Qualification of Class 1E Equipment". 2.7 Supplement No. 2 to IE Bulletin 79-01B, September 30, 1980. 5.1 ORNL, "ORIGEN2, Isotope Generation and Depletion Code Matrix Fxponential Method", ORNL Report No. CCC-371.
- 5. 2 J. F. Perkins, U. S. Army Missile Command, Redstone Arsenal, Alabama, Report No.
RR-TR-63-11 ( July, 1963) . 5.3 Oak Ridge National Laboratory, "Modifications of the Point-Kernel QAD-PSA", ORNL-4131, July 196'
- 5. 4 ORNL, "ORIGEN-79, Isotope Generation and Depletion Code Matrix Exponential Method", ORNL Report No. CCC-217.
Kev. 2 ~ WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR ! SUPPLY SYSTEH ANALYSIS PROJECT /I2 Page 7-2 5.'5 EDS Report No. 01-0740-1138 Revision 0, "Source Term Report", December 1980.
~ ~ \'4 5~6 WPPSS, WNP-2 FSAR Section 15.6.5.5.1.1.
5 ' Regulatory Guide 1.3 "Assumptions Used f or Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors". 5 ~ 8 WPPSS, WNP-2 FASR Sections 6.7.2 and" 6 ' ' 5 ' 'tandard Review Plan SRP 15.6.5 "Loss of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary". 5.10 ORNL, "QAD-CG, A Combinitorial Geometry Version of QAD-P5A, A Point-Kernal Code", ORNL Report No. CCC-307. 6 ~ 1, Letter BRWP-RO-81-288, dated July 29, 1981, '"Forty-Year Integrated Dose for Radiation Zones in the Reactor Building". 6 ~ 2 Let ter BRWP-RO-81-181, dated September 29, 1981, "Forty-Year Integrated Dose for Radwaste and Turbine Buildings". B-1. The same as 5.4. B-2. The same as 5.3. B-3. The same as 2.2. 8-4. Letter EDSWP-81-015, dated February 18, 1981, "SGTS Filter Modeling Assumption".
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT I/2 Page 7-3 Murphy and Campe "Nuclear power plants Control Room Ventilation System Design", 13th AEC Air Cleaning Conference. 8-6 ~ Letter EDSWP-80-031, "Shielding Design Input Data" . B-7. Theodore Rockwell II I, "Reactor Shielding Design Manual:, USAEC, TID-7004, March, 1956. 9-8. David A. Slade "Meteorology and Atomic Energy", USAEC, July, 1960. 8-9. Record of Conversation, D.A. Wert to J.A. Ogawa, dated 9/22/80, "Shielding Source Term". C-3.. The same as 2.2. C-2. WPPSS, WNP-2 CESAR Section 15.6.5.5.1.1. C-3. The same as 5.6. C-4. The same as 5.7. C-5. The same as 5.14. C-6. The same as 5.3. C-j. EDS Nuclear Inc. Calculation 0740-004-004, Rev. 0. C-8. EDS Nuclear Inc. Calculation 0740-004-006, Rev. 0. D-1. WPPSS, WNP-2 FSAR, Section 15.6.5.
Rev. 2
!WASHINGTON PUBLIC POWER "
PLANT SHIELDING WASHINGTON NUCLEAR i SUPPLY SYSTEM ANALYSIS 82 'ROJECT Page 7-4 D-2 ~ Standard Review Plan, SRP-6.5.3, "Fission Product Control Systems", USNRC, June 1975. D-3. Regulatory Guide 1.3, "Assumptions Used For Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors" USAEC, Revision 2, June 1974. D-4. The same as S.3. E-l. The same as B-6 E-2. The same as 2.2 E-3. - E-4 ~ The same as B-4 ORNL "ORIGEN-79, Isotope Generation and Depletion Code Matrix Exponential Method", ORNL Report No. CCC-2I7. David A. Slade "Meterology and Atomic Energy", USAEC, July, 1968.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR
,SUPPLY SYSTEM ANALYSIS PROJECT '/2 Page A-1 APPENDIX A A basic assumption to the plant Shielding Unisolated Leak Analysis is that the, Reactor Building path Report isolates such that there is no radiation leakage path to the outside. A leakage path investigation was done to verify the above assumption. While performing'his investigation, the total number of lines (69) penetrating the RB boundary, the associated system components and interface systems were reviewed.
The assumption eliminating the con-sideration of leakage is consistent with NUREG 0737, Clarification 2. This investiga-tion assumed that containment isolation
,occurred prior to the egress of highly radioactive fluid. Additionally, it assumed that all safety-related equipment was available, and that all safety systems were pressurized. Therefore, at any interface, such" as a heat exchanger, no potential leakage was considered if the non-radioactive system was at a higher pressure than the l
radioactive system. This investigation has not considered leakage from equipment seals, closed valves or pipe rupture, except in the evaluation of the equipment and floor drain systems. The systems considered are tabu-lated by drawing number in Table A-l.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT /j2 Page A-2 Table A-1 System Flow Diagrams Employed to perfo*rm the Review Drawing i)umber Revis ion Drawing Number Rev is ion M501 10 M536 12 M502 17 M537 25 M503 5 M538 9 M504 25 N539 28 M505 14'3 N540 15 N506 M541 13 N507 27 N54;? 4 H508 25 N543 17 M509 10 N544 10 M510 30 M545 15 M511 15 N546 10 i415 12 8 M547 9 M513 33 N548 14 M514 138 N549 14A M515 17C N550 9 t45 16 20 N551 8 M517 25 N552 12 M518 14 H553 10 H519 M520 18 15 N554 H555 ll 7 M521 20 N556 10 H522 6 N557 4 N523 29 H607 Sheet 1 7 H524 19 N607 Sheet 2 5 H525 19 N607 Sheet 3 3 H526 25 M527 18 M528 15 M529 21 N530 18 M531 24 M532 20 M533 Sheet 1 1 H533 Sheet 2 1 t4533 Sheet' 1 H534 16 H535 Sheet 1 26 M535 Sheet 2 21
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82
'Page B-1 APPENDIX B The ma j or tools used in the developmen t o f SOURCE TERM source terms and parametric studies inside DEVELOPMENT Secondary Containment were the ORIGEN and AND PARAMETRXC QAD-P5A computer codes. Descriptions of the STUDIES IN codes are in References B-1 and B-2. ORIGEN SECONDARY CON- was used to compute the activities and TAINMENT energies of fission products released from the reactor core. The output of ORIGEN (the time-dependent energies and activity of radioactive fission products following LOCA) was used as input to calculate the airborne, shine, and direct doses for standard geometries as well as the basis of direct dose parametric studies.
B. 1 Radioactive The ORIGEN Computer Code (Ref. B-l) was used Source Terms to calculate the radioactive source terms in Secondary inside Secondary Containment for liquid-Con ta inm en t containing and gas-containing systems. The fission products at the end of fuel life (maximum burnup at power level of 3481 MWt for 1000 days) were assumed to be available for release immediately following the acci-dent. The concentrations of noble gases, halogens, and other f ission products released to the gaseous and liquid sources were com-puted. Subsequent fission product decay and daughter product generation were then calculated for twenty time periods, covering a total period of one year. The assumptions used in determining the ini-tial distribution and leakage of radioac-tivity in the Primary Containment air and liquid space are as follows:
- l. 100% of the noble gases and 254 of the halogens are distributed homogeneously
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT f/2 Page B-2 within the Primary Containment free volume immediately following the postu-lated accident.
- 2. 50% of the halogens and 1'A of the remaining fission products contained in the core are mixed instantaneously and homogeneously with the Primary Containment liquid space. The Primary Containment liquid space is defined as the sum of the suppression pool liquid and the Reactor =Coolant System (RCS) 1 iqui<l.
- 3. The fission products available for release are defined as the total inven-tory generated in the equilibrium core after 1000 days at reactor power of 3481 MNt.
Assumptions 1 and 2 are RC recommended Ni assumptions for defining radioactivity release fractions for the qualification of safety-related equipment (Ref. 8-3) and are detailed in Reference 8-9. Assumption 3 represents the maximum burnup level in the core prior to radioactivity release and is conservative. Table B-l shows the gamma ray activity con-centration at selected time periods for the 1 iquid-containing system, while Table 8-2 shows the airborne gamma activity. The results of Table B-l and 8-2 were used as input in the dose parametric study, while the results of Table 8-2 were used in airborne
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 Page B-3 dose calculation. Due to rapid decay of the high-energy isotopes, the average gamma ray energy. for the gas-containing system varies from 0 8 Mev at the beginning of the accident
~
to 0.3 Mev at 1 year after the accident B.2 Airborne Dose The time-dependent post-I QCA activity levels in Secondary as calculated by the ORIGEN computer code Containment were used as input in the calculation of the airborne beta and gamma dose rates and integrated doses inside the cubicles in the Secondary Containment. The assumptions used in this analysis are as follows:
- 1. Activity that leaks into the Secondary Containment is homogeneously mixed with the Secondary Containment atmosphere prior to its exhaust from the-building with the Standby Gas Treatment System (SGTS) ~
- 2. The minimum SGTS flowrate of 1100 SCFM is assumed to be the flowrate of the effluent air.
- 3. Air that leaks out of the Primary Containment flows directly and totally into the Secondary Containment. Bypass leakage is not considered.
- 4. Geometric factors are used to convert the semi-infinite cloud gamma dose to a finite volume gamma dose.
- 5. Primary Containment activity leakage rate is 0.54/day.
PLANT SHIELDING WASHINGTON NUCLEAR WASHINGTON PUBLIC POMER ANALYSIS PROJECT /f2 SUPPLY SYSTEH Page B-4 Assumption NRC-.recommended l is consistent with the assumptions used for calcula-tion of doses inside Primary Containment (Ref. B-3) . Assumption 2 is" conservative because it represents the minimum flowrate of the SGT system .(with the SGT system running and flow-balancing dampers set at the minimum flowrate) (Ref. 8-4). Assumption 3 is conservative when considering dosage in the Secondary Containment, since it maximizes the buildup of rac3ioactivity in the Secondary Containment. Assumption 4 is based on the assumption used in Reference B-5, .and is based on an average gamma ray energy of 0.733 Hev. The effect of time dependence of average gamma ray energies has been proven to be negligible. Assumption 5 is consistent with the assump-tions established in Reference 0-6. A model of the primary and Secondary Containment atmosphere is shown in Figure B-l. The activity concentration of a certain isotope inside the containment is changing Rue,to the following three mechanisms:
- 1. Transport of activity, due to air leakage.
- 2. Depletion of activity due to radioactive decay.
- 3. Increases in activity levels due to daughter product generation from radioac-tive decays of other isotopes.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCI EAR SUPPLY SYSTEN ANALYSIS PROJECT f12 Page B-5 Because activity inside the containment is assumed to be homogeneously distributed, the rate of change of radioactivity concentration due to daughter product generation and radioactive decay. is independent of radioac-tivity 'transport. In other words, radioac-tivity would be transported at the same rate from the Primary Containment to Secondary Containment as if there were no decay. Therefore, the activity concentration inside the Primary and Secondary Containment can be expressed as Cl.(t) = F ~ R(t) Fl~(t) Cl ~ (0) (B-l) C2i(t) = FiR(t) F2<{t) Cli(0 (B-2) where:
'
R( xR t) Depletion factor of radioactivity concentration due to isotope decay and daughter product generation. Reduction factor of Primary Containment radioactivity due to transport of air through leakage and is constant for all isotopes. F V(t) Reduction factor of Secondary Containment radioactivity due to transport of air through leakage and is constant for all isotopes. (0) Airborne activity concentration in Primary Containment of a certain isotope at time (t=0).
WASHINGTON PUBLIC POWER PLANT SHIELDING MASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT i'I2 Page B-6 C2i (0) Airborne activity concentration in Secondary Containment of a certain isotope at time {t=o) k ORIAEN computer code calculates isotope decay and daughter product generation and is used to, compute FiR( t) Cli(0) . The method of
~
calculating Fl~(t) and F2>(t) is developed as follows: Ignoring activity decay and daughter product generation, the activity balance in Primary Containment is: (B-3) dt Initial conditions: at t = 0, C . = C {0) ( B-4) The solution of equation (9-3) becomes (0) - (Qll l) (B-5)
~p Total Activ ity balance in Secondary Containment:
dt 2i 2 l li .2 2i (R-6) Ini tial conditions: at t = 0, C 21 = 0
. (8-7)
The solution to equation {B-6) becomes:
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON iNUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page B-7 Q,/V, -(Q 1 /V1 )t ( B-8) 23. (Q2/V2-Ql Vl) exp exp ~2 2 j li Defiping: F ( t) = C2 ~
/Cl (0)
~
( t) Ql/V2 (Ql/Vl) t
(Q2/V2 Ql/Vl)
(Q2/U2) t exp (B-9) To calculate the airborne g'amma dose rate inside the Secondary Containment, the method as described in Reference 0-5 is used: D = N Z 3.= 1
'(3.'2' 0.25 E
- 23. (B-10)
X i = F {t).F. {t).C .(0) {3-11) D D Y P 00 (B-12) GF GF = 1173 VO. 338 { B-13)
N SUPPLY SYSTEM
'LANT WASHINGTON PUBLIC POWER SHIELDING ANALYSIS WASHINGTON NUCLEAR PROJECT r!2 Page B-8 Where:
D = Semi infinite gamma cloud dose rate ( Rads/sec) E . = Average gamma energy of the isotope (Mev) /d is X . = Activity concentration inside Secondary i Con ta Amen t ( Ci/m3 ) GF' Geometric factor used to scale the semi-infinite gamma cloud dose to a finite c lo ud dose . Dy = Volume Gamma of the finite cloud ( f t3) cloud dose rate in the center. of the compartment (rad/sec) F(t) = Dilution Factor Ql = Rate of air leakage from Primary to Secondary Containment (m3/sec) Time after accident Vl = Primary Containment air volume (m3) V2 = Secondary Containment air volume {m3) By taking F. Cl.(0) from ORIGEN output and using equation (B-5) ko calculate F2v(t); the total gamma dose in Secondary Containment can be computed by using equations (B-10) through (B-13).
