Text

Kansas State University, Safety Analysis Report Dated 12/21/2004
	ML052580517
Person / Time
Site:	Kansas State University
Issue date:	12/21/2004
From:	; Kansas State University
To:	; Office of Nuclear Reactor Regulation
	Witt K, NRC/NRR/DRIP/RNRP, 415-4075
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ML052620181	List: ML052580517; ML052580519; ML052590053;
References
	10 CFR 2.390(d)(1)
	Download: ML052580517 (315)
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KANSAS STATE UNIVERSITY TRIGA MARK II NUCLEAR REACTOR FACILITY LICENSE NO. R-88 DOCKET NO. 50-188 SAFETY ANALYSIS REPORT DATED 21 DECEMBER 2004 REDACTED VERSION*

IN ACCORDANCE WITH 10 CFR 2.390(d)(1)

Redacted text and figures blacked out or denoted by brackets

KSS . .

Department cf Mechanical and Nuclear Engineering 302 Ratbofine HIl on~lo. KS 66506.5205 1212212004 785532-5610 Fax 785532.7057 Paul M. Whaley Kansas State University Nuclear Reactor Facility Manager 112 Ward Hall Manhattan, Kansas 65D6-2506 Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pilke Rockville, MD 20852-2738 RE: Additional3 niormation Relative to License Renewal of the lansas State UniversilyNucleiar Reactor Facility (License R-88, Docket 50-188)

Following submission of information supporting renewal of the KSU TRIGA 11 facility operating license (including a power uprate to 500 kW) an on-site conference occurred to identify and resolve issues, and a Request for Additional Information (RAI) was issued for the proposed KSU Safety Analysis Report and the Technical Specifications.

The attached material includes (1) a revision to the proposed Safety Analysis Report (addrcessing both relevant informal recommendations and the Request for Additional Information) , and (2) a tabulation of revisions and a tabulation of how each RAI item was addressed.

Correspondence relating to this information should be directed to P. M. Whaley (address above).

Thank y P. M. Whaley I %erify under penalty orperjury that the forgoing is true and 1t '.1fi/t 'ft3

X c - S PAS CC: D. Hughes, USNRC Project Manager hOMatC Iafa~ / ~oo

Safety Analysis Report Kansas State University TRIGA Mark II Nuclear Reactor Facility AtdaI-t -' $-I

-- h -l'-,S-i V* - -Iaer--l-.:

License R-88 Docket 50-188 21 December 2004 Department ofMechanical and Nuclear Engineering K-State Nuclear Reactor Facility Kansas State University 110 WardHall 302 Rathbonc Hall Manhattan, KS 66506 Manhattan, KS 66506

SAFETY ANALYSIS REPORT This page intentionally blank K-State Reactor

KSU TRIGA II Safety Analysis Report Table of Contents Table of Contents The Facility

1. The Facility ..... 1-1 1.1 Introduction . . . . .** 1-1 1.2 Summary & Conclusions on Principal Safety Considerations . . .1-2 1.2.1 Safety Considerations . . . . .1-2 1.2.2 ConsequencesofNormnalOperations . . . . . 1-3 1.2.3 Consequences of Potential Accidents . . . . . 1-5S

a. Maximum Hypothetical Accident . . . .1-5

b. Loss of Coolant .... 1-6 C. Insertion of Excess Reactivity . . ...................... 1-6 1.3 General Description of the Facility . . . . . . 1-7 1.3.1 Geographical Location . . . . . 1-7 1.3.2 Principal Characteristics of the Site . . . . . 1-7 1.3.3 Principal Design Criteria, Operating Characteristics, & Safety Systems ........... 1-7 1.3.4 EngineeredSafety Features . . . . . 1-8 1.3.5 Instrumentation and Control (I&C) and Electrical Systems . . .1-8 'S

a. Reactor Control System . . ............................................... 1-9

b. Proccss Instruments. . . 1-9

c. Reactor Protection System .................. . . . . 1-9

d. Radiation Safety Monitoring Systems. . ........................................... 1-10

c. Electrical Power . . . . 1-11 1.3.6 Reactor Coolant and other Auxiliary systems . . . . . 1-11

a. Reactor Coolant System .. . . 1-12

b. Secondary Cooling System .. . . 1-12

c. Bulk Sbield Tank . . . .1-12

d. Makeup .... 1-12 1.3.7 Radioactive Waste Management and Radiation Protection ..... 1-12

a. Gaseous Waste ............................ .......... 1-12

b. Liquid Waste . . . ..................... 1-13

c. Solid Waste . . . ...................... 1-13 1.3.8 Experimental Facilities and Capabilities .................................... . . 1-13

a. Central Thimble . . . ................. ; 1-13

b. Rotary Specimen Rack ............... ..... 1-13

c. Pneumatic Specimen Tube . . . ................. 1-14

d. Themrnalffhermalizing Columns . . . ................ 1-14

e. Beim Tube Facilities . . . ........ t 1-14 1.4 Shared Facilities and Equipment . . . .. 1-14 1.5 Comparison with Similar Facilities . . . . . 1-15 1.6 Summary of Operations .. ................................ 1-17 1.7 Compliance with Nuclcar Wastc Policy Act of 1982 ..... 1-18 1.8 Facility Modifications and History. . . . .................. . 1-19 1.9 Bibliography . . ... 1-19

12/04 Site Characteristics

2. SiteCharacteristics.,._ 2-1 2.1 Geography and Dcmogmphy . . . ... 2-1 2.1.1 Sitc Location and Description ........... . . . . 2-1

a. Specification and Location . . . .. 2-1

b. Boundary and Zone Area Maps ..... ; 2-1 2.1.2 Population Distribution ...... 2-7 2.2 Ncarby lndustrial, Transportation, & MilitaryFacilitis ...... 2-9 2.2.1 Locations and Routes ...... 2-9 2.2.2. AirTraffic ...... 2-9 2.23 AnalysisofPotentialAccidents . . . . . .2-9 2.3 Meteorology... ..- _2-9 23.1 General and Local Climate .............. . . .. 2-9 23.2 Site Meteorology ............................. . . . . . . 2-10 2.4 Geology, Seismology, and Geotechnical Engineering ...... 2-12 2.4.1 Regional Geology .............. 2-12 2.42 - Site Geology . ..... 2-12 2.43 Seismicity ...... 2-12 2AA Maximum Earthquake Potential . ........ . .................................................... 2-15 2.45 Vibratory Ground Motion ...... 2-16 2.4.6 Surface Faulting ......... . .. .. 2-17 2.4.7 Liquefaction Potential ........... . . . . 2-17 2.5 Hydmlogy ...... 2-17 2.6 Bibliography ................... 2-17 Chapter 2 Appendix A: Color Plates Chaptcr 2 Appcndix Bi Population Distribution In Riley County, Kansas Chaptcr 2 Appendix C: Meteorological Frequency Distributions Dcsign of Structures, Systems, & Components

3. Design of Structures, Systems, & Components . ................................................. 3-1 3.1 Design Criteria ................................... ; . ................... 3-1 3.1.1 General Conditions . . . ............... 3-1 3.1.2 Architectural and Engineering Design Criteria . . ................................. 3-2 3.1.3 Structural System Design ofthe 1961 Building . . ................................. 3-6 3.1.4 Structural System Design of the 1972 Building . . ............................ . 3-7 3.1.5 Sanitary Sewer System . ................................... . 3-7 3.1.6 Ston Sewer ................................... . . . .3-8 32 Meteorological Dama.ge...................._._._...._..._3-8 3.3 Water Damage...... .... 3-8 3.4 Seismic Damage. _._.

.............. . -.. . 3-11 3.5 Systems and Components .. . . 3-12 3.5.1 Fuel System . . . ................. 3-12

a. Potential forZr-WYaterReaction . ........... . 3-13

b. Phase1Volume Changes ....... . ...... ......... _.

3-14

c. Internal Fuel Rod Pressure . . ................... 3-14

d. Conclusion ...................... _._.3-17 3.5.2 Shielding ................. . ; . .. .3-17 3.5.3 Control Rod Scram System _ ................ 3-17 Page ii of iix

12104 3.5.4 Confinement and Ventilation Systems ............................. 3-18 3.6 Bibliography .3-19 Rcactor Description

4. Reactor Description ......... 4-1 4.1 Summary Description .......... 4-1 4.2 Reactor Core........4-3 4.2.1 Rcactor Fuel ......... 4.4

a. Dimensions and Physical Properties ................................ 44

b. Composition and Phase Properties ................................ 4-5

c. Core layout . 4-6 4.2.2 Control Rods ......... 4-8

a. Control Function ........ 4-11

b. Evaluation of Control Rod System ....... 4-11 4.2.3 Neutron Moderator and Reflector ......... 4-12 4.2A Neutron Startup Source. .. . . ... . . 4-13 4.2.5 Corc Support Structure ......... 4-13 4.3 ReactorTank .......... 4-13 4A Biological Shicld ......... 4-14 4.5 Nuclear Design ......... 4-14 4.5.1 Design Criteria - Reference Cor.. .. . . . . . 4-15 4.5.2 Reactor Core Physics Parameters. _ . :.4-15 4.5.3 Fuel and Clad Temperatures ......... 4-16

a. HcatTransferModels .......... 4-16

b. Spatial PowerD istr bution.4-20

c. Steady-State Mode of Operation ........ ............... 4-23

d. Pulsed Mode of Operation ................... 4-24 4.6 Thermal Hydraulic Design and Analysis ....... 4-25 4.7 Safety Limit. .............. .4-27 4.8 Operating Limits ......... 4-27 4.8.1 Operating Parameters..........4-27 4.8.2 Limiting Safety System Settings ......... 4-27 4.8.3 Safety Margins ......... 4-27 4.9 Bibliography .4-28 Appendix 4-A: Post-Pulse Fuel and Cladding Temperaturc Reactor Coolant Systems

5. Reactor Coolant Systems........... ... 5-1 5.1 Summary Description ............. ... ............................. 5-1 5.2 Primary Coolant System . . . ............ 5-3 5.3 Secondary Cooling System.. . . . . . -54 5.3.1 Secondary Cooling System Flows -

S.....

5.3.2 Secondary Cooling Automatic Control System ..... 5-6 5.3.3. Secondary WaterQualie 7.....-

5.4 Primary Cleanup System . .. . 5-8 5.5 Makeup Water System. . . . 5-9 5.6 Nitrogen Control System . . . . .. 5-10 5.7 Auxiliary Systems Using Primary Coolant ......................... . . 5-10 5.8 Bibliography . . . . .. 5-11 Page iii of Hix

12/04 Engineered Safety Features

6. Engineered Safty Fcaturcs . ....................................................... 6-1 6.1 Bibliography. .... ** 6-1 Instrumentation and Controls Systems

7. Instrumentation and Controls Systems...... . . ... . . 7-1 7.1 Summary Descption....... .... 7-1 7.2 Design of Instrumentation and Control Systems............................................... 7-3 7.2.1 Design Crtra -.. 7.3 7.2.1 7.2.2 Design Design Basis equria ments............................................................. .................

Basis Requirements...7-3 7-3 7.23 System Description. . . . 7-5 7.2.4 System Performance Analysis. . 7-5 7.2.5 Conlson.- .... n 7-5 732. Conclusionto..................................................

Sytm.. . ....................................... 7-5 7-7 7.3 Reactor Coiitrol . .

73.1 Neutroinc Instruments (Reactor Power) . . ... ....... . . 7-7 732 Temperature..... 7-10 733 Water Conductivity. .... . .7-10 73.4 Control Rod Drives. . . . 7-10

a. Standard Control Rod Drives.........- .. .7-11

. b. TransientRodDrive . . . ............... 7-14

c. Interlocks........................................................................................... 7-15 7.4 Reactor Protection System........... ......... 7-17 7.5 Engineered Safety Features Actuation Syste.s... ..... .. 7-17 7.6 Control Console and Display Instruments.......... . . ...... 7-17 7.7- Radiation Monitoring Systems.....7-18 7.8 Bibliogaphy .......... . .... .... 7-22 Electrical Power Systems

8. Electrical Power Systems..... . . .. 8-1 8.1 Normal Electrical Power Systems ............ 8-2 82 Emergency Electrical Poser Systems..... ... 8-3 83 Bibliography . . 8-4 Auxiliary Systems

9. Auxiliary Sytems..... 9-1 9.1 Heating, Ventilation, and Air Conditioning System . . . 9-1 9.2 Handling and Storage of Reactor ueb. . ......... 9-I

93 Fire Protection Systems and Programs............................................................ 9-2 9.4 Comrnunications Systems.... . .......... 9-3 9.5 Possession & Use of Byproduct, Source & Special Nuclear Material. . ; 9-3.

9.6 Cover Gas Control in Close Primary Coolant Systems...... . 94 9.7 OtherAuxiliarysystems ............... -. .. .._. 94 9.7.1 Reactor Sump System ..... 9-4 9.7.2 ReactorBay Polar Crane .......... .. 94 9.73 Beam Facilities .. 94

a. Thermal Column Door . ......-. . 9-5 b Beamport Plug Handlng . 9-5

c. Bcam'Facility Vents and Drain.... ......... 9-5 Pame iv of iix

12/04

d. Pneumatic Transfer System (Rabbit) .......................................................... 9-5 9.7.4 Associated Laboratories .......................... 9-6 '

9.8 Bibliography .............................................................. 9-6 Experimental Facilities and Utilization

10. Experimcntal Facilities and Utilization . . . . . 10-1 10.1 Summary Description ..... 10-1 10.1.1 Expcrimcntal Programs ...... 10-1 10.1.2 Experiment Monitoring and Control . . . . . . 10-1

a. Leak Detection .............................................................. 10-2

b. Area Radiation Monitor. . . .................... 10-2

c. External Scram . . . ............ 10-3 10.1.4 Experiment Review and Approvals .... . .10-3 10.2 E:xperimcntal Facilities . . . . .10-3 10.2.1 In Core Facilities . . . . .10-3

a. Available Fuel Element Spaces . . . .10-3

b. Small Upper Grid Plate Penetrations . . . .10-3

c. CcntralfThimble . . . . 10-3 10.2.2 In Tank, Ex Core Facilities ... :........ 104

a. Thermal Column ..................... ..............

. . 10.4

b. Thermalizing Column._..

b~~~~~.ThrnlznCoun......................................... ..................... 10-5

c. Rotary Specimen Rack. .. . 10-5 10.2.4 Automatic Transfer Facilities . ... . ...................... 10-7 10.2.5 Bcam Ports . ............. 10-7 10.3 Experiment Review . . . ... 10-9 10.3.1 Planning and Scheduling of New Experiments ...... 10-10 10.3.2 Review Criteria . .......... .. .......... 10-10 10A Bibliography . ..... 10-11 Radiation Protection and Waste Management 11 Radiation Protection and Waste Management. . . . .11-1 11.1 Radiation 11-1 11.1.1 Radiation Sources ...... 11-1

a. Airborne Souce .s.11-2

b. Liquid Radioactive Sources.11-3

c. Solid Radioactive Sources . . . .114 11.2 Radiation Protection P rom .am.11-4 11.1.3 ALARA, Program ..... 11-6

a. Policy and Objectives . . .......................................................... 11-5

b. Implementation of the ALARA Program ..................... ..................... 11-7

c. Elements of the ALARA Review and Report .. .. 11-7

d. Review and Audit....... 11-8 I .1A Radiation Monitoring and Surveillanc .. . . . . 11-8

a. Surveillances. -. . . 11-8

b. Radiation Monitoring EquipmenL . . ................................................. 11-8 C. Instrument Calibratior. . . . 11-8 11.1. Radiation Exposure Control and Dosimetry .................................................. 11-9

a. Shielding ..... 11-9

b. Personnel Exposure... 11-10 C. Record Keeping.......11-12 11.1.6 Contamination Control. . . . . 11-12 Page v of iix

12/04 11.1.7 Environmental Monitoring . . .. 11-14

a. Area Radiation Monitors . . ............I........ 11-14

b. Airborne Contamnination Monitors . .................................................. 11-14 C. Pool Surface Monitor1...1-14

d. Additional Monitoring1... 1-14 11.2 Radioactive Waste Management.. . . . . 11-16 11.2.1 Radioactive Waste Management Program . ...................................... 11-16 11.2.2 RadioactiveWasteControls . . . . . 11-16 11.23 Release of Radioactive Waste . . . .11-17 113 Bibliography . . . ....................... 11-18 Appendix A: Radiological Impact of 4 ARAND 1N During Normal Operations A.1 Introduction I11.............

1.A-I A.1.1 Purpose .. 11A-1 A.1.2 Radiological Standards 1.A-2 I

A.13 KSU TRIGA DesignBases ,... 11.A A.2 Radiological Assessment of41Ar Sources .................................................. 11.A-4 A.2.1 Production of 41ArfromBeams .......................... .. I1.A4 A-2.2 Production of 4t Ar in Rotary Specineri Rack . ......................... 11.A-5 A.23 Production ofd4Ar from Coolant Water ..... 11.A5 A.2A Maximum Impact of4 Ar Outside the Operations Boundary . . 11.A-7 A.2.5 Radiological Assessment of 16N Sources ................................ .. 1.A-8 A.3 Bibliography..;A-IO Conduct of Operations

12. Conduct of Operations ..... 12-1 12.1 Organization......... . . 12-1 12.1.1 Structure .......... , 12-2 12.1.2 Responsibility ......... . . . . . . 12-4

a. Reactor Operations Line Management.... ....... 1.2.........

124

b. Environment, Safety, and Health Staff .............................................. 12-5 C. Principal Advisory and Oversight Committeces . . ............ 12-6 12.1.3 Staffing ..... 12-7 12.1A Selection and Training ofPersonnel . . . . . 12-7 12.1.5 Radiation safety ............... . . . _.,,

.. 12-8 12.2 ReviewandAuditActivities . . . . .................. 12-8 12.2.1 Reactor Safeguards Committee Composition and Qualifications ............... 12-9

a. Charter and Rules ...... . . . . 12-9

b. Review Function .: . . .................. 12-9

c. Audit Funtion.......12-10 123 Procedures ........ ,,,,,,,,,, 12:10 123.1 ReactorOperations ...... 12-11 123.2 Health Physics ... ............ . . . . 12-11 12.4 Required Actions ..... 12-11 12.4.1 ViolationofFacilitySafetyLimit ....... . . . . 12-11 12A.2 Occurrences Reportable to theU.S.NuclearRegulatory Commtiltee . .12-12 Page vi of fix

12104 12.5 Reports to the Nuclear Rcgulatory Commission . . ................. 12-12 12.5.1 Imrnediate Notification...12-13 12.5.2 14-Day Notification..12-13 12.5.3 Thirty-Day Notification...12-13 12.5.4 OtherRports...12-13 12.6 RccordRetention . . .12-14 12.6.1 Fivc-Year Retention Schedule...12-14 12.6.2 Certification Cycle...12-14 12.6.3 Life-of-the-Facility Records . . .12-15 12.7 Emergency Planning ............................ 12-15 12.8 Security Planning .12-15 12.9 Operator Training and Requalification .12-15 12.9.1 Requalification Program .12-16

a. Medical Certifcation..12-16

b. Proficiency .. 12-16

c. Examinations .. 12-16

d. Lectures. 12-17

c. Rccords..12-17 12.10 Medical Certification of Licensed Operators and Senior Operators .12-17 12.11 Bibliography .12-17 Accident Analysis

13. Accident Analysis .. 13-1 13.1 Accident Initiating Events and Scenarios .. 13-1 13.2 Accident Analysis and Determination of Consequences, .. . 13-1 13.2.1 Notation and Fuel Properties ................... ,,.,,.. , .. 13-1 13.2.2 Loss of Reactor Coolant .. 13-3

a. Initial Conditions, Assumptions, and Approximations .13-3

b. Core Geometry .............................. , . .............. 134

c. Decay Power .............................. , . .............. 13-5

d. Maximum Air Temperature .13-6

e. Fuel and Cladding Temperature Distribution .13-8 f Radiation Levels from the Uncovered Core .................................... 13-10

g. Conclusions . ........... , , .......... 13-13 13.2.3 Insertion of Excess Reactivity..13-14

a. Initial Conditions, Assumptions, and Approximations.......................... 13-14

b. Computauional Model for Power Excursions ............ ...................... 13-15 C. ,,,,,,,,,,...........................

Conclusions ................................. 13-16 13.2.4 Single Element Failurc in Air.13-16

a. Assumptions, and Approximations ......................... ,. ,.,. 13-17

b. Radionuclide Inventory Buildup and Decay ... 13-17 C. Data From Origen Calculations ..................... ,.,,.,,,,,.,,.,,.,. 13-18

d. Refrence Case Source Terms.13-19

c. Dcrived Quantities ., 13-19

f. Comparison with the DAC and the ALI .13-22

g. Comparison with the effluent concentration. 13-22

h. Potential downwind dosc to a member of the public. 13-22 Page vii of iix

12/04

i. Residual Activity from Fuel Utilization Prior to Receipt ..................... 13-25 I Conclusions .......................................................... 13-26 13.3 Bibliography .......................................................... 13-27 APPENDIX A: ORIGEN Input file for I tonne U-235 at 1 watt for 40 years APPENDIX B: ORIGEN Input Filc for 1 tonne U-235 at 1 watt 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per day for 5 days APPENDIX 13.C: ORIGEN Output File Extracts for 1 tonne U-235 at 1 watt for40 Ycars APPENDIX 13.D: Output File Extracts for.1 tonne U-235 at I W for 8 hld, 5 Days APPENDIX 13.E: Maximum Activity Availablc for Release One TRIGA Elemcnt at 86.42 W for 40 Years APPENDIX 13.F: Maximum Activity Available for Release One TRIGA Element at 31.125 kWY, 8 hld, 5 Days Technical Specifications

14. Published Separately ................................

Financial Qualific'ations

15. Financial Qualifications .. 15-1 15.1 Financial Ability to Operate the Reactor ............................... :. 15-1 152 Financial Ability to Decommission the Facility . .15-1 153 Bibliograpby .. 15-1 Appendix 15.A: Financial Statements Page viii of iix

1. The Facility 1.1 Introduction The Kansas State University nuclear research reactor is owned by Kansas State University (K-State), and operated by the Department of Mechanical and Nuclear Engineering. The reactor was obtained in 1960-1962 through a grant from the Unitcd States Atomic Energy Commission and is currently operated under Nuclcar Regulatory Commission License R-88 and the regulations of Chapter 1, Titlc 10, Code of Federal Regulations. The facility supports education and training, rescarch, and public service activities. The reactor facility is located on K-State campus in Manhattan, Kansas, a city of approximately 50,000 residents and 20,000 students, as described in Chapter 2, Site Characteristics.

This report is based on the Kansas State University TRIGA Mark II Hazards Summary Report (1961) for the initial operation of the reactor at 100 kW thermal power, the 1968 Safety Analysis Report and Safcty Evaluation Report for license amendment to allow 250 kW steady state thermal power (250 MW pulsing capability), and subsequent analyses supporting steady state operations to a maximum of 1,250 kW (pulsing to a nominal S3.00 reactivity insertion. A S3.00 reactivity -

insertion is expected to result in a peak thermal power of approximately 1,340 MWV). Based on proposed reactivity limits, the KSU reactor will only be able to achieve about '4 the proposed maximum power level for steady state operation; therefore thermal-hydraulic and consequence analyses arc conservative by at least a factor of 2.

The 1961 KSUTMIl Hazards Summary Report identified a set of potential hazards associated with operation of the reactor. The U. S. Atomic Energy Commission reviewed the report and concluded that there is "reasonable assurance that the reactor can be operated at the designated location without undue risk to the health and safety of the public." In 1968, a Safety Analysis evaluated changes in the original hazards analysis for operation a higher steady state power level and the addition of pulsing capability to support a license amendment for operation at 250 kWV with pulsing to 250 MW. The U. S. Atomic Energy Commission reviewed the requested amendment and concluded that there is "reasonable assurance that the reactor can be operated at the designated location without undue risk to the health and safcty of the public."

This report addresses safety issues associated with operation of the reactor at steady state power levels up to 1,250 kV, and increased pulsing capabilities. The maximum excess reactivity permitted by Technical Specifications cannot achieve a continuous steady state power level greater than about 500 kW; therefore analysis performed for steady state operations at 1,250 kV is extremely conservative in evaluating consequences and characteristics of normal and accident scenarios. This report reflects the as-built condition of the facility, and includes experience with the operation and performance of the reactor, radiation surveys, and personnel exposure histories related to operations to a maximum of 250 kW steady-state power. Where appropriate, radiological characteristics have been extrapolated to reflect operation to a maximum of 1,250 kW. The conscquence of routine generation of radioactive cffluent and other waste products from.

steady state operation to a maximum of 1,250 k1V is addressed in Chapter 11. Radiation worker and public doses from radiation associated with routine operations are well within the limits of Title 10, Code of Federal Regulations, even under extremely conservative scenarios. The consequence of accident scenarios from operation at 1,250 kW steady-state power and pulsing is K-State Reactor 1-1 Original (12104)

Safety Analysis Report

THE FACILITY presented in Chapter 13. The consequences of accidents postulated to occur under extremely conservative conditions are wvell within limits. Therefore, analysis demonstrates that there is still a "reasonable assurance that the reactor can be operated at the designated location without undue risk to the health and safety of the public."

The description of the reactor core and thermal hydraulic analysis presented in Chapter 4, the Secondary Cooling System in Chapter 5, and the Reactor Control System in Chapter 7 are based on 1,250 k-V operations.

1.2 Summary & Conclusions on Principal .Safety Considerations Design basis parameters of the KSUTMI1 are (1) power level, (2) fuel temperature, and (3) fuel loading required to achieve desired power. Limits on the amount of fuel loaded in the core and on the maximum power level ensure the KSU TRIGA Mark II nuclear reactor is an inherently safe reactor.

1.2.1 Safety Considerations As of July, 1999, there were over 70 TRIGA reactors in use or under construction at universities, government and industrial laboratories, and medical centers in 24 countries. Historically, analysis and testing of TRIGA fuel has demonstrated that fuel cladding integrity is not challenged.

as long as stress on the cladding remains within yield strength for the cladding temperature.

Elevated TRIGA fuel temperatures evolve hydrogen from the zirconium matrix, with concomitant pressure buildup in the cladding. Therefore, the strength of the clad as a function of temperature establishes the upper limit on fuel temperature. Fuel temperature less than limiting values will ensure clad integrity (as evaluated in NUREG 1282) and therefore contain radioactive materials produced by fission in the reactor core.

As a natural-convection cooled system, heat removal capacity is well defined as long as the primary coolant is 'sub cooled, restricting potential for film boiling. Limiting the potential for film boiling assures fuel and clad temperatures are not capable of challenging cladding-integrity.

The maximum beat generated within a fuel element and the bulk water temperature determines the propensity for film boiling. The design basis analysis in Chapter 4 indicates that steady state operation at power levels greater than 1,250 kW in natural convective flow will not lead to film boiling. Analysis indicates that transition boiling may occur during the maximum pulse, but that this condition will not evolve to film boiling.

Negative fuel temperature feedback inherently limits the operation of the reactor. Increases in fuel temperaturc associated with operation at power regulate maximum possible steady state power, as described in Chapter 4. This coefficient of feedback is a function of the fuel composition and core geomety within established core systems, the negative temperature coefficient is rather constant with temperature, as described in chapter 4. Excess fuel (above the amount required to establish a critical condition) is required to overcome the negative temperature fccdback as operation at power (or pulsing) causes the fuel to heat up. Consequently, maximum possible power using TRIGA fuel is controlled by limiting the amount of fuel loading.

Limits on total K-State Reactor 1-2 Original (12/04).

Safety Analysis Report

CHAPTER 1 fuel loading and excess reactivity ensure that the maximum power level will not Icad to conditions under which design basis temperatures are possible. A limit on the maximum pulsing reactivity ensures pulsed operations do not lead to conditions under which design basis temperatures are possible.

1.2.2 Consequences of Normal Operations -

As indicated in Chapter 11, radiation sources arc discharged from the reactor facility in gaseous (airborne), liquid or solid form. These forms are treated individually in subsections of Chapter

11. Airborne radiation sources consist mainly of Argon-4l1, Nitrogen-16 and Tritium, with Argon 41 the major contributor to off site dose. Limits on Argon-l and Tritium are tabulated below, with Cesium 137, the other significant isotope of interest for the KSU reactor.

IOCFR20: Appendix 13to Part 20-Annual Limits on Intake (ALrs) and Derved Air Concentrations (DACs) or Radionuclides for Occupational Exposurc- Effuent Concentrations; Conccrntrations ror Release to sewerage Table 1 Table 2Tal3 Occupational Values Concentratns Releases to AtomiRadionucflde class ol. Col. 2 Co 3 Col. 1 Col. 2 No. Oral Inhalation I Monthly Ingestion AU DAC i Air Water Average AU jftmCi/ml)l (pCI/ml ) (pCVmI) Concentration 18j.0ro.i ). --(Pci/mi) is Argon-41 Submersion' 3E-6 ll1E-8__

Water, DAC _-

1 Hydrogen-3 Includes 8E+4 l2E-5 1+4 IE-7 1E-3 IE-2 absorption 55 1 Cesium-137

_____137 0 all copound -, 1E+2 223+Z 6E-8 2E-10 IE-6 I-Gas (HU or T2) Submersion a: Use above values as HT and T2 oxidize In air and ia the body to HTO A general limit on off site doses from gaseous cffluents is also contained in § 20.1101 Radiation protection programs:

d) To implement the ALARA requirements of § 20.1101 (b), and notwithstanding the requirements in § 20.1301 of this part, a constraint on air emissions of radioactive material to the environment, excluding Radon-222 and its daughters, shall be established by licensees other than those subject to § 5034a, such that the individual member of the public likely to receive the highest dose will not be expected to receive a total effective dose equivalent in excess of 10 mrem (0.1 mSv) per year from these emissions.

Argon 41 is the major contributor to radiation exposure incident to the operation of the K-State -

reactor. Argon 41 is attributed to neutron activation of natural argon (in air) in the reactor bay atmosphere, rotary specimen rack adjacent to the core, and dissolved in primary coolant. Argon 41 has 1.8 h half-life. Calculations based on 1,250 kW steady state continuous operations show that doses in the reactor bay remain below inhalation DAC. Using extremely conservative K-State Reactor 1-3 Original (12104)

Safety Analysis Report

THE FACILITY assumptions of operational conditions in concert with the worst-case wind stability class, the off site dose from Argon 41 is slightly less than 10% of the 10 nrenilyear limit. A summation of all relative frequencies for winds under Pasquill stability category A (Table B-3) indicates frequency less than 0.6%, i.e., the contribution to off site doses from Argon 41 produced during a year of full power, steady state operations accounts for less than 0.6% of the total dose. All other atmospheric dispersion calculations show that the off site dose from Argon 41 is well within limits, and doses in the reactor bay are below the levels requiring controls of an airborne radioactivity area. Chapter 11 Appendix A shows peak off-site activity concentration during normal operations would be about 4.5 x IO pCi/mL at 53 m downwind under extremely unstable atmospheric conditions, less than the effluent limit of 0.01 pCi/mL. A full year exposure to equilibrium argon concentration for 1,250 kW operations under normal atmospheric conditions would lead to an cffective dose of less than 7 mrem, wcll within applicable limits.

Nitrogen 16 is the major contributor to radiation fields directly over the reactor pool during operation. Nitrogen 16 is produced by a fast neutron reaction with oxygen (as a unatural component of water in the core). Nitrogen 16 has a 7:1 second half-life, and consequently does not remain at concentrations capable of contributing significantly to off-site dose. Chapter I I shows very conservative calculations lead to an expected exposure rate of slightly less than 100-mrelhr at one meter above the center of the reactor tank during sustained operation at 1,250 kW thermal power. The 22-foot level has radiation monitors directly above the pool and at the rail surrounding access to the pool. Measured exposure rates directly above the pool surface are about 20-30 mR/h at 250 kW operations, and measurements at the rail approach 2 mR/hr. During normal, steady state 500 kW operations dose rate can be expected to achieve 40-60 mremnhr, and during steady state operations at 1,250 kW the area directly above the pool surface may become a high-radiation area. Therefore, radiation dose rates directly above the reactor pool during expected operations at levels up to 500 kW-are within required levels for a radiation area as defined in I OCFR2O, and additional administrative controls for access to the area directly above the reactor pool 500 kW to the maximum license power level of 1,250 kW may be required.

Installed monitoring systems provide information necessary to identify appropriate access controls.

Tritium is generated by sequential activation of hydrogen (in water) in the core area. Measured tritium specific activity in primary coolant is less than 5 x 10' pCi/g. If the reactor bay atmosphere were saturated with this water at 301C, the water concentration in the air would be less than 3 x 10'5 g/mL and the activity concentration in the atmosphere 1.5 x 1I jiCilznl, well below the DAC limit and well below the atmospheric effluent limit with the dilution factor of 200 for discharge from the top of the react'orbay. Even under the extremely conservative assumption that the complete tritium inventory of the reactor pool is released into the reactor bay atmosphere, the tritium concentration would be within limits for an unrestricted area.

No liquid radioactive material is routinely produced by the normal operation ofrthe KSU TRIGA reactor except for miscellaneous neutron activation product impurities in the primary coolant.

Non-routine liquid radioactive contamination may be produced during decontamination, maintenance activities (such as resin'changes), or occasional level adjustments in the reactor tank or bulk-shicld tank. Most releases occur because of condensation in the air-handling unit during surmmer months. Liquids in the reactor bay floor drains are collected in the reactor bay sump, along with condensate from the air conditioning system. Quantities arc small, and these liquids K-State Reactor 1-4 Original (12104)

Safety Analysis Report

CHAPTER 1 arc released to thc sanitary sewerage system after assay and filtration.

Most of the impurities produced in the primary cooling system are deposited in the mechanical filter and demincralizer rcsins. Therefore, these materials are dealt with as solid waste. The only radionuclides observed are tritium and trace quantities of '"CS. Typically there are three releases of liquids annually, each amounting to 2.5 mi3. Even without dihiltion, concentrations of these isotopes arc well below IOCFR20 Appendix-B cffuent conecntration limits and monthly sewerage limits. Even unfiltered, untreated primary coolant would meet the liquid effluent limit without further dilution.

1.2.3 Consequences of Potential Accidents Safety Analysis, Chapter 13, recognizes three classes of accidents for which analysis is required.

7hc maximum hypothetical accident is a fuel element failure with maximum release of fission product inventory, from which the radioactive materials can migrate into the environment. _

Complete loss of coolant from the reactor pool is the second accident analyzed. The final accident is an insertion of the maximum available positive reactivity. Analysis demonstrates the consequences of reactor accidents are acceptable.

a. Maximum Hypothetical Accident.

Source quantities of radioactive noble gases and iodine are computed and tabulated in Chapter 13 for a maximum hypothetical accident involving cladding failure in a single TRIGA fuel clement and the escape of the radionuclides into the environment Two -

limiting cases of operation are considered. For short-lived radionuclides, source terms arc computed for element failure subsequent to eight hours full-powcr operation per day for five days. For long-lived radionuclides, source terms are computed for clement failure subsequent to continuous operation for 40 years at the average power experienced by the reactor over its first 33 years of operation. Also examined arc residual sources still present in fuel, but generated in reactor operations prior to local receipt of the fuel in -

1973. Potential consequences of radiological releases are examined.

Under extremely conservative assumptions, potential release and dispersion of a few -

radioiodincs and radioactive strontium inventories exceeds activity levels that permit normal occupancy of the reactor bay. In a 100% release of the inventory of these species, the Derived Air Concentrations for these radioisotopes in the reactor bay could require worker access to be controlled to prevent exceeding the Annual Limit of Intake.

The K-Statc Reactor Emergency Plan controls responses to accidents involving fuel element failure as well as recovery and reentry operations. Unrestricted, uncontrolled, and unmonitored access to the reactor bay following a fuel handling accident is not permitted.

Under extremely conservative assumptions, only a few radionuclides exceed permissible concentrations for relcase to unrestricted areas, even within the reactor bay. Even in the cxtremcly unlikely event that 100% of the radionuclide inventory is released from a damaged fuel element to the outside atmosphere, very conservative calculations reveal that radionuclides inhaled by persons downwind from the release would lead to organ K-State Reactor 1-5 Original (12104)

Safety Analysis Report

THE FACILITY doses or effective doses very far below regulatory limits.

Even in the maximum hypothetical accident, no workers or members of the public are at risk of receiving radiation doses in excess of limits prescribed in federal regulations.

b. Loss of Coolant Although total loss of reactor pool water is considered to be an extremely improbable event, calculations have been made to determine the maximum fuel temperature rise that could be expected to result from such an event taking place after long-term operation at full power of 1,250 kW. Even under extraordinarily conservative assumptions and approximations, the maximum fuel te crature reached in a loss of coolant accident is less than 3001C, well below any safety limit for TRIGA reactor fuel.

Radiation doses from loss of coolant accident under extremely conservativc assum tions are computed and have been tabulated in Chapter 13. plr

c. Insertion of Excess Reactivity T vo reactivity accident scenarios are presented. The first is the insertion of 2.1% (S3.00) reactivity at zero power by sudden removal of a control rob The second is the sudden removal of the same reactivity with the core operating at a power level equivalent to the remaindei of the core excess reactivity. Analysis shows that peak fuel temperatures in the first case does not reach fuel temperature limits, with a maximum temperature less than 7501C at the peak in the hot channel for conditions where initial steady state power level is regulated only by the balance of core excess reactivity, while cladding temperature remains below 5001C. In the second case, maximum fuel temperature is calculated at a maximum of less than 870CC at the peak in the hoi channel, agaii with cladding temperature less than 500oC. _ IeI K-State Reactor 1-6 Original (12104)

Safety Analysis Report

CHAPTER 1 1.3 General Description of the Facility 1.3.1 Geographical Location The reactor is located on the campus of Kansas State University, in the City of Manhattan, in Riley County, Kansas. The licensee (through the University Police Dcpartment) controls access to Kansas State University facilities and infrastructure. County, city, and university maps are supplied in Chapter 2. The reactor is located in the north wing of Ward Hall. Latitudc and longitude, Ward Hall building plans, universal Transverse Mercator coordinates, population details, etc. arc provided in Chapter 2.

The operations boundary of thc reactor facility encompasses the reactor room and control room.

The site boundary encompasses the entire building and adjacent fenced areas controlled by management of the facility.

1.3.2 Principal Characteristics of the Site The site is in the Flint Hills uplands of northeast Kansas, characterized by glacial sediments. Soil borings reveal modest topsoil, varying levels of silt and clay loams overlying bedrock, limestone, and shale. Groundwater is encountered in sand or gravel layers 18 to 35 feet below existing grade.

The reactor site is located on high ground in the northwest sector of the university campus.

Climate is temperate, with typically 32 inches of rain annually. Storm drainage is excellent and sanitary sewerage from the reactor building is collected by a system serving the entire university. _

The site is in a seismic risk zone 2. Liquefaction potential of local soils is minimal.

1.3.3 Principal Design Criteria, Operating Characteristics, & Safety Systems The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool. The cactor is fueled with heterogeneous elements clad with stainless steel, consisting of nominally nriched uranium in a zirconium hydride matrix. In 1968, the KSUTMII was licensed to operate at a steady-state thermal power of 250 kW with a pulsing thermal power limit of 250 NW. Application is made concurrently with license renewal to operate up to a maximum steady state power level of 1,250 kNV steady-state thcrmnal powers and pulsing to $3.00 -

(nominal 1,340 MW eak power). Reactor cooling is by natural convection. The 250-kW core consists typically olflem ents (a minimum oflolanned for the l,250-kNV core), each containing as much asrms of 235U. The reactor core is in the form of a right circular cylinder ab6ut 9 in. (23 cm) radius and 15 in. (38 cm) depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.) diameter and 6.25 m (16 AL) depth. Criticality is controlled and shutdown margin assured by control rods in the form of aluminum or stainless-steel clad boron carbide or borated graphite. The 250 kW core originally used three control rods, the 1,250-kW core will be controlled by four. The reactor tank is surrounded on the side and at K-State Reactor 1-7 Original (12104)

Safety Analysis Report

THE FACILITY the base by a biological shield of reinforced concrete at least 8.2 it (2.5 mn) thick. The tank and shield arc in a 4078 m3 (144,000 ft3) dynamic confinement building made of reinforced concrete and structural steel, with .composite sheathing and aluminum siding. Sectional views of the reactor are shown in Figures 1.1 and 1.2, with a floor layout in Figure 1.3 showing thc Ofoot, 12-foot and 22-foot levels of the facility.

1.3.4 Engineered Safety Features The design of the KSU TRIGA Reactor, licensed in 1962, and the power upgrade to 250 kW in 1968 imposed no requirements for engineered safety features. As discussed in Chaptcr.13, and from previous analysis, neither forced cooling flow nor shutdown emergency core cooling is required for operation at steady state thermal power as high as 1900 kW, a large margin over the 1,250 kW steady state operations.

1.3.5 Instrumentation and Control (1&C) and Electrical Systems Instruments and controls are described in Chapter 7, with the electrical power system described in chapter 8. The reactor instrument and control systems include the reactor control system, process instruments, reactor protection system, and radiation safety monitoring systems. As previously noted, there are no engineered safety features at the KSUTMII and therefore no associated instrumentation.

aSkEAftI

.2g Figure 1.1, Cutaway Viewof the K-State Reactor.

K-State Reactor 1-8 Original (12104)

Safety Analysis Report

CHAPTER 1 The bulk of the reactor I&C systems arc hard-wired analog systcms (primarily manufactured by Gencral Atomics) widely used at various NRC-liccnscd facilities. A comprehensive upgrade of the rcactor's original rcactor control system and reactor protection system was accomplished during control room modifications in 1993 and 1994. The original console and vacuum tube instruments were replaced by a solid-state console (previously used in the U.S. Geological

.Survey's TRIGA Miark I reactor). New General Atomics neutronic measuring channels (N-1000 series) were installed. These channels have optically isolated outputs, allowing other devices to utilize the ncutronic data. Installation of a fourth control rod is planned for the near future, with design and hardware supplied by GA. Improvements to air monitoring were accomplished during a modification to the rcactor bay HVAC system in 1998.

a1. Reactor Control System The reactor control system includes the mechanical and electrical systems for control rod drives, and instruments that monitor control rod position. Each control rod can be independently manipulated by pushbutton console controls. One control rod can be operated in an automatic mode to regulate reactor power according to a manual setpoint, indicated power on the linear power level monitoring channel and a wide range power level monitoring channel (period) fecdback. The wide range power lcvcl monitoring channel of the reactor protection system provides interlock signals and actions to the reactor control system. The reactor control system is also interconnected to the reactor protection system through a manual scram bar above the control rod drive switches (allowing the reactor protection system to be actuated manually) and the automatic mode control (as described above).

b. Process Instruments Primary water temperature is measured in the water box and displayed on the console. A manometer indicates flow rate through the cleanup loop locally. Conductivity probes measure water purity at the entrance and exit of the cleanup loop. A level alarm/switch alerts the operator when the reactor pool water level is low. Primary and secondary flow rates arc indicated on local, monitors. The makeup water system is instrumented with a flow meterftotalizer to measure water added to the reactor pool or bulk shield tank. Lcvel switches provide indication for low secondary surge tank water level and high reactor bay sump levels. Fuel temperatures for two clemcnts can be monitored on the reactor control console and on an auxiliary panel; the fuel temperature indicator on the auxiliary panel provides input to the reactor protection system.

c. Reactor Protection System The reactor protection system is designed to ensure reactor and personnel safety by limiting parameters to operation within analyzed operating ranges. Process parameters that can automatically initiate reactor protection system actions include neutron level, rate of rise (period) and fuel temperature. Circuit provisions allow additional, external scrams to be installed when personnel, facility, or experiment protection might require rapid shutdown based on instrument-monitored parameters. A bar above the control rod drive switches allows the scram system to be actuated manually by the reactor operator at the controls.

K-State Reactor 1-9 . Original (12/04)

Safety Analysis Report

THE FACILITY 212 F 10 IXL lCOMM"

.JL ra U<3

._mbA- KVW4AL Figure 1.2, Top View of K-State Rcactor. '

Three neutronic instruments measure reactor power separately: a wide-range logarithmic channel, a multi-range linear channel, and a percent power channel. These provide at least two indications of reactor power from source range to power range. The nuclear instruments of the reactor protection system are integrated into the reactor control system through the automatic power level control system and through rod control interlocks. If a reactor pulse is performed another channel is added to the central thimble to record pulse data. One fuel temperature indicator has a trip relay built into the meter movemcnt for a high fuel element temperature trip. Most of the components of the reactor control system are located within the same enclosure as sections of the reactor control system, although the reactor safety system fuel temperature indicator is mounted on the auxiliary control panel. Sincc the core is cooled by natural convection, no engineered safety features are necessary for safe reactor shutdown.

d. Radiation Safety Monitoring Systems Radiation monitbrs are installed to monitor radiological conditions at the facility. One monitor is stationed on the top of the reactor, with a local, high range indicator and alarm (at 5 R/hr) to initiate evacuation of the reactor bay. One monitor is stationed at the control room door to the reactor bay, with a 2.5-mrnrerhr-alarm setpoint. Electrical connections are installed near each beani port, permitting control room and local K-State Reactor 1-10 Original (12104)

Safety Analysis Report

i

'I I

CHAPTER 1 JI I

indication of radiation levels near an open beam port. The remaining radiation monitors _,

have indicators and alarms both locally and at a central location in the control room.

e. Electrical Power Primary clectrical power is provided through the Kansas State University power grid, supplied by an on-campus plant and commercial generators. Main power lines traverse underground tunnels, inhibiting tampering. Loss of electrical power will de-cncrgizc the control rod drives, causing the rods to fall by gravity into the core and placing the reactor in a subcritical configuration. Since the core is cooled by natural convection, no emergency power is required for reactor cooling systems. Loss of electrical power does not represent a potential hazard to the reactor. Backup battery sysfems arc provided for cmergency lighting and the security system.

1.3.6 Reactor Coolant and other Auxiliary systems The reactor coolant and auxiliary systems arc very simple in design and operation. These systems arc required for operation, and not for safety. Many of these systems have been upgraded in recent years to permit extended full power operation of the reactor. Detailed descriptions of the -

coolant and auxiliary systems equipment and operation arc provided in Chapters 4, 5 and 13 of this report.

K-Stale Reactor 1-11 Original (12104)

Safety Analysis Report

THE FACILITY

a. Reactor Coolant System.

During full power operation, the nuclear fuel elements in (he reactor core are cooled by natural convection of the primary tank water. To remove this bulk heat to the environment, the primary water is circulated through a heat exchanger where the beat is transferred to a secondary cooling loop. A cleanup loop maintains primary water purity with a filter and denineralizer to minimize corrosion and production of long-lived radionuclides that could otherwise occur. The primary coolant provides shielding directly above the reactor core.

b. Secondary Cooling System The sccondary cooling system provides the inteifacc for heat rejection from the primary coolant system to the environment. The secondary system is an open system, with the secondary pump discharging through a primary-to-secondary heat exchanger, then through a forced-draft cooling tower. Water returns to an open surge tank (located in the reactor bay) by gravity.

c. Bulk Shield Tank.

The bulk shield tank provides radiological controls for an experimental facility, the thermalizing column. Irradiated fuel elements can be stored in the bulk shield tank, or alternately in dry fuel storage pits.

d. Makeup A distillation unit in a room adjacent to the reactor bay provides makcup water through a filter demineralizer unit to the reactor pool or bulk shield tank.

1.3.7 Radioactive Waste Management and Radiation Protection Operation of the K-State TRIGA reactor produces (low concentration) routine discharges of radioactive gases, periodic batch.discharges of (sometimes) slightly contaminated water to sewerage, and small quantities of solid waste. Details of the waste management and radiation protection procedures at the KSU TRIGA reactor are provided in Chapter 11 of this report.

a. Gaseous Waste Maintaining negative pressure controls concentrations of radioactive gases in the reactor bay during operations. An exhaust fan in the roof of the reactor bay, directly over the reactor pool, maintains negative pressure in the reactor bay to ensure that discharges are controlled under conditions of analysis.

K-State Reactor 1-12 Original (12104)

Safety Analysis Report

CHAPTER 1

b. Liquid Waste Liquid sources arc limited generally to tritium-bearing condensate water from the facility air handling system, and occasional rclcascs of tritium-bearing primary coolant from lcvel adjustments in the reactor tank or bulk-shicld tank. All reactor bay floor drains and the HVAC condensate drains discharge to a reactor bay sump. Contents of the reactor bay sump are sampled and assayed to assure limits for discharge arc met prior to discharge. Sump cffluent is filtered prior to discharge to meet NPDES requirements for discharge to campus sewerage. Liquid wastes are released through the sanitary sewerage system after filtration and assay for beta, gamma, and alpha activity.

c. Solid Waste Solid waste is very limited in volume and specific activity. Solid wastes include ion-cxchange rcsin used in reactor-watcr cleanup, contaminated tools, lab-warc, samples and sample handling material for completed cxperiments, and anti-contamination clothing associated with reactor experiments and surveillanec or maintenance operations.

Shipments of solid waste to commercial disposal facilities are made infrequently. Solid wastes shipments are coordinated with the University Radiation Safety Office, Division of Public Safety.

1.3.8 Experimental Facilities and Capabilities Standard experimental facilities in the KSU TRIGA reactor, as supplied by the vendor, General Atomics, include the central thimble, rotary specimen rack, pneumatic specimen tube, thcrmal/thermnalizing columns, and four beam tubes. Experimental facilities arc described in Chapter 10, with auxiliary systems supporting beam tubes in Chapter 9.

a. Central Thimble The reactor is equipped with a central thimble for access to the point of maximum flux in the core. The central thimble consists of an aluminum tube that fits through the center holes of the top and bottom grid plates terminating with a plug below the lower grid plate. The tube is anodized to retard corrosion and wear. The thimble is approximately 20 f (6.1 m) in length, made in two sections, with a watertight tube fitting. A removable screen covers the top end of the tube to allow gas relief and to prevent objects from falling into the reactor tank. Although the shield water may be removed to allow extraction of a vertical thermal-neutron and gamma-ray beam (not currently done at the KSU facility at the time this report was completed), four 0.25-in (6.3-mm) holes arc located in the tube at the top of the core to prevent expulsion of water from the section of the tube within the reactor core.

b. Rotary Specimen Rack A 40-position rotary specimen rack (RSR) is located in a well in the top of the graphite radial reflector. A rotation mechanism and housing at the 22-ft level of the reactor allows the specimen to be loaded into indexed positions and also rotation of samples for more uniform exposure across the set of co-irradiated samples. The RSR allows large-scale K-State Reactor 1-13 Original (12104)

Safety Analysis Report

THE FACILITY production of radioisotopes and for activation and irradiation of multiple material samples with neutron and gamma ray flux densities of comparable intensity.

c. Pneumatic Specimen Tube A pneumatic transfer system, permitting applications with short-lived radioisotopes, rapidly conveys a specimen from the reactor core to a remote rcceiver. The in-core terminus is normally located ini the outer ring of fuel-element positions.

d. ThermalTlhermalizing Columns The }ZSU TRIGA Reactor has two graphitc moderated experimental facilities for spectrally tailored experiments requiring well thermalized neutrons. The thermal column, shown in Figures 1.1 and 1.2, has a concrete door acting as shielding. The thernal column is accessible by winching the'door, mounted on rails. The space occupied by the rails is covered ivith steel plate when the .thermal column is not in use, shown as the rectangle attached to thereactor pedestal in the 0-foot section of Figure 13; towvirds the top of the figure. The other facility is the thermalizing column, shown in Figures 1.1 and 1.3. Shielding forthe thermalizing colurmn is provided by water in the bulk shield tank.

e. B3eam Tube Facilities The KSU TRIGA Rcact2r is provided with four beam tubes. Beam-tube sleeves are welded to the outside surface of the tank and extensions (on axis) arc welded to the inside surface. The beam tubes provide beams of neutrons and gamma rays for a variety of experiments. They also provide irradiation facilities for specimens as large as 6 in. (15.2 cm) in diameter. Tree of the beam tubes are aligned radial with respect to the center of the core. One of the radial beam tubes is aligned with cylindrical void in the reflector graphite, while the remaining radial beam tubes terminate at the outer edge of the reflector assembly. The fourth beam tube is oriented tangentially with the outer edge of the core.

1.4 Shared Facilities and Equipment K-State Reactor 1-14 Original (12104)

Safety Analysis Report

CHAPTER 1 1.5 Comparison with Similar Facilities Thc design of the fuicl for the KSU TRIGA is similar to that for fuiels used in 70 reactors in 24 nations (General Atomics July 1999 data). Of total number of rcactors, 45 are currently in operation or under construction with 40 rated for steady-state thermal powers of 250 kW or grcater, 22 at 500 kW or greatcr, and 20 at I MW or greatcr. Nine of thc larger power reactors are TRIGA Mark 31. The TRIGA Mark 11 design is a substantial fraction of the 70 reactors using TRIGA fuc1 world-wide.

In the United States, there have been 26 TRIGA reactors built, with 19 currently in operation (5 TRIGA facilities and 3 non-TRJGA reactors converted to operate with TRIGA fuel at power levels greater than 1,000 k}W, as indicated in Table 1.1).

Table 1.1, U.S. TRIGA REACTORS T 500 kV OR GREATER. I I1 Pullman, W~ashing~ton Statecovrin 100 20096 Washington Univcrsity Convcrsio l_,00_2000__96 Madison, University of Wisconsin Wisconsin Convcrsion 1,000 2,000 1967 CollUgc cexas AiM Mtation, Tcxas University Conversion 1,000 2,000 1968 Bethesda, Armed Forces Maryland Radiobiology Res. Mark F 1,000 3,300 1962 Inst. (AFRRI),

Denvcr, U.S. Geological Colorado Survey Mark I 1,000 1,200 1969 Corvallis, Oregon State Mark Il 1,000 3,200 1967 Oregon University University Pennsylvania Statc Park, University Mark III 1,000 2,000 1965 Pennsylvania Austin, Texas ivcrsity of Mark II 1,100 1,600 1992 California__ TCxas .M k1 201009 Sacramento, UC Davis Mark 11 2,300 1,200 1990 C allifo mnia __ _ __ _I_ _ _ _ I__ _ _ I__ _ _ I__

K-State Reactor 1-15 Original (12104)

Safety Analysis Report

vl1 THE FACILITY Table 1.2, U.S. MARK 11TRIGA REACTORS.T Urbana Illinois University of Illinois 1,500 6,500 Shutdown 1960 Ithaca, Newv York Cornell University 500 . 250 Operating 1962 Corvallis, Oregon UrniveorStyat ,000 3,200 Operating 1967 acramento, - McClellan AFB 2,300 1,200 Operating 1990 California III usin,Texas Pniversity of Texas I1,100 1,600 Operating 199 Major design paramctcrs for the KSU TRIGA are given in Tablc 1.3. Fuel for the KSU reactor is standard TRIGA fuel having^,of uranium, by weight, enriched up toJin the 23 5U isotope. TRIGA fuel is characterized by inherent safety, high fission product retention, and the demonstrated ability to withstand water quenching with no adverse reaction from temperatures to I 150'C. The inherent safety of TRIGA reactors has been demonstrated by extensive experience acquired from similar TRIGA systems throughout the world. This safety arises from the large prompt negative temperature coefficient that is characteristic of uranium-zirconium hydride fucl-nioderator elements used in TRIGA systems. As the fuel temperature increases, this coefficicnt immediately compensates for reactivity insertions. This results in a mechanism whereby reactor power excursions are limited/terminated quickly and safely. Table 1.2 indicates research reactors at 500 IW or above using TRIGA fuel in the U.S..

Maximum nulse reactivity .

Maximum excess reactivity Minimum shutdown margin Integral fuel-moderator material U.&I-11.,, IFltcrn-afl oderator I Lght water -I K-State Reactor 1-16 Original (12/04)

Safety Analysis Report

CHAPTER 1 1.6 Summary of Operations Reactor utilization since 1980 is summarized in Table 1.4.

Table 1.A, Tivo-decade operatino history for the KSU TRIGA Reactor Facility.

.kT 61xp -,

1981 398 30 8 62 1982 440 47 8 52 1983 501 58 10 51 1984 401 36 8 55 1985 410 36 8 20 1986 469 52 9 22 1987 441 43 9 18 1988 308 31 6 25 1989 257 26 5 27 1990 370 26 7 24 1991 411 35 8 38 1992 449 35 9 38 1993 299 34

37 1994 363 38 7 41 1995 394 27 8 29 1996 309 40 6 33 1997 388 28 8 36 1998 458 26 9 26 1999 346 25 7 10 UVilI:ation low in 1993 because ofremodeling, console replacement and coolingsystem replacement.

Since 1980, the following outside users madc direct use of the KSU TRIGA reactor services:

University of Kansas University of Ncbraska at Lincoln Purdue University Regional Kidney Disease Program -

0 Kansas Jr. Academy of Sciences National Transportation Safely Board Boeing Corporation 0 Wolf Creek Nuclear Operating Corp.

0 Armed Forces Radiobiology Res. Inst.

SE Kansas Agricultural Exp. Station Kansas Highway Patrol Washington University of Saint Louis State University of New York University of Chicago University of Nebraska at Lincoln K-State Reactor 1-17 Original (12104)

Safety Analysis Report

THE FACILITY Since 1980, the following outside university users made use of the KSU TRIGA reactor's neutron activation analysis services in cooperation with the Geology Department at Kansas State University.

0 University of Kansas 0

University of Southern California 0

University of Georgia S

Louisiana State University 0

Cincinnati University S

Nevada-Reno University aU Wichita State University Baylor Univcrsity S Carleton University I 1.7 Compliance with Nuclear Waste Policy Act of 1982 Compliance with Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 for disposal of high-level radioactive waste and spent nuclear fuel is cf~ccted through contract b'hacten Kansas State University and the U.S. Department of Energy. A copy of the fuel cycle assistance contract is found in Appendix B of this report.

Table ISMajor Facility Modifications.

__________Yai__ _l !.,tt'%nomja.. 4Vi4i!

1960 Authorization of Construction 1962 Initial Criticality, 100 kW 1968 License Amendment for 250 kWV and pulsing 1973 Replaced Mark 11 elements with Mark III 1975 Added new wing to Ward Hall Replaced secondary cooling system

-1993 Replaced reactor console Installed new power level detectors Enlargecd control room 1999 Changed reactor bay HVAC. from positive. pressure, distributed unit HVAC system to negative pressure confinement, recalculating HVAC 2001 Increased cooling towercapacity 2001_ _ Replaced heat exchanger 2004 Replace secondary pump 2005 (seduled) Install 4 control rod ZOOS_____________Conversion to 1,250 kW steady state, S3.00 pulsing license K-Stale Reactor 1-18 Original (12/04)

Safety Analysis Report

CHAPTER 1 1.8 Facility Modifications and History Criticality was first achicved on October 16, 1962 at 8:25 p.m. In 1968 pulsing capability was added and thc maximum steady-state operating power was incrcased from 100 kilowatts (kV) to 250 kW.

The original aluminum-clad fuel clements were replaced with stainless-stcel clad clcments in 1973. With support from the U.S. Department of Energy, coolant system replacement was complcted in 1993, as was replacement of the reactor operating consolc, and enlargemcnt and modernization of the reactor control room. All ncutronic instrumentation was replaced in 1994. -

thc secondary cooling system capacity was increased in 2001. Addition of a 4 control rod is scheduled for fall 2002. This description of major facility modifications is presented in tabular form (Table 1.5) to illustrate the timclinc. -

1.9 Bibliography Kansas State University TRIGA Mark}ll Reactor Hazards Summary Report, by R.W. Clack, J.R.

Fagan, W.R. Kimcl, and S.Z. Mikhail, License R-88, Docket 50-188, 1961.

Analysis of Certain Hazards Associated with Operation of the Kansas State University TRIGA Mark 1I Reactor at 250 kW Stcady State and with Pulscd Operation to $2.00, by R.W. Clack, ct .^

al., and thc Safety Evaluation by the U.S. Atomic Energy Commission Division of Reactor Licensing, License R-88, Docket 50-188, 1968.

NUREG-1282, "Safety Evaluation Report on fligh-Uranium Content, Low-Enriched Uranium-Zirconium Hydridc Fuels for TRIGA Reactors," U.S. Nuclear Regulatory Commission, 1987.

I K-State Reactor 1-19 Original (12104)

Safety Analysis Report

2. SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Site Location and Description

a. Specification and Location.

The reactor is located on the campus of Kansas State University, in the City of Manhattan, in Riley County, Kansas. It is located in the north Ying of Ward Hall, which faces onto 17th St., about 180 meters south of Clafim Rd. Latitude and longitude coordinates of the site are 3901 1'30" N, 96°352" W. The reactor site, in Universal Transverse Mercator coordinates, is Landmarks and nominal locations relativeto the reactor arc as follows:

Kansas River- 3 kan SE Tuttle Creek Reservoir- 7 km N Interstate Highway 70- 13 kn S State Capitol, Topeka - 82 km E Fort Riley Military Reservation- 12 km W Manhattan Airport - 9 km SW' U.S. Anny Marshall Field Airport- 22 km SW

b. Boundary and Zone Area Maps.

Manhattan is in Riley County, a northeast county in the State of Kansas.

The location of Manhattan relative to other counties in the State of Kansas is shown in Figurc 2.1 and Chapter 2, Appendix A, Figure 2A1, along with major bighway access to Manhattan.

Manhattan is in the southeast portion of Riley County, with panrs of the metropolitan area inside Pottowatomie County. The location of Riley County in Kansas is shown in Figure 2.1; with the location ofManhattan inside Riley County shown in Chapter 2, Appendix A, Figure A2.2.

The location of airports, waterways and public highways surrounding Manhattan is shown in Figure 2.2, and Chapter 2, Appendix A, Figure 2A3.

The location of the reactor near the center of the city of Manhattan is shown in Figure 23. Highways and aquatic features directly surrounding the city of Manhattan are shown in Figure 2A. The location of the reactor on the Kansas State University campus is shown in Figure 2.5

CHAPTER 2 Figure 2.1, Riley County in Relation to Other Counties in Kansas Figure 2.2, Manhattan and Surrounding Items or Interest.

K-State Reactor 2-2 Original (9102)

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CHARACTERISTICS Figure 23, Manhattan and Surrounding Highways, Streams, Rivers and Bodies of Water Figure 2.4, Location of the Reactor Within Central Manhattan.

K-State Reactor 2-3 Original (9102)

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IP

-I, CHAPTER 2 Kwsf. SIQa tritvmty Ca=ipa skp J1)

.J

-5

~11

-D

-5)

~j 19

~ji J

-5

-b J,

-5)

-5 Ij Figure 2.5, Reactor Facility Location within the Central Campus.

3 K-State Reactor 2-4 Original (9102) 31 Safety Analysis Report 3

no.

CHARACTERISTICS The first floor and basement plans for Ward Hall are displayed in Figures 2.6 and 2.7, with the "Reactor Bay" in Figure 2.8. For emergency planning purposes, the operations boundary encompasses the Reactor Facility, Room 110 of Ward Hall; the site boundary encompasses Ward Hall and any adjacent fenced areas, access to which and evacuation from which may be controlled by the management of the Reactor Facility.

I K-Stale Reactor 2-5 Original (9)02)

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CHAPTER 2 I

I -

K-State Reactor 2-6 Original (9102)

Safely Analysis Report

CHARACTERISTICS C

2.1.2 Population Distribution Manhattan, Kansas, home of Kansas State University and the TRIGA reactor, is in Riley County. Junction City, 29 km SW of Manhattan, is located in Geaty County. The town of Wamego, 22 km E of Manhattan, is located in Pottawotainie County. Population characteristics of the three counties are described in Table 2.1. A more detailed tabulation of data for Riley County is found in Chaptcr 2 Appendix B, based on year 1990 census data (year 2000 census data is only partially complete).

Chapter 2, Appendix A, Figures 2A.4 and 2A.5 identifies Ccnsus blocks and 1990 population densities in zones extending to i maximum of 12 kan and 8 km respectively from the TRIGA Reactor Facility. Thc census data do not allow for students rcsidcnt in dormitories on the university campus. As noted above, year 2000 census data is not completely analyzed for distribution but shows i slight decrease in population densities over areas closest to KSU. Population data, prepared for use in accident analysis and listed by radial sector and azimuthal sector, on the sixteen-point compass, arc listed in Table 2.2.

K-State Reactor 2-7 Original (9102)

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J

-I CHAPTER 2 Table 2.1, Population Characteristics of Riley, Geary, and Poltawatomic Counties.

Riley Geary Pottawatomic Families 13,450 (12,262) 8,191 (7,578) 4,390 (4,931)

Households 21,280 (22,137) 10,676 (10,458) 5,938 (6,771)

Urban population 49,743 21,287 3,849 Rural population 17,405 9,166 12,279 Total population 67,139 (62,843) 30,453 (27,947) 16,128 (18,209)

1990 U.S. Ccnsus Data, Database C9OSTFIC;Ycar 2000 Data in Parcnthcsis TABLE 2.2, Population Distribution Rladially from the Reactor Site.

IRadial distance from reactor (km)

Sector 0-1 1-2 24 4-6 6-8 8-10 10-12I Tntal N 91 0 182 107 372 42 0 794 NNE 9 485 1741 27 104 2 0 2368 NE 245 201 2032 0 41 180 6 2705 ENE 272 237 1187 271 372 749 503 3591 -

E 607 969 2238 179 156 388 1709 6246 ESE 439 2134 1863 287 325 88 150 5286 SE 412 2431 2036 14 41 10 68 5012 SSE 586 1900 558 0 0 35 0 3079 S 654 3331 682 77 49 0 0 4793 SSW 254 947 849 96 351 449 11 2957 SW 610 437 1412 101 1924 409 683 5576 WSW 645 448 158 0 57 663 0 1971 V 1601 2568 3542 546 33 611 0 8901 WNW 233 610 3741 388 25 271 360 5628 NWV 312 222 1096 11 17 161 78 1897 NNW 347 148 801 22 137 14 37 1506 Total 7317 17068 24118 2126 4004 4072 3605 62310 Source: 1990 U.S. census data, processed uising theArcVicev Graphical Information System.

K-State Reactor 2-8 Original (9102)

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CHARACTERISTICS 2.2 Nearby Industrial, Transportation, & Military Facilities 2.2.1 Locations and Routes Manhattan, Kansas, hosts light manufacturing and service industries, but no chemical plants or refineries, and no minifig or significant quarrying operations. There are no missile sites near the city and no docks, ports, or railroad yards. Natural gas pipelines are for local service, as are electrical distribution lines. The Manhattan airport, which provides general aviation and feeder-airline service is located 9 km from the nuclear reactor. Marshall field, on the Fort Riley military reservation, and 22 km from the nuclear reactor, is a base for rotary-wing army aircraft.

2.2.2 Air Traffic whilc the Manhattan airport is located less than 9 km from the nuclear reactor, actual and projected commercial or military aircraft movements arc far less than 16,000 annually.

2.2.3 Analysis of Potential Accidents Thcre are no nearby industrial, transportation, or material facilities that could experience accidents affecting the safety of the nuclear reactor.

2.3 Meteorology 2.3.1 General and Local Climate' Manhattan is located near the geographical center of the United. States, and the middle of the temperate zone. For the most part, the city is located along the north bank of the Kansas River, which flows in an easterly direction, and the west bank of the Big Blue River, which flows southerly into the Kansas River. The river valleys, two to four miles wide, are bordered by rolling prairie uplands. Flooding is always a threat along river valleys, but construction of a levee around the city of Manhattan, and tWe Tuttle Creek Reservoir, on the Big Blue River, has largely alleviated the threat. Kansas State University, site of the TRIGA reactor, is located approximately 25 meters in elevation above the rivers and has never been under threat of flood. Even during the 500-year flood of 1993,- which impacted the entire Midwest, the TRIGA reactor was never threatened. Flood waters never penetrated the reactor bay or the basement of Ward Hall.

Seventy percent of the annual precipitation normally falls during the months of April through September. The rains of this period are usually of short duration, predominantly of the thunderstorm type. They occur more frequently during the nighttime and early morning hours than at other times of the day. Excessive precipitation rates may occur.

1._

'Adapted from Local ClimatologicalDataforTopeka Kasna, supplied by the National Climatic Data Center, National Oceanic and Atmospheric Administration, Asheville. North Carolina, ISSN 0198-2192 (1995).

K-State Reactor 2-9 Origlnal (9102)

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CHAPTER 2 with warm-season thunderstorms. Rainfall accumulations of over eight inches in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months have occurred. Tornadoes have occurred in the area on several occasions and caused severe damage and numerous injuries.

Individual summers show wide departures from average conditions. Hottest summers may produce temperatures of I 000 F or higher on more than 50 days. On the other hand, 25 percent of the summers pass with two or fcwcr 100 0 F days. Similarly, precipitation has shown a wide range for June, July, and August, varying from under three inches to more than 27 inches during the three months. Summers are hot with low relative humidity and persistent southerly winds. Oppressive warm periods with high relative humidity are usually of short duration.

Winter temperatures average about 450 F cooler than summer. Cold spells arc seldom prolonged. Only on rare occasions do daytime temperatures fail to rise above freezing.

Winter precipitation is oflen in the form of snow, sleet, or glaze, but storms of such severity to prevent normal movement of traffic or to interfere with scheduled activity are not common.

In the transitional spring and fall seasons, the numerous days of fair weather arc interspersed with short intervals of stormy weather. Strong, blustery winds are quite common in late winter and spring. Autumn is characteristically a season of warm days, cool nights, and infrequent precipitation, with cold air inversions gradually increasing in intensity as the season progresses.

Nearly all crops of the temperate zone can be produced in eastern Kansas. Wheat and other small grains, clover, soybeans, fruit, and berries do evel, and the area supports extensive dairy and beef cattle operations.

Based on the 1951-1980 period, the average first occurrence of freezing in the fall is October 14, and the average last occurrence in the spring is April 21.

Historical data for severe weather phenomena during the 50-year period ending in 1995 are presented in Table 2.3. Except as indicated, data arc from the National Climatic Data Center, National Occanic and Atmospheric Administration.

2.3.2 Site Meteorology Multivariate frequency distributions for wind speed, wind direction, and atmospheric stability (Pasquill categories A through G) are listed in Chapter 2, Appcndix C. These data were processed from 1991-1996 data and were prepared especially for this report by the National Climatic Data Center, U.S. Occanic and Atmospheric Administration. They may be used in conjunction with population data presented in Section 2.1.2 to evaluate potential radiation doses associated with hypothetical accidental releases of radionuclides into the atmosphere. The wind-specd data arc summarized in a wind rose presented in Figure 2.9. The shading moves from calmest at center to strongest at the isobar lines near the edges of the construct. Using the scale and measuring from the perimeter of the Calm circle, fractional frequencies of occurrence for any direction on the sixtcen-point compass may be derived. Directions from which the wind arrives are shown.

K-State Reactor 2-10 Original (9102)

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CHARACTERISTICS Table 2.3, Severc Veather Phenomena for Topeka, Kansas.

Phenomenon Magnitude Date Mcan nind speed and direction 9.3 mph at 1800 Maximtim 2-minutc wind speed 44 mph at 340D June, 1984 Maximum 5-second wind speed 66 mph at 3500 June, 1984 Maximum 24-hour rainfall' .5.52 in. Junc, 1967 Maximum 24-hour snoWfall' 15.2 in. Fcbruary, 1971 Maximum snow depth 18 in. March, 1960

'Record-breaking incidences In Manhattan nre 6.28 in. rainfall In June, 1977 and 18.0 in. snowfall in February, 1900 [Goodin, ct aL, 1995J.

Topeka, Kansas 1992-1 996 N

w E Calm \

Ea 0-3 \

cm 4-6 7-10 11-16

> 16 knots 0 0.04 0.08 S

Figure 2.9, W~ind-Speed Frequency Data for Topeka, Kansas.

National Climatic Data Center, U.S. National Oceanic and Atmosphcric Adiministration K-State Reactor 2-11 Original (9102)

Safety Analysis Report

CHAPTER 2 2.4 Geology, Seismology, and Geotechnical Engineering Tcst drillings in the vicinity of the reactor reveal a thin layer of topsoil with varying levels of glacial deposits overlying bedrock, limestone and shale. Subsurface water is encountered 18 to 35 feet belowv existing grade in thin sand and gravel layers near the bedrock surface. The sand and gravel aquifer acts as a temporary trap that could, if charged with radioactively contaminated watcr, cventually discharge radionuclides into the water-bcaring veins underlying the Blue River valley. The city of Manhattan, as well as the university, draws water from wells in this same river valley. It is possible, but highly improbable, that a gross Icak of soluble radioactive materials into the ground at the reactor site could eventually deliver contamination to these wells. That such a leak could occur undetected, unchecked, and in sufficient magnitude to produce measurable contamination in a potable water supply appears to have a vanishingly small probability.

Operation, surveillance, monitoring, inspection, and auditing procedures in place at the Reactor Facility assure that encapsulated sources are regularly inspected and monitored, and that uncncapsulatcd, soluble radioactive materials are not held in inventory.

2.4.1 Regional Geology The general physiography of the region is illustrated in Chapter 2, Appendix A, 2A.6.

The city of Manhattan, in Riley County, is locatcd in the Flint Hills uplands. To the cast is a glaciated region, to the west the Smoky Ilills.

2.4.2 Site Geology As illustrated in Chapter 2, Appendix A, Figures 2A.7 and 2A.8, the university is located in a region of quajernary glacial-luvial unconsolidated sediment, which is characteristic of much of the city of Manhattan. To the cast and south, along the beds of the Big Blue and Kansas Rivers is also unconsolidated terrace deposits and alluvium beds of sediment.