WASHINGTON PUBLIC POWER PLANT SHIELDING VASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page B-9 The airborne semi-infinite cloud gamma dose rate is shown in Figure B-2. As can be observed f rom the f igures, 'the gamma dose- inside Secondary Containment reaches the peak at around three days after the accident, and decays slowly thereafter du< to the decay of radioactivity by radioactive decays and removal by the SGTS exhaust system. The geometric factor in equation (B-13) is developed in Reference B-5 for average gamma energies of Qe733 Mev. There has been a concern that this geometric factor may vary appreciably with time since the average gamma energy decreases with time due to the faster decay rate of the high energy isoto-pes. The average gamma energy during various time periods following the accident
'were computed and the results show that the average gamma energy varies from 0.3 Mev to 0.0 Mev. As discussed in Reference B-Q, the geometric factor changes by less than 5%
within that energy range. It is therefore concluded that the change in the geometric factor with time is negligible and that equation (B-13) can be used to calculate the finite cloud gamma dose inside the Secondary Containment. B-3 Parametric Studies The purpose of the parametric study was to for Direct Piping identify the parameters which have a signifi-Dose cant affect on the radiation dose rates. The computer code QAD-PSA was used to develop a correlation scheme for the significant para-meters such that a simplified procedure for calculating rad iation dose rates for complex source and receptor geometries can be dev'e-loped. The dose rate at a target distance'f 8 f t. radially outwards from the centerline
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY- SYSTEM ANALYSIS PROJECT //2 a Page B-10 of'an 9-inch Schedule 40 pipe, infinitely long (standard pipe) was first calculated and defineR as the standard dose rate. A para-metric'tudy was then performed to investi-gate the effects of the variation of parameters such as pipe length, pipe diameter, shield thickness, and target loca-tions on the dose rate. The results of this parametric study were then correlated as a set of correction factors to the standard dose rate. A simplified procedure was deve-loped to calculate the dose rates and cumu-late doses for complicated source-target configurations hy using these correction factors. B. 3. l Functional of The Gamma ray energy flux from a "S " to a detector point "P" (see line source Dependence figure B-3) Various is shown in Reference B-7 as: Parameters on Secondary BS a> Con ta inmen t L -b l Seco i~ g -b lSec0 Dose Ra tes 4 mr 0 exp d0- exp d6 0, (9) where: uncoil ided gramma ray flux (photons/cm~ sec) bl total attenuation through shield SL Source strength of line source (photons/cm sec)
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT jt'2 page B-ll B = Buildup factor Angle subtended by the length of the line=source,(see Figure B-3). The source strength "SL" is a function of the volume of liquid inside the pipe segments, which is also a function of the diameter and volume of the pipe. The angles " el" and " e2" are also functions of "a/r" and "b/r", respectively (see Figure R-18 for definition of "a/r" and "h/r" respectively). Therefore, the Eunctional dependence of gamma ray dose rates on the various parameters can be repre-sented by the following equation: (10) where: Base gamma ray flux for standard pipe FD = Pipe diameter Correction Factor FR = Radial Distance Correction Factor FL (a/r, hl) = Pipe Length Correction Factor B.3.2 Parametric The procedure Eor performing this parametric Study study is documented as follows: Procedures
- 1. Calculate the dose rate at a target distance of 8 ft. from the centerline of an 8-inch Schedule 40 pipe infini-tely long (standard pipe).
0 I
~~
h'ASHINGTON PUBLIC POWER PLANT SHIELDING MASHINGTOiV NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT //2 Page B-l2
- 2. P'erfo'rm parametric studies on the variation of dose rates with:
a ~ radial distance from the pipe centerline
- b. length of the pipe
- c. nominal pipe diameter t ime
- e. axial position along the pipe
- 3. Correlate the results of the parametric study by a set of geometric correction factors.
- 4. Develop a procedure for calculating dose rates by using the correction factors.
- 5. >1erify the correlation scheme by calcu-lating the dose rates at different target locations due to source piping of varied geometries through the use of QAD-PSA computer code, and compare the results to those obtained by using the procedure developed in step 4.
B.3 ' Direct Dose The Standard Pipe gamma dose rate and Parametric integrated dose curves for the clifferent Stubbly Results systems having different source terms (source Inside term assumptions defined in Section 5.5) are Secondary shown in Figures B-4 through B-ll. The Con ta inmen t various correction factors were calcu-lated by the following correlation.
l WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM AVALYSIS PROSPECT f/2 page 8-13
-- ~ Dose rate at a radial distance "r" from an inf initel long 8" sch 40 pipe FR(r) = Dose rate at a radial distance of 8 f t Crom an infinitely long 8" sch 40 pipe Dose rate at a radial distance of 8 Ct from an 8". sch 40 pipe of len th "2~"
F<(R)= Dose rate at a radial distance o 8 ft from an infinitely long 8" sch 40 pipe Dose rate at'a radial distance of 8 ft Crom an infinitely"d"long sch 40 oipe of nominal diameter FD(d)= Dose rate at a radial distance of 8 ft Crom an infinitely long 0" sch 40 pipe
'The above mentioned correction factors for liquid system source terms are shown in Figures 8-12, 8-13 and 8-14. The correction factor curves for gaseous source terms are shown in Figures R-15, 8-16 and R-17.
8.3.4 Correc t ion Using the parametric curves Crom Section Factor He thod 8.3.3, one obtains dose rates at varied o f Determining radial distances (between 2 f t to 40 f t) .Crom Direct Doses varied pipe diameters (between 2 in. to 24 in Secondary in.) of varied lengths (between 2 ft to Containment infinity) at any given time period within one year. The step-by-step procedure for calcu-lating direct dose is as follows:
- a. Identify a/r, b/r parameters and obtain pipe length correction factor FL from Figure 8-13 or 8-16, depending on the system being considered. (See Figure 8-18 for definition of "a/r" and II b/r II
)
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT (/2 Page B.-14
- b. Obtain the standard dose rate from the standard dose rate curve for time "t" desired ~
- c. Obtain the pipe diameter correction factor FD(d).
- d. Obtain radial distance correction fac-tor FR(r).
- e. The dose rate .for the given pipe segment can be computed by:
Dose Rate = (Standard Dose Rate) ~ FR HFDF> Table B-3 compares the results for dose rate of 17 different pipe geometry and target locations as calculated using the Correction Factor i1ethod to those calcu-lated by using the computer code QAD-P5A. It was observed that the biggest clif-ference in results between the two methods is less than 10%. -It is concluded that the Correction Factor Method is adequate for calculating direct dose
TABLE B-1 Gamma Energy Concentration in Liquid-Containing Systems Mean Gamma Energy Concentration (Mev/sec-cc) Energy (Mev-) 0 hr 1- hr 9 hrs 24 hrs 72 hrs 720 hrs 2880 hrs 4320 hrs 8760 hrs 0.30 3.76ES 2.52ES 2.58ES 2.30ES 1.58ES 1.54E7 9.26E5 7.52E5 4.80E5
,0. 63 4.34E9 2.84E9 7.71E8 4.29E8 ~
1.62E8 '.87E7 1.66E7 1.13E7 5.44E6. 1.10 2.43E9 1.48E9 5.23ES 1.45ES 1.87E7 1.81E6 6.53E5 5.66E5 4.09E5 1.55 2.40E9 5.66ES 1.55ES 5.50E7 2;82E7 6.44E6 3.22E5 2.49E5 1.80E5 1'. 99 2.43ES 1.91ES 7.33E7 1.67E7 1.76E6 5.04E5 1.89E5 1.61E5 1.04E5 2.38 2.27ES 1.74E7 1.74E6 1.39E6 1.29E6 .3.34E5 2.85E4 2.35E4 1.64E4 2.75 3.47ES 8.76E6 2.36E4 3.39E3 2.88E3 2.74E3 2.32E3 2.07E3 1.46E3 3.25 1.11E8 8.03E6 7.26E4 1.94E2 1.10E2 1.02E2 8.63E1 7.71El 5.43El
'.70 6.68E7 7.62E6 2.17E2 1.82E-2 1.82E-2 1.74E-2 l. 54E-2 1.43E-2 1.24E-2 4.22 1.13ES 5.71EO 1.31E-2 1.31E-2 1.30E-2 1.25E-2 1.10E-2 1.03E-2 8.90E-3 4.70 2.05E8 1.63EO .6.90E-3 6.88E-3 6.88E-3 6.59E-3 =
5.81E-3 5.43E-3 4.69E-3 5.25 1.69E6. 1.00EO 4.84E-3 4.84E-3 4.82E-3 4.63E-3 4.07E-3 3.82E-3 3.30E-3
TABLE B-2
. Total Gamma Activity of the Airborne Fission Products Mean Activity (Photons/sec)
Gamma Ray Energy hrs (mev) O.l hr 1 hr 9 hrs 24 hrs 72 hrs 720 hrs 2160 0.30 6.37E18 5.13E18 4.08E18 3.25E18 2.22E18 1.20E17 9.6E14 0.63 1.62E19 9;46E18 1.96E18 '9.38E17 2.49E17 9.26E15 1.83E14 1.10 3.86E18 2.69E18 8.08E17 2.09E17 1.45E16 1. 04E12 4. 79E10 1.55 2.42E18 1.81E18 1.70E17 2.97E16 4.36E14 1.20E13 4.65Ell 1.99 1.14E18 8.25E17 1.53E17 1.48E16 8.85E13 negl negl 2.38 1.04E18 9.43E17 7.75E16 1.89E15 1.12E12 negl negl 2.75 5.31E17 3.02E17 8.79E15 1.35E14 9.28ES negl 'egl 3.25 3.44E16 7.08E15 4.09E11 1.74ES negl negl negl 3.70 2.65E16 4.56E15 1.03E11 3.11E2 negl negl negl 4.22 4.50E16 1.11E11 negl negl- negl negl negl 4.70 8.57E15 1.92E10 negl negl negl negl negl 5.25 6.28E15 1.56E10 negl negl negl negl negl negl = negligible
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Figure B-l Hodel of the primary and Secondary Containment
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WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM AVALYSIS PROJECT P2 Page C-l APPENDIfi C: PROCEDURE FOR THE CAICULATION OF' ECONDARY CONTAINi~lENT RADIATION ZONE DOSES C.l Introduction Three Nile Island Lessons Learned Short Term Recommendations (NUREG-0578),
'ection '2-1.6.b, requires all nuclear power plant licensees to calculate post-LOCA accident environmental service conditions for all safety-related equip-ment. This procedure is specifically concerned with the definition of the Post-Accident radiological environments in the Secondary Containment of Washington Public Power Supply System Nuclear Project Unit 2 (iAJP-2), a Boiling tfater Reactor (BNR) .
The assumptions used in this procedure are based on a non-mechanistic LOCA sce-nario in which core damage is experienced at the beginning of the accident and
~
Primary Containment isolation is achieved prior to radiation transport. The radiation level at a given location inside the Secondary Containment of NNP-2 during and following such an accident is defined by the following major source contributors.
- l. Gamma ray dose from airborne radioactive sources inside Secondary Containment (Airborne Gamma Dose).
- 2. Gamma ray dose from radioactive sour-ces suspended in the drywell and the
I bfASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTE1 ANALYSIS PROJECT i/2 page C-2 wetwe3.1 inside Primary Containment (Containment Shine Dose).
- 3. Gamma ray dose from piping containing recirculating fluids (Direct Gamma Dose).
- 4. Beta ray dose from airborne radioac-tive sources inside Secondary Containment (Airborne Beta Dose). (See Appendix E)
The methods presented in this procedure make it ooss ible to ca lc ula te the wor s t case gamma ray dose due to the above-mentioned source contributors inside radiation zones (see Section C.2 for the definition of radiation zones) of the Secondary Containment of WNP-2. The radiation zone dose calculated by using this procedure is applicable solely for the purpose of environmental qualification of safety-related equipment. The following sections of this procedure describe the nomenclature, assumptions, and methods used in calculating radiation dose rates and cumulative doses. Section C.2 defines the terms and nomenclature found in this orocedure. The assumptions and approximations used in developing the dose rate calculation method, as well as limitations to this method, are stated in Section C.3. Section C.4 provides a step-by-step procedure for determining the worst case gamma dose rate and cumulative dose inside a particular radiation zone.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT 82 Page C-3 I The.calculation of airborne beta dose is defined in a separate calculation proce-dure and is not included in this proce-dure. (See Appendix E) C.2 Definition This section contains the definition of the o f Terms terms and symbols as used in this procedure: CIND: Cumulative Inte rated Dose ( Rads) Cumulative dose due to expo-sure to the decaying radioac-tive sources. D Airborne Gamma Dose Rate ( 4ds/hr ) Gamma Dose rate resulting from radioisotopes suspended in the atmosphere of the Secondary Containment. Direct Dose Rate (Shds/hr) Gamma dose rate resul ting from the radioactive fluid contained inside recir-culating pipes. D Shine Dose Rate
. ( Rods/hr) Gamma dose ra te in the Secondary Containment resulting from radioisoto-pes suspended and depo-sited inside Primary Containment.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM AVALYSIS PROJECT 82 j Page C 4 Total Gamma Dose Rate Gamma dose rate contri-buted by the sum of air-borne, direct, and shine components.
=D.'a +Dd+D d s GF: Geometric Factor Scaling factor used to convert semi-in f ini te airborne gamma dose to finite dose inside enclosed air soaces.
Da D GF GF = 1173 0.338 { Ref. C-8) Length Conversion Factor A scaling factor dependent upon the source pipe segment length and spatial orientation relative to a target {see Figure C-1 for the calculaton of this factor). F is used to convert the standard dose to the dose emitted by a pipe segmen,t of finite length.
0 i WASHINGTON PUBLIC POtNER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEii ANALYSIS PROJECT 82 Page C" 5 4w Jv w F D
)
Diameter Conversion Factor A scaling factor dependent upon the source pipe diameter. F is used to convert the standard dose to the close emitted by a pipe of specified d i arne te r. F Radial Distance Conversion Factor A scaling tactor dependent upon the ra;3 ia l d is tance of the targe t f rom the source oiping. F is used to convert the standard dose to the dose at a target of specified radial distance from the source piping. Total Dose Contribution Correction Factor A scaling factor used to con-vert the standard dose to the dose at a target from a pipe segment of soecified geometry and orientation. F = F ~ F ~ F F Sum of Dose Contribution s Correction Factor A scalang factor used to con-vert the standard dose to the radiation zone close due to all the significant pipe sources l in the ",.one.
!v'ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTOVi Vi UCLEAR SUPPLY SYSTEM ANALYSIS PROJECT r/2 Page C-6 n F = Z F 1=1 Radiation Zone: A region in the Secondary Containment defined to be such that gamma and beta radiation calculated in the zone bounds the magnitude of dose received by the pieces of safety-related equipment located in that zone. Source Term: The total radiated energy ( Y,B ) associatecl with a specifiecl quantity of radioactive material released froin the reactor as the result of a postulated accident. Special Sources: Radioactive source of such geometry or concentration that can-not be approximated by pipe segments of diameters 2 inches through 24 inches and containing contaminated liquid of acti-vity concentration established in Section C.3.1. This can he a heat exchanger, standby gas treatment filter, pump, etc. Standard Dose: Gamma dose at a target navzng a radial distance of 8 ft. from a source pipe centerline segment, infini-tely long, of nominal pipe diameter 8 inches, schedule 40 piping. Target: The point in space chosen to represent a location or object for which a dose rate and/or cumulative dose is being calculated.
I 4'ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTD1 AVALYSIS PROJECT ~'r'2 Page C-7 Worst Target: Location of the piece of radiation zone which will experience the highest gamma dose among all the pieces of, safety-related equipment in that zone. C.3 Assumptions, Approximations, and Limitations C.3. 1. Basic Assump- Gamma doses and 'dose rates inside radiation tions to be zones will be determined for three types of Used in the radioactive 'source distributions: Analysis Isotopes suspended in the atmosphere of the Secondary Containment ( a irborne gamma dose ) . Gamma irradiation f rom the primary Containment ( shine dose ) . Direct gamma irradiation from the II radioac tive fluid contained inside recirculating pipes (direct dose). The dose contributed by each of these sources is determined by the location of the equip-ment, the, time-dependent distribution of the source, and the <<ffects of shielding. The assumptions used in determining the ini-tial distribution and leakage of radioac-tivity in the primary Containment are as follows:
- 1. 100% of the noble gases and 25% of the halogens in the reac tor core will be
'.<ASHINGTON PUBLIC POWER PLANT SHIELDING h'ASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROSECT !/2 Page C-8 distributed homogeneously within the Primary Containment free volume imme-diately following the postulated accident.
- 2. 50% of the halogens and 1% of the remaining f ission products in the core will he mixed homogeneously with the Primary Containment liquid space instan-taneously. The Primary Containment liquid space is defined as the sum of the suppression pool liquid and the Reactor
'oolant System (RCS) liquid.
- 3. The core fission product source term is defined as the total product generated in the core after 1000 days at reactor power of 3481 MWt.
- 4. Primary Containment leakage of 0.50%
volume/day was considered.
,justification of Assumptions l. and 2 are NRC recommended As ump tions,' assumptions for def ining ra<3ioactivity release fractions for the qualif ication of safety-related equipment (Ref. C-1) and are con sistent with the accident analysis Reference C-2.
Assumption 3 represents the maximum hurnup
-
level in the core prior to radioactivity release and is conservative. Assumption 4 is consistent with the assump-tions e.,tablished in Reference C-3.
I
'.WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM'tf ANALYSIS PROJECT ~'/2 Page C-9 C. 3, 1. 1 Assumptions l. Activ'ity that leaks into the Secondary Used in the Containment is homogeneously mixed with Calculation the Secondary Containment atmosphere prior of Airborne to its removal from the atmosphere through Dose Rate the Standby Gas Treatment Syste:n (SGTS) .
Inside Secondary 2. The minimum SGTS flowrate of 1100 SCFtC is Containment assumed to be the flowrate of the ef fluent air.
- 3. Air that leaks out of the primary Containment flows directly into the Secondary Containment. Rypass leakage is not considered.
- 4. Geometric factors can be'sed to convert the semi-infinite cloud dose to a finite cloud dose.
Justification assumption 1 is consistent with the NRC o'ssumptions recommended assumptions used for calculation of doses inside Primary Containment (Ref. C-1) . Assumption 2 is conservative because represents the fninimum flowrate of the SGTS it system (Ref. C-4). Assumption 3 is conservative when considering dosage in the Secondary Containment, since it maximizes the buildup of radioactivity in the Secondary Containment. Assumption 4 is based on the results pre-sented in Reference C-5 and based on average gamma ray energy of 0.733 Nev. The effect of variation of this parameter due to differen-ces in gamma ray energies have been proven to
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTE1'1 ANALYSIS PROJECT i'r'2 Page C-10 be negligible (see Appendix 8 for justification). C. 3. 1. 2 As sump t ion s 1. 'Ao deple t ion of activity due to leakage is Used for the assumed. Calculation of Shine Dose 2. The airborne source is assumed to be uni-From Primary formly distributed in the drywell and in Containment the wetwell air space. The effect of the plate-out of iodine on the walls is not considered in Secondary Containment.
- 3. Activity in the wetwe11 water volume is assumed to be uniformly distributed in the sump wa te r.
- 4. The dosage at a point inside the region closest to the source is considered to be representative of the gamma dose in the region.
'Justification of: Assumption 1 maximizes the source activity Assumptions and is conservative.
Assumptions 2 and 3 are necessary because plate-out mechanisms are unknown. These assumptions are consistent with that con-sidered in Reference C-l. Assumption 4 maximizes the gamma ray dose at the region and is conservative.
tWASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT f!2 page C-ll C. 3. l. 3 Assumptions 1. No valve leakage is assumed. and Approxi-mations Used 2. Schedule 40 piping is assumed. in the Calcu-lation of 3. Heat exchangers and pumps can oe approxi-Direct Doses mated as pipe systems. The volume of. radioactive liquid in the component and its length are used to determine an equivalent volume of liquid. 4 Radioactive piping with diameters 2-1/2 inches or less was not modelled unless it was determined that such a pipe was a major source contributor. A major source contributor is defined as the only radioactive pipe in a target area or the radioactive pipe of closest proximity to the target. Justification of Assumption 1 is consistent with Reference 2.5, Assumptions Item IX 8.2,
~ Clarification (2) .
Assumotion 2 is a conservative simplif ication of the calculation process. Because the majority of the pipe segments considered are schedule 40 piping, and because increases in pipe schedule can only decrease the dose rates at the targets, this approximation is considered to be conservative and appropriate. Assumption 3 is a crude approximation for dose rates contributed by complex geometries. Because the pump and heat exchanger walls are thicker than the pipe walls of schedule 40 piping, this assumption is conservative. Assumption 4 is made because the dose contri-butions due to oipe segments of diameters
WASHINGTON PUBLIC POWER PLANT SHIELDING MASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT 82 Page C-12 less."than 2-1/2 inches are generally negliglible, unless they are major source contributors. C.3.2 Limitations The following limitations apply to the use of this orocedure for the calculation of radiation zone doses.
- 1. This procedure is only applicable to the calculation of radiation zone doses in the Secondary Containment of NNP-2.
- 2. The assumptions stated in Section C.3.1 are basic to the methodology used in this procedure. Changes in any of the assump-tions will affect the accuracy of tne results generated using this procedure.
- 3. The calculation of direct Roses using the generic curves in this procedure is limiteR to liquid sources in schedule 40 pipe segments or equivalent pipe segments with nominal pipe diameters ranging from 2 inches to 24 inches. Any deviation from these pipe geometries should be modeled as special cases. Note:
Schedule 40 piping is used because the majority of the oipe segments to be con-sidered are standard pipes (sch 40). Increases in the pipe schedule only introduces conservatism in the results.
- 4. The results for direct dose calculated using the generic curves were found to be accurate to within 10% (see Ref.C-G for error study).
I WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROSPECT '!2 page C-l3
- 5. Source piping located 40 f t or further from the target is generally an insigni-ficant close contributor. If its contri-bution is not found to be negligible, it should be considered as a special source.
C.4 Procedures For This proceclure describes the method The Calculation used in cal'culating the radiation doses of Secondary insicle radiation zones. Containment Radiation For equipment located inside a zone, the Zone Dose.. following three sources contribute to the total close level. Airborne dose (gamma and beta). Direct gamma dose from sources within pipes. Direct gamma shine close from drywell and wetwell, A step-by-step procedure is discussed in the following sections for the calculation of the maximum total gamma dose and dose rates for each zone. The calculation of the airborne beta dose and dose rate is discussed in a separate calculation and is not included in this procedure. (See Appendix E) C.4.l PROCEDURE A: The first. step in preparing a zone dose Radiation Zone calculation is to identify all the parameters Dose Calculation to be usecl. This inclucles the identification of all the potential sources and targets, both inside and outside the zone, and the identification of the climensions of the zone. Figure C-2 is a'tep-by-step flowchart of the
I '!WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTE>1 ANALYSIS PROJECT i72 Page C-14 calculation procedure. When'dentifying sources outside the zone, sources at the upper and lower elevations in the review pro-cess are included. home rough dose estimates are used to determine whether sources outside a zone are significant source contributors, For example, if. the closest pipe segment is a few feet away from a target, calculations will show that pipe segments outside the room thirty feet away are insignificant source contributors. Conversely, if a target is located near a wa31 with several pipes on the other side of a wall, then those pipes may become signif icant source contributors. C .4 .? PROCEDURE B: Because the semi-infinite airborne dose and Airborne Dose dose rates are already calculated and shown Calculation in in Figures C-6 and C-7, the only calculation Gecondary Con- involved in determining the airborne dose is tainment the conversion of the semi-inf inite cloud dose at Reactor Building concentrations to a finite cloud dose inside the cubicles'in which the radiation zones are defined. The first step in this calculation is to deter-mine the volume which defines the air space (or zone) of interest. An enclosed air space is defined as a cubicle, at least 95% shielded by concrete (or equivalent shielding) at least 1 foot thick. To convert a semi-inf.inite cloud dose (calculated in Ref C-7) to a finite cloud dose, a geometric factor is used. (t) D (t) GF (4-1)
WASHINGTON PUBLIC POWER 'PLANT SHIEL'DING WASHINGTON NUCLEAR SUPPLY SYST&l ANALYSIS PROJECT 82 page C-15 where GF = 1173 0.338 (Ref. 'C-8)(4-2) GF = geometric factor (dimensionless) V = Volume of the enclosed air space (ft ) Similarly, IND (t) CIND (t) (4-3) GF Figure C-3 i a step-by-step flowchart of tne procedure for cal.culating airborne gamma doses. C. 4 ~ 3 PROCEDURE C: Containment shine doses are calculated Prxmary Con- using the QAD-P5A computer code. Guidelines Shine
'ainment for preparing input parameters are documented Dose Calculation in Proceclure E and Reference C-6. The modelling procedure and the accuracy of the results are highly dependent on the geometry to be modeled, specification of the source volume, and the selection of a buildup factor.
Figure C-4 is a step-by-step procedure for-calculating containment shine doses. C. 4. 4 PROCEDURE D: The'first step in the direct dose calculation Direct Dose (from Ref C-8) is the iclentification of the = Ca lcula t ion "worst" target. Normally, the worst target is the piece of equipment that is closest to the major, source piping and can be selected by inspection. 'owever, if situations arise such -that the ~orst case target cannot be choser by simple inspection, order-of-magnitude calculations are performed for each potential worst case target in the zone. These calcu-
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROZECT i/2 Page C-l6 ~ ! I lations.wre illustrated in Steps 3a through 3c of Figure C-5. The next step is to.identify specia'l sources. Special sources are defined as source geometries that cannot be represented by liquid pipe segments between 2 and 24 inches in diameter. Example special sources are: SGTS f ilters, RCIC steam pipe, turbines, and heat exchangers larger than 24 inches diameter. Other components such as pumps and small heat exchangers should be modelled as pioes. The pipe cross-sectional area is calculated by dividing the total fluid volume by the effective length of the component. The contribution due to sources with shield walls is investigated next. Figure C-13 is used for this evaluation. If these sources are determined to be significant contribu-tors, special QAD-P5A modeling procedures as described in Procedure E are followed. It is unlikely that all sources under con-sideration will contribute significantly to the dose at a specific target. If all source contributions were to be calculated, the time involved in performing the calculation would be unnecessarily long without making a substantial improvement in the accuracy of the resul ts. Hence, as the sources are being identified, good judgment is used to distinguish between sources which contribute signifi-cantly to the target dose and those sources which do not.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT i)2 page C-17
~4 ~ NIW An, insignif icant source is determined by comparing its dose contribution to the source making the largest dose contribution. The comparison is facilitated by arranging sour-ces in decreasing order of importance and assigning rank numbers to the sources. The largest dose contributor is given a ranking number of l. The largest dose contributor is determined by inspection of the sketches and drawings being used. The largest dose contributor is generally the longest segment with the large-t pipe diameter and the least amount of intervening shielding between the target and source. All sources which are in the radiation zone and have been assumed to be insignificant contributors are listed as such to indicate that those sources have been considered.
Equations Used in The following procedure is followed for the Calculation calculation of correction of. Dose Rates of Dose Rates factors of Dose Rates (Step 9 through Step 12 of Figure C-5):
- l. Identify from the radial distance of the pipe the target; read F from segment Figure C-ll.
If the target is in contact with the source piping, read F from Table C-1 and set F and F equal t8 l. (Note: dose rate Ys not 9 function of pipe length and radial distance). If the target 'is geometrically in line with the source pipe segment, as shown in configuration 3 of Figure C-l, set F =1 and read F and F from Figures C-14 and
NUCLEAR WASHINGTON PUBLIC POWER SUPPLY SYSTEM ANALYSIS'ASHINGTON PLANT SHIELDING PROJECT I/2 Page C-18 C-15, respectively. (Note: F is defined "here because dose rate is not kensi'tive to pipe length variation.)
- 2. Identify the pipe diameter; read F from Figure C-10.
- 3. Determine F from Figure C-12; use equations ih Figure C-1 to calculate this factor.
- 4. The total dose contribution factor for a given pipe segment (I) is given as Ft(I) = FD(I) . FR(I) . FL(I)
- 5. When all the significant contributions have been calculated, sum the total dose contribution factors.
n F = E F ( I) n=l
- 6. To determine if =a source is negligible, the following test should be'erformed:
When N source segments are being con-sidered and the dose contribution of ranking I is less than 1/10 of tne dose rate calculated from the largest source divided by (N-I), the sources remaining should not contribute more than 10% to the total source contribution. This level of accuracy should be adequate for most calculations.
I
'NASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEi~f 'NALYSIS PROJECT i/2 Page C-19 The total integrated direct dose and dose rate can be calculatecl.