The Flint Hills surrounding the city arc composed of limestones and shales of the Pcrnian Systcrn, shales interleaved with limestone, with regions of unconsolidated sediment along streams.

2.4.3 Seismicity As seen in Figure 2.10,OManhattan, Kansas is located in seismic risk zone 2, so classified because of several modified Mcrcali VII-VIII earthquakes that have occurred in the past.

Thirty felt earthquakes with epicenters in Kansas have been documented since 1867.

These are illustrated in Figure 2.11.

Earthquakes of modified Mcrcali intensity greater than IV arc identified in Figure 2.11.

It appears that quakes of intensity VI or greater occur irregularly at intervals of20 to 40 years.

K-State Reactor . 2-12 Original (9102)

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CHARACTERISTI CS Figure 2.10, U.S. Seismic Risk Map Source: 1997 Uniform Building Code.

K-State Reactor 2-13 OrigInal (9102)

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CHAPTER 2 j

EARTHQUAKES IN KANSAS 7 1; i-I4-

[;aIw z az ,a S

Zan a *m

- * *- , ~ ~ I

- S S S ~b Q 1 *o 0s i fA Explanation a 1867 VII f 1906 VII k 1927 V p 1929 V u 1933 V b 1875 V 9 1907 IV 1 1927 VI q 1929 V v 1942 IV c 1881 III h 1919 IV m 1928 IV r 1929 V w 1948 IV d 1903 11 1 1919 IV n 1929 V s 1931 VI x 1956 VI e 1904 IV 1 1926 ? o 1929 V t 1932 VI y 1961 V Location and dates of earthquakes In Kansas during the past 110 years. The number following the date Is the earthquake intensity on the Modified Mercalli Scale.

Figure 2.11, Rccorded Earthquakes in Kansas.

Source: DuBois and Wilson, 1978.

K-State Reactor 2-14 Original (9102)

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CHARACTERISTICS 2.4.4 Maximum Earthquake Potential According to Steeples et 2L. [1988], the most serious of the recorded earthquakes in Kansas were the intensity Vil events that occurred in 1867 and 1906 in the vicinity of Manhattan. Although the structures responsible for these two earthquakes have not been positively identified, the source of seismic activity appears to be movement on the Nemaha Ridge, a buried Precambrian granitic uplift, or the Humboldt Fault along the eastern boundary of the Ncmeha Ridge. The Midcontinent Geophysical Anomaly (MGA) is also known to be an important structural feature in the region: The MGA extends from Lake Superior southwestward through central Kansas into Oklahoma.

Surface structures associated with the MGA are present in the Manhattan area. These are the Abilene Anticline and the Riley County kimbcrlite intrusives. Work is under way by' the Kansas Geologic Survey, begun under the support of the Nuclear Regulatory

Commission, to gain a better understanding of the seismicity of the region and the link between the scismicity and the tectonic setting of the region. DuBois and Wilson [1978]

describe the most severe earthquake recorded in Kansas in Figure 2.11; an isoseismal map of the earthquake is illustrated in Figure 2.13 Intenslty ?lM VII-VIII Time 2:30 p.m., 24 April 1867 Epicenter Lat 39°1D', Long 96a18'near Alma, Kansas, 32 km SE orManhattan Felt area 300,000 sq. mi.

Figure 2.12, Ground-Acceleration Map for the Rcgion.

Source: Underwood, 1990.

K-State Reactor 2-15 Original (9102)

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-3 ii 31 CHAPTER 2 ISOSEISMAL MAP OF THE APRIL 24,1867 EARTHQUAKE IN KANSAS

'i 100° 98' 96 O

100 200 Ms J 0 100 200 300 Km J Figure 2.13, Mlap of the IMTost Severe Earthquake Recorded in the Region.

2.4.5 Vibratory Ground Motion Maximum accclerations havc bccn cstimated using methods of Algcrnnisscn and Perkins

[1976]. Figurc 2.13, presented by Underwood [1990] from the data of Algcnnisson and Perkins, is a map of potential maxima in thc region. Contour lincs represent a probability of no more than 10 percent in 50 years of exceeding in an acceleration expressed as percentagc of the acceleration of gravity.

K-State Reactor 2-16 Original (9102)

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CHARACTERISTICS 2.4.6 Surface Faulting There arc no known faults within 8 km of the TRIGA reactor site. Historically reported earthquakes in the region are previously enumerated.

2.4.7 Liquefaction Potential The phenomenon of soil liquefaction is associated primarily with medium to fine grained saturated cohesionlss soils [Das, 1993]. As indicated in the soil map presented in Chapter 2, Appcndix A, Figure 2A.8, such conditions are not met at the site of the TRIGA reactor. Soils are loams, silty loams, and clay loarns. Though the reactor is located some 25 mctcrs above the levels of local rivers, and though groundwater depths do vary, sandy soils in the saturated zone are not expected at this site; hence, the liquefaction potential of local soils is quite minimal.

2.5 Hydrology This topic was addressed in the Hazards Surmmary Report [Clack et al., 1961] reviewed by the Atomic Energy Commission prior to their granting the original 40-ycar license for operation of the reactor. Extracts from that report are repeated here.

The reactor site is located at an elevation of 1,082 feet above mean sea level, approximately 65 feet above the highest recorded flooding in the area. Average annual rainfall amounts to about 32 inches in the Manhattan area, with April, May, and June ordinarily being the wetter months.

Terrain features arc illustrated in Chapter 2, Appendix A, Figure. 2A.8.- Soils in the region are primarily silt and clay loamns of various classifications. The reactor site is in a region of the university campus thai is convex upward, a circumstance that minimizes the probability of local flooding. Surface water not absorbed drains into storm sewers on the campus. The storm sewer system runs through the city of Manhattan and discharges into the Kansas River south of the city. The Kansas River flows eastward through Topeka and Lawrcncc, joining the Missouri River at Kansas City.

2.6 Bibliography

Algermissen, S.T., and D.M. Perkins, 'A Probabilistic Estimate of Maximum Acceleration in Rock in the Contiguous United States," Open-file Report 76-416, U.S.

Department-of the Interior Geological Survey, 1976.

Clack, P.W., J.R. Fagan, W.R. Kimel, and SZ. Mikhail, "Kansas State University TRIGA Mark 11 RcactorfHazards Summary Report," LiccnscR-88, Docket 50-188, 1961.

Das, Brajma M., "Principles of Soil Dynamics," PWS-Kent Publishing Co., Boston, 1993.

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Safety Analysis Report

I

=1 i

CHAPTER 2 i

DuBois, S.M., and F.W. Wilson, "List of Earthquake Intensities for Kansas, 1867-1977,"

Environmental Geology Series 2, Kansas Geological Survey, Lawrencc, Kansas, 1978.

Goodin, D.G., J.E. Mitchell, M.C. Knapp, and R.E. Bivens, "Climate and Weather Atlas of Kansas," Educational Report Series 12, Kansas Gcological Survey, Lawrence, Kansas, 1995.

Steeples, D.WV., G.M. Hildebrand, B.C. Bennett, RD, Miller, Y. Chung, and R.W.

Knapp, "Kansas-Nebraska Seismicity Studies Using the Kansas-Nebraska Microcarthquake Network," Report NUREG/CR-5045 RA, US. Nuclear Regulatory Commission, Washington, DC, 1988.

NCDC, "Local Climatological Data for Topeka Kansas," National Climatic Data Center, National Oceanic and Atmospheric Administration, Ashevillc, North Carolina, ISSN 0198-2192, 1995.

Smith, B.D., and A.W. Archer, "Geologic Map of Riley County, Kansas," Map M-36, Kansas Geologic Survey, University of Kansas, Lawrence, Kansas, 1995.

Underwood, I.R., "Seismic Risk in Kansas and the Midwest," in "The Loma Pricta Earthquake: Observations and Implications," by Albert N. Lin, Report 217, Kansas Engineering Experiment Station, Kansas State University, 1990.

UBC1 997, "Structural Engineering Design Provisions," Uniform Building Code, Vol. 2..

International Conference of Building Officials, 1997.

USDA, "Soil Survey Geographic (SSURGO) Data Base," Publication 1527, United States Department of Agriculturc, Natural Resource Conservation Service, National Soil Survey Center, Fort Worth, Texas, 1995.

K-State Reactor 2-18 Original (9102)

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CHAPTER 2 Appendices A. Color Plates B.. Census Data C. Multivarate Meteorology Frequency-Distributions

Chapter 2 Appendix A: Color Plates Figure 2A.1, M1ajor Roads in Kansas.

K-State Reactor 2.A-1 Original (9102)

Safety Analysis Report

tI Figure 2A.2, 1Iighivay access to Manhattan from Topeka and Kansas City.

flBE, . 1 Y, %0 DW~l, I 02002 MapQuest.com, Inc. rI , J1i Is Figure 2A3, The city of Manbattan, Kansas State University, the TRIGA reactor and ncarby transportation features and waterways Persons per sq. km 20-100 an *100-30D 300-500 -

500-700 MX >700 1,2-km radial zones W+E S

K8.'.

a4 A 8 Ilbmeters f.t.M 1._

K-State Reactor 2.A-2 $ Original (9102)

Safety Analysis Report

Figure 2A.4, Population Density to About 12 Kilometers from the Facility.

J1 Persons per sq. km b0-10 E;j 10-20 gM 20-1 00 100-300 M 300-500 MM 500-700

-MMO >700 1,2-km radial zones N

S I Y/Il / a3.#..

1 0 1 2 Klbmetes _ v 0 Figure 2A.5, Population Density to About 4 Kilometers from the Facility.

MM adiedIogba o C~zftnam am Lc~oa~ad IM smo na;s Mr $MAYaI.tM.Iu EJ loR H0 ONQh MI ad"g C.afe.

.s Hb AM d w IO f Figure 2A.6, Gcneral Physiography of Kansas.

Source: Goodin et aL, 1995 K-State Reactor 2.A-3 Original (9/02)

Safety Analysis Report

."1.

p.

Figure 2A.7, Pbysiography ofRilcy County.

Source: Kansas Gcologic Survey Pottawatomle County Ciatermiry Systema ffbesse Ord Plelstocoe 3erles Pecst-iam bidn Wd older Eabriet PrwLllan System WiItCOMPlAh Stries P.ndylraslas System VirgIllaanSeries

-M Adntre CV E V&SaieOr" Figure 2A.8: Physlography of Pottavatorile County.

Source: Kansas Geologic Survey 4

K-State Reactor 2A-4 Original (9102)

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Reacdor site 1,2-km rings Solis Smolon sit loam aWymn ore silly day loam j lvaniA(ennebec sill bam Readng sll ba.m I Chse taly day boam M;3 Eudora sill ben bMit siltloam -

E;;21 Clm eSogn eom plex Tully Silly day loam Hayrie samdy borm Carr.Spy complex Slony eteep lnd Beni ekl.Florence com pleja Other looms Miscellhneous flWater

'E S

MI.

6"IbSurf0 S 0 5 ICI Klomeers Figure 2A.8, Terrain Surrounding the Facility.

I K-State Reactor 2A-5 Original (9/02)

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f .

Chapter 2 Appendix B Population Distribution in Riley County, Kansas 1990 US Census Data, Databasc: C90ST3B ZIP Code 66502: ZIP=66502 PERSONS Universe: Persons Total ......................... 50229 rAN5LIES--

Universe: Families Total ....................... 10598 HOUSEHOLDS Universe: Households Total ........................ 17977 DR2AN AMN RURAL Universe: Persons Urban:

Inside urbanized area ........................ .0 Outside urbanized area ............... . . . ..... 37744 Rural:

rarm ... 302 Nonfarm ................ 12183 SEX:

Universe: Persons Hale ... ,.26004 Female . . . 24225 RACE.

Universe: Persons White ... .......... 45178 Black ......................... 2296 American Indian, Eskimo, or Aleut . . ................. ..... 303 Asian or Pacific Islander . ................................. l99 Other race ............ ;553 AGE Universe: Persons Under 1 year .. ....................... . 645 1 and 2 years ...... 1306 3 and 4 years . . . ........... . 1163 5 years ........................... . . . . ....................................

546 6 years ...... 689 7 to 9 years .. ..... 1537 10 and 11 years .. .. .. 1095 12 and 13 years ...... 1032 14 years ...... .465 15 years ...... 467 K-State Reactor 1.Bt Original (9102)

Safety Analysis Report

16 years .... 429 17 years .... 395 18 years . . . .1............ 486 19 years . . . ............. 3133 20 years . . . . .......... 3277 21 years .... 3176 22 to 24 years ........ . . . 5531 25 to 29 years .... 5114 30 to 34 years .... 3749 35 to 39 years .... 3178 40 to 44 years...........................................................2771 45 to 49 years . . . . .......... 1867 50 to 54 years . . . . .......... 1187 55 to 59 years .... 1146 60 and 61 years. . . . .569X 62 to 64 years .............. 660 65 to 69 years .............. 1186 70 to 74 years............................................................850 75 to 79 years .............. 648 80 to 84 years ........................................................ ... 538 85 years and over ......... 394 SCHOOL ENROLLMENT AMD TYPE OF SCHOOL Universe: Persons 3 years and over Enrolled in preprimary school:

Public school ....................................... 634 Private school ....................................... 206 Enrolled in elementary or high school:

Public school ....................................... _5S7B7....

Private school ....................................... 219 Enrolled in college:

Public school ....................... 7438 Private school...................... 277 Not enrolled in school..................................................23717 EARNINGS IN 1989 Universe: Households =

With earnings .......................... ,5831 1

No earnings .......................... 2146 WAGE OR SALARY INCOME IN 1989 Uni verse: Households With wage or salary income.............................- 15494 No wage or salary income..: ...............................................2483_

AGGREGATE WAGE OR SALARY INCOME IN 1989 Universe: Households Total ... 393 136419 PER CAPITA INCOME IN 1989 Universe: Persons Per capita income in 1989 ......................... 10718 K-State Reactor 2.B-2 Original (9102)

Safety Analysis Report

Chapter 2 Appendix C Mctcorological Frequency Distributions Table B-I. Wind rose data excluding calms, for Topeka, Kansas, 1992-1996.

Wind speed (knots)

Direction 0-3 4-6 7-10 11-16 17-21 21-N 0.00740 0.01608 0.02543 0.03054 0.00698 0.00125 NNE 0.00523 0.01359 0.02044 0.01683 0.00299 O.OOOD NE 0.00515 0.01060 0.01633 0.00723 0.00099 0.00000 ENE 0.00693 0.01695 0.01609 0.00785 0.00025 0.00000 E 0.01956 0.02830 0.02556 0.00897 0.00000 0.00000 ESE 0.01822 0.02294 0.02469 0.00997 0.00012 0.00000 SE 0.01371 0.02880 0.02942 0.01247 0.00000 0.00000 SSE 0.00649 0.01309 0.02019 0.01296. 0.00050 0.00000 S 0.01333 0.02244 *0.04189 0.04837 0.00835 0.00050 SSW 0.00998 0.01396 0.01820 0.02368 0.00349 0.00087 SW 0.01018 0.01084 0.01010 0.01334 0.00174 0.00050 WSW 0.00664 0.00848 0.00685 0.00473 . 0.00050 - 0.00000 W 0.00798 0.01521 0.01296 0.00984 0.00187 0.00075 WNW 0.00921 0.01421 0.01434 0.01234 . 0.00212 0.00037 NW 0.00553 0.00910 0.01571 0.01209 0.00124 0.0000D NNW 0.00415 0.00810 0.01433 0.01259 0.00411 0.00175 Source: National Climatic Data Center, U.S. National Occanic and Atmospheric Administration, Station 13996, Topeka, Kansas, 1992-1996.

Table B-2. Wind rose cumulative data including calms, for Topeka, Kansas, 1992-1996.

Wind specd (knots)

Direction 0-3 4.6 7-10 11-16 21 21-N 0.01338 0.02946 0.05489 0.08543 0.09241 0.09366 NNE 0.01121 0.02480 0.04524 0.06207 *0.06506 0.06506 NE 0.01113 0.02173 0.03806 0.04529 0.04628 0.04628 ENE 0.01291 0.02986 0.04595 0.05380 . 0.05405 0.05405 E 0.02554 0.05384 0.07940 0.08837 0.08837 0.08837 ESE 0.02420 0.04714 0.07183 0.08180 0.08192 . 0.08192' SE 0.01969 0.04849 0.07791 0.09038 '0.09038 0.09038 SSE 0.01247 0.02556 0.04575 0.05871 0.05921

0.05921 S 0.01931 0.04175 0.08364 0.13201 0.14036 0.14086-SSW 0.01596 0.02992 0.04812 0.07180 0.07529 0.07616 SW 0.01616 0.02700 0.03710 0.05044 0.05218 00.05268 WSW 0.01262 0.02110 0.02795 0.03268 0.03318 0.03318 W 0.01396 0.02917

0.04213 0.05197 0.05384 0.05459 WNW 0.01525 0.02946 *0.04380 0.05614 0.05826 .0.05863 NW 0.01151 0.02061 0.03632 0.04841 0.04965 0.04965 NNW 0.01013 0.01823 0.03256 0.04515 0.04926 0.05101 Source: National Climatic Data Center, U.S. National Oceanic and Atmospheric Administration, Station 13996, Topeka, Kansas, 1992-1996.

K-State Reactor 2.C-1 Original (9102)

Safety Analysis Report

Table B-3. Rclativc frequency for winds under Pasquill stability category A, for Topeka, Kansas.

Wind spccd (knots)

Direction 0-3 4-6 7-10 11-16 17-21 21- J N 0.00045 0.00012 0.00000 0.00000 0.00000 0.00000 NNE 0.00013 0.00025 . 0.00000 0.00000 0.00000 0.00000 NE 0.00000 0.00000 0.00000 0.00000 0.00000 0.00000 ENE 0.00013 0.00025 0.00000 0.00000 0.00000 0.00000

£ 0.00013 0.00025 0.00000 0.00000 0.00000 0.00000 ESE 0.00077 0.00037 0.00000 0.00000 0.00000 0.00000 SE 0.00033 0.00062 0.00000 0.00000 0.00000 0.00000 SSE 0.00000 0.00D00 0.00000 0.00000 0.00000 0.00000 S 0.00019 0.00000 0.00000 0.00000 0.00000 0.00000 SSW 0.00007 0.00012 0.00000 0.00000 0.00000 0.00000 SW 0.00026 0.00012 0.00000 0.00000 0.00000 0.00000 WSW 0.00000 0.00000 0.00000

° 0.00000 0.00000 0.00000 W 0.00007 0.00012 0.00000 0.00000 0.00000 0.0000 WNW 0.00026 0.00012 0.00000 0.00000 0.00000 0.00000 NW 0.00026 0.00012 0.00000 0.00000 0.00000 0.00000 NNW 0.00020 0.00037 0.00000 0.00000 0.00000 0.00000 Source: National Climatic Data Ccnter, U.S. National Oceanic and Atmospheric Administration, Station 13996, Topeka, Kansas, 1992-1996.

Tablc B4. Relativc frequency for winds under Pasquill stability category B, for Topeka, Kansas.

Wind speed (knots)

Direction 0-3 4-6 7-10 11-16 17-21 21-N 0.00089 0.00100 0.00062 0.00000 0.00000 0.00000 _

NNE 0.00039 0.00087 0.00062 0.00000 0.00000 0.00000 NE 0.00102 0.00087 0.00100 0.00000 0.00000 0.00000 ENE 0.00045 0.00112 0.00100 0.00000 0.00000 0.00000 E 0.00122 0.00287 0.00150 0.00000 0.00000 0.00000 ES E 0.00108 0.00112 0.00125 0.00000 0.00000 0.00000 SE 0.00143 0.00125 0.00137 0.00000 0.00000 0.00000 SSE 0.00067 0.00175 0.00112 0.00000 0.00000 0.00000 S 0.00065 0.00187 0.00262 0.00000 0.00000 0.00000 SSW 0.00067 0.00137 0.00062 0.00000 0.00000 0.00000 SW 0.00038 0.00025 0.00050 0.00000 0.00000 0.00000 4SW 0.00036 0.00137 0.00037 0.00000 0.00000 0.00000 W 0.00067 0.00075 0.00037 0.00000 0.00000 0.00000 WNW 0.00102 0.00087 0.00112 0.00000 0.00000 0.00000 _

VNW 0.00039 0.00087 0.00075 0.00000 0.00000 0.00000 NNW 0.00032 0.00062 0.00075 0.00000 0.00000 0.00000 Source: National Climatic Data Center, U.S. National Oceanic and Atmospheric Administration, Station 13996, Topcka, Kansas, 1992-1996.

K-State7Reactor 2.C-2 Original (9102)

Safety Analysis Report

. @@s Table B-5. Relative frequency for winds under Pasquill stability category C, for Topeka, Kansas.

- Wind speed (knots)

.Direction 0-3 4-6 7-10 11-16 17-21 21-N 0.00059 0.00137 0.00337 0.00112 0.00025 0.00000 NNE 0.00029 0.00125 0.00511 0.00062 0.00012 0.00000 NE 0.00022 0.00062 0.00312 0.00037 0.00012 0.00000 ENE 0.00027 0.00224 0.00362 0.00050 0.00000 0.00000 E 0.00091 0.00175 0.00386 0.00087 0.00000 0.00000 ESE 0.00040 0.00100 0.00362- 0.00112 0.00000 0.00000 SE 0.00068 0.00212 0.00449 0.00100 0.00000 0.00000 SSE 0.00043 0.00125 0.00461 0.00137 0.00000 0.00000 S 0.00040 0.00212 0.00898 0.00337 0.00112 . 0.00000 SSW 0.00035 0.00175 0.00374 0.00162 0.00037 0.00000 SW 0.00054 0.00100- . 0.00249 0.00112 0.0b012 0.00000 WSW 0.00037 0.00075 0.00162 0.00037 0.00000 0.00000 W 0.00031 0.00137 0.00337 0.00062 0.00012 0.00000 WNW 0.00054 0.00212 0.00337 0.00037 0.00000 0.00000 NW

0.00015 0.00125 0.00187 0.00050 0.00012 0.00000 NNtV 0.00028 0.00112 0.00137 0.00050 0.00012 0.00000 Source: National Climatic Data Center, U.S. National Oceanic and Atmospheric Administration, Station 13996, Topeka, Kansas, 1992-1996.

Table B-6. Relative frequency for winds under Pasquill stability category D, for Topeka, Kansas.

Wind speed (knots)

Direction 0-3 4-6 . 7-10 11-16 17-21 21-N 0.00121 0.00711 0.01770. 0.02942 0.00673 0.00125 NNE 0.00081 0.00424 0.01184 0.01621 0.00287 0.00000 NE 0.00088 0.00362 0.00997 0.00686 0.00087 . 0.00000 ENE 0.00124 0.00598 O:01097 0.00735 0.00025 0.00000 E 0.00194 0.00910 0.01870 0.00810 0.00000 0.00000 ESE 0.00114 0.00636 0.01683 0.00885 0.00012 0.00000 SE 0.00132 0.00686 0.01970 0.01147 0.00000 0.00000 SSE 0.00096 0.00299 0.01047 0.01159 0.00050 0.00000 S 0.00097 0.00449 0.02082 0.04500 0.00723 0.00050 SSW 0.00036 0.00237 0.00898 0.02206 0.00312

0.00087 SW 0.00024 0.00112 : 0.00424 0.01222 0.00162 0.00050 WSW 0.00012 0.00125 0.00299 0.00436 0.00050 0.00000 W 0.00042 0.00299 0.00436 0.00922 0.00175 0.00075 WNW 0.00059 0.00337 0.00648 0.01197 0.00212 0.00037.

NW 0.00035 0.00224 0.01072 0.01159 0.00112 0.00000 NNW 0.00030 0.00324 0.00997 0.01209 0.00399 0.00175 Source: National Climatic Data Center, U.S. National Oceanic and Atmospberic Administration, Station 13996, Topeka, Kansas, 1992-1996.

K-State'Reactor 2.C-3 Original (9102)

Safety Analysis Report

Table B-7. Rclativc frequency for winds under Pasquill stability category E, for Topeka, Kansas.

Wind speed (knots)

Direction 0-3 4-6 7-10 11-16 17-21 21-N 0.00000 0.00212 0.00374 0.00000 0.00000 0.00000 NNE 0.00000 0.00287 0.00287 0.00000 0.00000 0.00000 NE 0.00000 0.00237 0.00224 0.00000 0.00000 0.00000 ENE 0.00000 0.00362 0.00050 0.00000 0.00000 0.00000 E 0.00000 0.00648 0.00150 0.00000 0.00000 0.00000 ESE 0.00000 0.00474 0.00299 0.0000D 0.000D0 0.00000 SE 0.00000 0.00760 0.00386 0.00000 0.00000 0.00000 SSE 0.00000 0.00274 0.00399 0.00000 0.00000 0.00000 S 0.00000 0.00536 0.00947 0.00000 0.00000 0.00000 SSW 0.00000 0.00262 0.00486 0.00000 0.00000 0.00000 SW 0.00000 0.00274 0.00287 0.00000 0.00000 0.00000 WSW 0.00000 0.00112 0.00187 0.00000 0.00000 0.00000 W 0.00000 0.00337 0.00486 0.00000 0.00000 0.00000 WNW 0.00000 0.00162 0.00337 0.00000 0.00000 0.00000 NW 0.00000 0.00150 0.00237 0.00000 0.00000 0.00000 NNW 0.00000 0.00150 0.00224 0.00000 0.00000 0.00000 Source: National Climatic Data Ccnter, U.S. Nalional Oceanic and Atmospheric Administration, Station 13996, Topeka, Kansas, 1992-1996.

Table B-8. Rclative frequency for winds under Pasquill stability catcgory F, for Topeka, Kansas.

Wind speed (knots)

Direction 0-3 4-6 7-10 11-16 17-21 21-N 0.00108 0.00436 0.00000 0.00000 0.00000 0.00000 NNE 0.00162 0.00411 0.00000 0.00000 0.00000 0.00000 NE 0.00144 0.00312 0.00000 0.00000 0.00000 0.00000 ENE 0.00126 0.00374 0.00000 0.00000 0.00000 0.00000 E 0.00302 0.00785 0.00000 0.00000 0.00000 0.00000 ESE 0.00329 0.00935 0.00000 0.00000 0.00000 0.00000 SE 0.00318 0.01035 0.00000 0.00000 0.00000 0.00000 SSE 0.00125 0.00536 0.00000 0.00000 0.00000 0.0000 S 0.00316 0.00860 0.00000 0.00000 0.00000 0.00000 SSW 0.00176 0.00573 0.00000 0.00000 0.00000 0.00000 SW 0.00159 0.00561 0.00000 0.00000 0.00000 0.00000 WSW 0.00101 0.00399 0.00000 0.00000 0.00000 0.00000 W 0.00133 0.00661 0.00000 0.00000 0.00000 0.00000 WNW 0.00168 0.00611 0.00000 0.00000 0.00000 0.00000 NW 0.00159 0.00312 0.00000 0.00000 0.00000 0.00000 NNWV 0.00066 0.00125 0.00000 0.00000 0.00000 .0.00000 Sourcc: National Climatic Data Ccntcr,.U.S. National Oceanic and Atmospheric Administration, -

Station 13996, Topeka, Kansas, 1992-1996.

K-State Reactor 2.0-4 Original (9102)

Safety Analysis Report

1*

Table B-9. Relative frequency for winds under Pasquill stability category G, for Topeka, Kansas.

Wind speed (knots)

Direction 0-3 4-6 7-10 11-16 17-21 21-N 0.00318 0.00000 0.010000 0.00000 0.00000 0.00000 NNE 0.00199 0.00000 0.C10000 0.00000 0.00000 0.00000 NE 0.00159 0.00000 0.C10000 0.0Q000 0.00000 0.00000

.ENE 0.00358 0.00000 D.CO00D 0.00000 0.00000 0.00000 E 0.01234 0.00000 0.0 10000 0.00000 0.00000 0.00000 ESE 0.01154 0.00000 D.010000 0.00000 0.00000 0.00000 SE 0.00677. 0.00000 0.0 D0000 0.00000 0.00000 0.00000 SSE 0.00318 0.00000 O.C 10000 0.00000 0.00000 0.00000 S 0.00796 0.00000 QO10000 0.00000 0.00000 0.00000 SSW 0.00677 0.00000 D.COJ0OO0 0.00000 0.00000 0.00000 SW 0.00717 0.00000 0.010000 0.00000 0.00000 0.00000 WvSW 0.00478 0.00000 0.010000 0.00000 0.00000 0.00000 W 0.00518 0.00000 0.0 10000 '0.00000 0.00600 0.00000 WNNW .0.0518 0.00000 0.0 0000 0.00000 0.ODDO 0.00000 NW- 0.00279 0.00000 0.0 10000 0.00000 0.00000 0.00000 NNW 0.00239 0.00000 0.0)0000 0.00000 0.00000 . 0.00000 Source: National Climatic Data Center, U.S. National Oceanic and Atmospheric Administration, Station 13996, Topeka, Kansas, 1992-1996.

Table B-10. Relative frecquency of occurrence of Pasquill stability categorics, for Topeka, Kansas.

Pasquill stability category Direction A B

C D E F G N 0.00057. 0.00251 0.00670 0.06342 '0.00586 0.00544 0.00318 NNE 0.00038 0.00188 0.00739 0.03597 0.00574 0.00573 0.00199 NE 0.00000 0.00289 0.00445 0.02220 0.00461 0.00456 0.00159 ENE 0.00038 0.00257 0.00663 0.02579 0.00412 0.00500 0.00358 E 0.00038 0.00559 0.00739 0.03784 0.00798 0.01087 0.01234 ESE 0.00114 0.00345 0.00614 0.03330 0.00773 0.01264 0.01154 SE 0.00095 0.00405 0.00829 0.03935 0.01146 0.01353 0.00677 SSE 0.00000 0.00254 0.00766 0.02651 0.00673 O.OD661 O.OD318.

S 0.00019 0.00514

0.01599 0.07901 0.01483 0.01176 0.00796 SSW .0.00019 0.00266 0.00783 0.03776 0.00748 0.00749 0.00677 SW 0.00038 0.00113 0.00527 0.01994 0.00561 0.00720 0.00717 WSSW 0.00000 0.00210 0.00311 0.00922 0.00299 0.00500 0.00478 W 0.00019 0.00179 0.O0579 0.01949 0.00823 0.00794 0.00518 WNW . 0.00038 0.00301, 0.00640 0.02490 0.00499 0.00779 0.00518 NW 0.00038 . 0.00201 0.00389 0.02602 0.00387 0.00471 0.00279 NNW -0.00057 0.00169 0.00339 0.03134 0.00374 0.00191 0.00239 Total 0.00608 0.04501 0.10632 0.53206 0.10597 0.11818 0.08639 Calms 0.00212 0.00611 0.00324 .0.00686 0.00000 0.01795 0.05934

'Relative friquence of calms distributed within stability category. . z Source: National Climatic Data Center, U.S.National Oceanic and Atmospheric Administration, Station 13996, Topeka, Kansas, 1992-1996.

K-State Reactor 2.C-5 Original (9102)

Safely Analysis 1Report

3. DESIGN OF STRUCTURES, SYSTEMS, &

COMPONENTS This chapter describes the principal architectural and engineering design critcria for the structures, systems, and components that arc required to ensure reactor facility safety and protection of the public.

The KSU TRIGA Mark II Nuclear Reactor Facility, which houses the TRIGA reactor, is located in a building constructed for that purpose. The building was constructed in two phases. The first phase was built in 1961 and was identified formally as "Nuclear Science and Engineering Laboratories." The building was named WVard Hall, to honor the late Professor Henry T. Ward, Head of the Chemical Engineering Department from which, in 1958, the Department of Nuclear Engineering was evolved. Construction was completed in 1972 for a major addition to the nuclear science and engineering facilities, known formally as the "Addition to Ward Hall." The original building and the addition are now identified collectively as a single building, namely, Ward Hall.

Building areas include office space, laboratory space, shop facilities, utility service areas, and classrooms, many of which support the activities of the Nuclear Reactor Facility. The TRIGA reactor itself establishes the fundamental requirements for two specific rooms, Room 110, the Reactor Bay, and Room 109, the Control Room. The geographic placement of Ward Hall and the TRIGA reactor are described in Chapter 2. Figures 3.1 and 3.2 illustrate layout of rooms in Ward Hall. The northern wing (Figure3.1) is the 1961 structure; the southern wing the 1972 addition.

3.1 Design Criteria 3.1.1 General Conditions The basic design goal of a TRIGA reactor is integrity of the fuel by cladding, acting as a physical containment system for fission products. Fuel design prevents the release of radioactive fission products during routine reactor operation and potential accident conditions. Limits on the amount of fuel loaded in the core'(i.c., reactivity) establish a maximum steady state and transient power levels. Maximum possible power levels limit maximum fuel temperatures, which arc the basic design constraints for the fulc. Design constraints are described in 35.1, Fuel System. Fuel design is detailed in Chapter 4, Reactor Description.

The reactor control system maintains safe shutdown conditions. Since operational limits prevent achieving conditions that could lead to fuel element failure, control system response speed is not significant to protection of fuel integrity. However, to ensure the control system is functioning normally, a test circuit measures rate of drop from full out to full insertion. Design constraints are discussed in 3.5.2, Control Rod Scram System; system design is discussed in Chapter 4,

Reactor Description and Chapter 7, Instrumentation and Control Systems.

Facility design also controls personnel exposure to radiation associated with use of the fuel in reactor operation, and the release of effluents such as radioactive gases during normal operation or potential accident conditions. Design of the reactor bay as an air confinement system protects K-State Reactor 3-1 Original (12104)

Safety Analysis Report

CHAPTER 3 operating personnel and the general public against any operational hazards such as release of air activation products, namely, 'Ar during normal operation and fission products, namely, halogens and noble gases, during accident conditions. . Design Constraints arc. discussed in 3.5.3, Confinement and Ventilation System. Release criteria are based on 10 CFR 20. Chapter 11, Radiation Protection Program and Radiological Waste Managemcnt, describe control of radiation levels.

Building and structure design for meteorological, hydrological, and seismic cffects are discussed in the following sections.

3.1.2 Architectural and Engineering Design Criteria The reactor vendor vas General Atomics Division of General Dynamics Corporation, San Diego, California. Architect-Enginecr was UeL C. Ramey and Associates, Wichita, Kansas. Reactor constructor was Holmes and Narvcr, Inc, Orange, California. The building code for the State of Kansas at the time of the construction of the original building was the National Building Code, latest edition, as adopted by the Kansas Legislaturc (according to the State Architect 196011961).

The Uniform Building Code (UBC) gradually replaced the National Building Code in the late 1960's through the early 1970's. The Design Architect of Record has identified the UBC as the code used to construct the 1972 addition to the original building. The UBC is still the primary code used by the State of Kansas. Table 3.1 lists all codes currently used by the State of Kansas.

All state property is exempt from all other local codes per state statute.

TIhe University Architect in oi0ice during construction of both the 1961 and 1972 structures has identified practice during the period of construction as developing building programs for each project; the programs outlined in broad terms the spaces required to be designed and any special requirements needed to comply with the applicable regulations (in this case, the Atomic Energy

Commission). Copies of the'programs are no longer available for review in the State archives.

Both the Architect-of-Record and the then-University Architect state that the State Architct's office did not have a detailed manual of its expectations of design firms during that time.

The design services provided by both the 1961 and 1972 design teams were those typically delivered by architects and engineers after WWII. The design criteria from the period were eventually codified in the National Building Code, the Uniform Building Code, etc. with augmentation through University programs for construction-on campus. However, formal design constraints of the 1961 structure are not available (the architect firm has ceased business, and the Architect-of-Record for the 1961 structure has since died). The Structural Engineer-of-Record for the 1961 building is no longer available; although his firm still performs design services in Kansas. A building program from the University described the design criteria to be used on the addition, according to the Architect-of-Record for the 1972 addition. The Structural Engineer-of-Rdcord for the 1972 addition has retired and closed his practice.

However, some of the structural design constraints were incorporated into building plans, and are therefore available forfreview. These design constraints include such items as the soil boring log, the allowable maximum soil pressure, drilled pier conditions and various guides for placement and splicing of reinforcing steel design stresses and loads for concrete and structural steel.

Selected design constraints applying to the 1961 structure, including the reactor bay and control K-State Reactor 3-2 Original (12/04)

Safely Analysis'Report

DESIGN OF STRUCTURES. SYSTEMS, AND COMPONENTS room are listed in Table 3.2. Dcsign constraints applying to the 1972 addition are available for inspection. A comparison bctwcen drawings for the 1961/1972 constructions and other morc reccnt campus constructions show striking similarities (as noted in section 3.4).

Table 3.1: Codes and their editions used by the State or Kansas.