D (t) = D (t) ~
""- + D (t) (Special Sources)
CIND (t) = CIND Do (t) . F s
+ D (t,)
(Special Sources) where D Do (t) and CINDDo (t) are dose rates
.and cumulative doses Cor standar.l pipe, segments and are found on Figures C-8 and C-9.
.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHLtGTON NUCLEAR SUPPLY, SYSTD1 ANALYSIS PROJECT i72 Page C-20 Table C-1 Diameter Correction Factor (FD) for Targets in Contact with the Source Piping
-Nominal Pipe Diameter Pipe Diameter Correction Factor (in) (FD) 18.4 24 ~ 4 54 ~ 6 33 F 3 35.3 12 35 '
14 35 ' 33 ' 32 ' 24 29 '
CONFIGURATION 1 SOURCE PIPING Lr)-~( Lr) 2a 2b F ( L 2
<<TARGET CONFIGURATION 2 SOURCE PIPING I
I I 2a )+F I I
~( 21
( ) I L 2 I I
+~TARGET CONFIGURATION 3
~TARGET PIPING 0 SOURCE F = 1 Figure C-l Calculation of Length Correction Factor
START STEP 1 IDEHTIFY DIHENSIONS OF lONE LOCATE TARGKTS IHSIDE STEP 2 lOHE ON THE lOHE SKETCH TDENTIFY ALL DOSK CONTRIBUTING SOURCES INSIDE STEP 3 lONE AND LOCATE ON lONE SKETCH IDEHTIFY ALL POSSIBLE DOSE STEP II CONTRIBUTING SOURCES OUTSIDE lONE CALCULATE AIRBORNK GAHPIA STEP 5 DOSK RATK AND INTEGRATED DOSK (REFER TO PROCEDURE B CALCULATK SHIHE GA.'VIA DOSK STEP 6 RATE AHD IHTKGRATED DOSE (REFER TO PROCEDURE C) CALCULA TK DIRECT DOSE STEP 7 CA.'V1A DOSE RATE ANO INTEGRATED DOSE (PROC,O) CALCULATE TOTAL G~~ DOSE Ar D DOSE RATE STEP 8 DC ~ Da + Dd + De ClhDa ~ CINDA i CINDd i ClNDs STEP 9 COHPLETE TABLE 1 PLOT TOTAL DOSE RATE AND STEP 10 INTEGRATED DOSE PROBLEM CO".PLK TED Figure (:-2 PROCEDURE A: PROCEDURE FOR CAI.CULATING RADIATION ZONE DOSES
FROW STEP 5 OF PROCEDURE A IDENTIFY SHIELD HALL STEP 1 THICKNESS AHD VENT OPEH INGS SURROUNDING ZONE IS CRITERIA FOR IDENTIFY DIMENSIONS OF THE HO STEP 2 AIR SPACK STEP 2A KNCLOSED ENCLOSED AIR SPACE SATISFIED? IN HHICH THE ZONE IS LOCATED YES CALCULATE VOLUHE OF STEP 3 ENCLOSED AIR SPACE CALCULATE GEOHETR IC FACTOR STEP 4 GP ~ 1173 88 USING FIGURES (6) AHD (7) CALCULATE AIRBORNE GAHNA DOSE RATE AHD INTEGRATED DOSE STEP 5 D (t) ~Dt GF ClllD ( t) CI ND ~t RETURN TO STEP 6 STEP 5 OF PROCEDURE A Figure C-3 PROCEDURE B - PROCEDURE FOR CALCULATING AIRBORNE GANA DOSE RATE AND INTEGRATED DOSES
FROM STEP 6 OF PROCEDURE A SELECT POINT IHSIDE ZONE STEP 1 WHICH IS EXPECTED TO HAVE THE HIGHEST SHINE DOSE MODFL THE PRIMARY CONTAINMEHTg P ER T I HEN T REACTOR BU I LD I NG STEP 2 HALLS AND FLOORS ACCORDING TO THE SCHEMES ESTABI.ISHED P O R D I SCR ET I ZE 1'HE AIRBORNE AHD STEP 3 SUMP SOURCES INSIDE CONTAINMENT PEFORM SENSITIVITY STUDY OH SPEC I F ICATION OF VOLUME STEP II SOURCE POINTS UNTII. RESULTS CONVERGE TO HITHIH 5X RUN GAD- P5A TO ESTABLISH STEP 5 TME TIME DEPENDEHT SHINE DOSE RETURH TO STEP 6 OF PROCEDURE A Figure C-4 Procedure C: Procedure for the Calculation of. Containment Shine Dose
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CALCULAT( r <I) r <>> a Aao TL<>> tr Vslhs TASLC 2 STEP 9 Aao r leva(s (10), Ill) Aho (12) (aon) CALCVLAT( TOTAL Dost racToa Il'l cch'Ia<tvfloa coat<<flow I A<II.ql STEP 10 SVI Vr ALI Doll Coartl ~ IIT<oa coat<<~tea racloas STEP 11 ra ra ~ rs<II 1<<>>(o.s~s trT I~ I ~ I a I STKP 12 Its CALCULAfl \la< Nr<VNVT Dl ~ (CT D05t I ~ I( AAD STEP 1) lallsaaf<D Dos( sr s<rarlho hof(' ~ raastT 15 Ia cowfacl wish rhl sovac(, ALL DOSt CohlalsV'Iloa
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ll'1';i t" le 0 10 10 10 TINE AFTER ACCIDENT I)IRSI FigUre C-9 Gamma Integrated Dose at a Target 8 Feet Away From Standard Pipe
3.2 3.0 2.8 2.6 2.4 2.2 2.0 1.8 1.6 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 6 8 10 12 14 16 18 20 22 24 Pipe Diameter (Inches) Figure C-10 Pipe Diameter Correction Factor
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Figure C-12 Pipe Length Correction Factor
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I .WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SYST&i 'UPPLY ANALYSIS PROJECT g2 Page D-1 APPENDIX D The SGTS filters are located in the Reactor Calculation of Building {elevation 572'0") of NÃP-2 ahd the Secondary function to process the radioactive gaseous Containment eCCluent from the primary and Secondary Radiation Doses Containment. In the event of a LOCA in the Due to the Standby Primary Containment, the SGT system vill be Gas Treatment System actuated, and the gaseous iodine, the methyl iodine as I-131 that leak out of the primary Containment, will be absorbed in the charcoal absorber filters while the particulates will be absorbed in the prefij.ters and HHPA filters. Depending on the radioactive source distribution and the Primary Containment leakage rate, the radioactive iodine con-centration in the filters will increase with time. The purpose of this study is to evaluate the time dependent gamma radiation level exposing, safety-related equipment located near the SGTS filters and in adjacent rooms folio<<ing a LOCA. The time-dependent buildup of activity in each of the filters is <<irst calculated. The time arid energy-dependent gamma activity levels on the SGTS filters is developed by a combination of computer runs and hand calcu-lations and is used as input to the QAD-P5A computer code to calculate the gamma radiation levels Cor the pieces of safety-related equipment located in the room. A discussion of the analysis is given as follows. D.l Description of Figure D-1 is a" drawing of the SGTS filter the SGTS Filters train. The SGT syst m consists of two Cully redundant filter trains, each oC which con-sists of the following components in series:
rASHINGTON PUBLIC POP/ER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT N2 page D 2
~I p a) A demister to eemove entrained water par-ticles in the incoming air stream.
b) Two banks of electrical coil heaters designed to limit the humidity of the incoming air to 704 at design flow during
. post-LOCA conditions.
c) A bank of prefiltees to eemove large par-ticulates Crom the air stream. d) A bank of high efficiency particulate air (HEPA) filtees to eemove 99.9% par-ticulates Crom the airsteeam. e) Two four inch deep banks of charcoal absorbers arranged as shown in Figure D-2, are designed to absorb the gaseous elemen-tal 'and organic halogens from tne airstream. 'he dimensions of the charcoal il f ters 'are shown in Figure D-3. f) A second bank of HEPA filters, identical to that described in d) above. The function of this second HEPA filter hank is to cap-ture charcoal dust which may escape from the charcoal filters. Both SGTS filter units are located on ele-vation 572 of the Reactor Building and are automatically actuated and become fully operational within 34 seconds in the event of any of the three isolation signals.
- a. High radiation in the Reactor Building ventilation exhaust duct.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT rl2 Page D-3
- b. High drywell pressure.
- c. Low water level in the reactor vessel.
D.2 Calculation of The analysis of the time-dependent= transport Time-Dependent of the radioactivity from the primary Filter Activity filters filter is thebasedacti-Containment to the SGTS and Concentration vity concentration on each on the following assumptions:
- 1. The SGTS filters are assumed to be loaded by halogens at the rate of 0.73% Pritnary Containment free volume per day. This is composed of 0.5% from Primary Containment leakage and 0.234 from MSIVLCS.
- 2. Straight exhaust through the filters, with no mixing or holdup. in the Secondary Containment atmosphere, is assumed.
- 3. The released halogen fraction is 25% of the core halogen inventory. This halogen fraction is assumed to be composed of 91%
elemental, 4% organic and 5% particulate halogens.
- 4. The particulate halogens will be homoge-nously distributed within the Prefilters and the HEPA filters, while the elemental and organic halogens will he homogeneously distributed within the two charcoal f ilters of the f ilter train.
- 5. MSIVLCS discharges directly to the inlet of the operating SQTS filter unit, Therefore, there is no secondary cloud dose associated with the MSIVICS d ischarge.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT >'/2 Page D-4
~ et -4 IO Assumption l is based on the primary Containment rated leakage flowrate and is consistent with the assumptions used in the radiological con'sequence analysis (Ref . D-l) .-
Assumption 2 is a NRC recommended assumption for the analysis for Fission product Control Systems (Ref. D-2). Assumption 3 is recommended by the NRC for use in radiological. consequence analysis (Ref. D-3) . Assumptions 4 and 5 ar. necessary because the time dependent absorption and leakage of halogens in the filters is unknown. The homogeneous assumption is considered-appropriate and conservative. The time and energy dependent activity il gamma concentration in the SGTS f ters was f irst investigated. As discus ed in Section 5.5.4, this analysis was performed by a combination of computer analysis and hand calculations. The activity concentration of. a halogen iso-tope inside a SGTS filter is changing with time due to the following three mechanisms:
- 1. Transport of activity from the primary Containment and deposition on the filters due to ai r leakage.
- 2. Depletion of activity due to radioactive decay.
- 3. Increases in activity levels due to daughter product generation from radioac-tive decay of other isotopes.
MASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS, PROJECT i(2 Page D-5 Because activity is assumed to be transported directly to the SGTS filters and that acti-vity inside the Primary Containment and inside a filter is assumed to be homoge-neously distributed, the rate of change of radioactivity concentration due to daughter product generation and radioactive decay is independent of radioactivity transport. In other words, radioactivity would be transported at the same rate from the Primary Containment to the SGTS f ters as no decay. il if Therefore, the activity con-there
'ere centration inside the fiLters can be expressed as:
- 11 = F R(t) FlV(t) Cl (0) (D-l) C (t) = F ~ R(t) F ~(t) Cl. (0) (D-2) where: Depletion factor of radioac-tivity concentration due to
'iR'"'1~( isotope decay and daughter pro-duct generation and. is indepen-dent of transport.
t) Red uc t ion fac tor of Primary Containment radioactivity due to transport of air through leakage and is constant for all isotopes. F (t) Buildup factor of radioactivity in the n-th SGTS filter due to Primary Containment leakage and is constant for all isotopes.
WASHINGTON PUBLIC POWER PLANT SHIELDING MASHINGTON NUCLEAR ! SUPPLY SYSTEH AVALYSIS PROJECT //2 Page D" 6 1 C lx (0)
. = airborne activity concentration in Primary Containment of a certain isotope at time (t=0).
ORIGEN computer code calculates isotope decay and daughter product generation and is used to compute F.RC1'.(0). 1R 11 The method of calcu-lating Fl<( t) and F3<( t) is developed as f o1lows Ignoring radioactivity decay and daughter oroduct generation, the activity balance in Primary Containment: d dt li 1 Ql li (D-3) Initial conditions: at t=0, Cli = C li(0) ( D-4) The solution of equation (D-3) becomes: Cli(t) = Cli(O)e -(Q/~ )t ( D-5) Total activity balance on SGTS filters ( Ignor ing daugh ter product genera tion and radioactivity decay): d (CN VN) ( D-6) Initial conditions: at t=0, C ni =0 (D-7)
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page D-7 The solution to e~guation (D-6) becomes (D-8) Defining: C P NU (t) = Ni
- c. (o)
IVl -(gl/V ) t nv (t) F = ' e (D-9) V n Where C li t)
. ( = Gamma activity concentration in primary Containment of the i-th energy level (photons/sec-cm ) .
C ni.(t) Gamma ac t iv i ty concen tra t ion of the i-th energy lev. 1 in n-th SGTS f ilteg (photons/sec-cm ). gl In Halogen fraction to be Gepo-s' d in the n-th SGTS f ilter. Air leakage rate from primary Containment (cm~/sec) . Time a f ter accident ( sec) . Uolume of Primary Containment (cm~).
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT I2 Page D-8 I V n Volume, o f n- th SGTS f i 1 ter segment (cm3). Subs ti tu ting equation (D-2) and defining ( D-9 ) into equation
-
Xli(t) = Fi~( t) Cl i(0) ( D-10) Equation (0-2) becomes: C Ni (t)
= N 1 N
1-exp 1 1 ( X li
.(t)
D-11) The ORIGFN computer code was used 'to calcu-late Xl.lx (t). The ORICFN run result is shown in Table D-l. D~ 3 Calculation of Af ter the activity concentration in each Activity Con- filter segment is determined, the gamma centration in radiation dose for safety-related equipment the SGTS Filters located in the SGTS f ilter room is determineR by the use of Computer Code QAD-P5A (Ref. 0-4) . The QAD-PSA modelling procedure as described in Appendix C is followed for this analysis. The following modelling assump-tions were used:
- l. Self shielding of the filters are conser-vatively neglected.
- 2. Shielding due to the sheet metal housing are conservatively neglected.
filter Assumption 1 is made because the density of the charcoal dust or the wire mesh (Prefilter il il and HEPA f tees) in the f ters is low.