1. Uniform Building Code, 1997

2. Uniforn Building Code Standards, 1997

3. Uniform .cchanical Code, 1997

4. Uniform Plumbing Code, 1997

5. National Electric Codc, 1999

6. Kansas State Boiler Code, 1988

7. American wvith Disabilities Act Accessibility Guidelines for Buildings and Facilitics (ADAAG), without the clevator excmption, published in the Federal Register 7-26-91 (KS.A. 58-1301 ct seq.). Uniform Fcderal Accessibility Standards apply to agencies covered by Section 504 of the Rehabilitation Act of 1973

8. ASI-IRAE/IES Standard 90.1-89

9. Kansas Fire Prevention Code

10. Undcnvritcr's Laboratories Fire Resistance Directory, 1997

11. National Fire Protection Association, National Fire Codes and Standards, Latest

Edition at Date of Original Contract

12. American Welding Society, AWS D-10.12-89, AWS D-1.1-96, AWS A-5.8-92, AWS D-10.80

13. Amcrican Institute of Steel Construction-Ninth Edition, as

14. required

15. American Concrete Institute, ACI Standards 318-95

16. Safity Code for Elevators And Escalators, ASME A17.1 Code, 1996

17. ASME Boiler and Pressure Vessel Code, 1995 K-State Reactor 3-3 Original (12104)

Safety Analysis Report

CHAPTER 3 A1 I

I ,W K-State Reactor 34 Original (i2104)

Safety Analysis Report

i DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS I K-Slate Reactor 3-5 Original (12104)

Safety Analysis Report

CHAPTER 3 Table 3.2: Selected design guides for the 1961 structure.

(G. Hartvell & Co*, Structural Engineers)

1. Allowable maximum soil pressure =4000 psf. Footings balance for DL

2. Piers shall be drilled to material capable of supporting 4000 psf. Bottom of piers shall be belled to indicated sizes and any loose material removed from bottom of excavation before placement of concrete.

3. Grade beams, wall footings, & tunnels may be poured to lines of next excavation.

4. Design stresses: concrete - 3000 psi min. at 28 days; reinforcing and structural steel -

20,000 psi min.

5. Design loads: roof LL-3D psf; floor LL- 100 psll loading dock -7000 lb.

concentrated load; crane capacity -7000 lb.

6. Concrete columns supporting roofare 12 in. x 12 in. w/4 #5 vertical and #2 tics @ 12 in.

centers.

NOTE (1): Nhls value Isextremely conservafve; she currentUniversyvArchitect hasperformedanIndependent calculationwing currentmethodolok showing afloorloadingof opproxlmately 350 psfis acceptable 3.1.3 Structural System Design of the 1961 Building The structural system of the 1961 building is primarily poured-in-place concrete except for the structural steel octagonal shaped dome over the actual reactor. The concrete foundation in the reactor bay has poured-in-place concrete walls l'-O' thick setting on a continuous footing of 2'-6" W x 1'-0" D. The floor slab is 4' thick, wire-mesh reinforced concrete. The main 1961 building has 30" diameter drilled piers that have a 66" diameter bell resting on bedrock. Connecting each pier is a grade beam upon which rest the remainder of the poured in place columns and beams supporting the main floor and roof structure. The 1961 building was inspected in 1999 for degradation, with no sign of structural movement or damage to the primary structured system of the building and reactor bay. The 1961 building is typical in its detailing and requirements to the poured-in place reinforced concrete structural systems still being designed in the year 2000. There are no evident signs of any special structural design constraints placed on the 1961 building by the Atomic Ener Commission.

The design of the exterior walls are masomy block infitling the area between the concrete columns with a limestone exterior face extending up to the roof. In the reactor bay the masonry, stone and window wall is 7-0" tall with the remainder of the walls and roof of the octagonal dome being clad in a metal insulated panel. Visual inspection of these components of the building indicate no signs of structural movement or damage. Except for the normal aging of the exterior stone, windows and related components exposed to the elements, the 1961 building in its entirety is in excellent structural condition.

K-Stale Reactor 3-6 .Original (12104)

Safety Analysis Report

- -i DESIGN OF STRUCTURES. SYSTEMS, AND COMPONENTS J1 3.1.4 Structural System Design of the 1972 Building The structural system of the 1972 building is poured-in-place concrete throughout the building. 3 The building has a full basement and one story above grade. The structural design criterion for indicates the building was designed to support additional floors that have never been construetcd.

The building, like its 1961 predecessor, is supported on piers that have been drilled into thc _

bedrock. These piers vary in diameters from 2'-' to 4'-O" depending upon location and loading condition. The extcrior foundation walls are generally 1'4" thick and extend approximately 15'-O" J below grade to the drilled piers. The foundation walls are heavily reinfiorced, as are the basement J level bcams and columns to support the anticipated loads of a three story (above-grade) building.

The basement floor is 6" thick and also heavily reinforced. A 28'-O" square area of the basement J

,was designed to function as a "hot cell' area with the sidewalls varying in thickness from 2-O" to J 5'-0" thick. Additionally a poured-in-placc roof structure above the "hot cell" was cast-in-place and measures S-O" thick. In the center of the "hot cell" room is a 5-0" diameter by 15'-O0 deep J well with a 1'-O" diameter by I0'-O stainless steel auxiliary well extending below the foundation of the first welL The lowest point of the auxiliary well is approximately 25'-O" below the basement floor levcl. The exterior of the foundation walls and underside of the basement floor were provided with damp-proofing and a foundation underdrain system. Ground water problems have not occurred in the main basement area. However, ground water is sometimes found in the main well and auxiliary wcll structures to varying depths depending upon subsurface water conditions.

The first floor and roof framing system is poured-in-place concrete. beams, columns and reinforced floor slabs. One limited area of structural steel beams, columns and barjoists exists in the small areas where the 1972 structure abuts the 1961 building. Visual inspection of the 1972 building in 1999 indicates no sign of structural movement or damage to the primary structural system of the building. This building like it's predecessor, is typical in its detailing and requirements to the poured-in-place reinforced concrete structural systems being designed in the year 2000. There arc numerous signs of special design constraints being placed on the 1972 building by the Nuclear Regulatory Commission. The construction documents contain extensive _

details on the construction of the 'hot cell" area as well as the immediately adjacent basement level laboratories for the "hot cell" lab, neutron generator lab, fuel processing lab, nuclear chemistry lab and radioisotope application lab. The design of the exterior walls matches those in _

the 1961 building-masonry block infihling the area between the concrete columns with a limestone exterior face extending up to the roof. Visual inspections of these components of the building indicate no signs of structural movement or damage.

3.1.5 Sanitary Sewer System Ward Hall and the TRIGA. Reactor are at the high end of a branc -sanitary sewer lini which serves the northern portion of the Kansas State University CamuL Mi Jhr, tjon the primary City o Manhattan sanitary sewer and exits campus. The City of Manhattan sanitary sewer continues on an cast/southeastcrly direction for approximately 7 miles exiting at the new water K-State Reactor 3-7 Original (12104)

Safety Analysis Report

CHAPTER 3 treatment facility constructed by the City of Manhattan in 1995. At this point, after treatment, the water is placed in the Kansas River. A map of the system appears in Figure 33.

3.1.6 Storm Sewer Ward Hall and the TRIGA Reactor are located along some of the higher land of the Kansas State University Campus. Storm sewers do not serve the building. All rain and storm water flows run on grade to the west and north, flowing onto adjacent campus streets. Storm water flows along these streets for approximately 114 mile to the north then 112 mile to the cast/northeast where they enter Campus Creek, an open year around creek flowing through campus. The creek flows in a southeasterly line until it reaches the border of campus where it goes underground into the City of Manhattan storm sewer system. The City of Manhattan storm sewer system extends underground approximately 6 miles to the Kansas River. A map of the system appears in Figure 3A.

3.2 Meteorological Damage The available design criteria of both the original building and its addition do not provide any specific insight into how historical data factors on wind velocity, gust factors, recurrence intervals, tornado loading or other factors may have entered into the original design team's consideration. However, in the 30-year life of the 1961/1972 buildings and the direct experience of the current University Architect the entire building complex has an excellent history in

withstandingmeteorological damage. In the decade ofthe 90's the building has withstood a snow in excess of 18 inches, rains that came at the equivalent level ofone 1000-year, two 500-year and many 100-year rainfalls, wind gusts in excess of 110 mph, nearby lightning strikes and severe hail. Except for the minor interior damage that would come from aging roofing materials, no effect to the building structure or any of its infrastructure system was noted. The current University Architect assesses that "It can be reasonably assumed based on the KISU TRIGA REACTOR building's performance that the orijinal design of the 1961 and 1972 structures were more than adequate for their intended use.'

3.3 Water Damage The University Architect notes that the niid 90s were exceptionally wet years culminating in the floods of 1993 when many areas of Manhattan were under water. This building is located on.

some of the highest ground in the city and campus. Only the reactor bay of the original 1961 structure was designed to be 12'.0w below grade as the remainder of the 1961 building was constructed 'on grade." However, the 1972 structure was designed to have a full and occupiable basement. During this period, when all of the Midwest was flooded for weeks on end, no difficulties were noted with water entry to the building from any point, at or below grade, except for the minor damaged areas in the roofing previously addressed. The University Architect believes that while the water table in the area of the building was higher during the period of the flood the location of the building on high ground helped keep underground water pressure from building up and seeping into the 1961 reactor bay and the 1972 foundation walls. Additionally, the existing sloped grade around both buildings allows surface water to run off quickly.

K-State Reactor 3-8 Original (12104)

Safety Analysis Report

il

'I) ii j1)

DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3) 3)

-I)-

A1)

J) j j,

3

-3 3

-J J

J K-State Reactor 3-9 Original (12104)

Safety Analysis Report

I ..

4.

CHAPTER 3 E

l I K-State Reactor 3-10 Original (12104)

Safety Analysis Report

DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS ii

.1 3.4 Seismic Damage J The structures associated with the TRIGA reactor were designed in accordance with codes and standards applicable for the seismic zone designation at time of construction. This ensures that the reactor can be returned to operation without structural repairs following an earthquake likely J to occur during the lifetime of the plant. Failure of the reactor tank and loss of the coolant in the event of a very large earthquake has been considered in Chaptcr 13 and the consequences found acceptable for the standpoint of public safety.

Boring logs for the 1961 structure and the 1972 structure arc available for inspection, as are logs from the 1999 Ackert Hall Addition site immediately Northwest of this building and slightly downhill. In the view of University Architect (Gerald Carter), the logs arc remarkably consistent.

All of these soil borings indicate a modest amount of topsoil, varying levels of glacial deposits overlying bedrock, limestone and shale. The 1999 Soils Report notes that, 'Cohcsionless materials are generally encountered at depths near the bedrock surface." The 1961 boring log notes this level near the bedrock surface as being sand and gravel, the 1969 boring logs make the same comparison. Typically, this is the level *where subsurface water is cncountcrcd. The 1999 Soils Report notes that ground water was encountered at depths ranging from 18'-O" to 34'-6" below the existing grade. This reflects the ground water level typically found in this area of campus.

The 1999 Soils Report details the reconmmendations for design of a major multistory addition to the primary biology teaching and research facility, Ackert Hall, on campus. It details the most current requirements for design and construction of structural systems that are typical to those used in the design and construction of the original 1961 building and the 1972 addition. With the sole exception of a 1999 recommendation to use a below grade vapor and drainage system for the foundation floors and walls, a preventative measure now being used around many major structures after the 1993 floods, there are few differences to be noted. The similarity of the 1999 requirements to the 196111972 design criteria and actual design is striking to Mr. Carter. The original structural designs of the G. Hartwcll and Joseph C. Weakly firms are remarkably similar to the current structural designs of the Ackert Hall Addition. Chief among these similarities is the use of drilled piers extending down to the limestone bedrock. In order to absorb the type of live and dead loads typically found in major structural systems most Kansas State University buildings have utilized drilled piers being cut into the limestone bedrock.

This resolves not only the building loads but also the additional requirements placed on structures for earthquake loads and potential liquefaction problems. However, the potential for liquefaction is so remote in this area that it is not mentioned in the 1999 soil report, nor in any other buildings soil report on the main campus property.

The design recommendations for spread footings, such as those found in the reactor bay are also very similar to those used in the Ackcrt Hall Addition project as well as other adjacent projects.

The only area of difecrencc noted is that current guidelines call for 18' of fill materials is now placed under basement slabs instead of the 2" to 6" typically used in the 1950's and '60's. This difference reflects both the experiences with the recent floods plus the knowledge that some soils in the immediate area are subject to volume change (shrink/swell) with variations in moisture content. It does not appear that the Hartwell firm considered expansive clay soils of its 1961 K-State Reactor 3-11 Orginal (12104)

Safety Analysis Report

CHAPTER 3 design. However, in 39+ years there is no evidence of distress or movement caused by expansive soils anywhere around the original building or reactor bay. Given the recent climnatological conditions of drought, flood, and then drought again there has been ample time for any soils problem to present itself.

3.5 Systems and Components

The reactor facility design uses a defense in depth concept to reduce and control the potential for exposure to radioactive material generated during reactor operation. Fuel cladding -is the principal barrier to the release of radioactive fission products. Shielding (including biological shielding, reactor pool water and bulk shield tank witer) controls exposure of personnel to radiation associated with operation of the reactor (during operations, and also to activated material). The control rods assure safe shutdown conditions are maintained when reactor operation is not required. If radioactive material releases associated with reactor operations occur, a dynamic confinement system controls release to the environment.

Cladding integrity is ensured by the fuel system (fuel rod and core design). Fuel cladding surrounding individual fuel elements is the primary barrier to the release of radioactive fission products. The fuel system maintains cladding integrity through interrelated limits on temperature, reactivity, and power to ensure cladding integrity is not capable of being challenged.

Shutdown reactor conditions are initiated and maintained by the control rod scram system. Since inherent shutdown mechanisms of the TRIGA prevent unsafe excursions, the TRIGA system does not rely on speed ofcontrol as paramount to the safety~of the reastor. The control system ensures reactor shutdown conditions, and controlling power level during operation.

The confinement and ventilation system maintains airflow from the reactor bay to induce a negative air pressure in the bay. A confinement exhaust fan located in the center of the confinement dome normally operates during operation to provide the negative air pressure.

3.5.1 Fuel System Production of hydrogen from a high temperature zirconium-water reaction is a well-known phenomenon. Zirconium hydride does not exhibit the same chemical reactivity as zirconium, and this reaction is not an issue for TIGA fuel.

The reactor design bases are predicated on the maximum operational capability for the fuel elements and configuration described in this reporL The TRIGA reactor'system his three major and interrelated areas that are used to defne the reactor design bases:

a. Fuel temperature,

b. Prompt negative temperature coefficient,

c. Reactor power.

Of these three, fuel temperature is the most amenable basis for establishing design constraints fuel temperature is a limit in both steady state and pulse-mode operation. A summary is presented below of the conclusions obtained from the reactor design bases described in this chapter.

K-State Reactor 3-12 Original (12/04)

Safety Analysis Report

DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS Fuel growth and dcformation can occur during nornal operations at steady state fuel temperatures _1 of 750 0C or greatcr, as described in Gencral Atomics technical report E-1 17-833. Damagc J mcchanisms includc fission recoils and fission gases, strongly influenced by thermal gradients.

Operating with maximum long-term, steady state fucl temperature of 750°C does not havc I significant timc- and tempcrature-dcpcndcnt fucl growth. Since the KSU reactor will not be operated in the regime vulncrablc to this degradation, the damage mcchanism is not applicablc.

Thc fuel temperature limit for pulsed mode operation is related to outgassing of hydrogen from the fuel and the subsequent stress produced in thc fuel clement clad material at elevated temperatures. The strength of the clad as a function of temperature can set the upper limit on the fuel temperature. Fuel temperature limits of 1150°C (with clad < 5000 C) and 950°C (with clad >

5000C) for U-ZrH (HIZr,.&s) have bccn set to preclude the loss of clad integrity (NUREG 1282).

The basic parameter that provides the TRIGA system with a large safety factor in steady-state operation and under transient conditions is the prompt negative temperaturc coefficient that is rather constant with tempcrature (-0.01% AkMkC). This cocfficicnt is a function of the fuel composition and core geometry. As power and temperature increase, matrix changes cause a shift in the neutron cncrgy spectrum in the fuel to higher energies. The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the power increase.

Thc KSU TRIGA reactor, operati at 500 ckW thermal power, is designed to use stainless-steel clad TRIGA fucl clcmcnts, with fperccnt by weight uranium in a ZrHIA matrix, the uranium enriched up to crcent in 35 U. Element outside diameter is'1i. cmn) and cladding thickness isin. _ m ). Details of the fuel system are presented in Chapter 4, as is fuel thermatchanictcris!ics during steady-state and pulse-mode operations. Fuel system behavior in accident conditions is addressed in Chapter 13.

Therefore, Mhc basic safcey limit for the TRIGA reactor is lhe limit on the fuel temperature for both steady state and pulsed-modc operations. Tcmperature is limited to prevent fuel expansion through phase changes, and to prevent gas pressure buildup.

a. Potential for Zr-WVater Reaction Among the chemical properties of U-Zrll and ZrH, the reaction rate of the hydride with water is also of interest. Since the hydriding reaction is cxothermic, water will rcact more readily with zirconium than with zirconium hydride systems. Zirconium is frequently used in contact with vatcr in reactors, and the zirconium-water reaction is not a safety hazard. Experiments carried out at GA Technologies show that the zirconium hydride systems have a relatively low chemical reactivity with respect to water and air.

These tests have involved the quenching with water of both powders and solid specimens of U-ZrH after heating to as high as 850°C, and of solid U-Zr alloy after heating to as high as 1200'C. Tcsts have also been made to determine the extent to which fission products arc removed from the surfaces of the fuel elements at room temperature.

Results prove that, because of the high resistance to leaching, a large fraction of the fission products is retained in even completely unclad U-ZrH fuel.

K-State Reactor 3-13 Original (12104)

Safety Analysis Report

1..

CHAPTER 3 Therefore, temperatures and chemical reactivity of TRIGA fuel matrix ensure that a zirconium water reaction will not occur at magnitudes that could cause hazard to the reactor. Additionally, a large fraction of fission products will be retained in the matrix.

b. Phasc/Volumc Changes Two limiting temperatures are of interest, depending on the type of TRIGA fuel used.

The TRIGA fuel that is considered low hydride, i.e. with an H/Zr ratio of less than 1.5, has a lower temperature limit than fuel with a higher H/Zr ratio. Figure 3.5 indicates that the higher hydride compositions are single phase, not subject to the large volume changes associated with the phase transformations at approximately 5301C in the lower hydrides.

Also, it has been noted [Merton 19623 that the higher hydrides lack any significant thermal diffusion of hydrogen. These two facts preclude concomitant volume changes.

The important properties of delta phase U-ZrH are given in Chapter 4, Reactor Description, Table 4.1.

For the rest of the discussion of fuel temperatures, we will concern ourselves with the higher hydride (H/Zr > 1.5) TRIGA fuel clad with 304 stainless steel (0.020 in. (0.5083 mm) thick, or a cladding material equivalent in strength at the temperatures discussed. At room temperature the hydride is like a ceramic and shows little ductility. However, at the elevated temperatures of interest for pulsing, the material is found to be more ductile.

The effect of very large thermal stress on hydride fuel bodies has been observed in hot cell observations to cause relatively widely spaced cracks which tend to be either radial or normal to the central axis and do not interfere with radial heat flow. Since the segments tend to be orthogonal, their relative positions appear to be quite stable.

Therefore, volume and other physical changes associated with phase change or mechanical deformation of the fuel at high temperatures for the fuel in use at the K-State reactor do not have the potential to mechanically challenge the cladding or affect the ability of the fuel to transfer heat.

c. Internal Fuel Rod Pressurc The limiting effect of fuel temperature is hydrogen gas overpressure. Figure 3.6 relates equilibrium hydrogen pressure over the fuel as a function of temperature for H/Zr ratio of 1.65.

The hydrogen gas over pressure is not in itself detrimental; however, if stress produced by gas pressure within the fuel exceeds ultimate strength of the clad material, a rupture of the fuel clad is possible. While the final conditions of fuel temperature and hydrogen pressure in which such an occurrence could come about are of interest, the mechanisms in obtaining temperatures and pressures of concern are different in the pulsing and steady-state mode of operation, and each mechanism will be discussed independently of the other. In this discussion it wivll be assunied that the fuel consists of U-ZrH (H/Zr 1.65) with the uraniumbeingv,63 %, and further that the cladding can is 304 stainless steel. The clad thickness i vith an inside clad diameter of

_ 9 Ihese fuel parameters have been chosen since they represent the nominal specifications for TRIGA fuel elements. Figuri 3.7 shows 304 stainless-steel yield and K-State Reactor 3-14 Original (12104)

Safety Analysis Report

I A

DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS ultimate strengths as a function of tcmpcraiure. In determining hoop stress on the cladding from the internal hydrogen gas pressure the following equation applies [larvey 1974]:

S= rP I

I I-H:Zr Aom Raio Figure 3.: Zr-Hydride Phase Diagram in which S is the stress, P is the internal pressure (in thesame units as the stress), r is the (inside) radius of the cladding, and I is the cladding thickness. Thus, for the TRIGA cladding, S 36P.

It is of interest to relate the strength a f the clad material at its operating temperature to the stress applied to the clad from the internal gas pressure associated with the fuel temperature. Figure 3.8 illustrates the stress applied to the clad as a result of hydrogen dissociation for fuel having a H/Zr ratio of 1.6 to 1.7 as a function of temperature.

There are several mitigating factors that would cause gas pressure to be lower for transient conditions as compared to (predicted) equilibrium values. For cxamplc, the gas diffusion rates are finite; surface cooling is believed to be caused by endothermic gas emission, which tends to lower the diffusion constant at the surface; re-absorption takes place concurrently on the cooler hydride surfaces away from the hot spot; there is K-Slate Reactor 3-15 Original (12104)

Safety Analysis Report

CHAPTER 3 evidence for a low permeability oxide film on the fuel surface; and some local heat transfer does take place during the pulse time to cause a less than adiabatic true surface temperature.

25Iff W TMUIPUTEII - at Figure 3.6: Pressure versus Temperature for ZrHis

The limiting design parameter for TRIGA fuel is therefore related to buildup of pressure through disassociation of hydrogen in the matrix. This free hydrogen is heated by fuel temperature, causing a buildup of pressure on the internal surfaces of-the cladding. As illustrated in Figure 3.8, cladding temperatures exceeding about IODD0C (based on a H/Zr ratio of 1.6 to 1.7) has potential to lead to pressures that exceed ultimate tensile strength.

K-State Reactor 3-46 .Origlnal (12104)

Safety Analysis Report

DESIGN OF STRUCTURES, SYSTEMS. AND COMPONENTS

d. Conclusion Calculations show that a fully bonded fuel element (i.e., cladding temperature at fuel temperature) will not fail at fuel temperatures below about 1000°C. Thereforc, design limits for the TRIGA fucl are based on fuel temperature.

to, I

V

( A0 I-o I am

_ID*00 IN 1_00 I:

TEMPERATURE rci Tb[WRAWC I*')

Fig. 3. 7: Temperature dependent stress vs. Figure 3.8: Stress Versus Temperature UTS for type 304 SS cladding At Hydrogen Compositions 3.5.2 Shielding Design bases for TRIGA shielding derive from General Atomic shielding design analysis for the I .MW reactor, which is similar construction and dimensions as the KSU TRIGA reactor. Design basis radiation levels at the core level and at I MW, are as follows:

= 1 Mrad/h at the core boundary

40 rad/h at the tank boundary

< l mremlh outside the biological shield Design requirements allow access through the shielding to experimental areas, and permit extracting beams of radiation form the shielded volume into the reactor bay.

3.53 Control Rod Scram System The KSU TRIGA reactor, operating at 500 kW thermal power, is designed to be operated with three standard and one transient (pulsing) control rod. Standard rods are nominally 0.875 in.

(2.22 cm) outside diameter, the pulse rod is nominally 1.25 in. (3.18 cm) outside diameter. Both K-State Reactor 3-17 Original (12104)

Safety Analysis Report

CHAPTER 3 are 20 in. (50.8 cm) long and are clad with 30 mil (0.0762 cm) aluminum. The control material is either boron carbide or borated graphite. During operation, standard rods are held in place by electromagnets, the pulse rod by air pressure. All are manually withdrawn or inserted by motor-driven gear mechanisms. .Upon a scram signal, power to electromagnets is interrupted and air pressure is vented, with all control rods descending by gravity into the core. Standard rods have a maximum drop time of I second and the pulse rod has a comparable drop time. Details of the control rod scram system are addressed in Chapters 4 and 7.

3.5.4 Confinement and Ventilation Systems The confinement and ventilation systems are intended to control the level of airborne radioactive contaminants in the restricted area, and to release reactor bay air in unoccupied area at the top of the confinement structure.

The reactor bay is surrounded by a roughly dome shaped building, as shown in Figure 3.9. As described in Chapter I1, Appendix Ai, the reactor bay dimensions are approximately as follows:

The reactor bay (Ward Hall Room 110) is approximated as a right circular cylinder 36 ft (10.973 m) high and 36.68 ft ( 1.18 m) radius. The reactor vessel structure is approximated as a right circular cylinder, co-axial with the bay, 22 ft (6.706 mn) high and 11 t (3.3528 m) radius. The free volume is 144,000 fO (4078 fn3 ). P-

_bc The reactor bay is maintained at a slight negative pressure; the measurement is obtained across the separation of the control room and the reactor bay. __

_t Figurc 3.9: ReactorBay Confinement Structurc K-Stale Reactor 3-18 Original (12104)

Safety Analysis Report

DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS J j

In addition to artificial lighting, a series of plate glass windows surround the reactor bay, allowing I ambicnt light into the reactor bay. A security fence rcstricts access to the windows from the outside.

__ I The ventilation system for the reactor bay exhausts from a roof vent approximately II m above J grade. As discussed in Chaptcr 11, Appendix A, this discharge during operation maintains a slight negative pressure on the reactor bay, and controls 4"Ar concentrations within the bay and at the site boundary to within all applicablc limits.

The control room has a separate fan system that can be used on demand for ventilation. The control room has a window to the environment that may be opened at discretion of the facility staff.

3.6 Bibliography Wuclekar Science andEngineeringLaboratories,Kansas State University, 1961. " Architectural Plans and Specifications for construction of a building prepared by Uel C. Ramey, & Associates, Associate Architcct, Ramey, UcI C., Architcect-ot-Rccord IauclearScienceandEngineeringLaboralories,KansasStale University, 1961,N Structural Plans and Specifications for construction of a building prepared by G. Harvrcll & Co.Baxter, Leslie K, Structural Engincer-of-Rccord

'NuclearScience and EngineeringLaboratories,KansasState University, 1961, " Office of the Statc Architect. Representing Statc of Kansas in all Design and Construction projects, circa 1961.

Canole, James, State Architect,

.4dditIon to JMardHall, KansasState University, Dec. 31, 1969, "Architcctural Plans and Specifications for construction of a building addition prepared by Brown Slcmmons Krueger, Associate ArchitcciKrucger, Louis, Architcct of Record, Addition to IardHall, KansasState University. Dec. 31, 1969," Structural Plans and Specifications for construction of a building prepared by Joseph C. Weakly, Consulting Engineer, Weakly, Joseph C., Structural Engineer-of-Record "Addition to IWard Hall, KansasState University, Dec. 31,1969,"Officc of the State Architect.

Representing State of Kansas in all Design and Construction Projects circa 1969-70. Hale, William R. State Architect Cool, Vincent J., University Architect, Kansas State University, circa 1958-1987.

t The main room is 11I'-3" by 22'. The foyer is 6'-5" by 6'.

K-State Reactor 3-19 Original (12104)

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CHAPTER 3 Building Code of the State ofKansas, KSU Campus SanitatySewver Map andKSU Campus Storm Sewer Map, February2000.Carter,Gerald R. Director of Facilities Planning/University Architect, Kansas State University. 1993 to date.

"GeotechnicalEngineeringReport, Ackert HallAdditionA-8542, Kansas State University, Manhattan,Kansas, November 10, )999, HTeracon; Pxeparcd by Michael D. Barnet, PG on behalf of the firm.

E-117-833, "The U-ZrHlA4Ioy: JIIProperfiesand Use in TRIGA Fuel, "GA Project No. 4314, M.

T. Simnad, Fcbruary 1980

. . J..

K-State Reactor 3-20 Original (12104)

Safety Analysis Report

4. Reactor Description 4.1 Summary Description 31 The Kansas Statc Univcrsity (KSU) Nuclear Reactor Facility, operated by the Dcpartment of Mechanical and Nuclear Engineering, is located in Ward Hall on the campus in Manhattan, Kansas. The Dcpartment is also thc home of the Tatc Neutron Activation Analysis Laboratory. J Thc TRIGA reactor was obtained through a 1958 grant from the United States Atomic Energy Commission and is operated under Nuclear Rcgulatory Commission License R-88 and the J regulations of Chapter 1, Title 10, Codc of Federal Regulations. Chartered functions of the J Nuclear Rcactor Facility are to serve as: 1) an educational facility for all students at KSU and nearby universities and colleges, 2) an irradiation facility for researchers at KSU and for others in the central United States, 3) a facility for training nuclear reactor operators, and 4) a J demonstration facility to increase public understanding of nuclear energy and nuclear reactor systems.

The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operatcd in an open pool and fueled with heterogeneous elements consisting of nominally prcent enriched uranium in a zirconium hydride matrix and clad with stainless steel. Principal experimental features of the KSU TRIGA Reactor Facility are:

Central thimble

Rotary specimen rack

Thermalizing column with bulk shielding tank

Thermal column with removable door

Beam ports c Radial (2) 0 Piercing (fast neutron) (1)

Tangential (thermal neutron) (1)

The reactor was licensed in 1962 to operate at a stcady-statc thermal power of 100 kilowatts (kW). The reactor has been licensed since 1968 to operate at a steady-state thermal power of 250 kW and a pulsing maximum thermal power of 250 MW. Application is made concurrently with license renewal to operate at a maximum of 1,250 kV, with fuel loading to support 500 kW steady state thermal power with pulsing to $3.00 reactivity insertion. All cooling is by natural convection. The 250-kW core consists of 81 fuel elements typically (at leas lanned for the 1,250-kWV core), each containing as much asprams of 2"U. The reactor core is in the form of a right circular cylinder about 23 cm (approximately 9 in.) radius and 38 cm (14.96 in.) depth, positioned with axis vertical near the base of a cylindrical water tank 1.98 m (6.5 ft.) diameter and 6.25 m (20.5 R.) depth. Criticality is controlled and shutdown margin assured by control rods in the form of aluminum or stainlcss-stecl clad boron carbide or borated graphite. Reactivity requirements (i.e., minimum shutdown margin with the most reactive rod fully withdrawn and maximum excess reactivity) can be met for 250 kV with three control rods, but reactivity required to compensate for fuel temperature and rission products for operations at power levels of K-State Reactor 4-1 Original (12104)

Safety Analysis Report

CHAPTER 4 500 kW requires four control rods to meet reactivity requirements. A biological shield of reinforced concrete at least 2.5 m (8.2 ft) thick provides radiation shielding at the side and at the base the reactor tank. The tank and shield are in a 4078-ms (144,000 Ai?) confinement building made of reinforced concrete and structural steel, with composite sheathing and aluminum siding.

Sectional views of the reactor are shown in Figures 4.1 and 4.2.

Criticality was first achieved on October 16, 1962 at 8.25 pm. In 1968, pulsing capability was added and the maximum steady-state operating power was increased from 100 kW to 250 kW.

The original aluminum-clad fuel elements were replaced with stainless-steel clad elements in 1973. Coolant system was replaced (and upgraded in 2000), the reactor operating console was replaced, and the control room was enlarged and modernized in 1993, with support from the U.S.

Department of Energy. All neutronic instrumentation was replaced in 1994.

22 Figurc 4.1, Ycrtical Section Through the KSU TRIGA Reactor.

K-State Reactor 4-2 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION 4.2 Reactor Core The Gencral Atomics TRIGA reactor design began in 1956. The original design goal was a completcly and inherently safc reactor. Completc safcty means that all the available excess reactivity of the reactor can be instantaneously introduced without causing an accidcnt. Inherent safcty means that an increasc in the temperaturc of the fuel immediately and automatically results in decreased reactivity through a prompt negative temperature coefficient. These features were accomplished by using enriched uranium fucl in a zirconium hydridc matrix.

N. . a u~t 3 PTjkM A- -ac^ .1

c*, TI s_,

_~~~~U F ; riaTOCOF Issoszin

-;twne

?1

93 FT a al XU tIX>T rtAE/

West 2-

-D rV A. . - East Figure 4.2, Horizontal Scetlon Through the KSU TRIGA Reactor.

The basic parameter providing the TRIGA system with a large safety factor in steady state and I transient operations is a prompt negativc temperature coefficient, relatively constant with I temperature (-0.01% Ak/k&C). This coefficient is a function of the fuel composition and core geometry. As power and temperature increase, matrix changes cause a shift in the neutron energy spectrum in the fuel to higher energies. The uranium exhibits lower fission cross sections for the higher energy neutrons, thus countering the-power increase. Therefore, fuel and clad tempcrature automatically limit operation of the reactor.

K-State Reactor 4-3 Original (12104)

Safety Analysis Report

CHAPTER 4 It is convenient to set a power level limit that is based on temperature. The design bases analysis indicates that operation at up to 1900 kW (with an%clement core and 1207F inlet water temperature) with natural convective flow will not allow film boiling; therefore high fuel and clad temperatures capable of causing loss of clad integrity cannot occur. An Wdelement core distributes the power over a larger volume of beat generating elements, and therefore using" elements in analysis results in a less favorable, more conservative thermal hydraulic response.

4.2.1 Reactor Fuel' TRIGA fuel was developed around the concept of inherent safety. A core composition was sought which had a large prompt negative temperature coefficient of reactivity suchthat if all the available excess reactivity were suddenly -inserted into the core, the resulting fuel temperature would automatically cause the power excursion to'terminate before any core damage resulted.

Zirconium hydride was found to possess a basic mechanism to produce the desired characteristic.

Additional advantages were that ZrH has a high heat capacity, results in -relatively small core sizes and high flux values due to the high hydrogen content, and could be used effectively in a rugged fuel element size.

TRIGA fuel is designed to assure that fuel and cladding can withstand all credible eivironmental and radiation conditions during its lifetime at the reactor site. As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation.

The fuel temperature limit stems from potential hydrogen outgassing from the fuel and the subsequent stress produced in the fuel clement clad material. The maximum temperature limits of I I50 0C (with clad < 5001C) and 9501C (with clad > 5001C) for U-ZrH (H/Zr:. 65 ) have been set to limit internal fuel cladding stresses that might challenge clad integrity REG 1282). Thesc limits are the principal designb.ases for thi safety analysis.

a. Dimensions and Physical Properties.

The ]CSU TRIGA reactor is fueled by stainless steel clad Mark III fuel-elements. Three.

instrumented aluminum-clad Mark II elements arc still available for use in the core.

General properties of TRIGA fuel arc listed in Table 4.1. The Mark lII elements are illustrated in Figure 43. To facilitate hydriding in the Mk III elements, a zirconium rod is inserted through a 0.635 cm. (1/4-in.) hole in the center ofthe active fuel section.

Instrumented elements have three chromel-alumel thermocouples embedded to about 0.762 cm (03 in.) from the centerline of the fuel, one at the axial center plane, and one each at 2.54 cm. (I in.) above and below the center plane. Thermocouple leadout wires pass through a seal in the upper end fixture, and a leadout tube provides a watertight conduit carzying the leadout wires above the water surface in the reactor tank.

'Unless otherwise indicated, fuel properties are taken from the General Atomics report of Simnad [1980]

and from authorities cited by Simnad.

K-State Reactor 44 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION ji Graphite dummy elemcnts may be used to fill grid positions in thc core. The dummy i elements are or the same general dimensions and construction as the fucl-moderator clements. They are clad in aluminum and have a graphite length of 55.88 cm (22 in.).

Table 4.1, Nominal Properties of Mark ll and Mark 111 TRIGA Fuel Elements J in use at the KSU Nuclcar Reactor Facility.

Dimensions J!k_-' .16M &IEIIii P i^H-! TI r s Outside diamctcr, D, = 2r0 I Inside diameter, D,= 2r1 [* 1 Overall length Length of fucl zone,_ J Length of graphite axial reflectors 4 in. (10.16 cm) - 3.44 in (8.738 cm)

End fixtures and cladding aluminum 304 stainless steel Cladding thickness C Burnable poisons Sm %vafers None Uranium content NWcight percent U 235U enrichment percent 235U content am Physicalpropertiesoffuet e-rcluding cladding HILZr atomic ratio 1.0 1.6 Thermal conductivity (W cm71 K71) 0.18 0.18 Heat capacity [T 20°C3 (J cm73 KI) 2.04 + 0.00417T M6echanicalproperfiesofdelta phase U-Zrl' Elastic modulus at 20°C 9.1 x 106 psi Elastic modulus at 650°C 6.0 x 10' psi Ultimate tensile strength (to 650°C) 24,000 psi Compressive strength (20'C) 60,000 psi Compressive yield (20°C) 35,000 psi

'Source: Texas SAR [19911.

b. Composition and Phase Properties The Mark III TRIGA fuel elcment in use at Kansas State University contains nominally fbby weight of uranium, enriched tolfsU, as a fine metallic dispersion in a zirconium hydride matrix. The HEZr ratio is nominally 1.6 (in the facc-centered cubic delta phase). The equilibrium hydrogen dissociation pressure is governed by the composition and temperature. For ZrHj6, the equilibrium hydrogen pressure is one atmosphere at about 760°C. Thc single-phase, high-hydride composition eliminates the problems of density changes associated with phase changes and with thermal diffusion of the hydrogen in earlicr designs. Over 25,000 pulses have been perfiorned with the TRIGA fuel elements at General Atomic, with fuel temperatures reaching peaks of about I 1500C.

K-State Reactor 4-5 Original (12104)

Safety Analysis Report

, I CHAPTER 4 The zirconium-hydrogen system, whose phase diagram is illustrated in Chapter 3, is essentially a simple eutectoid, with at least four separate hydridc phases. The delta and cipsilon phases are respectively face-centered cubic and face-centered tetragonal hydride phases. The two phase delta + epsilon region exists between ZrHM4 and Z;rHI 74 at room temperature, and closes at ZrH,.7 at 455C. From 4551C to about 1050 0C, thc delta phase is supported by a broadening range of H/Zr ratios.

C.

I I

STAINLESS T

3.44*

STEEL BOTTOM END-FIXTURE 1 * .:

.. 4- .

Figure 43, TRIGA Fuel Element.

K-State Reactor 4-6 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION

c. Core Layout A typical layout for a KSU TRIGA II 250-kW core (Core 11-18) is illustrated in Figure 4.4. The layout for thc 1,250-kV core is expected to be similar, cxccpt that the graphite elemcnts will bc replaced by fuel elcments, one additional control rod will be added, and control rod positions will bc adjusted..

J LI I Thc additional fucl elements arc required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod is required to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions will be different to allow a higher worth pulse rod (the 250 kW pulse rod reactivity worth is S2.00, the 1,250 kW core pulse rod reactivity worth is S3.00), balancing the remaining control rods wcrth's to mect minimum shutdown margin requirements, and meeting physical constraints imposed by the dimensions of the pool bridge K-State Reactor 4-7 Original (12104)

Safety Analysis Report

.. .. I CHAPTER 4 4.2.2 Control Rods Control rods are 50.8 cm. (20 in.) long boron carbide or borated graphite, clad with a 0.0762 cm.

(3D-mil) aluminum sheath. The pulse rod is 3.175 cm. (125 in.) diameter. Other rods are 2.225 cm (718 in.) diameter.

The control rod drives are connected to control rod clutches through three extension shafts. The clutch and upper extension shaft for standard rods extend through an assembly designed with slots that provides a hydraulic cushion (or buffer) for the rod during a scram, and also limits the bottom position of the control rods so that they do not impact the bottom of the control rod guide tube (in the core). The buffers for two standard rods are shown in the left hand picture below (slotted tubes on the right hand side) along with the top section of the pulse/transient rod extension. The pulse rod drive clutch connects to a solid extension shaft through a pneumatic cylinder, the dimensions of the cylinder limits bottom travel.

Figure 4.S, Control Rod Upper Extension Assemblies The lower extension of the pulse rod is shown on the left hand side of Figure 4.5. The upper extension shaft is a hollow tube, the middle extension is solid. Abe upper extension shaft is connected to the middle extension shaft with lock wire or a pin and lock wire for standard rods, with a bolted collar for the pulse rod (the nmcchanical shock during a pulse requires a more durable fastener). Securing the upper control rod extension to the middle extension at one of several holes drilled in the upper part of the middle extension (Figure 4.6) provides adjustment for the control rods necessary to ensure the control rod full-in position is above the bottom of the guide tube.

* . .~.. S.:JS K-Stale Reactor 4-8 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION Figure 4.6, Middle Extension Rod Alignment Holes The middle (solid) extension is similarly connected to the lower extension. The lower extension is hollow, the middle extension fits into the lower extension and a hole drilled in the overlap secures the lower extension to the middle extension. Typically the lower extension has a tighter fit than the upper extension because the lower and midldc extension arc not separated for inspections and because the interface with upper extension is used to set the bottom position of the control rod. Pictures of the lowcr connector for the pulse rod and one standard rod are shown -

at the left in Figure 4.7..

I I

Figure 4.7, Standard & Pulse Rod Lowver Coupling The bottom of the lower extension attaches directly to the control rod. Pictures of the control rods taken during the 2003 control rod inspection are in Figure 4.8. The rods move within control rod guide tubes, shown in Figure 4.9. The guide tubes have perforated walls. Alignment pins in the bottom end fitting of the guide tube fit into holes in the lower guide plate. The alignment pins have a small metal wire in the tip that fits into the lower grid plate; a setscrew inside the bottom of the guide tube pushes the wire against the lower grid plate to secure the guide tube.

I K-State Reactor 4-9 Original (12104)

Safety Analysis Report

. 4 Io. :

I CHAPTER 4 **

Pulse Rod ShimRod RegRod Figure 4.8, Control Rods During 2003 Inspection

. 0 Figure 4.9, Control Rod Guide Tubes K-State Reactor 4-10 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION

3. Control Function J While three control rods were adequate to mect Technical Specification requirements for reactivity control with the 100 kW and 250 kW cores, reactivity limits for operation at a maximum power level of 1,250 kW requires four control rods (three standard and one I transicnt/pulsing control rod). The control-rod drives arc mounted on a bridge at the top of the reactor tank. The control rod drives arc coupled to the control rod through a connecting rod assembly that includes a clutch. .The standard rod clutch is an I electromagnct; the transient rod clutch is an air-operated shuttle. Scrams cause the clutch to release by de-cncrgizing the magnetic clutch and venting air from the transient rod clutch; gravity causes the rod to fall back into the core. Interlocks ensure operation of the control rods remains within analyzed conditions for reactivity control or limit potential for accident scenarios, while scrams operate at limiting safety system settings.

A detailed description of the control-rod system is provided in Chapter 7; a summary of interlocks and scrams is provided below in Table 4.2 and 43. Note that (1) the high fuel temperature and period scrams are not required, (2) the fuel temperature scram limiting sctpoint depcnds on core location for the sensor, and (3) the period scram can be prevented by an installed bypass switch.

Table 4.2, Summary of Control Rod Interlocks

!r '.lNEERLOCTPttl tTUISE O TX -4vXI F IO O PURPSE4!P(.

Inhibit stndard rod motion if nuclear Soure Interlock2 cs insanrment startup channel reading is less u than Instrument scrstivitylcnsure nuclear instrumnent startup channel is operating Pulse Rodt Inteslock Pule rod inserted Prevent applying power to pulse rod unless rod insertedfptevent inadvertent pulse Multiple Rod Withdrawal Withraw signal. Prvent withdrawal of mom than I rodl/mit

_ more than I rod

_ __ maximum reactivity addition rate PulsMode switch in Hi Prevent withdawing standard control rods in Pulse pulse mode Pulse-Power Interlock PuI -oerI!elokIW10 JkW Prevent kW pulsing If power lewl is greater than I

NOTe: Pulse-Power Interlock nonnally set at I kW

b. Evaluation of Control Rod System 7 The reactivity worth and speed of travel for the control rods are adequate to allow complete control of the reactor system during operation from a shutdown condition to full power. The TRIGA system does not rely on speed of control as significant for safety of the reactor, scram times for the rods are measured periodically to monitor potential degradation of the control rod system. The inherent shutdown mechanism (temperature feedback) of the TRIGA prevents unsafe excursions and the control system is used only for the planned shutdown of the reactor and to control the power level in steady state operation.

K-State Reactor 4-11 Original (12104)

Safety Analysis Report

.. 1 :I I4 CHAPTER 4 Table 43 Summary of ReactorSCRAMs Linear Channel High 110% N/A 104%

Power _

Power Channel 110% N/A 104%

Hih Rower _

Detector High 90% 90% 90%

Voltage__ _ _ __ _ _ _ _

600C B Ring element High Fuel 555°CC Ring element 4500 C Temperature 4800 C D Ring element 3800C E Ring element 3500 C PeriodIA I N/A 3 sec NOTE: Period trip and temperature trip are not required The reactivity worth of the control system can be varied by the placement of the control rods in the core. The control system may be configured to provide for the excess reactivity needed for approximately 500 kW operations for eight hours per day (including xenon override) and will assure a shutdown margin ofat least $0.5*.

Nominal speed of the standard control rods is about 12 in. (30.5 cm) per minute (with the stepper motor specifically adjusted to this value), of the transient rod is about 24 in. (61 cm) per ruinute, with a total travel about 15 in. (38.1 cm). Maximum rate of reactivity change for standard control rods is specified in Technical Specifications.

4.2.3 Neutron Moderator and Reflector Hydrogen in the Zr-H fuel serves as a neutron moderator. Denineralized light water in the reactor pool also provides neutron moderation (serving also to remove heat from operation of the reactor and as a radiation shield). Water occupies approximately 35% of the core volume. A graphite reflector surrounds the core, except for a cutout containing the rotary specirien rack (described in Chapter 10). Each fuel element contains graphite plugs above and below fuel approximately 3A in. in length, acting as top and bottom reflectorsl The radial reflector is a ring-shaped, aluminum-clad, block of graphite surrounding the core radially. The reflector is OA57-m (18.7 in.) inside diarneter3 1.066-m (42 in.) outside diameter, and 0559-m height. Embedded as a circular well in the reflector is an aluminum housing for a rotary specimen rack, with 40 evenly spaced tubular containers, 3.18-cm (1.25 in.) inside diameter and 27.4-cm (10.8 in.) height. The rotary specimen rack housing is a watertight assembly located in a re-entrant well in the reflector.

-~ . , . ; ,,

K-State Reactor 4-12 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION The radial rcflector assembly rests on an aluminum platform at the bottom of thc rcactor tank. -J Four lugs arc provided for lifting the assembly. A radial void about 6 inches (15.24 cm) in J diameter is located in the rcflector such that it aligns with thc radial picrcing beam port (NE beam port). Thc rcflector supports the core grid plates, with grid plate positions set by alignment fixtures. Graphite inserts within thc fuel cladding provide additional reflection. Inserts arC placed at both cnds of the fuel meat, providing top and bottom reflection.

4.2.4 Neutron Startup Source J A 2-curic americium-beryllium startup source (approximately 2 x 106 n sr) is used for rcactor startup. The source material is encapsulated in stainless steel and is houscd in an aluminum-cylinder source holder of approximately the same dimcnsions as a fuel clmcent. The source J holder may be positioned in any one of the fuel positions dcfincd by the upper and lowcr grid plates. A stainless-steel wire may be threaded through the upper end fixture of the holder for usc in relocating the source manually from the 22-ft level (bridge level) of the reactor.

4.2.5 Core Support Structure The fuel elements arc spaced and supported by two 0.75-in. (1.9 cm) thick aluminum grid plates.

Thc grid plates have a total of 91 spaces, up to 85 of which arc filled with fuel-moderator elcments and dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids. The bottom grid plate, which supports the weight of the fucl clements, has holes for receiving the lower cnd fixtures. Space is provided for the passage of cooling water around the sides of the bottom grid plate and through 36 experiment penetrations. The 1.5-in. (3.8 cm) diametcr holes in the upper grid plate serve to spacc the fuel elements and to allow withdrawal of the clements from the core.

Triangular-shaped spacers on the upper end fixturcs allow the cooling waler to pass through the upper grid plate when thc fuel elements are in position. The reflector assembly supports both grid plates.

4.3 Reactor Tank The K}SU TRIGA reactor core support structure rests on the base of a continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum concrete thickness of approximately 249 cm. or 8 ft 2 in), as illustrated in Figures 4.1 and 4.2. The tank is a welded aluminum structure with 0.635 cm. (114-in.) thick walls. The tank is approximately 198 cm (6.5-ft) in diameter and approximately 625 cm (20.5-fl) in depth. The exterior of the tank was coated with bituminous material prior to pouring concrete to retard corrosion. Each experiment facility penetration in the tank wall (described below) has a water collection plenum at the penetration. All collection plenums arc connected to i leak-off volume through individual lines with isolation valves, with the leak-off volumes monitored by a pressure gauge. The bulk shield tank wall is known to have a small leak into the concrete at the thermalizing column plenum, therefore a separate individual Icak-off volume (and pressure gauge) is installcd for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak from the pool K-State Reactor 4-13 Original (12104)

Safety Analysis Report

CHAPTER 4 through an experiment facility, pressure in the volume will increase. Isolating individual lines allows identification of the specific facility with the leak.

A bridge of steel plates mounted on two rails of structural steel provides support for control rod drives, central thimble, the rotary specimen rack, and instrumentation. Ihe bridge is mounted directly over the core area, and spans the tank. Aluminum grting with clear plastic attached to the bottom is installed that can be lowered over the pool. The grating normally remains up to reduce humidity at electromechanical components of the control rod drive system and to prevent the buildup of radioactive gasses at the pool surface during operations. The grating can be lowered during activities that could cause objects or material to fall into the reactor pool.

Four beam tubes extend from the reactor wall to the outside of the concrete biological shield in the outward direction. Tubes welded to the inside of the wall extend toward the reactor core.

Three of the tubes (NWV, SW, and SE) end at the radial reflector. The NE beam tube penetrates the radial reflector, extending to the outside of the core. Two penetrations in the tank allow neutron extraction into a thermal column and a thermalizing column (described in Chapter 10).

4.4 Biological Shield The reactor tank is surrounded on all sides by a monolithic reinforced concrete biological shield.

The shielding configuration is similar to those at other TRIGA facilities operating at power levels up to 1 MW. Above ground level, the thickness varies from approximately 249 cm. (8 ft 2 in.) at core level to approximately 91 cn. (3 ft.) at the top of the tanrd The massive concrete bulk shield structure provides additional radiation shielding for personnel working in and around the reactor laboratory and provides protection to the reactor core from potentially damaging natural phenomena.

4.5 Nuclear Design The strong negative temperature coefficient is the principal method for controlling the maximum power (and consequently the maximum fid temperature)forTRIGA reactors. This coefficient is a function of the fuel composition, core geometry, and temperature. For fuels withI enrichment, the value is nearly constant at 0.01% AkMc per IC, only weakly dependent on geometry and temperature.

Fuel and clad temperature define the safety limit. A power level limit is calculated that ensures that the fuel and clad temperature limits will not be exceeded. The design bases analysis indicates that operation at 1,250 kW thermal power with arftlement across a broad range of core and coolant inlet temperatures with natural convective {low will not allow film boiling that could lead to high fuel and clad temperatures that could cause loss ofclad integrity.

Increase in maximum thermal power from 250 to 1,250 kW does not affect fundamental aspects of TRIGA fuel and core design, including reactivity feedback coefficients, temperature safety K-State Reactor 4-14 PrIgInal (12104)

Safety Analysis Report

REACTOR DESCRIPTION limits, and fission-product release rates. Thermal hydraulic performance is addressed in Section 4.6.

4.5.1 Design Criteria - Reference Core The limiting core configuration for this analysis is a compact core defined by the TRIGA Mk 11 grid plates (Section 4.2.5). The grid plates have a total of 91 spaces, up to 85 of which are filled with fuel-moderator elements, and graphite dummy elements, and the remaining spaces with control rods, the central thimble, the pneumatic transfer tube, the neutron source holder, and one or more voids in the E orF (outermost two rings) as required to support experiment operations or limit excess reactivity. The bottom grid plate, which supports the weight of the fuel elements, has boles for receiving the lower end fixtures.

4.5.2 Reactor Core Physics Parameters The limiting core configuration differs from the configuration prior to upgrade only in the addition of a fourth control rod, takng the place of a graphite dummy element or void experimental position. For this reason, core physics is not affected by the upgrade except for linear scaling with power of neutron fluxes and gamma-ray dose rates.

For comparison purposes, a tabulation of total rod worth for each control element from the K-State reactor from a recent rod worth measurement is provided with the values from the Comell University TRIGA reactor as listed in NUREG 0984 (Safety Evaluation Report Related to the Renewal of the Operating license for the Cornell University TRIGA Research Reactor).

Table 4A., 250 kW Core Parameters.

1 (effective delayed neutron frction) 0.007 t (effectivencutronliretime) 43 PS

-SO.017 EC' C'Tf (PMfpttempaeutfre coeflicient) 250kW -275EC Cv (void coefficient) -0.003 1%' void 40o.006 Mr' to -

ap(owcr temperturecoefricient-weightedaec) - SO.0l kW' I

Table 4.5, Comparison of Control Rod Worths.

UrgRl -- n'j ; ai: F C-3, Shim S2.88 D-16, Shim S2.20 D-16, Regulating _ S1.5 D-4, Safety SI.99 D-10, Pulse S1.96 D-10, Transient I S1.88

-TBD TBD E-1, Regulating l $0.58 K-State Reactor 4-15 Original (12104)

Safety Analysis Report 1%

CHAPTER 4 The pulse rod is similar to a standard control rod, and the worth of the pulse rod compares well with the comparable standard control rods in similar ring positions. A maximum pulse is analyzed for thermal hydraulic response and maximum fuel temperature.

4.5.3 Fuel and Clad Temperatures This section analyzes expected fuel and cladding temperatures with realistic modeling of the fuel-cladding gap. Analysis of steady state conditions reveals maximum heat fluxes well below the critical heat flux associated with departure from nucleate boiling. Analysis of pulsed-mode behavior reveals that film boiling is not expected, even during or after pulsing leading to maximum adiabatic fuel temperatures.

Chapter 4, Appendix A of this chapter reproduces a commonly cited analysis of TRIGA fuel and cladding temperatures associated with pulsing operations. The analysis addresses the case of a fuel element at an average temperature of 10000C immediately following a pulse and estimates the cladding temperature and surface heat flux as a function of time after the pulse. The analysis predicts that, if there is no gap resistance between cladding and fuel, film boiling can occur very shortly after a pulse, with cladding temperature reaching 4700 C, but with stresses to the cladding well below the ultimate tensile strength of the stainless steel. However, through comparisons with experimental results, the analysis concludes that an effective gap resistance of 450 Btu hr fr 2C,.I (2550 W m7 2 K') is representative of standard TRIGA fuel and, with that gap .resistancc, film boiling is not expected. This section provides an independent assessment of the expected fuel and cladding thermal conditions -associated with both steady-state and pulse-mode operations.

a. Heat Transfer Models The overall heat transfer coefficient relating heat flux at the surface of the cladding to the difference between the maximum fuel (centerline) temperature and the coolant temperature can be calculated as the su* of the temperature changes through each element from the centerline of the fuel rod to the water coolant, where the subscripts for each of the 1T's represent changes between bulk water temperature and cladding outer surface, (bro), changes between cladding outer surface and cladding inner surface (roro, cladding inner surface and fuel outer surface - gap (g), and the fuel outer surface to centerline (ricl):

Td=7+A T +A,,,+AT +AT,d, (1)

A standard heat resistance model for this system is:

K-State Reactor 4-16 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION

.J D T.= T. + q" I ( )+ r r 1 (2)

-h k, rTh, *2kJ and heat flux is calculated directly as:

q =UAT= suit~ (3) q1 rn(r1r)+ rO +b 7;'

h k, rA, h.2 k in which r. and r, are cladding inner and outer radii, h, is the gap conductivity, h is the convective heat transfer coefficient, and kf is the fuel thermal conductivity. The gap conductivity of 2840 W rn 2 K" (S00 Btu h' ft *2CV) is taken from Appendix A. The convcctive beat transfer coefficient is mode dependent anud is determined in context.

Parameters are cross-referenced to source in Table 4.6.

Table 4.6: 71nrmod lame Valucs 2.ParametcrW;;M XSymbol Value6 iZ i=, fReferenceO 1--

Fuel conductivity kr 18 Wnfm K Table 13.3 14.9 W m I K71(300 K1 Table 13.3 Clad conductivity ks 16.6 W m' K (400 K) Table 13.3 19.8 W m [K'(600 K) Table 13.3 Gap resistance h. 2840 W mZK' AppendixA Clad outerradius 31 0.018161 M Table 13.1 Fuel outer radius rj 0.018669 M Table 13.1 Active fuel length L - 0.381 M Table 13.1 No. fuel elements N I N/A Chap 13 Axial peaking factor APF I____--I N/A Table 13.4 General Atomics reports that fuel conductivity over the range of interest has little temperature dependence, so that:

2k =5.1858E-04-&K (4) 2k, w Gap resistance has been experimentally determined as indicated, so that:

. .. . .(5)

-h =3.6196E-04 jW I%

K-State Reactor 4-17 Original (12104)

Safety Analysis Report

CHAPTER 4 Temperature change across the cladding is temperature dependent, with values quoted at 300 K, 400 K and 600 Y, Under expected conditions, the value for 127 0 C applies so that:

r. Inl m' (6)

A' =3.103e_5mK Table 4.7, Cladding Heat Translcr Coefficient JI Temp (0K) MiTemrnp (O .K fivM2 lv 300 27 3:457e-5 400 127 3.103e-5 600 327 2.601e-5 It should be noted that, since these values are less than 10% of the resistance to heat flow attributed to the other components, any errors attributed to calculating this factor are small.

The convection heat transfer coefficient was calculated at various steady state power levels. A graph of the calculated values results in a nearly linear response function.

Convection Heat Transfer Coefficient

. ......... a..... .... I a .. a I Fnmo~

IE a

Y E&So AC U m no 1w 11M w ._ .

D Paw~tLaw! LXW Figure 4.10, Convection Hear Transfer Coecifient versus Power Level I = I (7) h 0.0326P(vatlrs)+16985 K-State Reactor 4-18 Original (12104)

Safety Analysis Report

REACTOR DESCRIPTION The reference core contains 83 fuel elements; temperature calculations using the reference core are conservative because at least 83 elements are required for steady state l 500 kW operations, while analysis assumes 1.25 MW operation. Actual heat production will be less than heat calculated in analysis, so temperatures will be lower. More than 83 elements will distribute heat production across a larger number of fuel elements so that heat flux and temperatures will be lower than calculations based on the refcrcnce core.

Average heat flux per fuel rod is therefore: N q power P (8) area 2nrmL; With the maximum heat flux of:

83*22L, 2 332r.L, O.423m'P (9) _

Therefore, core ccnterline temperature for the fuel rod producing the maximum beat as a function of power can be calculated.as:.

T = T +0.423 P[ PI 985 +3.103 e-5+3.620 e-4+5.186 e-4]

- * ~0.0326 P + I68 For the purposes of calculation, the two extremes of cladding thermal conductivity were assumed (300 K value and 600 K value) to determine expected centerline temperature as a function of power level. Calculations using both vales are provided graphically, and shows the effect of thermal conductivity changes are minimaL The graph also shows that fuel temperature remains below about 750 IC at power levels up to 1900 kW with pool temperature at 27 IC (300 K), and 1700 kW with pool temperatures at 100 "C.

Finally, temperature calculations for the hottest location in the core were made assuming 1.25 MW steady state power at 200 C and 1000 C with the following results:

Table 4.8, Calculated Temperature Data for 1,250 kNV Operacion

IAuelC FueIGap IGap*Clad%4 'Clad/Water- Water IC Centerline -crntrficc 00: Islntcrface la WInterface IQe.x 503.2 229.0 37.7 21.2 20.0 582.0 307.8 116.4 100.0 100.0 K-State Reactor 4-19 Original (12/04)

Safety Analysis Report

CHAPTER 4 Hot Fuel-Rod Centerline Temperature at Power (Temperature Bevation over Pool WaterTemperature) 700 PF 600-

-iEl50 r- 400 I.

300 -

iL 4200

a U10 100 300 500 700 SW0 1100 1300 1500 1700 1900 Reactor Power (kW)

Figure 4.11, Hot Fuel-Rod Ccntcrline Temperature For subcooled boiling, -with water at amriient temperature Tb, the critical heat flux is calculated by (Ivey and Morris 1978)

(q )n4 = (faf Chf)F+ 0.1(- d.)] (11) in which cy is the heat capacity of the coolant. At the depth of the reactor core in the KSU TRIGA, static pressure is 0.153 MPa and coolant ambient temperature is taken to be 271C. Thermodynamic and physical properties of water under these conditions are tabulated in Chapter 4 Appendix B.

b. .Spatial Powcr Distribution The following conservative approximations are made in characterizing the spatial distribution of the power during steady-state operations.

The hottest fuel element delivers twice the power of the average.

K-State Reactor 4-20 Original (12104)

Safety Analysis Report I .

REACTOR DESCRIPTION Classically, the radial hot-channel factor for a cylindrical reactor (using R as the <.A physical radius and H4 as the physical radius and the extrapolation distance) is JJ given 2 by:

1.202* 5R) (12) jJ J,[2Ao.48*(,R)]

with a radial peaking factor of 1.93 for the KSU TRIGA 1I geometry,. However, TRIGA fuel elements are on the order of a mean free path of thermal neutrons, and there is a significant change in thermal neutron flux across a fuel element.L Calculated thermal neutron flux data 3 indicates fthat the ratio of peak to average neutron flux (peaking factor) for TRIGA cores under a range of conditions \

(temperature, fuel type, water and graphite reflection) has a small range of 136 j to 1A0.

Actual power produced in the most limiting actual case is 14% less than power calculated using the assumption; therefore using a peaking factor of 2.0 to determine calculated temperatures and will bound actual temperatures by a large margin, and is extremely conservative.

The axial distribution of power in the hottest fuel element is sinusoidal, with the peak power a factor of a12 times the average, and heat conduction radial only.

The axial factor for power produced within a fuel element is given by:

g(z)= 4*co *2 * ) (13) in which t = L / 2 and 1,X is the extrapolation length in graphite, namely, 0.0275

m. The value used to calculate power in the limiting location within the fuel element is therefore 4% higher a power calculated with the actual peaking factor. v Actual power produced in the most limiting actual case is 4%1 less than power calculated using the assumption; therefore calculated temperatures will bound -

actual temperatures.

At full power, the thermal power ofP. = 1,250 kV is distributed over 83 fuel elements, with a maximum heat flux given by.

Elements of Nuclear Reactor Design, 2"d Edition (1983). J. Weisman, Section 63 3GA4361, Calculated Fluxes and Cross Sections forTRIGA Reactors (8/1411963), B.

3.West K-State Reactor 4-21 Original (12/04)

Safety Analysis Report

CHAPTER 4 (14)

- = 9*rL =2.414e6W~m' 83q*D*

The radial and axial distribution of the power within a fuel element is given by q"'(r,z) = qSf(r)g(z), (15) in which r is measured from the vertical axis of the fuel element and z is measured along the axis, from the center of the fuel element. The axial peaking factor follows from the previous assumption of the core axial peaking factor, but (since there is a significant flux depression across a TRIGA fuel clement) distribution of power produced ac'ioss tlihradius of the fuel the radial peaking factor requires a different approach than the previous radial peaking factor for the

core.

The radial factor is given by:

a-+cr+ er2 (16) l+br+dr2 '

in which the parameters of the rational polynomial approximation ire derived from flux-depression calculations for the TRIGA fuel (Abrens 1999a). Values are: a = 0.82446, b = -0.26315, c= -021869, d = -0.01726, and c = -0.04679.

The fit is illustrated in Figure 4.1 1.

1.3 . . . . . .. .. . - . . . . ..

1.2 1.1

1. LO . . . . _._. . . . ._. . . . . . .

0.90 0.0

__7 O.0 0.20 040 0."0 0.A0 LO .2 14 16 1o 2.0

. (-)

Figure 4.12, Radial Variation ofPower Yithln a TRGIA Fuel Rod.

(Data Points from Monte Carlo Calculations [Ahrens 1999a])

K-State Reactor 4-22 Original (12J04)

Safety Analysis Report

REACTOR DESCRIPTION

c. Steady-State Mode of Operation Table 4.9 tabulates critical heat flux as a function of coolant temperature from the assumed inlet temperature to the saturation temperature. Also shown in the table is the I,,

CHFR, i.e., the ratio of the critical heat flux and the maximum heat flux at full poweri Table 4.9: Critical Heat Flux and CHFR ror 1.25 MW Operation at Selected Coolant Temperaturcs.

.9E bC qd r (MW ld 2 )1 CHF 27 84.9 6.02 5.8 30 81.9 5.86 5.6 40 71.9 5.33 5.1 60 51.9 4.26 4.1 100 11.9 2.13 2.0 111.9 0.0 19 1.4 It is clear from the table that there is a very wide margin between the operating heat flux and the critical heat flux even to unrealistically high pool water temperature, so that film boiling and excessive cladding temperature is not a consideration in steady-state operation. A parametric variation of this calculation for various power levels shows margin to DNBR up to 100C pool temperature for power levels greater than 1.9 MW.

Critical Heat Flux Ratio 1

. hw--- &W-0-tmm 1 I.M-.--1.miA-- MUW I.

a U

a 11 t

FS s

a I I I I .I . I I

.I I .

I . I I I M" 0 5 0 4a 0 a 0 a" "0 I U I 10 101 II0 As Bulk PodoTmpera (C)

Figure 4.13, Ratio of Departure from Nucleate Boiling to Critical Heat Flux at Various Power Levels K-State Reactor 4-23 Original (12104)

Safety Analysis Report I

CHAPTER 4

d. Pulsed Mode of Operation Transient calculations have been performed using a custom computer code TASCOT for transient and steady state two-dimensional conduction calculations (Alirens 1999). For these calculations, the initial axial and radial tempcrature distribution of fuel temperature was based on Eqs. (6) and (7), with -the peak fuel temperature set to 746 TC, i.e., a temperature rise of 719 DC above 27 C ambient temperature. The temperature rise is computed in Chapter 13, Section 13.2.3 for a 2.1% ($3.00) pulse from zero power and a 0.7% (S$.00) pulse from power operation. In the TASCOT calculations, thcrnal conductivity was set to 0.18 W cal}Ks (Table 4.1) and the overall heat transfer coefficient U was set to 0.21 W cmr' KI'. The convective beat transfer coefficient was based on the boiling heat transfer coefficient computed using the formulation (Chcn 1963, Collier and Thome 1994) q"=h*(T.-TW)=h(T.-?k)- (17) 1000 800 0C, 600 a,

I-Q 0 400 E

V 200 64s 0 ' ' ' ' ' I I I I I I I . I I I I I I l o.0 0.20 0.40 0.60 0.80 *1.0 12 1.4 1.6 1.8 2.0 22

Radius (cm)

Figure 4. 14, Midplane Radial 'Variatlon of Temperature Within the Fuel Subsequent.lo a S3.00 Pulse.

The boiling heat transfer coefficient is given by the correlation (Forster & Zuber 1955)

K-State Reactor 4-24 'I Original (12104)

Safety Analysis Report

-0REACTOR DESCRIPTION kAOl9 *c04 5 *AOI5 hb = 0.00122* v- ) * ?JT5

( - i)0 (18) in which 7. is the cladding outside temperature, T, the saturation temperature (111.9 C),

and Tb the coolant ambient temperature (27'C). Fluid-property symbols and values are given in Appendix B. Subscriptsf and g refer respectively to liquid and vapor phases.

The overall heat transfer coefficient U varies negligibly for ambient temperatures from 20 o 60 C, and has the value 0.21 W cmfl KI at j - 27 DC.

Figure 4.14 illustrates the radial variation of temperature within the fuel, at the midplane of the core, as a function of time after the pulse. Table 4.8 lists temperatures and heat fluxes as function of time after a 2.1% (S3.00) reactivity insertion in a reactor initially at zero power. The CHFR is based on the critical heat flux of 1.49 MW m-l from Eqs. (3) and (4) and from Table 4.2 for saturated boiling. Figure 4A.3 ofAppendix A, using the Ellion data, indicates a Leidenfrost temperature in excess of 500CC Thus transition boiling, but not fully developed film boiling might be expected for a short time after the end of a pulse.

4.6 Thermal Hydraulic Design and Analysis A balance between the buoyancy driven pressure gain and the frictional and acceleration pressure losses accrued by the coolant in its passage through the core determines the coolant mass flow rate through the core, and the corresponding coolant temperature rise. The buoyancy pressure gain is given by bP, =p.JJ0 oATgL, (19) in which pa and P.b are the density and volumetric expansion coefficient at core inlet conditions (270C, 0.15285 Mpa), g is the acceleration of gravity, 9.8c m2 s I, AT is the temperature rise through the core, and L is the height of the cote (between gridplates), namely, 0.556 m. The frictional pressure loss is given by A th= 2 jL (20) 2 2A Dxp,'

in which A is the coolant mass flow rate (kg sl) in a unit cell approximated as the equivalent annulus surrounding a single fuel element, A is the flow area, namely, 0.00062 me, and Dh is the hydraulic diameter, namnel!7, 0.02127 rn. The friction factorf for laminar flow through the annular area is given by 100 Re (Shah & London 1978), in which the Reynolds number is given by Dkhv I Au. in which pa is the dynamic viscosity at core inlet conditions.

K-State Reactor 4-25 Original (12/04)

Safety Analysis Report

CHAPTER 4 Tnble 4.10, Heat Flux and Fuel Temperatures Following a S3.O0 Puls from Zero Power, with 27PC) Coolant Ambient Tcmperature.

0 953 -

1 3.57 x105 4.2. 781 224 2- 7.34 xI05 2.0 683 432 4 8.52 x105 1.7 574 498 8 7.54 x105 2.0 461 443 16 5.71 x105 2.6 344 342 32 3A6 x105 4.3 224 218 64 1.04 x105 14.4 100 84 Entrance of coolant into the core is from the side, above the lower grid plate (see Section 42.5),

and the entrance pressure loss *wouldbe expected to be negligible. The exit contraction loss is given by

-AK (21)

A,2p0 A2 The coefficient J is calculated from geometry of an equilateral-triangle spacer in a circular opening, for which

rA 2 r3R2 60 o0°

[]A [3 sin6O cos6O]= 0171, (22) where R is the radius of the opening in the upper grid plate. Equations (11) through (13), solved simultaneously yield the mass flow rates per fuel element, and coolant temperature rises through the core listed in Table 4.9.

Table 4.11, Coolant Flow Rate and Temperatuic Rise forNatural-Convcction Cooling the TRIGA Reactor During Steady-State Operations.

50 0.047 3.1 100 0.061 4.7 200 0.077 7.5 300 0.090 9.6 400 0.100 11.5 500 0.108 13.3 750 0.125 17.2 1000 0.139

20.6 1250 0.150 23.8 K-State Reactor 4-26 Original (12/04)

Safety Analysis Report

REACTOR DESCRIPTION 4.7 Safety Limit d As described in 3.5.1 (Fuel System) and NUREG 1282, fuel temperature limits both steady-state and pulse-mode operation. Ile fuel temperature limit stems from potential hydrogen oulgassing from the fuel and the subsequent stress produced in the fuel element clad material by heated hydrogen gas. Yield strength of cladding material decreases at a temperature of 5000 C; C consequently, limits on fuel temperature change for cladding temperatures greater than 500C. A maximum temperature of 11501C (with clad < 500S ) and 9501C (with clad > 5001C) for U-ZrH (H/Zri.6s) will limit internal fuel cladding stresses that might lead to clad integrity (NUREG 1282) challenges.

4.8 Operating Limits 4.8.1 Operating Parameters The main safety consideration is to maintain the fuel temperature below the value that would result in fuel damage. Setting limits an other operating parameters, that is, limiting safety system settings, controls the fuel temperature The operating Parameters established for the KSU TRIGA reactor are:

Steady-statepowerlevel

Fuel temperature measured by thermocouple during pulsing operations

Maximum step reactivity insertion of transient rod 4.8.2 Limiting Safety System Settings Heat transfer characteristics (from the fuel to the pool) controls fuel temperature during normal operations. As long as thermal hydraulic conditions do not cause critical heat flux to be exceeded, fuel temperature remains well below any limiting value. Figure 4.13 illustrates that critical heat flux is not reached over a wide range of pool temperatures and power levels: As indicated in Table 4.9, the ratio of actual to critical beat flux is at least 2.0 for temperatures less than 1000 C bulk pool water temperature for 1.25 MW operation. Operation at less than 1.25 MW ensures fuel temperature limits are not exceeded by a wide margin.

Limits on the maximum excess reactivity assure that operations during pulsing do not produce a power level (and generate the amount of energy) that would cause fuel-cladding temperature to exceed these limits; no other safety limit is required for pulsed operation.

4.8.3 Safety Margins For 1,250 kW steady-state operations, the crtical heat flux ratio indicated in Table 4.9 ranges from 15.3 for pool water at room temperature (201C) to 10.2 at 60 IC (pool temperatures are controlled to less than 480 C for operational concerns). Even at pool water temperatures K-State Reactor 4-27 Original (12104)

Safety Analysis Report

CHAPTER 4 approaching boiling, the margin remains above 3. Therefore, margins to conditions that could cause excessive temperatures during steady state operations while cladding temperatures is below 5000 C are extremely large.