I 'ASHINGTON PUBLIC, POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTB1 ANALYSIS . PROJECT f!2 Page D-9 tVeglec ting the se 1 f sh ie ld ing e ff ec t o f the filters will not add toomuch conservatism to the results. Assumption 2 is made because of computer code stability considerations. The shielding effect of the thin sheet metal filter housing is negl ig ible. Five representative targets wer, chosen for direct dose analysis in the SGTS filter room. The safety-related equipment chosen are:
- 1. SGT-TC-1A1 (TC-2-1) 2 ~ SGT-NO-3A1 3 ~ SGT- AO- 2A SGT- DV-182 5 ~ SGT- FT- lA1 These targets are chosen according to their proximity to the SGTS filters. The relative locations of the targets are shown in Figure D-l.
The time and energy dependent gamma ray acti-vity concentration as calculated using the method described in Section D.2 was used as input to the QAD-P5A model-described above. The dose rate results as an output of this analysis were integrated numerically to give time-dependent integrated dose. Table D-2 shows the direct gamma dose rate and integrated dose results of the five targets.
Table D-1 Total Gaaaaa Activity of the Released Airborne Halogens Hean Gaaaaa Activity (Photons/sec) Energy Hev 0 hr 0.0333 hr O.l hr 0.333 hr 1.0 hr 3.0 hr 9.0 hr 24 hr 72 hr 216 hr 1440 hr 4320 hr 8760 hr 0.30 1. 71E18 1. 55E18 1.42818 1.36E18 1.28E18 1.27E18 1.38E18 1-25E18 8.61E17 5.00E17 6.06E15 4.22E11 5.48E07 0.63 1 ~ 14E19 1 ~ 13E19 1.10E19 9-92E18 7.54E18 4.05E18 1.83E18 9 26E17 2.49E17 5.71E16 6.90E14 6.02E10 3 '3E10 1.10 3.59E18 3.16818 2.90ElS 2.69E18 2.30E18 1.62E18 8.08E17 2.09E17 1 ~ 45E16 1.24E14 2.58E07 5.21E06 4.42E05 1.55 2.56E18 9.15E17 8.50E17 7 '5E17 5.91E17 3-41E17 1.39E17 2-76E16 4.01E14 1 ~ 27E13 4.63E02 1.99 1.87E17 1.87E17 1.85E17 1.78E17 1 63E17 1-25E17 6.21E16 1.27E16 8.83E13 3.00E07 0 2.38 1 ~ 36E17 4.70E16 6.90E15 1 ~ 21E15 7 'OE14 1.81E14 2.87E13 6.98E11 4.82E06 1. 59E-9 0
- 2. 75 2.15E17 7.69E16 2.27E16 1.22E16 5.39E15 5.94E14 1.13E13 5.21E10 3.43E05 1.13E-10 0 3.25 2.86E16 1.05E16 1.42E15 1 ~ 20E13 9 '8812 1.04E12 4.61EOS 1.78EOO 3.70 2.97816 1.66816 1.19E16 8.62E15 3.59E15 2.62E14 1.03E11 3.11E02 F 22 4.16E16 9.52E15 4 '3E14 1.19E10 S.54E-4 4.70 7.29E16 1.67E16 8.09E14 2*OSE10 1.48E-3 5.25
Table D-2 Direct Gamma Dose Rate and Integrated Dose Results For Targets In 'ihe SGTS Room TARGET TARGET TARGET TARGET TARGET
'IC 2 - 1 HO-3A1 SGT-AO-2A SGT-DV-182 SGT-FT-lAl Dose Integrated Dose Integrated Dose Integrated Dose Integrated Dose Integrated Time Rate Dose Rate Dose Rate Dose Rate Dose Rate Dose hr Rad r Rads (Rad r) Rads Rad r (Rads Rad r Rads) Rad r Rads 0.0 0.1 6.40E3 3-27E2 3.25E2 1.6681 6.9681 3.57EO 1.33E1 6.82E-1 1.85E2 9.4580 1.0 4-62E4 2.5284 2.34E3 1.28E3 5.02E2 2.74E2 9.60E1 5.24E1 1.3383 7 2782 3.0, 8.47E4 1-5685 4.30E3 7.93E3 9.23E2 1.7083 1.76E2 3.25E2 2.45E3 4 '1E3 9.0 '.31E5 8.03ES 6.65E3 4.08E4 1.43E3 8.74E3 2.73E2 1.6783 3.78E3 2.32E4 24.0 1.67E5 3.03E6 8.46E3 1. 54E5 1.81E3 3 '0E4 3.47E2 6.32E3 4.8183 8.76E4 72.0 1.83E5 1.14E7 9.30E3 5. 81ES 1.99E3 1.24ES 3.81E2 2 38E4 3.30E5 5.2983'.84E3 216.0 2.37ES 3.94E7 1.20E4 2.00E6 2.58E3 4.29E5 4.93E2 8.20E4 1.14E6 120.0 1.17E5 1.29E8 5.95E3 6.53E6 1.27E3 1. 40E6 2.4482 2.68ES 3.3883 3.71E6 1440.0 1. 60E4 1.7688 8.14E2 8 97E6 1.74E2 1.92E6 3.33El 3.67ES 4.6282 5.09E6
'2160. 0 1.7183 1.83E8 8.68E1 9.29E6 1.86E1 1.99E6 3.5680 2.81ES 4.9381 5.28E6 4320.0 2. 40EO 1. 85E8 1.22E-l 9.38E6 2.618-2 2.01E6 5.008-3 3.84ES 6.93E-2 5.33E6 8760 1.85E8 9.38E6 2.01E6 3.84E5 5.33E6-
(A)MOISTURE (C) PREFILTER (E) DEEP BED (F) DEEP BED EL IMINATOR CARBON CARBON FILTER FILTER (B) IYIAIN (D) HEPA FILTER (G) HEPA FILTER HEATER SGT- MO-3Al II i II I I 9.4x10 6 II 2 I II TC-2d I I II IhI I II (2 G T-TC.1 A1) p- SGT-AO-2A (2 I I )I 1.8x10G I I 2.0x 106 g LaLa ~ SGT- DV.1B2 ~ 41x105 APRT,OX. 40 46-3 SGT- FT-1A1 5.4x106 Figure D-l Standby Gas Treatment Filter
FOUR FILTERS 24 ll Note A II 24 Prefilter 8" HEPA Filter llew" Figure D-2 Geometry Of Prefilters And HEPA Filters
II 42 Not to Scale Figure D-3 Geometry of Charcoal Filters
Rev WASHINGTON, PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR r SUPPLY SYSTEM ANALYSIS PROJECT fl2 Page E-l. APPENDIX E: The source volume used for the beta dose ana-lysis in Secondary Containment is a sphere surrounded by a shell of sufficient thickness to stop all outside beta particles from entering the source volume. This spherical source volume is conservative for any genera-lized source volume shape (the dose at the center of the sphere is higher than the dose at any point of any generalized source of equal total volume). Justification of The assumptions used for the beta analysis in Assumptions Secondary Containment are presented in Section 5 ~ 5~2 ~ Assumption 1 is conservative because there will be some actual delay in transport of the gaseous fission products into the equipment. Assumptions 2 and 3 are based on NUREG-0588, Revision l. (Ref. E-l) .
"Assumption 4 is conservative and was required by the use of ORIGFN as a source code.
Assumption 5 is consistent with the assump-tions established in Reference E-2. Assumption 6 is conservative because it represents the minimum flowrate of the SGT system (with the SGT system running and the flow-balancing dampers set at the minimum flowrate) (Ref. E-3). Assumption 7 is consistent with the NRC-recommended assumptions used for the calculation of doses inside Primary Containment (Ref. E-4).
,WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT //2 Page E-2 Assumptions 8 and 9 are conservative. The first portion (Step l) of the calculation
"
is the determination of the air dose at the center of the spherical source as a function of the volume of the sphere. The second por-tion (Step 2) of the calculation is the determination of the equipment dose as a function of the air dose. Step l: The variation of beta dose rate from a typi-cal beta energy distribution in a one-dimensional absorbing medium can be approximated by the formula: D( X) = A exp (-v EX) (E-l) where D(X) is the dose at a point X A is a constant is the position in the material wE is a parameter that depends on beta energy. This relationship holds approximately up to the point where all beta particles are absorbed. This point is called the range of the beta particles. The range of a beta par-ticle is dependent upon the energy of the beta particle and is denoted rE. Both of the parameters u> and rE may be determined by empirical 7ormulas given below (based on the maximum energy of the beta
I WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT f/2 Page E-3 I particles) and approximately independent of the absorbing medium:
= 17p (E max) -1.14 pE ( E-2) and rF ='0.412/p) En for .t)1<F43 rE = (0.530E- .106)/p for 2.3<E<20 where p is material density (in g/c;q3)
E is energy of beta particle (in HeV) 0 PEis in cm rF is in cm n is defined as 1.265 .0954 LnF. The dose at a given point from a single beta source is now transformed into a dose from a uniform concentration of airborne sources which extend from radius zero to radius r. Th is re%'a t ion is found to be: D(r) = K(l-exp (pEr) ) ( E-5) where K is a constant This relationship is valid for r< rF. At rF, none of the beta particles originating beyond rF reach the target point. Hence, at this radius, an effective inf inite medium for
I 'NASH INGTON PUBLIC PONER PLANT SHIELDING NASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT >/2 Page E-4 airborne beta radiation has been reached. The dose* From a volume such that r>rE is equal to the dose from an infinite volume, which is denoted p. The dose as a function of volume radius is thus found to be given by the dual relation: (1-exp (-u r) ) D(r) = D (1-exp(-p Z 0<r< rE rE) ) This relation may be transformed to a func-tion of volume by noting that V = 4> r3/3. Since v< and rE vary for each beta energy, this equation cannot be solved analytically for the case of a mixture of many beta energies which is the case at hand. However, since D ~ for each beta energy is known (from the calculation of the semi-
~
infinite source), DE(v) for each beta energy at a given volume may be determined. All contributions to the total dose at a given volume are then added together. The largest volume to be considered in this calculation was determined by finding the volume of sphere of radius equal to the beta particle with the largest range. This led to a maximum volume of 1011 cc. A minimum volume of. 10 cc was arbitrarily chosen as the lower limit to the volumes to oe calculated. Volumes of 103, 105, 107, and 109 cc were also chosen. The dose rates and integrated six-month doses for volumes of these sizes were calculated, and the results are plotted in F'igures E-1 through E-6. These figures
I WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT >'/2 Page ~ are the reduction in ai r;lose f rom the semi-
~
infinite medium air:lose. Step 2: The absorbed beta dose within a physical t targe t is no always equal to the be ta close at a mathematical point in air at the surface of that piece of equipment; The beta ioniza-tion energy (dose) deposited on the surface of a solirl object is distributed in a thin I surface layer to a depth equal to the beta range in the material. The absorbed dose rate within the affected layer is equal to the average dose rate within the layer. The ratio of the absorbed dose to the air dose at the equipment surface is found to be:
"ErE D(abs)/D(0) = (1-e . )/r<p<
The value of this ratio is constant for a given energy group. For the beta energy distribution found'n this calculation, this ratio is found to lie within the values .137 to .204. Thus, air doses (Figures E-7 and E-8) are conservatively multiplied by the factor 0.204 to obtain heta doses to equip-ment. Airborne beta doses to equipment enclosed in small volumes are plotted as a function of volume size in Figure E-9. The dose at a given point from a single beta source is now transformed into a dose from a uniform concentration of airborne sources which extend from radius zero to radius r.
I WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT t2 Page E-6 This dose at a given point from a dif feren-tial source volume will be: dD(r) = C 4mr 2 Where C is a constant conversion factor to be def ined later. The first term indicates the dispersion of beta energy from a point source without absorption. The second term (D(r)/D(0) ) indicates the effect of distance for shielding. Integrating gives (for a point at the center of a spherical volume) with dV=r2drd ~: D(r) =- C Si 0-i
~4m,r 0-
-y
4 1 2 D(r) D(0) dAr 2 dr Substituting the formula for D(r) gives:
,4m,r
'
D(r) = C Si 0 ) dA O', A exp (-p r) 4m A exp (-p o) dr 4m r D(r) = C Si 0 > dQ 0~ exp (-uEr) dr 4m Solving gives: D(r) = C Si (1-exp (-pEr) ) D(r) = K (1-exp (->Er) ) ( =--10) where K is a constant and r is the extent of the source volume.
0 WASHINGTON, PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR ! SUPPLY SYSTEH ANALYSIS PROJECT 82 PaSe E 7 The absorbed beta dose within a physical target is not always equal to the beta dose at a mathematical point in air at the surface of that piece of equipment. The beta ioniza-tion energy (dose) deposited on the surface of a solid object is distributed in a thin surface layer to a depth equal to the beta range in the material. The absorbed dose rate within the affected layer is equal to the average dose rate within the layer. The ratio of the absorbed dose to the air dose at the equipment surface is derived as follows:- l = D.r.i1 D.= D. (abs)
- 3. 3.
(8-ll) where: E. = total energy deposited in the affected layer D. = average dose rate at a point in the a f fee ted Layer 1 beta range in the materiaL i = beta energy group D. 1 (abs) = absorbed dose within the target
I .'WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYST&f ANALYSIS PROJECT f12 Page E-8 We also know that: 1 D.(0) Vi (l-e 'i i) where: D. (0) = dose rate at the air-equipment boun-dary D.(0) = dose rate in air Substituting the first equation into the second gives: D.(abs) 1 r.i = (D.(0)/p 1
.) (1-exp(-p 1.r.))
1 i (E-13) and D.(abs)/D.(0) 1 1
= (1-exp(-u l.r.))/p 1
.r.
1 1
~ ~
Figure I.'.1 Six Honth Airborne Integrated Beta Dose rain ai W>te QQ U IS@) V I I IJII III
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infinite Air Dose ctor Build UI Vg
.3 Ci l0.0 tOQO (000.0 Wl~ (h~)
Infinite Ai:r Dose Containment Ol a0 'a Q 4J cd CC Q 4J Time (Hrs) ci>'<,'re E 9 Integrated Air Beta Doses Inside Containment
- i'ASHINGTON PUBLIC POWER PLANT SHIELDING VASHI~JGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page F-1 APPENDIX F- PRIMARY CONTAINMENT ANALYSES For safety-related equipment located inside the primary containment, the most severe post-accident radiation environ-ment is a postulated LOCA which comple-tely depressurizes the orimary system.