Normal pulsed operations initiated from power levels below 10 kW with a $3.00 reactivity insertion result in maximum hot spot temperatures of 746TC, a 34% margin to the fuel temperature limit. As indicated in Chapter 13, pulsed reactivity insertions of $3.00 from initial conditions of poweroperation can result in a maximum hot spot temperature of 8691C. Although administratively controlled and limited by an interlock, this pulse would still result in a 15%

margin to the fuel temperature safety limit for cladding temperatures below 500WC.

Analysis shows that cladding temperatures will remain below 5000 C when fuel is in water except following large pulses. However, mechanisms that can cause cladding temperature to achieve 5001C (invoking a 9500 C fuel temperature limit) automatically limit fuel temperature as heat is transferred from the fuel to the cladding.

Immediately following a maximum pulsed reactivity additions, heat transfer driven by fuel temperature can cause.cladding temperature to rise above 5001C, but the heat transfer simultaneously cools the fuel to much less than 9501C.

If fuel rods arm placed in an air environment immediately following long-term, high power operation, cladding temperature can essentially equilibrate with fuel temperature. In worst-case air-cooling scenarios, cladding temperature can exceed 500C, but fuel temperature is significantly lower than the temperature limit for cladding temperatures greater than 5000 C.

4.9 Bibliography "TASCOT: A 2-D, TransientandSteady State Conduction CodeforAnalysis of a TFJGA Fuel Element, "Report KSUNE-99-02, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999. Ahrens, C.,

"Investigationof the RadialVariation ofthe Fission-HeatSource In a TRUGA Mart)i! Fued Element Using MCNP, " Report KSUNE-99-01, Department of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, Kansas, 1999a. Ahrcns, C.,

-A CorrelationjorBoilingHeat Transferto SaturatedFluidsIn Convective.Flow, "ASME Preprint 63-HT-34, 6th National Heat Transfer Conference, Boston, 1963. Chen, J.C.,

Kansas State University TRIGA MANl ReactorHazardsSummaryReport," License R-88, Docket 50-188, 1961. Clack, R.W., JR. Fagan, %V.R.Kimel, and S2Z Mikhail Convective Boiling andCondensation,3rd ed., Oxford Press, New York, 1994.Collier, J.G., and JR. Thome,

'Bubble Dynamics andBoilingHeat Tranfer,"AlChE Journal 1, 532 (I955). Forster, H.K, and N. Zuber, K-State Reactor 4-28 Original (12)04)

Safety Analysis Report

REACTOR DESCRIPTION Theory andDesign ofModem Pressure Vessels, 2d. cd., Van Nostrand Reinhold, New York, 1974. p. 32. Harvey, J.F.,

"On tJie Relevance of the Vapour-LiquidExchange AMechanismfor Sub-CooledBoiling Heat J Transferat High Pressure." Report AEEW-R-137, United Kingdom Atomic Energy Authority, Winflith, 1978. Ivey, H. 1.and D. J. Morris "On the prediction of the Minimum pool boiling heat flux," J. Heat Transfer, Trans. ASME, 102, 457-460 (1980). Lienhard, 1.H. and V. K. Dhir, Thermal MigrationofHydrogen In Uranium-ZirconiumAlloys, General Dynamics, General Atomic Division Report GA-361 8, November 1962. Mernen, U., ct aL, MNRC. McClellanNucearRadiation Center FacilitySafety Analysis Report, Rev. 2, April 1998.

NUREG-1282,, "SafetyEvaluation Report on High-Uranium Content, Low-Enriched Uranium-Zirconium HydrideFuelsfor TRIGA Reactors," U.S. Nuclear Regulatory Commission, 1987.

"LaminarForcedConvection in Ducts, "p. 357, Academic Press, New York, 1978. Shah, R.K.,

and A.L. London, he U-Zr-HxAlloy: Its Propertiesand Use In TRIGA Fuel,"Report E-I 17-833, General .

Atomics Corp., 1980. Simnnad, M.T.

"SafetyAnalysis Report, TRIGA ReactorFacility,NuclearEngineeringTeachingLaboratory, University of Texas atAustin, Revision 1.01, Docket 50-602, May, 1991.

K-State Reactor 4-29 Original (12/04)

Safety Analysis Report

Appendix 4-A Post-Pulse Fuel and Cladding Temperature This discussion is reproduced from Safety Analysis Reports for the University of Texas Reactor Facility (UTA 1991) and the McClellanNuclearRadiation Center (MNRC 1998).

The following discussion relates the element clad temperature and the maximum fuel temperature during a short time after a pulse. The radial temperature distribution in the fuel clement immediately following a pulse is very similar to the power distribution shown in Figure 4A.1. This initial steep thermal gradient at the fuel surface results in some beat transfer during the time of the pulse so that the true peak temperature does not quite reach the adiabatic peak temperature. A large temperature gradient is also impressed upon the clad which can result in a high heat flux from the clad into the water. If the heat flux is sufficiently high, film boiling may occur and form an insulating jacket of steam around the fuel elements permitting the clad temperature to tend to approach the fuel temperature. Evidence has been obtained experimentally which shows that film boiling has occurred occasionally for some fuel elements in the Advanced TRIGA Prototype Reactor located at GA Technologies [Coffer 1964]. The consequence of this film boiling was discoloration of the clad surface.

Thermal transient calculations vcre made using the RAT computer code. RAT is a 2-D transient heat transport code developed to account for fluid flow and temperature dependent material properties. Calculations show that if film boiling occurs after a pulse it may take place either at the time of maximum heat flux from the clad, before the bulk temperature of the coolant has changed appreciably, or it may take place at a much later time when the bulk temperature of the coolant lias approached the saturation temperature, resulting in a markedly reduced threshold for film boiling. Data obtained by Johnson et al. (19611 for transient beating of ribbons in 100°F water, showed burnout fluxes of 0.9 to 2.0 Mbtu ft hW' for e-folding periods from 5 to 90 milliseconds. On the other hand, sufficient bulk heating of the coolant channel between fuel elements can take place in several tenths of a second to lower the departure from nucleate boiling (DNB) point to approximately OA Mbtu flu lu'l. It is shown, on the basis of the folloWinig analysis, that the second mode is the most likely; i.e., when film boiling occurs it takes place under essentially steady-state conditions at local water temperatures near saturation.

A value for the temperature that may be reached by the clad if film boiling occurs was obtained in the following manner. A transient thermal calculation was performed using the radial and axial power distributions in Figures 4A.land 4A.2, respectively, under the assumption that the thermal resistance at the fuel-clad interface was nonexistent. A boiling heat transfer model, as

shown in Figure 4A3, was used in order to obtain an upper limit for the clad temperature rise.

The model used the data of McAdams 119543 for subcoolcd boiling and the work of Sparrow and'

- Cess [1962] for the film boiling regime. A conservative estimate was obtained for the minimum heat flux in film boiling by using the correlations of Speigler et al. [1963], Zuber [1959], and Rohsenow and Choi [1961] to find the minimum temperature point at which film boiling could occur. This calculation gave an upper limit of 7600 C clad temperature for a peak initial fuel temperature of 10000 C, as shown in Figure. 4A.4. Fuel temperature distributions for this case are shown in Figure 4A.5 and the heat fluc into the water from the clad is shown in Figure 4A.6. In this limiting case, DNB occurred only 13 milliseconds after the pulse, conservatively calculated K-State Reactor 4A-1 OrIginal (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A assuming a steady-state DNB conrelation. Subsequently, experimental transition and film boiling data were found to have been reported by Ellion 19] for water conditions similar to those for the TRIGA systenL The Ellion data show the minimum heat flux, used in the limiting calculation described above, was conservative by a factor of 5. An appropriate correction was made which resulted in a more realistic estimate of 470°C as the maximum clad temperature expected if film boiling occurs. This result is in agreement with experimental evidence obtained for clad temperatures of 400°C to 500°C for TRIGA Mark F fuel elements which have been operated under film boiling conditions [Coffer et al. 1965].

2 "IDIUS fix.)

Figure 4A.I. Representative Radial Variation of Power Within the TRIGA Fuel Rod 1.10 a I a I I . _

I'-

0.8

o. 0.3 _

c.6 0.5 n I I L £ s 7 a AXIAL DISTAENCE FRH 1I4D-FLAE OF FUEL ELEMT (I1.)

Figure 4A.2, Rcpresentativc Axial Variation of Power Within the TRIGA Fuel Rod.

K-Stale Reactor 4.A-2 Original (9102)

Safety Analysis Report

..REACTOR DESCRIPTION I

-a a-a-

I t 103 10 a02 103 tog lr'TSAT ("f)

Figure 4A.3, Subcooled Boiling Heat Transfer for water.

C I

V P.

I Z

E talus {ad.}

Figure 4A.4, Fuel Body Temperature at the Midplane

. . . -. of a Well-Bonded Fuel Element After Pulse..

K-State Reactor 4A-3 Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A 10' I.r ia 3

.M 0.1 1.5 twSt1 TlstI fur tto Sonut (SEt Figure 4A.5, Surface Heat Flum at the Midplane of a Well Bonded Fuel Element After a Pulse.

-lo F

5 too G.1 1. C Ica tUnsto tlhFCno two of Puts MOtC Figure 4A.6, Clad Temperature at Midpoint of WVcll-Bonded Fuel Element.

K-State Reactor 4.A-4 Oidginal (9102)

Safety Analysis Report

REACTOR DESCRIPTION The preceding analysis assessing the maximum clad temperatures associated with film boiling assumed no thermal resistance at fuel-clad interface. Measurements of fuel temperatures as a function of steady-state power level provide evidence that after operating at high fuel temperatures, a permanent gap is produced between the fuel body and the clad by fuel expansion.

This gap exists at all temperatures below the maximum operating temperature. (See, for example, Figure 16 in the Coffer report 11965].) The gap thickness varies with fuel temperature and clad temperature so that cooling of the fuel or overheating of the clad tends to widen the gap and decrease the heat transfer rate. Additional thermal resistance due Co oxide and other films on the fuel and clad surfaces is expected. Experimental and theoretical studies of thermal contact resistance have been reported [Fenech and Rohsenow 1959, Graff 1960, Fenech and Henry 19623 which provide insight into the mechanisms involved. They do not, however, permit quantitative prediction of this application because the basic data required for input are presently not fully known. Instead, several transient thermal computations were made using the.RAT code. Each of these was made with an assumed value for the effective gap conductance, in order io determine the effective gap coefficient for which departure from nucleate boiling is incipient. These results were then compared with the incipient film boiling conditions of the 10000C peak fuel temperature case.

For convenience, the calculations were made using the same initial temperature distribution as was used for the preceding calculation. The calculations assumed a coolant flow velocity of 1 fi per second, which is within the range of flow velocities computed for natural convection under various steady-state conditions for these reactors. The calculations did not use a complete boiling curve heat transfer model, but instead, included a convection cooled region (no boiling) and a subcooled nucleate boiling region without employing an upperDNB limit. The results were analyzed by inspection using the extended steady-state correlation ofBemath 11960]

which has been reported by Spano [1964] to give agreement with SPERT II burnout results within the experimental uncertainties in flow rate.

The transient thermal calculations were perfomned using effective gap conductances of 500,375, and 250 Btu ft 2 hf' o7'. The resulting wall temperature distributions were inspected to determine the axial wall position and time after the pulse which gave the closest approach between the local computed surface heat flux and the DNB heat flux according to Bernath. The axial distribution of the computed and critical beat fluxes for each of the three cases at the time of closest approach is given in Figures 4A.7 through 4A.9. If the minimum approach to DNB is corrected to TRIGA Mark F conditions and cross-plotted, an estimate of the effective gap conductance of 450 Btu fW hi' IF" is obtained for incipient burnout so thatthe case using 500 is thought to be representative of standard TRIGA fuel.

-The surface beat flux at the midplane of the element is'shown in Figure 4A.10 with gap conductance as a parameter. It may be observed that the maximum heat flux is approximately proportional to the heat transfer coefficient of the gap, and the time lag after the pulse for which the peak occurs is also increased by about the same factor. The closest approach to DNB in these calculations did not necessarily occur at these times and places, however, as indicated on the curves of Figures 4A.7 through 4A.9. The initial DNB point occurred near the core outlet for a local heat flux of about 340 kBtu fr2 hrI OF1 according to the more conservative Bernath correlation at a local water temperature approaching saturation.

K-State Reactor 4.A-5 Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A J-I This analysis indicates that after operation of the reactor at steady-state power levels of I MW(t), or after pulsing to equivalent fuel temperatures, the heat flux through the clad is reduced and therefore reduces the likelihood of reaching a regime where there is a departure from nucleate boiling. From the foregoing analysis, a maximum temperature for the clad during a pulse which gives a peak adiabatic fuel temperature of 1000 0 C is conservatively estimated to be 470C.

As can be seen from Figure 4.7, the ultimate strength of the clad at a temperature of 4701C is 59)000 psi. If the stress produced by the hydrogen over pressure in the can is less than 59,000 psi, the fuel element will not undergo loss of containment. Referring to Figure 4.8, and considering U-ZrH fuel with a peak temperature of 10000 C, one finds the stress on the clad to be 12,600 psi. Further studies show that the hydrogen pressure that would result from a transient for which the peak fuel temperature is I 150TC would not produce a stress in the clad in excess of its ultimate strength. TRIGA fuel with a hydrogen to 2irconium ratio of at least 1.65 has been pulsed to temperatures of about 1150TC without damage to the clad [Dee et al. 1966].

7 I 4 I

\ ELAPSED TIME FROM

\END) OF PULSE - 0.247 SEC 6

I-C4 ACTUAL HEAT FLUX

. . 5

, I CRITICAL HEAT FLUX \

.U 4

2.

I.

3 7 8 9 10 11 12 13 DISTANCE FROM BOTTOM OF FUEL (IN.)

Figure 4A.7, Surface Hcat Flux Distribution for Standard Non-Capped (h..pw I..

500 Btulh ft *F) Fuel Element Aftcr a Pulse.

K-State Reactor 4A-6 Original (9102)

Safety Analysis Report

.REACTOR DESCRIPTION i.'

T .

e z1 s

7 a * 'to 11 1 13 IS

)ISTANCE rOM 9oTIOn OF NUZL(IR.)

Figure 4A.8, Surface Hcat-Flux Distribution for Standard Non-Gapped Fuel Element (hP- 375 Btulh ft2 'F) After a Pulse.

7 . *CflITlC~L HEAT FLUX.

AI-ELAPSED TIME FROM ENi OF PULSE IS 0.44D SEC

< 3.H 2_

7 8 9 10 11' 12 .13 14 IS DISTANCE FROM BOTTOt OF FUEL (IN.)

Figure4A.9, Surfatec Heat-Flus Distribution for Standard Non-Gapped Fuel Elemcnt (hip=

250 Btulh ft2 F) Aftcr n Pulse.

K-State Reactor 4.A-7 Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A

106 p-I-

IL. lot 0.02 0.1 I.0

LAPSED TIME FROM END OF PULSE (SEC)

Figure 4A.10, Surface Heat Flui at Mildpoint vs. Time for Standard Non-Gapped Fuel Element After a Pulse.

Is K-State Reactor 4A-8 Original (9102)

Safety Analysis Report

REACTOR DESCRIPTION Bibliography "A Theory ofLocal Boiling .Burnoutandlts Application to EristingData " Heat Transfer -

Chemical Engineering Progress Symposium Series, Storrs, Connecticut, 1960, v. 56,No.

20Bernath, L, Research in Improved TRIGA ReactorPerformance,FinalReport, General Dynamics, General Atomic Division Report GA-5786, October 20,1964. Coffer, C.O., et al.,

CharacteristicsofLarge Reactivity Insertions in a High PerformanceTPIGA U-ZrH Core, General Dynamics, General Atomic Division Report GA-6216, April 12, 1965.Coffer, C. 0., et al.

Annular Core Pulse Reactor, General Dynamic, General Atomic Division Report GACD 6977, Supplement 2, 1966.Dce, J. B., T. B. Pearson, J. R. Shoptaugh, Jr., M. T. Simnad, Temperature J'ariation,Heat Transfer. and Yoid Volume Development In the Transient Atmosphere Boiling of Water, Report SAN-1001, U. Cal., Berkeley, January, 1961. Johnson, H.A., and V.E. Schrock, ct al.,

A Study ofihe Mechanism ofBollingHeat Transfer, JPL Memorandum No. 20-88, March 1, 1954Ellion, M-E.,

Thermal ConductanceofMetalfic Surfaces In Contact, USAEC NYO-2130, May, 1959;Fcnecb, H., and W. Rohsenow, An Analysis ofa Thennal ContactResistance,-Trans.ANS 5, p. 476,1962.Fenech, H., and IJ.

Henry, "ThernalConductanceAcrossMetalJoints, " Machine Design, Sept. 15,.1960, pp 166-172.

Grafr, NJ.

Heat Transmission,3rd Ed., McGraw-Hill, 1954McAdams, -W.H..

MNRC, McClellanNuclearRadiationCenterFacili SafetyAnalysisReport, Rcv. 2, April 1998.

Heat, Mass andMomentum Transfer, Prcntice-Hill, 1961, pp 231-232.Robsenow, W., and H.

Choi, "QuarterlyTechnicalReport SPERT.Project,April, May, June, 1964," ISO 17030. Spano;A.

H.,

"he Effect ofSubcooledLiquidonFilm Boiling,"Hcat Transfcr 84,149-156, (1962).Spaxrolv, E.M. and R.D. Cess, K-State Reactor 4A-9 . Original (9102)

Safety Analysis Report

CHAPTER 4 APPENDIX A "Onset ofStable Flm Boiling and the Foam Limit," Int. J. Heat and Mass Transrcr 6,987-989, (1963). Speigler, P., et al.,

UTA, University of Texas atAustin TRIGA ReactorFacilitySafetyAnalysisReport, Docket 50-602, Rev. 1.01, May 1991.

"HydrodynamkcAspects ofBolfing Heat Transfer," AEC Report AECV-4439, TIS, ORNL, 1959.

Zuber, W.

K-State Reactor 4A-10 Original (9102)

Safety Analysis Report

Appendix B Water Properties at Nominal Operating Conditions The following data are from the NBSINRC Steam Tables, byjL. Haar, J.S. Gallagher, and GS.

Kell, Hemisphere 1984.

Saturated-Water at 0.153 MPa TEMPERATURE, T 111.9 c PRESSURE, P 0.153 MPa HEAT OF VAPORIZATION, 2. 2.22E+03 kJ kg7' SURFACE TENSION. cI 5.66E-02 J mr

p. density (kg mn3) 9;50E+02 8.78E-01 C, heat capacity (kl kg` K ) 3.71E+00 1.55E+00 C,5 heat capacity MkIk. WK 4.24E+00 2.09E+00 s, entropy (kl kg. K') lA14Et+0 7.22E+00 i, cnthalpy (kIJ kg') 4.70E+02 2.69E+03

u. internal energy (kJ kg') 4.69E+02 2.52E+03 sonic speed (m s") 1.53E+03 4.79E+02 k, thermal conductivity (W in 1K) 6.82E-0I 2.65E-02 1z. dynamic viscosity (kg mi's') 2.50E-04 1.27E-05 v, kinematic viscosity (m2 s') 2.64E-07 -IA5E-05 S thermal diffusivity (mr'1)2 1.70E-07 1.44E-05 Pr, Prandtl Number 1.55E+00 1.00E+00 fi volumetric expansion coefficient 8.14E-04 2.86E-03 Subcooled Water at 27 0C, 0.153 MPa TEMPERATURE (I) 27 eC PRESSURE (P) 0.153 MPa DENSITY (p) . 9.97E+02 kg m C, heat capacity (IJ kg"' K7") . aE C,* heat capacity (QJ ' ICK) s, entropy (J kg" I) i, enthalpy (k kg')

u, internal energy (kl kg')

sonic speed (m s')

k, thermal conductivity (W ni'IC')

p, dynamic viscosity (kg m's")

v, kinematic viscosity (m2 s') -

a, thermal diffusivity (m 2 s'")

Pr, Prandtl Number

, volumetric expansion coefficient K-State Reactor 4.8-1 . Revised 12123104 Safely Analysis Report

5. REACTOR COOLANT SYSTEMS The reactor coolant systems are very simple in design and operation. These systems are required from an operational standpoint, and not safety. Potential hazards associated with the primary cooling system are minimal. Many of these systems have been upgraded in recent years to permit extended full power operation of the reactor. A general overview of the reactor coolant system is follows.

During full power operation, the fucl elements in the reactor core are cooled by natural convection of the primary tank water. To remove bulk heat to the environment, primary water is circulated through a heat exchanger where the heat is transferred to a secondary cooling loop.

Water passing through the-sccoridary side of the heat exchanger is then passed through a forced-draft cooling tower to transfer beat to outside air. . x A cleanup loop maintains cleanliness of the primary water to minimize production of long-lived radionuclides and minimize corrosion. Radioactive contamnination of the primary water does not present a hazard to workers tin fact, activities are low enough to be directly released to sanitary sewerage). Primary coolant does provide shielding directly above the reactor core, and loss of this shielding would elevate radiation levels in the reactor bay, especially directly above the reactor tank.

Loss of primary water would also deprive the reactor fuel of its principal means of cooling.

However design analysis of TRIGA fuel shows that it may be cooled by natural convection in air, without risk of fuel failure. In the event of a water loss, makeup water for the primary system is available from a still or the bulk shield tank. If repairs of the primary tank require draining, fuel elements can be stored in either dry storage pits or the bulk shield tank.

The primary and secondary cooling system flows are isolated by a plate-type heat exchanger. A %

rupture in the heat exchanger could cause mixing of the coolant streams. Therefore, although the primary coolant is of little radiological significance, the secondary coolant is monitored periodically for radioactivity to detect even a small breach between the two systems.

5.1 Summary Description The reactor cooling system serves five major functions:

I. Remove and dissipate heat generated in the reactor

2. Provide radiation shielding from the core area

3. Control primary water conductivity (to minimize corrosion of reactor components, particularly the fuel elements)

4. Control primary water radioactive contamination (by removing nearly all particulate and soluble impurities)

5. Maintain optical clarity of the primary water

K-State Reactor 5-1 Original (12104)

Safety Analysis Report

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CHAPTER 5 V

C 2

L Abe primary system contains de-ionized water and is open to atmosphere. The reactor core is cooled by natural convection alone. To permit extended operation at full licensed power, bulk heat is transferred by forced convectiorilcioss a heat exchanger to the secondary cooling system.

The secondary cooling systeni then transfe=s the beat to the environment via a cooling tower, using service water trcatcd for corrosion and biological growLt.

This cooling system comibination provides enough heat removal for continuous full-power operation. In addition to the cooling system, the reactor is provided with a bulk-shielding taik 1This 6500-gallon (25 kL) tank contains distilled water, and can be used to supplement make-up wvater for the primary tank (using a makeup wvater system independent or the above drawing)'or provide temporary fuel storage. Maikeup water for both systems is provided by a steam-powered stiL -l In normal (automatic) control of the cooling systems, a single backlit pushbuttoa switch on the control console energizes the primary and secondary pump, and the cooling tower fan control system. Individual controls for system components are located on the 0-foot level. In addition to control systems, several cooling system parameters can be monitored in the control room.

Measurement functions available to thereactor operator are presented in Table 5.1.

K-State Reactor 5-2 . Original (12104)

Safety Analysis Report

REACTOR COOLANT SYSTEMS Table .1: Control Room lnstrumentation *

Measurement Location Device I Output Primary 'ater Conductivity Cleanup Loop Inlet Platinum Probe l Meter Cleanup Loop Outlet Platinum Probe 1 Meter Primary WVater Temperature Water Box RTD /Meter Heat Exchanger Inlet Transducer/ Computer Heat Exchanger Outlet Transducer / Computer Reactor Tank (x 2)

2 Transducers I Computer Primary Flow Rate Orificc 2 Transducers I Computer Primary Radioactivity Water Box G-M Tube I Mctcr Pool Surface G-!MTube/Meter Secondary WaterTemperature Heat Exchangcr Inlet Transducer I Computer Heat Exchanger Outlit Transducer I Computer 5.2 Primary Coolant System IJ Principal functional requirements of the primary coolant system are to (1) transfer heat from the reactor core to the secondary cooling system, and (2) provide radiation shielding directly above the reactor core. Although natural convection cools-the reactor core, pi.-jnary bulk water temperature should be kept below 1300 F (48.9 IC). This limit could easily be reached during extended operation at full licensed power, hence the need for a beat transfer loop. The second requirement involves the 16-feet (4.9 m) of shielding that the water provides.directly above the reactor core.

The primary coolant system layout is shown in Figure 5.1. The system consists principally of a reactor tank, a centrifugal pump, a pressure orifice, one side of the heat exchanger, and a cleanup 1%

loop (described in Section 5.4). :Since the prinmay coolant is de-ionized water, system components and piping are constructed of either aluminum or stainless steel. Most of the piping is aluminum of nominal 2.5-in (6A cm) dianmeter.

Cooling system inlet piping has two suctions, one just under the water surface and a second suction through a skimmer, which collects foreign particles on the pool surface. An installed valve may be used to isolate the skimmer. The tank itself provides 22-ft (6.7 in) of head (9.5 psig). This water is pulled into a self-prining, direct-coupled, centrifugal pump located on the 0-foot level of the reactor bay. The pump boosts vater pressure to about 60 psig (410 kPa), with a flow rate of approximately 110 gallons per minute (6.94 L sl).

From the primary pump, the flow is split, with about 10 gallons per ininute (0.63 L sl) diverted through a cleanup loop and the remainder passing through an orifice to the beat exchanger, also located on the 0-foot level. The orifice provides the necessary pressure drop to flow water through the cleanup loop. Exit pressure on the orifice.(and hence the entrance pressure to the heat exchanger) is approximately 30 psig (210 kPa).

A plate-type compact heat exchanger is used to remove heat from the primary coolant (see Figure 5.2). The heat exchanger consists of sandwiched stainless steel plites alternately carrying primary and secondary cooling water. .The heat exchanger has a transfer capacity of 682 kW K-State Reactor 5-3 Original (12104)

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CHAPTER 5 (2,327,080. BUT hI) under normal conditions. The unit was originally pressure tested at 130 psig (0.89 MPa). Four temperature transducers are located in both inlet and outlet streams, as well as two in ihe reactor tank, for performance monitoring of the heat exchanger. Additional plates may be added to the heat exchanger to provide expand cooling capacity. The compact heat exchanger and temperature probes were installed in-l993.

In addition to cooling functions, the primary tank pr6vides 16 ft (4.9 m) of shielding directly

above the reactor core. Water level is normally kept within a few inches of the top of the reactor tank. To prevent loss of prim4ary.coolant during maintenance operations, four valves allow isolation of the heat exchanger or reactor tank. In the event of a rupture in the primary piping, a siphon break is located about one loot below the water surface of the tank. The siphon break is a small hole in the pipe that allows air into the pipe if %waterlevel drops below the break point, breaking the siphon. The beam ports, when closed, are sealed on the outside by a gasket. A system of pipes connects the beam ports to the manifold with i pressure gauge, where any rupture would be indicated as increased pressure.

if a major loss of coolant were to occur, there are three level sensors that would illumifnte lights on the control panel. Two sensors are located in the reactor bay sump, activating when the sump level is high. Since all floor drains in the reactor bay connect to the sumip, any lcaks would accumulate-there. A third sensor is located at the top of the tank, activating if the tank level drops a few inches below normal operating levels. If these indicators fail, loss of coolant would be apparent by increased radiation readings on the remote area-monitors. The increased radiation levels directly above the reactor core could present a ridiological hazard on the 22-foot level.

The primary reactor tank serves as a shutdown cooling pool and spent fuel pool, therefore the impact of loss of'primary water on cooling fuel elemients was evaluated. General Atomics calculations for aluminum-clad fuel show that after shutdown from infinite operation at 250 kW that the maximum temperature at the fuel cladding interface would be less than 150 0C. .For these conditions, the pressure exerted by trapped air.and fission products is about 660 psi (4.6 MPa),

3.vhereas the yield stress for the aluminum cladding is >5000 psi (34 MPa) at 150C. The * .

Division of Reactor Licensing validated these findings in the 1968 Safety Evaluation. Current fuel inventory has stainless steel cladding, with two spare aluminum clad instrumented elements.

Stainless steel has greateryield strength, hence greaterresistance to cladding failure. Technical Specification G2 allows for spent fuel'elements to be stored in air or water.

5.3 Secondary Cooling System .

The secondary cooling system is designed for continuous operation at approximately 2,437,500 BTU h' (723 kW). The secondary cooling system circulates water from the heat exchanger through a cooling tower. The system uses service wvater that is treated to minimnze corrosion and bacteriological growth. The system consists of a surge tank, centrifugal pump, heat exchanger, and cooling tower.-- The system is regulated by an automatic'control system consisting of several temperature sensors, a pneumatic thrce-way valve, and electrical controls. With the eception of the cooling tower, aU equipment is located on the 0-foot level of the reactor bay.

K-Slate Reactor 54 Original (12/04)

Safety Analysis Report

REACTOR COOLANT SYSTEMS 5.3.1 SecondaryCooling System Flows.

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Water is drawn into the secondary cooling system from a surge tank. The tank has a volume greater than that of the cooling tower and stores this water when the tower is not in use. The tank has a sight gauge that is visible from the control room; and an indicator on the control console, which illuminates on low water leveL Service water makeup can be initiated from the control room by resetting a five-minute timer (to prevent overfilling) that energizes an electric valve connected to potable water. The' inakeupline terminates above'the'tahk; providing an air-gap for protection of potable water.

The water from the surge tank is then drawn into the system by a direct-coupled, self-priming, centrifugal pump(see Figure 53). When the controls are in a normal configuration, the pump is K-Stale Reactor 5-5 Original (12104)

Safety Analysis Report

S C CHAPTER5 energized (with the primary pump and cooling tower fans) from a backlit push button switch in the control room. Normal flowratc throughbthe system is250 gallons pcr minute (15.8 L.sl).

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Figure 53: Secondary Cooling Pump After leaving the pump the water passes through the beat exchanger, receiving beat fronm the primary loop. The temperature transducers on the heat exchanger provide temperature data in the control roonL Several manual temperature gauges are also located at various locations in the reactor bay.

Normally, water leaving the heat xchanger is then sent to the cooling tower (Figure 5.4) and returned to the surge.tank. . However during cold weather conditions (when outside air is less than -10 IF, -23 IC), the water bypasses the cooling tower at a three-way valve and is returned directly to the pump. *Any water in the cooling tower then drains into~the surge tank. The cooling tower is an induced-draft, crossflow design constructed of galvanized steel with a 146ton cooling capacity. It is located outside the reactor building, outside a fenced area at about the 12-foot level of the reactor. The tower contains 142 gallons (538 L) of water during operation. It has a two-speed fan controlled by the sccondary automatic control system. ITe tower was replaced along with the surge tank in 1991. ITe towcr wa7s upgraded in 2001 to permit 500 kV operations.

5.3.2 Secondary Cooling Automatic Control System The secondary automatic control system has three basic control functions:

1. *To maln'tai the primary waier it a set temperature.

2. To control the cooling tower fan speed to maintain secondary water at a set temperature.

3. To prevent cooling water freeze-up during cold weather operation (less than -10 0J (-23 IC).

The secondary automatic control system performs these functions by pneumatically controlling the three-way valve and electrically controlling the cooling tower fan. The secondary automatic control system is normally initiated by energizing the primary cooling pump. The three-way K-State Reactor 5-6 Original (12104)

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REACTOR COOLANT SYSTEMS J1 valve allows secondary water to flow to the cooling tower under normal conditions. If the outside ji air temperature is less than -10 IF (-233 IC), then the three-way valve stops cooling tower flow. j This outside air temperature limit is bypassed when the primary water temperature exceeds 110 IF (43.3 0C) thus reestablishing cooling tower flow. The temperature of the secondary water j

returning from the cooling tower controls cooling tower fan speed. At 70 IF (21.1 IC:), the J1 cooling tower fan starts at low speed and switches to high speed when the temperature reaches 901F (32.2 IC.). Manual control of fan speed is available at the O-root level breaker.

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5.3.3 Secondary Water Quaiity.: "

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Secondary water chemistry analysis is typically performed twice a month. Items tested are pH, chlorine, conductivity, and total alkalinity. Suggested manufacturers limits on pH are 7.6 to 8.6, and our typical values are 7.5 to 8.0. Chlorine levels and conductivity should be less than 4 times the maximum city concentration. Chlorine levels and conductivity of the city water are 2-4 ppm at point of use and OA to 0.7 mS cm i. Upon reaching a conductivity value 4 times that of service water, the cooling wvater is replaced. Total alkalinity should be less than 270 ppm, although this limit has never been reached during normal operation.

To detect possible leaks in the heat exchanger, the secondary water is tested monthly for radioactivity. Since the primaiy tank has 22-feet (6.7 m) of static head, as opposed to 3 to 9-fcet (0.9 to 2.7 m) in the secondary (depends on surge tank level), a breach in the beat exchanger would result in flow from the primary to secondary cooling system when the cooling system is secured. Leaks while the cooling system is operating arc unlikely; the heat exchanger was K-State Reactor Original (12104)

Safety Analysis Report

Fiiji CHAPTER 5 pressure tested to 150 psig, and the maximum pressure differential across the heat exchanger (primary pump running, secondary pump oft) is 30 psig.'A small breach in the heat exchanger would be evidenced by tritium contamination of the secondary water. A larger breach would be indicated by.loss of primary coolant-from the reactor tanL. Primary water typically remains

.within IOCFR20 limits -for release to-sanitary sewers, and is not a hazard even if a leak were to occur.

5.4 Primary Coolant Cleanup Sypte~m Technical specification requires average monthly water conductivity be kept below 2 pS cm' to minimize corrosion of reactor components and production of radioactive materials, as well as maintaining optical quality of the coolant. To maintain this low conductivity the primary coolant system was constructed with an integral cleanup joop, located on the 0-foot level of the reactor bay. As shown in Figures 5.1 and 5.5, this cleanup loop consists ofa water box, a fiber cartridge filter, a mixed-bed resin demincralizer, a flow meter, two conductivity probes, a RTD temperature probe, and a Geiger tube. Conductivity. measurement equipment was replaced in 1990.

Connecting piping is 1;in (2.5 cm) aluminum.

. : .I-WVater is drawn out of the primary system after leaving the primary. pump and returns to the system with the water exiting the beat exchanger. A flow orifice (Figure 5.6) is installed in the main primary loop before the heat exchanger to provide the necessary pressure drop to force water through the cleanup loop. Upon entering the cleanup loop the water passes through the water box, where 'temperature, conductivity, and radioactivity arc measured. Normal inlet conductivity is 0.8 to 1.3 pS cm1'.

The water is then passed through a filter assembly that contains two replaceable cartridges of at least 5-micron rating. Pressure gauges are supplied on either side of the filter to indicate clogging. In addition to improving the optical clarity of the water in the reactor tank, the removal of solid particles extends the life of the demineralizer resin. .

The water then passes through a mixed-bed demineralizer to remove soluble impurities. The typical resin in use is Amberlitc ARN-150. After the demineralizer, the water passes a second conductivity probe. Typical exit conductivity is 0.05 to 0.2 PtS cm 1l. Flow rate through this loop is regulated by a plug-type flow meter with a 0 to 28 gallon per minute range. A manual valve accomplishes regulation. Normal flow rate through this loop is 10 gallons per minute. Normal pressure in the loop is 55 to 60 psig.

Primary coolant is sampled every mnibth for radioactivity. The only nuclide of significance that is normally detected in the coolant is tritium arid is usually in therange of 200 to 1500 pCi rnt'.

In the event of a fuel cladding failure, the cleanup loop could be used to remove radioictive contaminants.

K-State Reactor 5-8 Original (12104)

Safety Analysis Report . .

'ii dl REACTOR COOLANT SYSTEMS dl Figure 5.5: Primary Coolant Cleanup Loop -

5.5 Makeup Water System i-.

Makeup water for the pinmary system and the bulk shield tank is provided by a still located in the basement adjacent to the reactor bay (Figure 5.7). The still is a stcam-powered unit, with steam supplied by the University's power plant. The steam is distributed in underground passages making service relatively secure from ordinaTy hazards of weather. The still is rated to produce 50 gallons distilled water per hour (3.2 L min ) and has a storage capacity of 80 gallons (300 L).

Input water to the still is conditioned by a water sollener to minimize scaling in the boiler section.

In addition to the still, the bulk shield tank contains 6500 gallons (25 kL) of distilled water, which could be used as makeup water.