Such an event maximizes tne integrated dose due to the orimary containment atmosphere and olate out source terms. This is added to the 40 year operation dose. basic radiation source term, i.e., I'.l Core Source Term The core inventory at time of the accident and during the following six month time period is being calculated with ORIGEN
- 2. A maximum nuclide core inventory was calculated using 105% full power (3481 M<'Jt) and an operating time of 1,000 days.
Calculation of the basic source term will be done by generating the decay rate spectrum for. solids, halogens and noble gases as a function of time following the postulated lOCA ev. nt. The fine time mesh interval spacing shown in Table F.l was used. The decay rates, for each of seven energy grouos, were integrated and an average value determined. Zn addition, a beta decay rate, averaged over six months, was calculated for each source tyoe, i.e., solids, halogens, and noble gases. The results are presented in Table F.2. F.2 Radiation Dose It was assumed that the orimary radiation From Liquid Systems source in liquid systems, in the Due to Normal drywell, is due to the N-16 activity in Operation the reactor water coolant.
!WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT 82 l page F-2 From Table I I. 1-4 o f the PSAR, the N-16 activity (max. ) equals 40 Ci/gm. For a pipe segment of length L, the average source strength, s, is given by S = S. i ti-exp(-xaL)]/(xaL) where L = length of pipe section (ft) a coolant transit rate ( in pipe section) (Sec/Ft) N-16 decay constant = 0.09677 sec-1 Si = activity at entrance to pipe section with S. = S exp (-AZ aL) where S = initial N-16 activity E aL 0 = sum over all previous pipe sections. F. 2.1 Liquid Systems The liquid systems which contribute the Considered major radiation dose, inside primary containment, are RNCU and connecting RRC piping NS lines, and RHR lines during their period of operation. For the iUfCtJ, RRC and HS lines the ana-lysis neglected the contributions from
Rev. 2 WASHINGTON PUBLIC POWER PLANT 'SHIELDING WASHINGTON NUCLEAR: SUPPLY SYSTEM ANALYSIS PRO~ECT /t2
'Page F-3 f ission products, corrosion products and other sources-since these dose contribu-tions are small compared to the N-16 contribution, during normal operation.
For the RHR lines the radioactive source was the coolant source terms given in the FSAR Table 12.2-5. For conser-vati.sm, no decay was assumed beyond 4 hours after reactor shutdown. F. 2. 2 Resul ts Dose rates from various RWCU and RRC pipe segments are shown in Figures F-1 to F-14. Dose rates are presented for different pipe lengths, since different detector (equipment) locations will.
'"see" different pipe lengths.
Figures F-15 to F-17 present the dose rate from RHR piping. In calculating the integrated dose during the expected 40 year plant life, it is assumed that the RHR lines contribute 20% of the t ime.
'Figure F-18 presents the dose rate from various lengths of a 26'ain Steam line due to N-l'6 at a level of 50 pCi/gm. Due to the relatively short transit time, decay was neglected, resulting in a con-servative value.
F.2 ~ 3 Dose Contribution In addition to the dose rate from the due to Genera'1 various systems, liquid and'gaseous, Radiation En- containing radioactive sources there is vironment in a general radiation environment due to Primary Con- normal reactor operation and the tainment resultant neutron leakage from the reac-tor core. A fraction of the neutron
4 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR ~ SUPPLY SYSTEH ANALYSIS PROJECT P2. Page F-4 core leakage flux manages to penetrate the. reactor pxessure vessel, escaping into the reactor cavity. Some will tra-verse vertically while others will manage to penetrate the sacrificial shield wall.. In addition to the resultant neutxon dose rate, the neutrons interact with the material along their, path generating secondarv gamma rays. The ANISN one-dimensional discrete ordi-nates computer code was used.to calcu-late the transport of neutrons, generation of secondary gamma rays and their transport. The material and geometric configuration was that at core mid-plane. The total dose rate* just beyond the outer steel linex of the sacrificial shield wall is 79 Rads/hour; 5 Rads/hour due to neutrons and 74.Rads/hour due to capture, prompt and fission product gam-
~
mas. An estimate, based on geometric and material attenuation factors, was made** to determine the axial variation of the dose rate. The approximate dose rate reduction factor as a function of distance from core mid-plane is shown in Table F.3.
- B6 R Calc. 5. Ol. 20
"'* B& R Calc. 5. 01. 97A
V Rev. 2 I !lWASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR I t SUPPLY SYSTEM ANALYSIS PROJECT //2 Page F-5 F.3 Dose Rate due to This section presents the methodology LOCA (Depressurized of analysis used to calculate the six-Primary System) month integrated dose following" a postu-lated loss-of-coolant accident (LOCA) in which the primary system is depressurized. In this scenario the radiation from the core was assumed to be immediately released into the con-tainment atmosphere. A detail description of the source and geometric analytical model assumptions follows in the next sections. The sources considered, were containment atmosphere (gamma ray & beta) drywe ll wetwel1 air space plate-out (gamma ray & beta) drywell surfaces piping & equipment suppression pool water F.3.1 Source To insure the maximum radiation environ-Assumptions ment for all relevant positions within containment, a non-mechanistic accident scenario was postulated. Conservative source terms, as per the appropriate Reg. Guides & NUREG's, were
Kev ~ I ItdASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYS7S 82 'ROJECT Page F-6 chosen for each contributing system, independently. However, in calculating the radiation dose at a particular location, not necessary to assume it was that all source distribution assumptions were conser-vative simultaneously. Instead, a set of mutually compatible assumptions were used which gave the maximum dose for the location being considered. Thus the contribution of 50% core iodines in con-tainment atmosphere plus 50% core halo-gens in the suppression pool were not added when only a maximum grand total of 50% of the core halogens is released following a ?OCA, (TID-14844, NUREG-0737 a II.B.2.(4)a). F.3.2 Containment Atmosphere Source (1) A nonmechanistic instantaneous release from the core inventory into the Primary Containment free volume of 100% noble gases and 50% halogens was assumed.
-(2) Nhen considering drywell and wetwell atmosphere sources, it was assumed that the activity released (in the drywell) was uniformly mixed bet-ween drywell plus wetwell air volumes .
{3) It was assumed, per NUREG 0588 Rev 1, that the iodine released to con-tainment atmosphere begins dif-fusing to be deposited on
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROZECT 82 Page F-7 ! containment surfaces (plate-out
...source). This iodine plate-out source was used in containment atmosphere dose calculations.
(4), Assumed that 1$ remaining solids are immediately contained within the suppression pool water. (5) Containment atmosphere source con-tains 100% of the core noble gases and 50% of the core halogens. (5) The maximum nuclide core activity was calculated using an operating time (operation at full power) of 1,000 days. F.3.3 Suppression (1) Liquid in the suppression pool was Pool Source assumed to contain the following percents of core inventory 0% noble gases, 50% iodines, 1% solid fission products This assumption is from Reg. Guide 1.7 (Table 1.7 (Table D-1) based on TID-14844 (See NUREG-0588, Appendix D, S6). (2) Source of (1) was assumed to be released into suppression pool. (3) Uniform source distribution in suppression pool volume plus liquid volume of the reactor coolant system (RCS) was assumed.
MASHINGTON PUBLIC POMER PLANT SHIELDING MASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page F 8 4 F.3.4 Plate-Out Source (1) It is assumed that 95.5% of released
.halogens diffuses to the primary containment surfaces.
(2) Plate-outoccurs by diffusion in a ccordance w i th NUREG/CR-0009 . (3) All plate-out sources are retained, through the 6 month time period. (4) Source densitv (Nev/cm2-sec) was given by total source of (1) divided by plate-out surface, (5). (5) Plate-out occurs on containment surfaces within drywell (i.e., incl. top and bottom of drywell, innex surface of biowall, inner and Oi outex surfaces of sacrificial wall and outer surface of reactor pressure vessel insulation) uni-formly. (See Figure F-19). (6) Fox dose locations near oiping (within drywell), plate-out on piping surface was considered. Source strength, oer unit area, as given by (4) was used. F.3.5 Drywell The dose rate inside the drywell, due to Atmosphere containment atmosphere, was calculated Dose Analysis using the source terms and weighted by source type percentages, i.e., 100% noble gases, and 50% halogens. These were input to the QAD point-kernel com-puter code. The truncated cone geometry is being mocked up by ~ 420 787X7 cubicles with the dose contribution from each section calculated independently, as a function of source energy.
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 Page F-9 For calculational purposes the drywell volume within the sacrificial shield was considered to be a solid steel cylinder. Scattering out this volume was neglected, as was albedo from sources outside the region. Material inside the drywell is being considered. To approximate the shadow shielding ef feet of the steel in the drywell, its mass in each 7X7X7 cubical is homogenized with the air to giv an averaged density. F.3.6 - Plate-Out The iodines that plate-out are assumed Dose Analysis to be on drywell walls, dome, floor, the inside and outside surfaces of, the sacrificial shield wall, and the exter-nal surfaces of the RPU insulation. The plate-out dose to equipment is calcu-lated from the surface activities taking credit for the intervening shadow shielding as described in Section F.3.5 above. F. 3. 7 Ne twe l 1 Dose "Dose rate contributions are being calcu-Analysis lated for wetwell atmosphere and suppression pool water sources for detector locations inside the drywell and wetwell. The general configuration is shown in Figure F-22. F.4 Dose Due to This section presents the methodology Systems After a and results of analyses used to calcu-LOCA Condition late the six-month integrated dose following a postulated LOCA,"in which the primary system is depressurized, due
WASHINGTON'PUBL'IC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTBf ANALYSIS PROJECT //2 Page F-10 to various systems. The systems con-sidered are: RHR HPCS LPCS RCIC RWCU MS RRC The accident source term, averaged over six months, as given in section F.l, Table F.2, was input to the QAD point kernel code. A multitude of computer calculations were made for each radioac-tive system as a function of pipe diameter. Figures F-23 through F-28 show the dose rate vs. distance from pipe surfaces for various pipe diame-ters, for the RWCU, RRC and RHR systems. Figures F-29 through F.30 show results for the LPCI system. Results for the. LPCS/HPCS systems are presented in Figure F.3l through F-36. Figure F-37 shows results for the NS system (26") piping. F.5 Dose to Specific The total gamma dose, i.e., the integrated Safety-Related 40 year normal operation dose plus the Equipment integrated six month dose following a postulated LOCA will be determined for all of the 1E/ill (safety-related) equipment
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT f/2 i Page F-ll located within the primary Containment, The equipment list and their elevation inside primary Containment are presented in Table 6.1. The results will be shown in Table 6.3. for gamma ray dose contributions due to drywell atmosphere ( LOCA) plate-out (LOCA) wetwell atmosphere (LOCA) suppression pool water (LOCA) systems (LOCA) systems normal operation
WASHINGTON PUBLIC POWER PLANT SHIELDING MASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT '/2 Page F-12 } } TA'BLE" F. 1 TIME MESH SPACING USED IN SOURCE CALCULATION Time (min. )
- Time (min. ) Time (min.)
1 ~ 12 + 03 1.080 + 05 0 + 0]*A 1 ~ 28 + 03 1.296 + 05 4.0 + 01 1 ~ 44 + 03 1.512 + 05 6 ~ 0+01 2.16 + 03 1 ~ 728 + 05 8.0 + 01 2.88 + 03 2.160 + 05 1 ~ 0 + 02 3.60 + 03 2.592 + 05 1.2 + 02 4.32 + 03 1~8 + 02 5.04 + 03 2 ' + 02 5.76 + 03 3.0 + 02 1.44 + 04 3 ' + 02 2.88 + 04 4.2 + 02 4 '2 + 04 4.8 + 02 5.76 + 04 6.4 + 02 7.20 + 04 8.0 + 02 8.64 + 04 9.6 + 02 After LOCA
** read 2.0 + 01 as 2.0 x 101
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PRODECT 82 Page F-13 TABLE F.2 BETA AVERAGE DECAY RATE (MeV/sec), 0 6 MONTHS AFTER LOCA BETA DECAY RATE (MeV/sec) Solids 1.29 + 19 Halogens 1.84 + 17 Noble Gases 1.54 + 17
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR t SUPPLY SYSTEM ANALYSIS PROJECT f12 I page F-14 I l I TABLE F.3 APPROXIMATE DOSE RATE REDUCTION FACTOR VS. DISTANCE FROM CORE MID-PLANE Distance (ft.) Reduction Factor 0 a.n 0 ' 10 0 F 01 15 o.noa
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 N Page F-l5 TABLE F ~ 4: to be issued at a later date
Rev. 2 !<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT //2 Page F-16 I TABLE F.5: to be issUed at= a later date
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT P2 Page F-17 TABLE F.6: to he issued at a 1atex date
Rev. 2 .MASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT //2 Page F-lB TABLF. F . 7: to be issued at a later d'ate
Rev. 2
!<ASHINGTON PUBLIC PONER PLANT SHIELDING MASHINGTON NUCLEAR SUPPLY SYSTEM AVv. ALYSIS PROJECT 'I2 Page F-19
't C TABLE F. 8: to be issued a t a la ter c3a te
Rev. 2 !WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEt1 ANALYSIS PROJECT //2 Page I'-20 TABLF F.9: to be issued at a late>> date
Rev. 2 h
'WASHINGTON PUBLIC POMER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTVf ANALYSIS PROJECT 82 Page F-21 TABLE F.10:
to be issuecl at a later <Pate
Rev. 2 !WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT i/2 Page F-22 I TABLE F.ll: to be issued at a later date
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT f/2 I Page F-23 TABT E F. 12: to be issued at a later date
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTF2i ANALYSIS PROJECT (f2 Page F-24 l TABLFl F. l3: to be issued at a later date
KeY ~ C NASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 Page F-25 l TABL E F.14: to be issued at a later c3ate
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT iI2 Page F-26
\
~ A. ~
TABLE F.13: WETWELL ATMOSPHERE DOSE Detector Location Dose Rate R (cm)* 2 (cm)** (R/hr ) In Wetwell: 10 632 2.44 + 04 650 632 2..l.0 + 04 650 900 2.81 + 04 650 12'00 2.90 + 04 650 1555 2.1.2 + 04 In Drywell: 815 1642 6 ~ 04 815 2000 6 ~ 20 730 2500 4.42 610 610 3000 3500 3 '4 2.69 500 4noo 2.20 From core center ~f:rom bottom oF pooT
WASHINGTON PUBLIC POtKR PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page F-27 TABLE F 14: SUPPRESSION POOL MATER DOSE Detection Location Dose Rate R (cm)* Z (cm)** (r/hr ) In Netwell: 10 632 2.77 + 03 650 650 632 900 5 '2 2.13 +
+ 02 03 650 1200 1.51 + 03 650 1555 F 02 + 03 In Dr@well:
, 815 S15 730 1642 2000 2500 2.74 2 '5 1.68 01 01 01 610 3000 1.21 01 610 500 3500 4000 9 '3 0.28 02 02 e from core center
** from bottom of pool
I6" RRC(I)-4S, V CC CL V7 I 4-aRC( ) 4.'Rwcu 4) 6"Rwcu(s) 4'~wcu(4 4."Roc(4) V IA IB CC (a" PWCU(3) GOMTAWMENT BOUNDARY z" ave(Si) Figure F.l Node Point and Line Identification RHCU and RRC Systems
3000.O 2000.0 lCOO.O 10.0
.PIPE LE &TH FT.PIPE LE CTH FT. P<PKLE 8TH
'O FT PiPE L CTH ul i.0 ul PIPE %TH
T PIPE LE Cia 0.)
'ZOO 400 600 800 IOOO DISTANCE FROM PIPF GURFAJ= crt PiRure F.2 Dose Rate at Fipe Hid-Flane vs. Distance from pipe Surface 2 in. RRC (51)-4 Node 9-~8 4.56 x 106 ii(eV/cc-sec
. 40FT. PIPE LENGTH 25FT. Pi> E LENCTH l5FT. PiPE, I ENl TH
. IOFT, PIPE,LENl TH L,
ul 5FT- PIPE I EMSTH i.o
~
2Y< I=T. PIPE LEE'S~ ZQO cfQQ ~o DISTANCE FROM PlPF SURFACE cr Pigure P.3 Dose Rate at PiPe Hid-Plane vs. Distance from Pine Surface 3 in. RRC(51)-4 Yode 8 -7 3.51 x 10 HeV/cc-sec
IOOO L>>
'Q FT. Pl PE LENGTH 25 FT. PIPE LENGTH I5 PT. PIPE'EHC7H 10 FT. PIPE LENCiTH 5 I T. PIPE LENGTH
~
2/p FT. P)PE LEUC'TH I-0 2.0O 400 DISTANCE FROM Pl PE SURFPCE
~ Sn
- CfA I QQO Figure F.4 Dose Rate at pine >jid-plane vs. Distance from pipe Surface 4 in . RRC (4) -4S Diode 18 -2B 3 . 64 x 106 HeV/ cc-sec
08300 IPOO. IOO. '
. 40 FT PIPE, LENGTH ul ~ 26 FT'IPE LENGTH 10.. 16 Fr, PIPE LEh!6TH
~
IO FT. PIPE LENCiTH 5 FT. PIPE LE.Ml'~H
. g~@PT. PiPE, Lames ~
0 Zoo 4'oo DSTAMCE FROM PIPF- BJRFACE-CTA Sn Fiaure F.5 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface 4 in. RRC(4) -45 Node lA ~2A 3.50 x 10 MeV/cc-sec U
IOOO I 5 f0 I IOO. IO.
. ~ R. PIPE LENGTH
~
W I-"T PIPE LEMSYH
~
ISR PIPE~CTH
~
IO FT. PIPE LB4CiTH Xl- I
~
5FT. PIPE ~ ZY~Fr. PIPE L 0 Zoo DISTANCE FROM PIPE MR~ bOO
<CO
-~ IOOO Fieure F.6 Dose Rate at Pine Hid-Plane vs. Disranre from Pipe Surface in. RRC(51)-4S and 4 in. RHCU(4)-4 ilode 7 ~ '. 8.09 x 105 HeV/cc-sec
loooO-lOOO- ~ N
~
too.
',f
~
wmvipa LC. Z5 FT. PIPE LG.
~
i5 FT. PlPK LG. lO- .
'O Fj plPK L4) h hl ~
6 FT. PlPE L&I
~
2Q FT PiPE Lk. 0 emcee.K Fem 200 PlI
~~wc'-a E,
(c00
.
Fipure F.7 Dose Rate at Pipe bifid-Plane vs. Distance fznm Pine Snrface 6 in. (RllCU(31-4 Node 28~3 2.42 x 106 i~feV/cc-sec
IPoo.. l00-'o.
~
OO FT'lPE l.G
'2677 PlpE LG
'B PT PlPE LG
~
lOFT PlPE LS l~
~
6 Pr PlPKm
~
2,+FT PlPE LGi O DISTANCZ ZOO FRnM 4.00 PlPE SoRRCE
~ -cm
@co lOOO Figure F.8 Dose Rate at Pine Hid-Plane vs. Distance from Pioe Surface 6 in. RttCU(3)-4 Node 2A -3 2.22 x 106 ~feV/cc-sec
iOQOO. IPOO-'OQ.
~
4OFT. PlPC i a-
~
ZS FT. PiPE Lc
~
i5 Ft: PiPB LG LX lo-lO FT. PIPE LSi ul uJ ZYqW. PiPe mi O 2OO 4OO iOOO DiSrauCE PRAM PlPE. SURFACE cr 7 I ollra 7. 9 T)nse Rare a t Pi ne Hid-Plane vs. Distance from Pine 6 in. RWCU(3)-4 Node 3 4 2.12 x 106 iNeV/cc-sec
0 100-0 IQ Q I 1.0 '
~
- 4) R; PIPEUC 25&7. PIPE LG.
I
~
15 FT, PlPE LC. lOFT. PlPE M. I .
. 6 PTPiPE it, E
LLl O.GI L 0 ZQO 400 DISTANCF FROM PIPE SUQRq~ -~ IOQO Figure F.10 Dose Rate at Pipe Hid-Plane vs. Distance from Pipe Surface 6 in. RHCU(3)-4 Node 6 --4 4.31 x 104 >1eV/cc-sec
I 5000 K)00'OOO-e0 ~ I - ~ IO.OI
~
40 FT PIPE LG 25 FT. PIPE L&. l5 FT. PIPE'S
~ tO I-T. PlPF lg
'/ziFT-'PE ~
0 200 400 400 GOO DlSTANCE P~ PlPE SURFACE -Crn Fipure F.ll Dose Rate at Pipe Hid- Plane vs. Distance from Pine Surface 6 in. R'HCU(3)-4 Node 4-~5 8.22 x 105 HeV/cc-sec
I f 40, 30 ZQooO 10000.' I ~ looO
'0 FT PIPE LG
~
Z5 FT. PIPF Lb. 100.
~
)0 FT. PlPE L8.
~
6 FT.11PE Lci 2YZW tolPE W. o. 0 ZOQ DISTANCE FROM 430 1o1PF SU~CE -Crn
~ 800 tOOO Figure F.12 Dose Rat'e at Pine Mid-pl~no vs. Distance from Pine Surface 12 in. RRp(l)-45 5 ~ 03 x 106 MeV/cc-sec
3QOOO 20,000'OOOO',
)OOQ
~
4O FT. PlPE. LC. 25 FT. PlPE, LS. l6 FT PlPK LG
-
ra ~lmLG, E ~ 6 FT. PCPE LS. I=7. PlPE, LC Q lOOO Dl >TAhlCE PROM PlPE SURFACE " l"-~ Fipure F.13 Dose Rate at p~n~ Mid-plane vs. Distance from Pine Surfaro 16 in. RRC(1)-4S 5.03 x 106 MeV/cc-sec
20,000. lOOO.,'oFT. ePELc.
- 25FT. PlPE LG-100..
- 6 FT PIPE uh.
~
2.4'. PlPF % 0 2oo 4oo lOOO DlSTANCE FROMM P(PE SURF~ a Figure F.14 Dose Rate at .'ipe Hid-Plane vs. Distance fram Pine Surface 24 in. RRC(1)-4": and 24 in. RRC(2)-4S 5.03 x 106 HeV/cc-sec
200. lOO.
'io.
~
4OFT. PIPE LG. l.O 26 ~ PlPF LG.
'l6 FT. PlPPLb.
lO l=T. PlPE. LS.
~
5 FT. PIPE LQ,
~
Ilgwu Firn 0 2') 400 bQO QOO OOOO DISTANCE FROM PIPE ERR CE cr Figure F.1S Dose Rate at Pipe ibid-Plane vs. Distance from Pipe Surface 12 in. RHR(1)-4 or 12 in. RIIR(1)-45
e 0
<OFT, PiPK tS, 25I=T. PIPE LC 1.0 hl
.Ol o mo en ~oo
-c DiexmCK FRcv RM suRR Figure F.16 Dose Rate at Pipe >jid-plane vs. Distance from Pipe Surface 14 in. RHR(1)-4 or 14 in. RHR(])-4g
200 lOO lO- " 40 FT. PlPE LQ-
~
2.5 FT.'lPE LCa. I.O, '
'2./Z FT. PIPE 'LG.
g'.l
-Ot o 2Go 400 ccoQ DISTANCE FPQtvl PIPE GORFRCF -CM Figure F.17 Dose Rate at Pipe Hid-Plane vs. Dist, !ce from Pipe Surface 20 in. RIIR(2)-4 or 20 in. RIIR(2)-46'
lo IO 0 IO '.
'6
~
.26 FT. PlPE LG.
Ie r-V.ZPE W, F7. PIPE LQ ZY2FT. PIPE 49 t o 2oo 4Qo GOO DiSTANt:K FROM DOSE POINT tQ OBEYER, PlPEMRF~-cw Figure F.18 Dose Rate vs. Distance from Surface of 26 in. i~iain Steam Pipe
. 0 0
~0P G4'A 2 l I $ T 5 f.EL, ~
cL,G '71-'2 E wbII tOIIIT TOII' jTJCL 4..4,&S-P4 ROOF T ILES
~
I'L5 TON dR.NG k CRAPPIE, C~ %VII~A~ +ILOlaR. PLATFORM
~aa(~
GIRDER DRyER 1-S'REFUEUt44 SEP.
~l 4
REACTOR. I'OOL CAVI
~ELL- ~
pic 5 OVb ICOS I PRIlCViLI4b 5a' Oleic EL ~ la 0 Cs '04 e sF a ~ EL. 5~II-CIIEL 5TD t Oc
. ',
IIOOI
~ 5 II 3 ~ l4 3l STOLA4 6 AREA P .S1Z-0
$ TML PRILIAgY COW TWIIJ W lIT l"
-R. "C
'~Z,=" iL veSSEL R,. P. V
'I 5
'- ~'I wb. SS I 'L ~ 54b-0 dlOL 6I C4 SWISLO ~+LL ','OLLM H 5 I ~ (
i EL col .O4 r 4t Ig'~a cl To K~/ ~ I r"'L. 5MM QRV ~ $ QL FLOOC S~L DRY WELL EL. 499 EL
-0 ~ '. . )ca 5oi'-o' ~o I 4K-Iles'UPP
~ EL 49k 6R R'E.5. POOL, 'L wATCC AVEL V v I > . 41I -0 A~.44 1
( g h )C ) 5Auv- COLS < v'i
<<ILLEP C'EhCT POCr.f. f e 4460
>e,oksTiL cL 435->'L OU<tC.) ~ I. ~ ~ 4Z8'34" SKID, INNER SkIRT+ L4 t.3 a , IN'5. DIM. F OUWOA TIVOLI IvCAT l t7'-0"a I43'o le'-0 SE CT lOLI L CIO~luG SOUT Figure F.19 Containment Cross Section
0 Rev. 2 I
- WASHINGTON PUBLIC POWER PLANT SHIELDING WASHII'tGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT f12 Page F-4 5 W~ \
FIGURE F 20
~
to be issued at a later date
Rev. 2 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT f12 Page F-4 6 FIGURE F 21
~
to be issued at a lat~ date
REFEREMCES: 4R Dwc. S f90 S7lb Saon R.P.Y DRY eELL AREA SAC. WALL 14'-\ 1k'1R EL 501 W" ~9~i ~ 497'-4" WET WELL SUPP'. KOL. WATER LEVEL. EL. ~'-4>4 (MO,X) 85-9" 1.D. EL. 44 '-0" hlOT TO SCALE Fir nr~ F.22 Hetwell Zone tiodel
10 lO l6 FT- PIPE LG, N i=t'. PIPE, uh 5 PT. PlPK L uJ Z/j,FT PiRE i.a lO O 200 4QO em DISTANCE FROM PiP<'SURFACE -C~
.
Figure F.2'3 Dose Rate at Pipe Hid-Plane vs. Distance from Pipe Surface RMCU, RRC and RllR Systems
0 '
10 . 4I I ~ io3 i m Fr. PiPE, LC . 26FT. 91PE 'LQa, 16 FT PIPe. Ls, . 10 FT. PlPF L6. ah', VV,'1va La 10 L,. 0 20o 400 60o DISTANCE FFCA/I PlPE SU&FAG,E C% Pigure F.24 Dose Rate at Pipe Hid-Plane vs. Distance from Pipe Surface RMCU, RRC and RHR Systems 3 in Pines-Sched. 160
0 1
~
4OFT. PIPE LG.
~
25FT. PiPE m.
- i5 FT. I ~PE W.
IO IO FT. PIPK LS. z IO ~ Q 200 4X) 600 800 (OOO DISTAhlCE FROM P(PF SURFACE urn Figure F.25 Dose Rate at Pipe i'Lid-Plane vs Distance from Pipe Surface RMCU, RRC and RIOR Systems 6 in . Pine~-S( bed, RO
(0
~
40 FT, PlPE, LG. QSF7, PlPK l.C.
~ l5 P. PlPE. LG, lOFT. PlPE L6.