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I ... a .. Figure 5.7: Makeup NYatcrSystczn 5 Ng. . e n 1 5.6 Nitrogen-16 Control System ;

The cooling systems return line to the reactor pool enters the pool through a diffuser. The diffuser is conitructed to induce a helical flow pattern in the rFactor tank. This extends transport time of the convection flow of water from the core to allow much of the niftogen-16 generated

during operation to decay before reaching the pool surface. 'A radiation monitor directly above the pool surface provides the control room operator with information to prompt exposure controls (generally energizing the primary cooling pump to initiate the helical flow for dccay, or limiting access to the area directly over the pool). Pool surfate monitor radiation measurements at 250 kW directly above the pool surface are typically 20 to 30 M&.1R! from ail sources with the primary cooling system operating, and is expected to be 90 mrR h1 at 500 kW operation, expected to be approximately 350 nR h'"at 1,250 kW operation (slightly less than 100 mnR-4 at I meter above the bridge). *-

'A radiation-monitor at the rail around.the pool provides the-control room operator with information to prompt exposure controls for personnel on the 22-foot level butnot directly over the reactor pool. At 250 kW, radiation levels at the rail are less than 2 mR/hr.f 5.7 Auxiliary Systems Using Primary Coolant

-*Although the bulk shield tank does not circulate with-the primary coolant, it can be used as a source of distilled water. Shielding of the thermalizing column by ihis tank is only needed for operation. The bulk shield tank water circulation syslem has the provision to allow for water to be pumped directly from the bulk shield tank into the primaly tank, providing 6500 gallons (25 kL) of makeup water. The bulk shield tank can also be used for temporary fuel storage during maintenance operations involving the primary tank. .Makcup water for the bulk shield tank is provided by the same still as used for the primary syiten K-State Reactor 5-10 Original (12104)

Safety Analysis Report

REACTOR COOLANT SYSTEMS 5.8 Bibliography GA 7860, Safety Analysls Reportfor TPRIGA Reactors usingAluminum-Clad Fuel, 6 March 1967 General Atomics/General Dynamics.

Safety Evaluation by the Division of Reactor LicensingforIncrease In Power Level, Skovbolt, Donald J. . 26 June 1968.

K-State Reactor 5-11 Original (12104)

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.' I K-State Reactor

6. Engineered Safety Features As discussed in Chapter 13, from previous analysis, and from experience at other TRIGA reactors, emergency core cooling is not required for opcrations at steady state thermal powers below 1900 kW. No engineered safety features are required for the KSU reactor because the steady state power limit is 500kW.

6.1 Bibliography KansasState University TRJGA Mark H1 ReactorHaxardsSummary Report, by R.W. Clack, J.R.

Fagan, W.R. Kimel, and S.Z. Mikhail, Licensc R-88, Docket 50-188, 1961.

Analysis of CertainHazardsAssociatedwith Operationof the Kansas State University TRIGA Mark1JReactorat 250 kiYSteady State andivith PulsedOperationto $2.00, by R.W. Clack, et al., and the Safety Evaluation by the U.S. Atomic Energy Commission Division of Reactor Licensing, License R-88, Dockcet 50-188, 1968.

NUREG-1282, "Safety Evaluation Report on High-UraniumContent, Low-Enriched Uranium-Zirconium Hydride FuelsforTRIGA Reactors,"U.S. Nuclear Regulatory Commission, 1987.

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StRc K-State Reactor

7. PPINSTRUMENTATION AND CONTROL ji SYSTEMS Much of the reactor's original instrumentation and control (I&C) systems were replaced during the control room modifications in 1993 and 1994. The original console and vacuum-tubc instruments were replaced by a surplus solid-state console obtained from U.S. Geological Survey's TRJGA Mark I reactor. This console was then outfitted with new N-I000 series _j neutronic channels from General Atomics. These channels have optically isolated outputs, allowing other devices to utilize the neutronic data.

7.1 Summary Description The bulk of the reactor I&C systems are hard-wired analog systems primarily manufactured by General Atomics and widely used at varibus NRC-licensed facilities. The general layout of these systems is shown in Figure 7.1.

.r REACTOR OPERATOR INPUTS : CONTROL CONSOLE OUTPUTS R,,aW 11 Iadirr' co ote8* I IjVmt h I1 Figure 7.1, Inter-connectivity Diagram.

The reactor control system (RCS) consists of the instrumentation channels, the control rod drive circuitry and interlocks, and an automatic flux controller. The RCS measures several key reactor parameters including power, fuel temperature, water temperature, and water conductivity. Three K-State Reactor 7-1 Original (12104) *_

Safety Analysis Report

vil CHAPTER 7 neutronic instruments measure reactor power separately: a wide-range logarithmic channel, a multi-range linear channel, and a percent power channel. These provide at least two indications of reactor power from source range to power range. Additionally, if a reactor pulse is performed another channel is added to the central thimble to record pulse data. Fuel temperatures can be monitored on both the console and on an auxiliary panel.. Primary water temperature is displayed on the console and measured by an RTD in the water box. Titanium electrodes at.the entrance and exit of the cleanup loop measure water.conductivity.

The control rod drives and their associated circuitry are simple in design. A rotary switch configures the primary mode of operation, namely automatic, steady-stale, or pulse mode.

Numerical indicators give drive position, with illuminated switches to manipulate the rods and to indicate rod and drive status. Several interlocks are incorporated to prevent unintentional rapid insertions of reactivity, except in pulse mode. An automatic control system links the RCS with the neutronic channels providing regulation of the power level.

The reactor protection system (RPS) is a component of the RCS instruments. The RPS will initiate a reactor scram if any of several'measured parameters in the RCS are outside of their limited safety system settings. The reactor scram effectively places the reactor in a subcritical configuration by releasing the control rods from their respective drives. Since the rods are no longer physically attached to the drive, they fall into the reactor core by gravity. High reactor power, high fuel temperaturc,.loss of detector high voltage, loss of building power, and short reactor period will automatically cause all -ofthe control rods to be dropped into the reactor corc.

A bar above the control rod drive svitcies allows this systemnto be actuated manually. Since the core is cooled by natural convection, no other engineered safety features are necessary for safe reactor shutdown.

The control console and display instruments are primarily housed in a control console, with auxiliary instruments located in a rack next to the console. At the console, the reactor operator has direct control over mode of operation, control rod drive positions, cooling system operation, opening of reactor bay doors, and manual scram of the reactor. Display instruments located in the control console provide measurements of reactor power, control rod positions, primary water temperature, and fuel temperature. Indicators in the console display scram information; lo* air pressure, low primary water level, high reactor sump water level, sump high water level, sump, overflow water level, secondary surge tank level low, source interlock status; reactor bay upper door open, reactor bay lower door open, thermal column door open, person on stairway, and rod drive status. Secondary surge tank makeup is controlled with n backlit pushbutton that indicates surge tank low level and surge tank makeup valve operation. An intercom system on the console provides communication to numerous locations around the reactor bay and staff offices. In the auxiliary rack, the operator can control pneumatic transfer system operation, actuate timers, and add water to the secondary cooling system. Display instrurnents located in this rack include, primary water conductivity, water activity; remote area radiation monitors, fuel temperature, and a strip-chart output of reactor power. Several audible alarms indicate high radiation levels in the primary-coolant and at various locations throughout the reactor bay. A breaker-box in the control room provides control over electrical devices in the reactor facility, including ventilation systems.

Radiation protection instruments arc distributed throughout the reactor bay. All instruments have visual indication of radiation level, visible alarm conditions, and audible alarm. Radiation area K-State Reactor 7-2 Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS monitors (RAM) strategically cover potential radiation areas throughout the reactor bay. A combined pool surface and primary water monitor indicated water activity. A 5 Rh l evacuation alarn is located on the 22-foot level. A continuous air monitor (CAIM) is energized during reactor operation.

The human-machine interface principles incorporated into control room design allow the reactor to be operated by a single individuaL All monitoring instruments are visible to the reactor operator at the console. The instruments and controls necessary for reactor operation are within reach of the operator, including an intercom and telephone. Surveillance instruments are located next to the console, with visual and audible alarms to signal the operator to abnormal conditions.

7.2 Design of Instrumentation and Control System 7.2.1 Design Criteria Reliability of essential equipment is ensured through redundancy. Multiple instruments and safety systems perform similar functions for all modes of reactor operation. The construction and installation of instruments was performed according to applicable regulations at the time of introduction. However, all crucial instruments arc checked daily for calibration and operability.

Testing and calibration procedures exist for repair and general service. Themajority of these I&C systems were manufactured by General Atomics, or other industrial manufacturers of nuclear equipment. Crucial systems to be considered include neutronic instruments, control rod drives, radiation monitors, and control systems.

Redundancy is designed into each of these systems. During steady state operation, a minimum of two ncutronic channels provide reactor power level indication, two of which provide high power level RPS actuation (scram). These neutronic instruments are tested prior to reactor operation for demonstration of scram capability. Two arc also tested for operability by internal calibration tests. There are two fuel temperature indications. The control rod drives drop their rods into the reactor core upon loss of power or RPS actuation, providing sufficient shutdown margin with even the most reactive rod stuck out. Multiple remote area radiation monitors cover important areas, including two directly above the reactor core and two monitoring primary coolant activity.

The maximum steady state power level for K.SU TRIGA Mark II reactor is proposed to be 500 kWV. Similar reactors operate up to 2 MW with 2 GW pulses. Therefore, the limited safety systems settings associated with reactor are extremely conservative when compared to the safety limits of the reactor. Thus the reactor has a considerable safety margin.

7.2.2 Design-Basis.Requirements The primary function of the RCS is to govern the manner in which reactivity is varied in the reactor core. The RCS system should prevent the reactor operator from unintentionally inserting large amounts of reactivity, through various interlock systems. The operator should only be able to remove one rod at a time from the reactor core, preventing large insertion rates. The pulse rod must not be able to be rapidly ejected from the core while in steady-state operation. Furthermore, the pulse rod should be the only rod that can be *withdrawn in pulse mode, preventing K-State Reactor 7-3 Original (12104)

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CHAPTER 7 supercritical pulses. There should also be an interlock to prevent startup without a power level signal above the minimum instrument sensitivity, preventing unmonitored or unanticipated criticality. Rod position indicators should show the rod position to 0.2% or total travel for accurate reactivity calculations.

Another primary function of the RCS is to provide the reactor operator with reactor status information. Reactor power, a crucial parameter, requires at least two instruments to provide confirmation of reactor power from shutdown to operating levels. Three instruments are used to cover this range: a wide-range logarithmic channel, a multi-range linear channel, and a percent power channel. Accuracy of measurement at full rated power increases accordingly with the refinement of scale. The log channel provides gross reactor power indication and is accurate to 20% of scale, the linear channel is accurate to 5% of scale, and the percent power channel is accurate to 3% of scale. The percent power channel wiU also display pulse parameters for large pulses. These instruments are calibrated annually and checked for operability at the start of each operating day. An additional channel is installed and calibrated in the central thimble to record pulse data. Fuel temperature must be monitored during pulsing operation.

The primary function of the RPS is to automatically insert the control rods into the reactor core when ceitain parameters deviate from limited safetysystem settings. Several scrams involve the neutronic channels in the RCS. If I I0% rated power level is exceeded in steady state mode, one of two trip-points will scram the reactor. Failure of the high voltage power supplies for operating neutronic channels will also cause a scram. Manual scram will be available in all modes of reactor operation. Rod drop times for the standard rods will bc measured regularly to ensure proper RPS function. No other ESF features are required in this design.

The primary function of the radiation snonitoring instruments is for personnel protection measures and emergency assessment actions. *The area monitors provide the reactor operator with information regarding the actual radiation environment inside the reactor bay. With this knowledge, reactor users can be informed of possible hazards. A 5 Rtb" monitor on the 22-foot level signals personnel to evacuate the reactor bay. A number of survey instruments (ion chambers, rem balls, G-M counters) are also available to personnel. Other instruments such as the constant air monitor, pool surface monitor, and.water box monitor indicate the presence of dispersible radioactive materials, an indication of possible fuel cladding failures.

The control room is designed so a single operator can manipulate all significant controls without leaving the room. The reactor operator should be able to de-energize all equipment and experiments in the reactor bay. The control room should provide sufficient ventilation to provide cooling of the reactor instruments.

K-State Reactor 7-4 Original (12104)

Safety Analysis Report

%J INSTRUMENT AND CONTROL SYSTEMS %J 7.2.3 System Description The overall system layout is depicted in Figure 72. The majority of the RCS is housed in a General Atomics (GA) console originally manufactured for the USGS reactor, which is shown with modifications in Figure 7.2. A detailed description of this figure is provided in Table 7.1.

Figure 7.3 shows a representative layout of the auxiliary instrumentation rack. Since the instrument racks are general use equipment, configuration may be changed to allow better utilization of space, installation of new equipment, support specific equipment modifications, etc.

without affecting function. The functions of each piece of equipment in this configuration are discussed in following sections with additional figures showing location and layout.

7.2.4 System Performance Analysis The system performance of the current I&C systems surpasses the original equipment Reliability has bcen high, with few unanticipated reactor shutdowns. Since daily checkouts are performed, any discrepancies would be observed and corrected in a prompt manner. The opto-isolated outputs of the neutronic channels allow the data to be utilized by other devices without concern over those devices affecting. the channels. A line conditioner provides regulated power to the instruments, protecting the equipment from electrical disruptions.

7.2.5 Conclusion The current l&C systems outperform the original equipment supplied with the reactor, while meeting all of the necessary design bases for the facility. The human design factors used in control room development allow the reactor to be operated by a single individual. Checkout and testing procedures ensure that all equipment is maintained in operational status.

Figure 7.2, USGS TRIGA Console with Mlodifications.

K-State Reactor 7-5 Original (12104)

Safety Analysis Report

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CHAPTER 7 Rod drop timer Not used Not used Pneumatic system controls Strip chart recorder Spare high voltage power supply Drawer unit Battery charger Figure 73, Instrumentation Rack.

K-State Reactor 7-6 Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS Table 7.1, Description of Fig ure 7.2.

Number Function Description -

I Console Power Push Button Switch 2 Magnet Power/ Scram Reset Key Switch 3 Control Rod Drive Position Push Button Switches 4 Apply Air to Pulse Rod Push Button Switch S Rod Position Indicators LED Displays 6 Modc Selector Rotary Switch 7 Removed Backup Range Switch 9 Automatic Power Demand Control 10 Turn Potentiometer 10 Manual Scram Bar Bar Covering Scram Switches 11 (not used) Backup Period Channel 12 Fuel #2 & NVater Temperature Display and Selector Switch 13 Scram Status, Source Interlock, Low Indicators and Control 11_

Air Pressure, Hi & Hi-Hi suimp level, Switches surge tank level & makeup, Upper &

Lower Doors, and Cooling System Power

14. (not used) Backup Count Rite Channel 15 (not used) Backup Log Channel 16-17 (not used) Backup Percent Power Channel 18 (not used) N/A 20 Wide Range Log Power Channel GA NLW-1000 Channel 21 Multi-Range Linear Power Channel GA NMP-1000 Channel 22 Percent Power and Pulsing Channel GANPP-1000 Channel 23 Source Interlock Override Key Switch 24 Period Scram Overide Key Switch 7.3 Reactor Control System The bulk of the reactor control system (RCS) is housed in the USGS console shown in Figure 7.2.

The remainder is contained in the auxiliary rack-mount panel next to the console, shown in Figure 7.3. The RCS consists of the instrumentation channels, the control rod drive circuitry and interlocks, and an automatic flux controller. These are shown in Figure 7.4. The RCS-measures several key reactor parameters including power, fuel temperature, water temperature, and water I.--

conductivity.

7.3.1 Neutronic Instruments (Reactor Power)

Three neutronic instruments measure reactor power separately: a wide-range logarithmic channel, a multi-range linear channel, and a percent power channel, as shown in Figure 7.5. Wiring diagrams and calibration procedures are found in the instrument maintenance manuals listed in the bibliography.

K-State Reactor 7-7 Original (12104)

Safety Analysis Report

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I e1.

CHAPTER 7 The wide-range log channel uses a fission counter for detecting thermal neutrons in the range of 1.4 to 1.4 x 105 nv, and provides approximately 0.7 counts-nVl. The detector has an aluminum case, an aluminum electrode, a U30s (>90%/ enriched in 2U) coating as the neutron sensitive material, and an argon-nitrogen mixture for a fill gas. A preamplifier is used to minimize noise and signal loss from the detector to the console, and it is located on the 12-foot level.

MME

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i.

I_ bf myg I 1- -

.Figure 7.4, Instrumentation Diagram. I.

. . I .

The remainder of the channel circuitry is located in the NLW-1000 unit in the central console.

. The NLW-1000 unit supplies the high voltage for the detector and power for the preamplifier.

The instrument switches from pulse mode operation to current mode as reactor power increases

.::. out of the source range, allowing the instrunent to measure reactor power in the upper ranges.

Three displays indicate reactor power, high voltage, and reactor period. The power signal is permanently recorded via an opto-isolated output to a strip-chart recorder located in the

instruwentation rackr he period meter has a snad i at 3 sel and there is a high voltage scram,

io of which are r bypassed Tbothin pulsem mode.

h cbannm s e also provides a protective interlock which prevents rod withdrawal when indicated neutron flux is < 2 eps, which is.also activated in pulse mode to prevent removal of the shin, safety and regulating rods. Another interlock K-State Reactor 7-8 '.Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS prevents pulsing when reactor power is above 10 kW (normally set at I kv). The unit has two calibration checks in pulse mode, two in current mode, and checks for the period and high voltage scrams.

J1

.1 cJi Figure 7.5, N-1000 Series Instruments.

The second channel provides multi-range linear power indication. This channel uses a compensated ion chamber for detection of thcrmal neutrons. The linear channel detector signal goes directly to the NMP-1000 unit in the center console, which in turn supplies high and compensation voltages. This unit features automatic or manual ranging to select the appropriate decade of power displayed. The instrument provides two indicators, power and high voltage.

The power signal is permanently recorded via an opto-isolated output to a strip-chart recorder located in the instrumentation rack In addition there is also a high power level scram (normally set for 104% nominal rated power) and a high voltage scram. The signal from the detector and the high voltage scram are bypassed in the pulsing mode. The unit has two calibration checks, an auto-ranging test feature, and checks for high power level and high voltage scrams.

Power range indication of neutron flux is provided by an uncompensated ion chamber signal, which indicates percentage. of power in the upper two decades of the power range. The.

uncompensated ion chamber is virtually identical in construction to the compensated ion chamber, but no gamma compensation is provided in the circuitry. The detector sends its signal to the NPP- 000 instrument in the center console, which provides a visual indication of reactor power, high voltage, Dv, and nvt measurements. The NPP-1000 supplies the high voltage for the detector. There is a high power scram (normally set for 104% of full power) and a high voltage scram. In pulse mode, the channel is designed to read off the maximum power and integral output of a reactor pulse. However, the pulse output readings are measured in reference to the 250 MW maximum. Hence an additional channel is added to the central thimble to permit recovery of data from pulses of various magnitudes. The unit has checks for high power and high K-Stale Reactor 7-9 Original'(12104)

Safety Analysis Report I'-

CHAPTER 7 voltage scrams.

An added pulsing channel consists of a small BF3 chamber, which can be inserted into the central thimble of the reactor core. A separate high voltage supply powers the instrument and a multi-range picoammeter reads the detector currenL A reference voltage output of the picoammeter is sent to a computer in the control room, which collects the pulse data. This channel is calibrated prior to pulsing operations and range selected in advance based upon the anticipated peak power.

7.3.2 Temperature Temperatuic indications for the primary water and specific B-Ring fuel elements are provided on the front section of the control panel and in the instrumentation rack. .The instrumented fuel elements have three -chromel-alumel thermocouples in the fuel elcnent that are used for temperature indication on -the console or in the instrumentation rack. The thermocouples are located 0.76 cm (03-in) below the fuel surface, spaced at the midpoint of the element and at i, 2.5 (I in.) cm from the midpointjan averaged value from all three thermocouples is typically used for instrument readings. The temperature in the primary cleanup loop is a nickel alloy thermistor, and is displayed on a console meter, which is shared with fuel temperature viaa rotary switch.

Aiother indication of fuel temperature is located in the instrumentation rack with the capability of initiating reactor scram if the measured fuel temperature exceeds a preset Value (normally 400 C...

Several other temperature measurements can be obtained from the computer in the control room.

The computer can read two additional fuel thermocouples from other fuel elements in various positions in the reactor core. Additionally ADS90 temperature transducers are located on the inlet and exit of both the primary and secondary sides of the heat exchanger to evaluate performance. Two other transducers are located in the reactor tank for bulk pool tempernturc measurement and high temperature alarm.

7.33 Water Conductivity Primary water conductivity is measured at the inlet.and outlet of the purification loop by titanium electrode cells that send signals to a bridge circuit in the instrumentation rack. The bridge circuit is automatically temperature compensated and nulled to provide good conductivity measurements over all reactor conditions. The inlet and outlet conductivities provide a good indication of the overall purity of the primary water and the effectiveness of the ion exchanger.

73.4 7-

. Control Rod Drives Four control rods are required for reactor operations at 1,250 kW to meet reactivity control requirements: a shim rod, a regulating rod, a transient rod, and a safety rod. The shim, regulating rand safety rods share identical control circuitry (Figure 7.7) and provide coarse and fine power.

control. Tvo of the rod drives tre original, analog syst'C's. One of the rod drives uses a stepper m6tor. Drive position is determined by.voltage drop across a potentiometer that is adjusted as the control rod drive is moved. The position indicator for the analog motors is attached to a shaft K-Stale Reactor 7-10 . Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS coupled to the drive motor shaft with a setscrew, while the stepper motor is connected to the 1 position indicator with a chain drive. The pulse rod is designed so that it can be rapidly ejected from the core to a preset height to initiate a reactor pulse. However, it still functions as a normal control rod in steady state mode. All rods can be individually scrammed without shutting down the reactor.

a. Standard Control Rod Drives JI The rod drive mechanism (see Figure 7.6) is an electric motor actuated linear drive, equipped with a magnetic coupler. Its purpose is to adjust the reactor control rod J position. In the analog drive motor, a I 0-V, 60-cps, two phase motor drives a pinion l gear and a 10-turn potentiometer. The potentiometer provides rod position indication.

The pinion engages a rack attached to the magnet drawtube. An electromagnet mounted Kl on the lower end of the drawtubc engages an iron armature that screws into the end of a long connecting rod which terminates (at its lower end) in the control rod.

The magnet, armature, and upper portion of the connecting rod are housed in a tubular J barrel that extends well below the reactor water line. Located part way down the connecting rod is a piston equipped with a stainless steel-piston ring. Rotation of the o motor shaft rotates the pinion, thus raising or lowering the magnet draw tube. If the magnet is energized, the armature and connecting rod will follow the draw tube so that the control rod is withdrawn from or inserted into the reactor core. In the event of a reactor scram, the magnet wilt be dc-energized and release the armature. The connecting C rod, piston, and control rod will then drop, thus reinserting the control rod into the reactor. Since the upper portion of the barrel is well ventilated, the piston will move _

freely through this range. However, when the connecting rod is within 2-in (5 cm). of the %

bottom of its travel,-the piston is restrained by the dashpot action of the restricted ports in the lower end of the barrel. This restraint cushions bottoming impact. Control rod drop times are measured semi-annually and must be less than one second.

The analog rod drive motor is dynamically braked and held by an electrically locked motor. In the static condition, both windings are energized with the same phase (see \J Figure 7.7), electrically locking the motor. Clockwise (up) or counter-clockwise (down) rotation is enabled by shifting the phase between the windings with a 1-PF capacitor, motor control switches allow the appropriate phase shift. The stepper motor operates using phase switched direct current power. The motor shaft advances 200 steps per revolution (1.8 degrees per step). Since current is maintained on the motor winding when the motor is not being stepped, high holding torque is maintained. A translator module diives the stepping motor.

Three microswitches limit and control the travel of the magnet drawtube. Actuation of the magnet up limit microswitch (S901) applies line voltage to one winding therefore allowing only the phase shift, which gives counter-clockwise rotation.

Actuation of the magnet down limit microswitch (S902) applies line voltage to the other winding therefore allowing only the phase shift that gives clockivise rotation. Actuation of the rod down microswitch (S903A) causes the phase shift for counter-clockwise K-Sltate Reactor . 7-11 - Original (12104)

Safety Analysis Report

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CHAPTER 7 rotation. Therefore, if the control rod drops, the magnet drawtube drives down until the magnet down limit microswvitch locks the rotor. Since the rod down microswitch drives the magnet draw tube down, then the rod down microswitch must be open before the magnet down microswitch during coupled withdrawal of the control rod.

Three lights indicate that, 1) the magnet drawtubc is full up, 2) the magnet drawtube is full down, and 3) the armature and magnet are coupled. When the magnet drawtube is full up, microswitch (S901) is actuated opening the short across the magnet up light (DS321). WVhen the magnet draw tube is full down, microswitch (S902) is actuated opening the short across the magnet down light (DS324). When the control rod drops, the .non-actuated magnet down rnicroswitch (S902) and the actuated rod down microswitch (S903B) short the contact light (DS317) indicating separation ofthe magnet and thlearmature.

Other features of the circuit are an adjustable bias resistor (P902), a 220-ohm surge resistor, 50-ohm current limiting resistors. The adjustable bias resistor compensates for the torque applied by the weight of the control rod and the magnet drawtube. The 220-ohm surge resistor limits the capacitor current surge during switching. The 50-ohm current limiting resistors limit the iturrents in the 12-volt indicating circuits when the indicating lamps are shorted.

The unconventional circuit employed in the rod-drive system minimizes the number of switch contacts required. Therefore, relays with their attendant reliability problems are not required. It should be noted that the rod drive units are identical both mechanically and electrically (with the exception that one unit uses a stepper motor) and they are, therefore, interchangeable.

The rod position indicators are three digit, LED display indicators that receive a variable DC voltage input from 10-turn potentiometers that are driven by the respective rod drive motors. The digital display is simply a voltmeter, since the voltage across the potentiometer is directly related to the control rod position. The position indicators have their own variable power supplies aiid are tliciefore completely independent. The indicator systems are located in the control console except for the 10-tum potentiometers on the drive and the associated wiring.

Normal rod motion speed is about 12-in. per minute. Using rod speed, rod position indication at UP and DOWN limit switch positions, and respective rod worth curves, the operator can determine the reactivity insertion rate for a given interval of rod motion.

K-State Reactor 7-12 Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS

-MAGNET WIRE CONDUIT

-MAGNET DOWN ADJUSTMENT SCREW MACNET DRAW I -MOTOR SIAS ADJUSIMENT

-CENTER SWITCH ROD DOWN WIlT SWI MOUNTING PLATE

-MAGNET WIRE CONDUIT PULL-ROD SPI -MACNET UP LIMIT SWITCH

,ADJUSTMENT SCREW

-POTENTIOMETER COVER MOTOR COiVER-

-PUSH ROD

-ELOCK PULL-ROD HOU!

-BARREL

-MACNET DRAW TUBE PULL RODC LOCK NUT-Standard rod drive mechanism Figure 7.6, Regulating, Safety and Shim Rod Drive.

K-State Reactor 7-13 Original (12104)

Safety Analysis Report

CHAPTER 7 MOTOR CONTROL P1321 SWITCHES . 0lic

)LO DOWN NOTELIUIT SWITCHES SHON WITH IO0 IN FUL LM POSITION.

Rod drive. metor control. and indicator lamp circuit Figure 7.7, Control Rod Drive Circuit.

b. Transient Rod Drive The transient rod-drive (Figure 7.8) is an electrically controlled, pneumatically operated, mechanically limited system. The transient rod and' aluminum extension rod are mechanically connected to a pneumatically driven piston inside a -worm gear and ball-screwv assembly. The system is housed on a steel support structure mounted above the reactor tank. A three-way solenoid valve mounted below the support controls air to the piston. The throw of the piston, and hence the amount ofreactivity inserted into the core during pulsing operations, is regulated by adjusting the worrn gear and ball-screw' assembly. Te-adjustnent is made from the central console by actuating a reversible motor drive, which is coupled to a worm gear and a 10-turn potentiometer for position indication. The operation ofthe position indicator is identical to that of the shim, safety and regulating control rods.. The drive circuitry is identical to the shim, safcty and regulating rods (Figure 7.7), except for that the motor is not continuous energized.

Relays in the drive unit allow it act similarly to the electrically locked motors. Since air is used to support the rod, there is no compensation for rod weight. The remaining differences involve the pneumatic relay controls.

K-State Reactor 7-14 Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS The solenoid valve is actuated by means of the console mounted transient rod fire and air release (scram) switches. When the rod fire switch is depressed, the solenoid valve opens, admitting air to the cylinder, coupling the piston and rod to the shock absorber. '!

Depressing the air release switch dc-energizes the solenoid valve, which removes air from the cylinder and vents air to the atmosphere. In the event of a reactor scram, the solenoid will be de-cnergized via the scram circuitry, which will allow the transient rod J to drop into the core after the air is removed. Micro-switches are used to indicate the J extreme positions, up or down, of the shock absorber. In steady state mode, an interlock prevents actuation of the rod fire switch if the drive is not in its fully down position.

In the pulse mode, a variable timer (usually six seconds) de-energizes the solenoid valve after the pulse is initiated. The shock absorber will remain in its preset position until the mode selector switch is taken to steady state. In the steady state mode of operation, the adjustable (normally six second) timer is disengaged and the cylinder remains pressurized. If the air supply for the pulse rod drive should drop below approximately 45 psig, an amber low air pressure.warning light will be actuated on the control console.

Loss of air pressure will cause the rod to fall into the core.

C. Interlocks Several interlocks are built into the control system of the reactor to prevent improper operation. These interlocks arc hard-wired into the control rod drive circuitry. They are N stated below:

1. No control rod withdrawal (shim, regulating and safety rods only) is possible unless the count rate neutron channel is indicating > 2 cps. This interlock prevents the possibility of a startup without a functional power level startup channel.

The low count rate interlock may be bypassed during fuel loading operations when core inventory is not high enough to multiply the source above 2 cps.

2. Air may not be applied to the pulse rod if the pulse rod shock absorber is above its full down position and the reactor is in the steady state mode. This interlock prevents the inadvertent pulsing of a reactor in the steady state mode.

3. There is no simultaneous withdrawal of two or more control rods when the reactor is in the steady state mode. This interlock prevents violation of the maximum reactivity insertion rate of the reactor.

4. The pulse rod is the only control rod that can be withdrawn when the reactor is in the PULSE mode (this does not prevent the scramming of any control rod). This.

interlock minimizes the possibility of pulsing a supcrcritical reactor. This interlock is provided by the source interlock, which is engaged when the log channel is placed in pulse mode.

K-State Reactor 7-15 Original (12104)

Safety Analysis Report

%1*I* .

.1 CHAPTER 7 SHOCK ABSORBER VENT EXTERNALLY -

THREADED CYLI NDER

UPPORT BEARII -WORM WORM t

-HOUS ING UP PISTON AIR SUPPLY/ N, VALVE HOSE

-LIMIT SWITCH-CYL DOWN

-LIHIT SWITCH:ROD DOWN I

I I

rnum BNeOH LIHIT/ l CON1R ECTION ROL ROD TO

-L .

Transient rod ddve mechardsm

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Figure 7.8, Transient Rod Drive.

Additionally, there is an interlock that prevents reactor pulses from being fired if the reactor power is above I OkW (normally set at I kW). lher.is also a key switch for bypassing the source interlock during fuel loading operations to check for criticality.

K-State Reactor 7-16 Original (12/04)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS 7.4 Reactor Protection System The reactor protection system (RPS) will initiate a reactor scram if any of several measured parameters are outside their limited safety system settings (LSSS). The reactor scram effectively shuts down the reactor by de-energizing portions of the shim, regulating, transient, and safety rod drives, causing the control rods to drop into the reactor core by gravity. The shim, safety and regulating rod drives utilize electromagnets to hold up their respective control rods. The pulse rod drive utilizes an electric solenoid to apply compressed air to the pulse rod. High reactor power, high fuel temperature, loss of detector high voltage, or short reactor period will automatically cause the control rods to be dropped into the reactor core. The reactor operator may manually scram the reactor as well by means of a scram bar on the console. If beam ports are in use in their open configuration, an additional reactor scram may be added to reduce radiation levels if personnel attempt to enter the beam area.

During steady state operation, the high reactor power scram, the high voltage power supply failure scrams for all neutronic channels, and the manual scram are required for operation.

Although the period scram is normally inr operation, it can be bypassed provided that the reactor operator calculate reactor period for each rod movement and that the calculated reactor period is greater than one second. The fuel temperature scram is active, but it is not necessary since the fuel temperature sctpoint (usually set for 4000 C) will normally not be reached during steady state mode.

In pulse mode, the mode selector switch is set to the HI PULSE position, interrupting detector signal to the linear channel WThen the pulse interlock is activated (to initiate the source interlock) to prevent withdrawal of the shim, safety and regulating rods, the detector signal to the logarithmic wide range detector is interrupted. Consequcntly, the period scram and the linear high power scrams are disabled. High voltage. scramns for the linear and log channels are also disconnected as large pulses produce detector currents that may temporarily overload the power supplies. The percent power channel increases its range to the pulse range (the percent power channel high voltage power supply scram remains active). The uncompensated ion chamber for the percent power channel is positioned further away from the reactor core, allowing for measurements of both steady state and pulsing power levels without excessive current from the high voltage supply.

7.5 Engineered Safety Features Actuation Systems There are no engineered safety features actuation systems. Control rod insertion is provided by gravity and core cooling is provided by natural convection in water or air. Therefore, ESF systems are not required in this design.

7.6 Control Console and Display Instruments. .

The control console and display instruments are shown in Figures 7.2-7.5, 7.9, 7.10 and 7.12.

Their layout (Figure 7.11) provides a single operator with all relevant reactor information. All push buttons required for general operation are located on the control console, within easy reach by a seated operator. All instruments on both the console and instrumentation rack arc visible K-State Reactor 7-17 Original (12/04)

Safety Analysis Report I

/

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CHAPTER 7 from this seated position. In addition to the various analog displays, a computer display on top of console can also be used to show relevant reactor status information on a single screen.

There are several additional pieces of equipment in the control room.

_ _ o interrupt power to electrical devices in the control room and reactor bay (see Figure 7.13a). A halon fire extinguisher is located next to the breakers for use in fighting electrical fires. Current core and -kcility configuration is shown in a displa cabinet (Figure 7.13.b).

A local radiation arma monitor (including indicator and alarrn) is thedoorinte controlroom to the reactor bay.

7;7 Radiation Monitoring Systems Radiation monitoring systems are employed throughout the reactor facility: G-M detectors at the reactor pool surface and cleanup loop, 7 remote area monitor channels (3 general area or process monitors, 4 channels for beam ports -11 beam port channels is currently instrumcnted, with the remainder scheduled for instrumentation near term), a 5 Rhl' evacuation alarm, several air activity monitors, and numerous portable radiation monitors, including those for contamination monitoring. Additionally, an independent monitor with visual and audible alarms is located above the door to the reactor bay.

Figure 7.9, Control Room Overall View.

K-State Reactor 7-18 Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS Figure 7.10, Control Console Operator's View.

B I.'

Figure 7.12, Instrumentation Rack Operator's View.

3 K-State Reactor 7-19 Original (12104)

Safety Analysts Report

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. CHAPTER7 Figure 7.13a, Control Room Behind Operator. Figure 7.13b, Corc Map.

The radiation monitor in the instrumentation rack provides an indication of radiation levels directly above the primary pool water surface. A meter on the right-hand section of the rack indicates a radiation dose rate of 100 mRl.h full scale, utilizing a GM tube detector measures coolant activity in the cleanup loop water box. The water box monitor has aii audible alarm, with a reset button.

Tfe remote area monitors utilize G-M detectors located throughout the reactor bay (typical unit illustrated in Figure 7.14). Permanent locations are: at the'top of the reactor tank, above the bulk shield tank, and near the ion exchanger in the prirary coolant system, and directly over the primary water tank. Each beam port has signal and power lines to support installing a beam port monitor.. The detectors feature an analog readout in both the control room and locally with visual indicators for normal, alert, and alarm conditions. The control room alarm has an audible signal as well.

I I

K-State Reactor 7-20 Original (12104)

Safety Analysis Report

INSTRUMENT AND CONTROL SYSTEMS Figure 7.14, Typical Area Monitor Instillation.

A 5 Rh l monitor on the 22-foot level serves an evacuation alarm. The alarm signals a 100+ dB audible alarm in the reactor bay.

There are numerous ion chamber and G-M portable survey instruments through the reactor bay and control room for garma and beta surveys. These instruments are calibrated semi-annually.

A rem-ball and a Bonner Sphere set are located in the reactor bay for neutron measurements. The rem-ball is calibrated annually. Low-level G-M counters in the reactor bay used for contamination monitoring. These are calibrated annually. Neutron and gamma sensitive pocket ion chambers are available in the control room for tours and personnel monitoring, and are calibrated semi-annually. Film and ring badges are issued to regular staff, which are read monthly and quarterly respectively.