I IQ lO gQQ DlsTANCE AMM APE,
~ ~FACE cm.
bOO lOOO
}'ivu> e i'. 'r>>i:.:, !~ul,, ~t pine .'iii(-p]ane vs. Dist;ance from pipe Surface i(4'I',ll, IU(: anti n)lR Svsr.ems r ~,. i), r.: Ci'b~i(I
0 10 5 FT. P1PK LJa 10 8 lO 0 ZOO 4OO 400 DISTANCE FRO& PIPE ~RFA~M. -cion Figure F.27 Dose Rate at Pipe Hid-Plane vs. Distance from Pipe Surface PJCCU, RRC and RHR Systems
10 lO
~
zAFr. ptpe. w t 10 0 400 600 2CO DtSTANCE FROM P!PE 'SURFACE t".rn I Figure F.28'ose Rate at Pipe Hid-Plane vs'istance from Pipe Surface RHCU, RRC and RHR Systems 24 in. Pipes-Sched. 80
I5 FT. PIPE I G. IO FT. PIPE Lk 3 IO ~-FY.-PK-E;M
. M~ FT'. P'IP~ La uJ io'
~ 4X)
FROM PI PE MRFACE BC'ISTANCE e~ Figure F 29 Dose Rate at Pipe Mid-Plane vs. Distance from Pipe Surface LPCI System, Schedule 80 D=12 in. n n=l2.7% >n. TD=ll.37" in
IO 4O F'T, PIPE LC, 25 FT PIPE LG. l5 FT PIPF IG 10 f=T, PIPE I 4'O E Ul 0 2OO 4aO <ooo FROM PIPE, SuRFACE +5'HTAblCE O
>>Bure F.30 'ose Rate at Pine ibid-Plane vs. Distance from Pioe Surface L,Pf:T Se~t s m. Srherl>>l ~ SO,D=l'4 in.
z<" (@TED) zo'(a~n) lO 12 "(STD)
~"& e) ~ e" (AH. 4o)
I "(~ ao I Io Q 2OO 400 lao DlSTANCE FRM BJRFACE,cm Fir.ure F.31 Dose Rate vs. Distance from Pioe Surface Pioe Length = 2.5 ft.
10 IO 20"- STD Ia" (Sm) I2"'{ST@) lO tel, e'st:H 4o) i "(see ee) 2'scH so) IO 0 400 400 2QO DISTANCE PRQhrl PlPE SURFACE ~ Figure F.32 Dose Rate vs. Distance from Pipe Surface Pipe Length = 5 ft.
lO zo" (sTD) m-(w D) I~ "(srD)
~ l2" STD n
lO L~ E hl (fl
)p+ L o
~ mo 4oo I" (SCH EQ) . z" {mHec)
DISrANCE vs@ PiPE SURFAcE~~ Figure F.33 Dose Rate vs. Distance from Pipe Surface Pipe Length = 10 ft.
'io io IO I
~ gscH, e) 0 200 iOOO DISTANCE PRO+ PiPESuRF~- c~
Figure F.34 Dose Rate vs. Distance from pipe Surface Pipe Lenpth = 15 ft.
10
~ ~
10 10 z" (so~ ao) t "(sew ea) lO L 0 2QO 43o bcQ GOO lOOO O~WWVCE FRY P)m MR~~- cm Figure F.35 Dose Rate vs. Distance from Pipe Surface Pioe I.eneth = 25 ft.
Figure F.36 Dose Rate vs. Distance from Pipe Surface Pipe Length = 40 Ft.
3= LCM d=SCm d= 5 Crn d= l5.24 Ce d=30.48 CW lO
'=Q.94 Crn.
'=.12).92 l-AI.
d= 243.54 cN. l I I d = 487.48 Crn. l lO d=+l4.4 Ce. \ GAOL
)o'.
E ILI 0 lO 20 source LENGTH (I=T.) Fin~re F.'37 Dose Rate vs. Pipe Source Length at Various Distances from Outer Pipe Surface
0 <ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT I/2 Page G-1 APPENDIX G: Beta This appendix presents the calculated Dose Contribution beta dose contribution due to a postu-in Primary lated LOCA. Table G-1 gives the average Containment six-month source, Hev/sec, for the three f ission product source types. For a containment atmosphere source con-taining 100% noble gases plus 25%, halo-gens, the source equal's 2.0 x 1017 Hev/sec. rJsing an infinite cloud model the beta 3EBA(Bad/Sec) dose rate is given by Dg0 2 where, average beta energy, and activity concentration, Ci/m3 The average beta dose rate of 7.84 x 105 Rad/hr corresponds to a total beta dose contribution due to containment atmosphere of 3.4 x 109 Rads. To determine the beta dose contribution due to plate-out of 25~ halogens, a plate-out surface area of 3.3 x 10 cm2 was conservatively assumed. For Eg 1 HeV, this corresponds'o a six month total beta dose contribution of 2.3 x 105 Rads.
WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 Page G-2 TABLE G-1: BETA SOURCE STRENGTH Radiation Source Source Streng th (Hev/Sec) 100% Particulates 1.29 x 1019 100% Halogens 1.84 x 1017 100% Noble Gases 1.54 x 1017
0 Rev. WASHINGTON PUBLIC POMER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT 82 Page H-1 APPENDIX H This appendix presents the methodology and VITAL AREAS AND ACCESS assumptions used to determine the ROUT ES ANALYSED FOR integrated dose to equipment and personnel POST-LOCA OPERATIONS for vital areas and access routes outside the Reactor Building during post-? OCA operations. The source term is the Reactor Building elevated vent with gaseous effluents being filtered by the Standby Gas Treatment System prior to discharge to the atmosphere. H.l Source of Two contributions were considered as the Radioactivity source of the radioactivity to the Reactor to the Reactor Building Elevated Vent. Building Elevated Vent o '. Leakage from the drywell to the Reactor Building and discharged. via the SGTS to the Reactor Building Elevated Vent was assumed at a rate of: 0.5%/day = 2.1E-4 Hr 1 Leakage from the assumed leaks on the Hain Steam Isolation Valves in the l1ain Steam Tunnel was assumed at a rate of: 0.22%/day = 9.2E-5 Hr Thus, the total leakage rate of activity from the primary system is assumed to be 0.72%/day = 3.0E-4 Hr 1
WASHINGTON PUBLIC POWER PLANT SHIELDING tlASHLNGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT I/2 Pave H-2 H.l.l Reactor All radioactivity considered outside the Building Air Reactor Building is assumed to discharge Discharge Rate via the Reactor Building Elevated Vent. The removal rate of the Reactor Building ventilat'ion can be determined as follows: Removal Rate = SGTS discharge rate Reactor Buxldzng volume SGTS discharge flow of 2430 Ft3/Min was assumed which represents one volume change per day. Reactor Building volume = 3.5E+6 Ft3 Thus, the Removal Rate is: Removal Rate = (2430 Ft~3 Min)(60 Min/Hr) 3.5E+6 Ft3 Removal Rate = 4.2E-2 Hr l This removal rate was used in the deter-mination of radiation levels outside the Reactor Building. H.2 Calculation A small computer program was written to Methodology complete the calculations for the 18 nuclides and f ive time periods and sum the results. The equation used to determine the dose is as follows: Dose(Rads) = DF(j) (~ Ql Q~Q 3600 (H-l)
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT //2 Page Dose j i.. Rads from jth nuclide for a ith time period Gamma Dose Factors for semi-infinite cloud Rad m3 for Ci - Hr jth nuclide. X/Ql sec/m3 for gaseous releases from the Reactor Building Uent to the atmosphere for the th time period. RF Removal fraction of activity via the standby gas treatment Transmission Fraction (1-RF) 0.01 for particulates and iodines (99% efficiency or RF) TF 1.0 for noble gases (MNP C2 FSAR Section 6.5). Qlj Integrated activity of jth nuclide over ith time period that was released via leaks in the main steam isolation valves (Curies/Hour). Integrated activity of jth nuclide over the ith time period that was released via leakage f rom the primary to secondary containment (Curies/Hour).
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT (/2 Page H'-4 3600" = Conversion from hours to seconds H.2.1 Assumptions The following equation from "Meterology and Used in Atomic Energy" (Reference E-1) was used to Calculation determine the X/Q values shown in Table H.l Methodology 2 Dilution, me;
= 2.22(M) (3.16 + 0.1 ~~~, )
Vex
= FB (Building wake factor)
= '1 if Intake and Exhaust same level 2 if Intane and Exhaust separated by one floor 4 if Intake cavity is in Building wate
= Shortest Intake Exhaust arc length Aex = Exhaust area V mean = mean approach flow Vex = mean exhaust flow The intake was assumed to be for category F Conditions with a V mean = 1 meter/sec.
Then x/0 = FB RR
Rev. 1 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 Page H-5 FB Building wake factor RR = Release Rate from Reactor Building Vent (m3/sec) Concentration in reactor vent CV ='/RR Q = Curies/sec. released. V Concentration 'at intake CZ = FB. CI also = Q( X/Q). Therefore: Ci = V = Q( X Q) FB ~Fg (~RR ( X/Q) = l = total dilution factor (F~) (RR) (DF) . An F class stability was assumed for atmospher conditions., Five percent meteorology was then applied for time periods from 0 to l80 days. The dilution factors decrease by the following ratios for the time periods indicated. Time(Hrs)0-2 2-8 8-24 24-96 96-4320 Ratio 1.0 0.35 0.04 0.02 0.0l, The dilution factors were multiplied by the fi percent meteorology ratios to determine the actual X/Q values used in these computations. (See Table H.l)
0 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM 'NALYSIS PROJECT f12 Page H 6 H~ 2 ~ 2 Integrated The time dependent activity of -=each nuclide Activity Equa- being released from the HSIV was analysed as tions used in follows: this Analysis a'Al 0072 dt P Ae0 (-4 +
.
224 t) ( H-3) where Fractional leak fyym HSIV per hour (9.2 E-5 H Ao Initial activity oC th nuclide in primary containment at = 0 hrs. Thus, the activity concentration over a time period of tl to t2 is:
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEM ANALYSIS PROJECT 82 Page - H P 2 -(:+ 3.0E-4) t tl PA o oz Ql (A + 0 3.0E-4) Le
-(A + 3.0E-4) tl (i + 3.0E-A)t'2
-e The integrated activity concentration Crom the primary to secondary containment leakage, Q2, was calculated as fol lows:
dA2 dt KA -L C 1 2 2 ( H-5) where
/
Fractional leak rate from pri-mary containment
. 005 = 2. 1E-4 Hr 1 24hr
= Activity in primary containment
-(ii .0072)t 24 1 o
Rev. 1 WASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTB1 ANALYSIS PROJECT /!2 Page Ao Initial activity (ci) at t = 0
.0072 Leakage removal rate from pri-24 mary containment per hour
~ 0072 3. OE-4 Hr 1 24 L2 Discharge rate from Reactor Building Vent via standby gas treatment = 2430 f t3/min(60 min/hr)
L2 = l.46E + 5 Ft3/hr C2 Activity concentration in secon-dary containment A2 Curies in secondary containment V2 Volume in secondary containment C2 = A2 = A2 V2 3.5E + 6t.'t3 Rearrang ing dA2 -(~+3.0E-4) t L2A at =KZ o e or L2 dA2 - (~ l) A2 dt, o
.
( H-5A) l dA2 = KA e 0 l -F 2 A2 dt, I
<ASHINGTON PUBLIC POVER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT 82 Page H-'9 where F 2
= Z- +
V2 and Fl = (X + 3.OE-4) A2 ~A, F
-F A2 + F 2 A2 = r{t),' r{t) = KA e (H 6) solving L(
KA,) ( F~ Fl ( H-7) at t=O, A=O
-i<A -2 'E- A F2-Fl 4.2E-2 0
.OOSA Thus, A2 (t) ,905A e "
(l e" (.O4 ) t) (H-8) and, Q2 = L2A2 (t) V2 t, -CZT l.4~ + 5 ft3/hr .405A e (1-e )
~.5E + 6 et
<ASHINGTON PUBLIC POWER PLANT SHIELDING WASHINGTON NUCLEAR SUPPLY SYSTEH ANALYSIS PROJECT //2
,
page H-10 where C2 = 0.042 thusg -C2t Q2 = 2.11E-< A e 0 To determine the integrated concentration: t2 Q2 ( t) = 2.',1E-4A dt tl (H-10) l Solving, Q2 = 2.1E-4A 0 e -e (X+ C2) (H-ll) The values of Ql and Q2 are substituted Ln ~ for each Nuclide and each time period. Then using equation II-l, the dose commitment for each nuclide and each time oeriod may be calculated, These results are oresented in Section 6.3.
PLANT SHIELDING WASHINGTON NUCLEAR <ASHINGTON PUBLIC POWER ANALYSIS PROJECT f/2 SUPPLY SYSTEM Page H-ll TABLE H.l: POST-LOCA X/Q VALUES USED FOR CALCULATIONS OF INTEGRATED DOSES OUTSIDE THE REACTOR BUILDING Area Time (in hours) 0-2 2W 8-24 24-96 ,96-4320 (180 days) Security Center 4.1-4 1.45-4 1 6-5 8 '-6 4 1-'6 Auxiliary Security Center 2 '-4 8 '-5 ~
- 9. 6-6 4.8-6 2
~
~ 4-6 Sample Analysis Area
( EOC) l. 5-4 5-3-5 6 '-6 3 '-6 1 ~ 5-6 Nitrogen Supply to Accumulators 5.3-4 1 ~ 8-4 2.1-5 1 ~ 1-5 5.3-6 Standby Service Water Pump Valves 2.4-4 8. 4-5 9 6-6 4 8-6 2/4-6 Remote Shutdown Room 5 '-4 1 8-4 '.1-5
~
1
~
1-5 5 '-6
'-6 Switchgear Room gl 5 '-4
'-4
~
1 8-4
~ 2 ~ 1-5 2.1-5 1
~
~ 1-5
,-1 1-5 5
'-6 Switchgear Room 52 1 8-4 5 Radwaste Control Room 5
5 '-4 ~ 1.8-4 F 1-5
~
1 1-5
~ 5 '-6 Battery Racks 5.3-4 l. 8-4 2.1-5 1.1-5 5.3-6 DC Battery Charger Motor Control Center 5 '-4 1 ~ 8-4 2. 1-5 .1 1-5
~ 5 '-6 Three Motor Control Center s/ 2.1-5 1.1-5 5.3-6 Three Switchgears b. 3-4 1. 8-4 DC Battery Charger and Rack 5 3-4 8-4 2. 1-5 1-5 5 3-6 Diesel Oil Tanks 5 '-4
~ 1 1
~
~ 8-4 2.1-5 1 ~
1.1-5 5 '-6
~
Solid Radwaste Control Panel 5 '-4 1 ~ 8-4 ,2 '-5 1 ~ 1-5 5 '-6 The standby service water pump valves are approximately 700 feet from the release point. This distance is too great to calculate' dilution based solely on a building wake factor. However, the conservative assumption will be made that the dilution at the valves is the same as at the auxiliary guard house which is only 420 feet.
- Read as 4.1 x 10 4 etc.}}