An air monitoring system (Figure 7.15) samples air over the reactor pool. These monitors should sense any changes in radioactive discharge from the reactor pool to the environment. Channels are provided for monitoring particulate, noble gas and iodine activity.

An independent air monitoring system, a continuous air monitor (Figure 7.16), is stationed on the 12-foot level. This monitor should sense any changes in airborne contamination in the reactor bay. The continuous air monitor has active background discrimination. One channel monitors radioactive contamination on a filter exposed to airflow from the reactor bay. A second channel monitors background radiation levels for a background subtraction Figure 7.15: Figure 7.16: Continuous Air Monitoring System Air Monitor K-State Reactor 7-21 Original (12104)

Safety Analysis Report

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CHAPTER 7 7;8 Bibliography WMarklJ.PulsinglReactorMechanicalMaintenance and OperationManualjor 250-kW TPJGA JKSU, General Atomics, GA-3399; I August.1962.

NMP-)OOD LinearPower Channel: OperationandMaintenanceManual General Atomics,,E 117-1017;Rev. 1; 1991.

NLJY-1000 Log Power Channel: OperationandMaintenanceManual, General Atomics, El 17-1019; 1992.

NPP-O000PercentPower Channel: OperationandMaintenaniceAManual, General Atomics,,

El 17-1010, Rev. 2; 1992.

TRIGA MarkliReactorinstrumentationManualfor USGS. GeneralAtomics, GA-9039; 1 I November 1968. (For console acquired from USGS)

TRJGA Mark;iPulsingReactorinstrmentationMaintenance ManualforKSU, General Atomics, GA-3 114, no date.

... .I K-State Reactor 7-22 Original (12/04)

Safety Analysis Report

Ii

8. ELECTRICAL POWER SYSTEMS Primnary electrical power is provided through the Kansas State University power grid, supplied by an on-campus plant and commercial generators. Main power lines traverse underground tunnels, thus inhibiting tampering. Loss of electrical power automatically places the reactor in a subcritical, secured configuration. Loss of electrical power will de-energize the control rod drives, causing the rods to fall by gravity into the core, and therefore does not represent a potential hazard to the reactor. Since the core is cooled by natural convection, no emergency power is required for reactor cooling systems. Backup battery systems are provided for emergency lighting, the security system, and the 22-foot level evacuation alarm.

8.1 Normal Electrical Power Systems The design basis for the normal electrical power systems is to provide sufficient current for normal operations. The original instrumentation and control systems were mostly vacuum-tube designs; hence there is more than adequate supply for the newer solid-state devices. Safe-shutdown is an automatic consequence of loss of electrical power and hence there are no major electrical backup systems. The reactor has no exclusive electrical supply and distribution, but derives from the building transformers. Supplied power is standard 60 Hz AC, available in 110 V single phase, 220 V single phase, and 480 V three phase configurations.

a=Wdd i aj L I J

_ -1

~I1 I[I I r- I l.c. . ._.

IC II

- -,

t-Figure 8.1, Electrical Distribution Layout.

Utility voltage (4160 kV) is delivered via underground cables to three oil-filled switches inside a locked, fenced area immediately outside the reactor bay, but within the facility boundary (Fig K-State Reactor 8-1 Original (9102)

Safety Analysis Report

I 3/4 e. I I CHAPTER 8 8.1). These switches connect power to step-down transformers for the original Ward hall (including control room power), reactor bay service loads, experiment services, the 1972 addition to Ward Hall, and Cardwell Hall (math and physics building, north of Ward Hall).

One transformer supplies building power and the control room breaker box. A second primary transformer supplies the breakers in the reactor bay for the cooling system and recirculation ventilator. A third dedicated service located in the reactor bay can be used to supply reactor experiments requiring large amounts of electrical power, which is normally disconnected at the transformer.

The control room breaker box supplies all electrical outlets throughout the facility, lightinga line conditioner, and the reactor bay crane. Fiom this source, a line conditioner in the reactor bay provides isolated, regulated power for reactor instruments and control systems, such that they are not affected by minor electrical fluctuations.. An interruption of electrical powver will cause the line conditioner to de-energize and it must be manually reset to resume operations. All standard reactor experiments utilizing the outlets in the reactor bay can be de-energized from the control room breakers.

Although the cooling system control breakers are supplied from a separate line, a relay network provides coIntrol from a switch located on the control console. Another breaker'supplies a recirculation ventilator inside the reactor bay, awhich cannot be dc-energized from the control room. A large electrical distribution center is located near the c6oing system breakers in the reactor bay that can be used to supply reactor experiments, although this system is normally disconnected at the transformer.

The bulk of electrical -wiring Is in shielded conduits as per commercial electrical codes.

Instrument wires from the reactor instrumentation and control system run through a sub floor conduit in the control room into a wire tray leading to the upper level of the reactor bay. A secondary tray runs the perimeter of the 12-foot level of the reactor. Most instruments use self-shielded cables. The signal cable for the log power channel, being sensitive to electrical noise,

runs inside a flexible metal conduit in the wire tray from the upper level of the reactor to the control console. Grounding straps are used to ensure common ground between the control room and instrument locations.

All electrical devices were installed according to the electrical codes in existence at the time of their introduction. During the control room modifications in 1993, the control room wiring was inspected and several systems were re-wired to meet current electrical codes.

8.2 Emergency Electrical Power Systems The design basis for emergency electrical power systems is to provide lighting and surveillance for emergency conditions and to maintain physical security. Consequently, battery backup power is used for emergency lighting, the University fire alarm system, the evacuation alarm, and the security system. All backup systems have regular maintenance schedules and are periodically tested for opicration.

Internal batteries supply emergency lighting at the upper level of the reactor and at all exits to the reactor bay. Exit signs are similarly illuminated. The fire alarm system has battery backup for K-State Reactor 8-2 Original (9102)

Safely Analysis Report

U!

U!;

ELECTRICAL POWER SYSTEMS UP sensors and pull-stations. These systems are maintained and periodically tested by KSU Fire UP Safety personneL -P Backup batteries supply the security systenm, but description of the system is limited to our Physical Security Plan as per IOCFR2.790.d. Maintenance and testing of the batteries is JI performed annually. UP 8.3 Bibliography Ward Hall Blueprints. Available from the Department of Facilities, Kansas State University.

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J1 Uj U

U JU j

K-State Reactor 8-3 Original (9102)

Safety Analysis Report

SAFETY ANALYSIS REPORT This page intentionally blank P. .

t K-State Reactor

9. AUXILIARY SYSTEMS ,

The systems covered in this chapter are not directly required for reactor operation, but are used in support of the reactor for normal and emergency operations.

9.1 Heating, Ventilation, and Air Conditioning Systems Heating and air conditioning of the reactor bay and control room is provided by either steam or chilled water from the University Physical Plant. Because of this design, heating and cooling cannot be simultaneously provided. Consequently, the system gets changed over from steam to chilled water and vice versa twice a year.

The reactor bay was specifically designed for handling radioactive materials. The reactor bay was originally equipped with four unit ventilators on the 12 foot level which drew intake from outside air and discharged into the reactor bay. These units have been disabled, removed and blocked. In-lcakage and a single exhaust fan at the top and center of the confinement dome provides outside air to the reactor bay. Prior to disabling the unit ventilators, a large unit ventilator was installed to support an experiment in the reactor bay with adjustable ducting between the 12-foot level and the top of the reactor bay. The unit has a coil connected to the HVAC lines and is now used for heating and cooling in the reactor bay. A thermostat was installed to permit temperature adjustments. The electrical disconnect for the unit is located near the cooling system breakers, behind the heat exchanger. The condensate line drains into the reactor sump. Routine maintenance and service of these systems is the responsibility of the KSU Department of Facilitiei. All ventilator filters in the reactor bay are surveyed for radionuclide deposition upon removal.

The control room features a single thcrmostat that can vary the temperature of the air coming from the main intake unit located in room I08, adjacent to the control room. The main intake unit supplies the bulk temperature change and secondary coils inside the control room duct allow zone control. The control room also features an exhaust fan, which ducts out of the roof immediately above the control room. A switch on the control room wall manually controls the exhaust fan. A single window to the outside may be opened from the control room; the window is covered with a metal barrier for security.

9.2 Handling and Storage of Reactor Fuel The majority of the reactor fuel is stored in the reactor tank, either in the reactor core or racks surrounding the reactor tank. Racks may be added to the bulk shield tank for fuel storage during fuel transfer operations. The racks are fabricated of aluminum and allow only for single row spacing of up to six elements, whose spacing in the rack is sufficiently far apart to prevent accidental criticalities.

Fuel may also be stored in 10 fuel storage pits which extend 3 m below the floor of the 0 foot level of the reactor bay. The fuel storage pits are constructed of 25A cm diameter Schedule 40 steel pipe, welded closed at the base with lockable covers. Due to water seepage into one of the K-State Reactor 9-1 Original (12104)

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I CHAPTER 9 pits, the pits are inspected annually. These pits are currently used to store elements that were damaged in handling or shipping. There are also two 25.4 cm beamport plug holders built into the walls of the reactor bay, one of which is used to store new fuel elements.

Fuel is individually handled by means-of a standard TRIGA element tool that grapples the end fitting on the element. The tool is 7.6 m long and requires two persons when moving fuel. One person is required to be a senior reactor operator, the other trained in fuil handling operations.

Additionally, a reactor operator is also required to be at the reactor console to monitor conditions in the reactor and reactor bay.

To facilitate transfers of activated fuel elements, a single fuel element cask is used. The reactor bay crane is used to move the cask. The fuel element is drawn in through the base of the cask and a shutter is slid under the element. The cask is steel over lead. The cask is 35.6 cm in diameter and 1 in in height, has a 5.1 cm internal cavity and weighs 1134 kg (2500 lbs).

For fuel inspection; there is a perisc6pe mounted in the reactor tank for visual inspection of cladding. The periscope is counterweighted, and can be lowered all the way to the reactor core level. It offers a 24° view in water, with a 48° apparent field of view. For fuel element gauging, a tool is permanently mounted to the side of the tank. A GO-NO-GO tube is used to check for swelling or bowing within 1.57 mm of manmfactured tolerances. The technical specification limitation is 1.59 mm. A dial indicator is used to measure element lengths to the nearest 0.03 mm against reference standards. Fuel elements elongation is limited by technical specification to 2.54 mnUL WVith a very small bum-up of fissile material per year, the KSU reactor has infrequent shipmenits of fule elements off-site. Commercial NRC-certified transport casks are used for off-site transport of used fuel, principally the Battelle BMI-l cask (to be replaced by the GE-2000 or other suitable cask). New fuel elements are delivered by General Atomics in their own licensed container.

9.3 Fire Protection Systems and Programs Fire protection systems arc maintained and serviced by the Campus Fire Safety, Department of Environmental Health and Safety, Division of Public Safcty. The building fire alarm system was installed in 1995 and is part of a campus-wide network. Fire alarm signals are sent to the campus police station via a line-monitored system, where the locations of the alarm as well as building maps are displayed on a computer terminal. The reactor has two pull-stations one in the control room and one in the reactor bay. There are two smoke detectors located about at the 20-foot level inside the reactor bay on the north and south walls. The reactor also has an additional system, consisting of seven smoke detectors in the reactor bay and one in the control room. This system is associated with thc reactorPhysical Security system and vill also notifythe campus police . .

There are eght fire extinguishers readily available to reactor personnel. A halon extinguisher is located in the control room for electrical fires. For general purpose fires, catbon dioxide fire extinguishers are located in the hallway outside'the control room, *on the 22 foot level of the reactor bay, oi.the 12 foot level, and on the 0 foot level. For equipment fares, dry chemical extinguishers are located on the 12 foot level, on the 0 foot stairs, and next to the cooling system K-State Reactor 9-2 Original (12/04)

Safety Analysis Report

AUXILIARY SYSTEMS pumps. Reactor personnel visually inspect all extinguishers monthly. Campus Fire Safety '1 performs pressure testing and general maintenance on an annual basis.

9.4 Communication Systems The reactor facility has an intercom system controlled by a 3M D-120 commercial unit, installed U in 1996. The control unit is mounted immcdiatcly above the control console, with a foot switch J underneatb the console for hands-free communication. Speaker locations include: Reactor Manager's Office, Reactor Operators' Office, control room door, lower level door to reactor bay, Neutron Activation Analysis laboratory, 22 foot level above tank, 22 foot level near pool surface, ;

and one next to each of the four beam ports.

Telephones at the facility share a common line. They are located in the Reactor Manager's office, the Reactor Operators' office, the control room, at the 22 foot level of the reactor bay, and at the 0 foot level. The NAA lab has a telephone with a separate line. The offices and control room also have high-speed Ethernet connections that can link with the reactor computer or outside machines.

9.5 Possession & Use of Byproduct, Source, & Special Nuclear Material Reportable quantities of radioactive materials are possessed under the University's State Radioactive Materials license, the Reactor Facility License, and a separate NRC special nuclear materials license. The reactor fuel is the property of the Department of Energy. Several radioactive sources are owned by Kansas State University. Radioactive materials, including special nuclear material (SNM) are inspected for contamination and inventoried on a quarterly basis. Several areas are designated for storage of these materials.

Byproduct material produced in the reactor for research purposes is transferred to the State License and recorded in a radioactive transfer log. The State license is maintained by the KSU Department of Environmental Health and Safety, Division of Public Safety and administered by the University Radiation Safety Committee. Only individuals listed under the license are permitted to receive materials. Normally, a member of reactor staff is also approved by the Committee to receive byproduct and special nuclear material under the state license. Possession limits are set by the State, and the University Radiation Safety Committee determines use limits.

Transfers off-campus to other licensees must first go through the Department of Environmnental Health and Safety, Division of Public Safety. The facility has several sources for reactor startup, research, and instrumentation calibration purposes that are possessed under this license. Low-level wastes generated under the State of Kansas license are transferred to the Department of Public Safety for disposal under the State license. Disposal of low-level wastes generated under the reactor license is coordinated with the Department of Environmental Health and Safety, Division of Public Safcty.

SNM inventory is reported to the Nuclear Assurance Corporation under Reporting Identification Symbol (RIS) ZKL. The reactor fuel comprises the bulk of SNM at the facility. This fuel is owned by the Department of Energy and possessed under the Reactor Facility license R-88. Also K-State Reactor 9-3 Original (12104)

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CHAPTER 9 under license R-88 are fission chambers containing SNM, owned by KSU. Possession limits are set by the license R-88 at 3.98 kg 235U in enrichments less than 20% and 0.02 kg 23'U in enrichments up to 100%. Small quantities of SNM for experiments arc owned by KSU under the State of Kansas SNM license 38-COI 1-01.

Storage locations for radioactive materials include the reactor bay, source cave, safe, and designated laboratories. Fuel storage locations are listed in Section 9.2. A shielded source cave, located in the northeast comer of the reactor bay, is uscd for storing large sources and low level wastes. A shielded, locked safe in the reactor bay can be used to store small sources..

Laboratories storing radioactive materials include the Panoramic Irradiation room, the Beta Shielding Laboratory, the Radiation Detection Laboratory, and the Neutron Activation Analysis Laboratory.

9.6 Cover Gas Control in Closed Primary Coolant Systems The*KSU reactor has an open primary coolant system and hence has -no cover gas control.

Nitrogen 16 is controlled as described in Chapters 5 and II by forcing convection cooling flow form the reactor core into a helical pattern (to enhance time delays for more decay). Using helium in the pneumatic system (instead of air) controls the possible inventory of radioactive argon.

9.7 Other Auxiliary Systems 9.7.1 Reactor Sump System All floor drains and the recirculating ventilator condensate line feed into the reactor surmp. The reactor sump is a square cavity covered by steel plates has a capacity of 3.8 kLd It is also connected to the sub-floor tracks used for moving the thermal colunm door, giving it an overall capacity of 4.5 kL. A sump discharge system was installed in 1997 that permits the sump to be recycled through filters and discharged through a separate filter to ensure that insoluble radioactive materials are not discharged to sanitary sewers. Before discharge, liquid samples are drawn for analysis of specific activity. Historically, the only isotope normally discharged was tritium from primary water, in quantities well below IOCFR20 limits.

9.7.2 Reactor Bay Polar Crane A polar crane in the reactor bay is used to manipulate loads of up to 3630 kg (8000 lb). Various lifting straps and attachments are available for handling varied loads. A breaker in the reactor control room supplies power to the crane. A lockable breaker on the west irall of the reactor bay permits securing power to the crane. A discconect on the crane permits personnel servicing the crane a local and positive control over power to the crane. A basket is attached to the outside edge of the crane for changing light bulbs in the ceiling of the reactor bay. Due to safety concerns, lighting was installed around the periphery of the riactor bay, reducing the need for the overhead lights and the basket.

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AUXILIARY SYSTEMS 9.7.3 Beam Facilities A complete description of the beam tube facilities is provided in Chapter 10. Auxiliary systems that support the beam tubes are described below.

a. Thermal Column Door The thermal column door rolls on railroad tracks, which are deeply recessed, into the floor. It is opened and closed by means of stcel cables and a winch. A key lock secures the winch. The stecl cables were replaced in 1997. Indicator tights on the control console indicating the position of the door were also added when the cables were replaced.

b. Beamport Plug Handling A special cask is available for use in removal and storage of beamport plugs It allows the plugs to be drawn directly from the port into the cask. A lead shutter can be closed over the open end, further reducing exposures.

c. Beam Facility Vents and Drains All air-filled spaces in beam facilities are connected by pipes to a manifold mounted on the outside of the reactor. Additional pipes near the beamport doors have valves that can be opened to allow insertion of a tool to open the lead shield doors or to vent the beamport. Bcamport doors are padlocked shut. Over the last several years, the void space near the thermalizing column has been collecting water. The source of the water is likely from the bulk shield tank, since the maximum observed pressures correspond to the level of the bulk shield tank and not the primary. With the bulk shield tank water level about 6 in. below the top of the tank, the water leakage is essentially secured. Regardless of the source, there was concern that this water might pass through the manifold and fill the other beam lines, or obscure evidence of leakage from other experimental facilities.

Hence in 1996, the thermalizing column beam port line was connected to a separate manifold to permit collection of fluids.

d. Pneumatic Transfer System (Rabbit)

The pneumatic transfer system or rabbit is used to rapidly transport samples between an in-core location and the Neutron Activation Analysis Laboratory. From commercial cylinders, compressed helium fills small tanks at either end of the system. Pressure is limited by release valves at 275 kPa (40 psi). The system is operated from the instrumentation rack (Fig 7.3) in the control room, where the reactor operator sets the direction of motion by. positioning vent valves and applies the helium by another valve.

Indicator lights show position of valves. A light indicates low helium pressure.

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CHAPTER 9 9.7.4 Associated Laboratories The C.C. Tate Memorial Neutron Activation Analysis Laboratory is located in the basement of Ward Hall. It features several germanium detectors and associated electronics for ganma my spectroscopy. There is also an alpha counter with beta and gamma discrimination. The Radiation Detection Laboratory is a student lab, but contains many additional pieces of detection equipment. Another room contains a panoramic irradiator for performing gamma-ay instrument calibrations. NBS-traceable alpha, beta, and neutron sources are also available.

9.8 Bibliography None

1.

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10. Experimental Facilities and Utilization 10.1 Summary Description 10.1.1 Experimental Programs The K-State reactor provides educational and training services to support the Department of Mechanical and Nuclear Engineering Bachelor of Mechanical Engineering with a Nuclear Option as well as Masters and Doctoral programs in Nuclear Engineering. Other K-State departments with nuclear, radiation or instrumental analysis components in other curricula are also supported.

The K-State reactor produces radioisotopes for research, including both tracer/gauging and radioanalysis applications. The K-State reactor performs extracted beam research.

10.1.2 Experimental Facilities Sectional views of the reactor are shown in Chapter 1. For orientation of experimental facilities with respect to the reactor tank and concrete shield, the reader is referred to Figs. 1.1 and 1.2 of Chapter 1. Components of experimental facilities located in the reactor pool are illustrated in Figure 10.1. Principal experimental features of the KSU TRIGA Reactor Facility include:

Central thimble

Rotary specimen rack

Thermalizing column (with bulk shielding tank)

Thermal column (with removable door)

Beam ports Radial (2)

Piercing (fast neutron) (1)

Tangential (thermal neutron) (1) 10.1.3 Experiment Monitoring and Control Monitoring requirements for individual experiments are identified in the applicable experiment procedure. Experiment monitoring which is a routine part of facility equipment includes a system to detect leakage from reactor pool-wall into the beam tubes or thermal/thermalizing column, and to minimize leakage into the reactor bay, capabilities for installing an area radiation monitor channel in the vicinity of an open beam port, and the capability for installing an external scram in the reactor protection system.

a. Leak Detection System The interface of experimental facilities (beam ports, thermal column and thermalizing column) and the reactor pool liner contains an open plenum with piping connected to a leak off volume.. The leak detection piping is connected to a single volume, except that a separate leak off volume and pressure gauge has been installed for the thermalizing column. If the pool leaks into the experiment facilities, the water will overflow the K-State Reactor 10-1 Original (12104)

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CHAPTER 10 plenum and fill the leak off volume. Pressure monitors in the leak off volume indicate When the volume is partially or fully filled. The lIdak off volumes and the pressure monitors arc located on the north wall of the biological shielding near the northwest (radial) beam port.

Watertight seals are installed at the interfacebetween the beam ports and the reactor bay.

With the original seal attached to the inside of the beam port door, the seals are designed to be used when beam ports arc not in use. If the pool wall fails, this seal will prevent complete loss of water from the pool.

S.

Figure 10.1: ExperimentFacility Components in the ReactorPool

. b. Area Radiation Monitor

Electrcal lines are installed to support area radiation monitors at eachbeamporL K-State Reactor 10-2 Original (12104)

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EXPERIMENTAL FACILITIES AND UTILIZATION

c. External Scram One scram relay has been designated as the "external scram channel." The external scram channel is not normally in use. In the event that potential consequences associated with an experiment requires protection for the facility, personnel or the experiment, sensors may be connected to cause a reactor scram based on significant parameters. I 10.1.4 Experiment Review and Approvals Experiments are reviewed and approved by the Reactor Safeguards Committee prior to performance as a reactor experiment procedure. The Reactor Supervisor or Nuclear Reactor Facility Manager may schedule an approved experiment for performance.

The Reactor Safeguards Committee approves reactor operations prior to performance via Reactor Operating Procedures. Operations supporting education, training and requalification without insertion of material in the experimental facilities is considered an operation.

10.2 Experimental Facilities The K-State reactor is a flexible, multi-use facility with irradiation facilities inside the core boundary, in the reflector, outside the reflector, and outside the biological shielding. One of the in-core facilities is a pneumatic sample delivery system capable of providing samples directly to the neutron activation analysis laboratory.

10.2.1 In Core Facilities Irradiation facilities within the core boundaries include available upper grid plate fuel element penetrations, a series of smaller penetrations in the upper grid plate, and the central thimble.

a. Available Fuel Element Spaces -

Experiments may be inserted in spaces designed for fuel elements. Core 18-11 has one empty space, two spaces occupied by graphite "dummy" rods, and one space occupied by a dry tube. Typical dry tube configuration uses a modified "S" bend to minimize streaming radiation at the pool surface. One dry tube is lined with cadmium to support v spectrally tailored neutron irradiations.

b. Small Upper Grid Plate Penetrations Tle upper grid plate was fabricated with a series of small (slightly larger than /4")holes to permit flux-mapping experiments. These penetrations may also be used to irradiate targets with suitable ge6metry.

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CHAPTER 10

c. Ccntral Thimblc The reactor is equipped with a central thimble for access to the point of maximum flux in the core. Procedures limit the maximuni power permitted with fueled experiments in the central thimble. A removable screen at the top end of the thimble allows gas relief and prevents objects from falling through the reactor tank covers.

The central thimble is an aluminum tube that fits through the centerholes of the top and bottom grid plates terminating with a plug at a point approximately 7.5 in. (19 cma) below the lower grid plate. The tube is anodized to retard corrosion and wear. Although the shield water may bc removed to allow extraction of a vertical thermal-nbutron and gamma-ray beam (not done at the KSU facility), four 0.25-in (63-mm) holes are located

. in the tube af the top of the core to prevent expulsion ofwater from the section of the tube

. within the reactor core. Dimensions of the tube are 1.5-in. OD (3.81 cm), with 0.083-in (2.0 nmm) wall thickness. The thimble is approximately 20 ft. (6.1 m) in length, made in two sections, with a watertight tube fitting. .

.10.2.2 -In Tank, Ex Core Facilities Experiment procedures authorize irradiations above and below the core (note that the fuel contain san internal reflector at the top and bottom of each element) and adjacent to the radial reflector.

iradiation is also authorized in the bulk shield tank, and at the outer face ofthe thermal column.

10.2.3 In Reflector Facilities . . .

The principal, traditional in-reflector facilities include a "thermal column" with dry irradiation space, and a thernalizing column provides irradiation space in the bulk Whield tank. A third facility is nested inside the radial reflector, the rotary specimen rack.

a.-. Thermal Column The thernal column is a large, boral-linid, graphite-filled aluminum container, with outside dimensions 4 ft (1.2 m) square in cross section and approximately 5.35 f (1.6 m) in length. The thermal-column liner is a seal welded container fabricated from 0.5-in.

(12.7-mm) aluminum plate. The outer portion is embedded in the concrete shield and the innerportion is welded to the aluminum reactor tank. The cxterior.surfaces-are coated with plastic for corrosion protection. The portion of the thermal column welded to the aluminum tank extends to the graphite reflector and matches the contour of the reflector over 1 00-degree angle. In a vertical plane, the column extends approximately 13 in. (33 cm) above and below the reflector, with the centerlines of the column and reflector coincident.

The aluminum container is open toward the. reactor room. Blocks of nuclear-grade graphite occupy the entire void except for a 2-in (0.79 cm) thick lead curtain located within the column. The individual blocks are approximately 4-in (10.2-cm) square in cross section, the longest being 50 in. (1.27 m) in length.

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EXPERIMENTAL FACILITIES AND UTILIZATION Twelve graphite blocks serve as removable foil stringers. These blocks are machined %J 1116 in. (1.6 mm) undersize for easy removal and insertion. The central block is aligned %

with a stringer-access plug in the thermal column door. To gain access to other than the central stringer, it is necessary to roll out the thermal-column door on its tracks.

Surrounding the graphite on the inside of the aluminum casing on all sides are 118-in. (3.2 mm) sheets of boral. This is done to reduce capture-gamma-ray production in the surrounding concrete shield.

b. Thermalizing Column The bulk-shielding experimental tank is 12 ft (3.7 m) deep, 8 ft (2A6 m) wide, and 9 ft (2.8 mn) long. The tank is waterproofed with an epoxy coating and is filled with water for shielding. The thermalizing column is similar to the thermal column, but smaller. Its outer section extends from the bulk-shielding experimental tank through the concrete shielding and to the aluminum reactor tank. The inner section of the column is welded to the tank, and extends to the reflector assembly and matches its contour. The division of the column into two sections allows for thermal expansion during reactor operation. The column is 2 ft (61 cm) square in cross section by approximately 433 ft (1.32 m) in length. It is fabricated from 0.5-in. (12.7-mm) thick scal-welded aluminum. The horizontal centerline is coincident with the centerline of the active fuel lattice. The exterior surfaces in contact with concrete are coated with plastic for corrosion protection.

An aluminum cover plate 0.625 in. (15.8 mmn) thick, held in place by twenty 518-in.

anodized aluminum bolts, keeps water out of the thermalizing column. A 0.25-in. (6.44-mm) neoprene gasket provides the scal. In the region of the concrete shield, the aluminum container is lined with 0.125-in. (32-mm) boral sheets that extend 3 ft. 2.125 in. (96.8 mm) inward from the bulk-shielding tank.

At the inner end, nearest the reactor core, the column.is filled with graphite blocks to an axial thickness of 8 in. (20.3 cm). All blocks are nuclear-grade graphite, 4-in (10.2-cm) square in cross section. This wall of graphite is backed by a 2-in. (50.8-mm) slab of lead.

In the outer portion of the container, where the aluminum is lined with boral, I in. (2.54 cm) thick polyethylene sheets line the boral. The graphite, in the form of 4-in (10.2-cm) square blocks, is stacked 24 in. (61 cm) thick from the outer edge of the column.

c. Rotary Specimen Rack A rotary, 40-position rotary specimen rack (RSR) is locatcd in a wcli in the top of the graphite radial reflector. The RSR allows large-scale production of radioisotopes and for activation and irradiation of multiple material samples with neutron and gamma ray flux densities of comparable intensity. Specimen positions are 1.25 in. (3.18 cm) diameter by 10.8 in. (27.4) cm depth. Samples are manually loaded from the top of the reactor through a water-tight tube into the RSR. The rack may be rotated (repositioned) manually from the top of the reactor. Figure 10.2 illustrates the design features of the RSR, and Figs. 103 and 10A are respectively photographs of the RSR during construction and the rotation mechanism and housing at the 22-frilevel of the reactor.

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CHAPTER 10 -

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I 0 K..4 Flgurc 10.2: Rotary SPcclzcn Rack FIgure 1OA: RSR Figure 103- RSR during construction.

rotation mechanism.

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EXPERIMENTAL FACILITIES AND UTILIZATION 10.2.4 Automatic Transfer Facilities %

A pneumatic transfer system, permitting applications with short-lived radioisotopes, rapidly conveys a specimen from the reactor core to a remote rcceiver. The in-core terminus is normally located in the outer ring of fuel-element positions. The sample capsule (rabbit) is made of polyethylene, approximately 14.2 mm ID and 100 mm length. It is conveyed to a receiver/sender station via aluminum and polyethylene tubing nominally 1.25-in. OD (3.18 cm) and at least 1.08 in. (27.4 mm) ID, with radii of curvaturc no less than 2 ft (61 cm). Figure 10.4 is a schematic diagram of the transfer system, as originally supplied. The gas supply has been replaced by compressed helium gas. The in-tank and in-core portion of the pneumatic transfer system is illustrated in Fig. 10.5.

Figure 10.5: In-tank and In-cre portions of the pneumatic transfer Xv.ktem.

10.2.5 Beam Ports The KSU TRIGA Reactor is provided with four beam tubes. Beam-tube sleeves are welded to the outside surface of the tank to allow extraction of neutrons and gamma rays for a variety of experiments, and irradiation facilities for speci iens as large as 6 in. (15.2 cm) in diameter. Three of the beam tubes are oriented radial with respect to the center of the core. The fourth is tangential to the outer edge of the core. All radial tubes terminate at the outer edge of the reflector assembly, but one is aligned with cylindrical void in the reflector graphite. In order that this void clears the rotary specimen rack in the reflector, all beam-tube axes are 2.75 in. (7.0 cm) below the centerline of the core.

The four beam tubes are in two sections within the concrete shield. The inner section is 6.065 in.

(15.2 cm) inside diameter. The sleeve has a 6315-in (16 cm) diameter.. A step is incorporated into the design, with the outer section 8 in. (203 cm) in diameter. The outer section is made of cadmium plated steel. A 0.5-in. (12.7-mm) line in the argon vent system leads from the outer section, thus permitting purge of accumulated gases. The inner sections or the beam tubes are aluminum, coated on the outside with plastic to prevent corrosion when in contact with concrete.

The gap between the inner tube and the sleeve prevents stresses resulting from thermal expansion in the aluminum tank K-State Reactor 10-7 Original (12/04)

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- I r-CHAPTER 10 A 4-in. (10.2-cm) thick steel shadow shield, 40-in. (102 cm) square, is placed around each.beam tube in the concrete shield to provide additional shielding for the area adjacent to the beam port.

The shield is placed around the inner tube, at its juncture with the outer. When beam tubes arc not in use, special shielding is provided along the tube axes. The original shielding configuration is fabricated in four sections: an inner concrete-filled plug, an outer wooden plug, a lead-filled shutter, and a lead-lined door. .

The inner section of the beam-tube shielding is a concrete-filled aluminurn plug (Fig. 10.6) approximately 48 in. (1.22 m) long. -It consists of a 42.5-in. (1.08-m) long 6-in. (152-cm) diameter inner portion, which is.rigidly joined to a 5-in (12.7 cm) long, 7.875-in (20-cm) diameter steel outer portion. The plug weighs approximately 180 pounds (82 kg). The shielding in the inner plug consists of 118 in. (32 mm) of boral on the inner end, followed by 4 in. (10.2 cm) of lead, 36 in. (0.92 m) of borated normal-density concrete, and the 5-in. steel outer portion.

This outer portion is equipped with a threaded hole for attaching the beam-tube-plug handling

'tool.

Figurc 10.6: Inner beam tube plug.

The outer beam-tube shielding (Fig. 10.7) is a wooden plug, 48 Mi.(1.20 m) long and 8 in.'(203 cm) diameter. The plug weighs about 45 pounds (20 kg) and has a handle for manual removal.

Experience has shown the outer wooden beam port plug is not required to control radiation levels at acceptable levels, and the use of wooden plugs in the beam ports is optional.

The outer end of the beam tube is equipped'with a lead-filled safety shutter and docr to provide limited gamma ray shielding when the plugs are removed. The shutter (Fig. 10.8) is contained in a rectangular steel box recessed into the outer surface of the concrete shield. Tbe shutter is 9 in.

(23 cm) square and 4.5 in. (I IA cm) thick. It is welded from 0.25-in (63-mn) steel Vlate and filled with lead. On one side or the shutter is welded a threaded socket foi connection of a push

--rod. . * * . * *: * * : -*

. . a . * . S K-State Reactor 10-8 Orginal (12104)

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EXPERIMENTAL FACILITIES AND UTILIZATION

...;%..f

1*

a* .

S.

S. * '** * .

-.r.'

Figure 10.7: Outcr beam tube plug.-

The shutter box is equipped with a door made of 0.375-in (9.5 mm) steel lined with 1.25 in. (3.2 cm) of lead for additional shielding. The door is hinged at one side and equipped with a rubber gasket and six clamps that permit the door to be sealcd against the possibility of rapid loss of shielding water in thecvent of a major beam-port ical. A valve mounted to the side of the steel box (not shown in the figure) provides access for the push rod operator (open position) and maintains the water seal on the safety shutter (closed position).

4-Figurc 10.8: Beam tube safety shutter.

:-e'.'*:

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10.3 Experiment Review ~'.<* '

A wide array of experiments have been documented and approved for execution in the operational history of the facility.'The experimcnt review and approval process is conducted in .f accordance with approved facility administrative procedures. If an experiment falls within the scope of an approved experiment, a request for operation is submitted to the Reactor Supervisor.

(or designated alternate). The Reactor Supervisor (or designated alternate) verifies that operation is within the scope of an approved experiment, and approves the request by signature so that the experiment may be scheduled. If it is determined that the proposed experiment does not fall with K-State Reactor 10-9 Original (12104)

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CHAPTER 10 LT

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s; n the scope of a previously approve experiment or if the expe'ridmcnt potentially involves an unreviewed safety question; thc experiment is considered a "new" experiment .

10.3.1 Planning and Scheduling of Niw Experiments.*

New experiments require approval.of the Reactor Safeguards Conimittee prior to implementation.

As noted in Chaptei 12, the Radiation Safety Officer must conctur for permissive decisions. To support 'Committee review, a written description of each proposed new experimeit must be prepared, with sufficient detail to-enable evaluation of experiment safety. The Committee shall make an evaluation as to whether new contemplated experiments, procedures, facility modifications (and/or chinges thereto) meet review criteria, and approve experimental operations (with or without changes or additional constraints) or prohibit the experiment from being performed. The following information is a minimum for the proposed experiment:

Purpose of the experiment'

Backgrcund (if appropriate)

Procedure - to include a description of the experimental methods to be used and'a

description of the equipment to be used. A sketch of the physical layout and a tabular list of equipment necessary for the experiment is recommended if appropriate:*

A summary of various effects that the experiment could cause, or that could interact with the experiment, or including: * .

- Reactivity Effects

- Thermal-Hydraulic Effects * . * *

- Mechanical Stress Effects * .

References. '.

I. .. . :

The Reactor Safeguards Committee may require additional information to determine that an experiment is acceptable; the experiment sall not be scheduled until the Committee has reviewed the proposed experiment, including any supplemental information requested by the Committee.

10.3.2 Review Criteria The Reactor Safeguards Committee shall consider new experiments in teims of effect onteactor operation and the possibility and consequences of failure, including, where significant, consideration of chemical reactions, physical integrity, design life, .proper cdoling, interaction with core components, and reactivity effects. Before approval, the Commnittee shall conclude that in their judgment the experiment by virtue of its nature and/or design will not constitute a significant hazard to the integrity of the core or to the safety bf personnel. Evaluation of the proposed experiment shall include (as a minimum, not limited to) that the likelihood of occurrences listed below are minimal or acceptable in both bornial and failure modes:

Breach orfission product barriers (which could occur through reactivity effects, thermal effects, mechanical forces, and/or chemical attack)

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Interference with reactor control system functions (which could occur through local flux perturbations or mechanical forces that can affect shielding or confinement)

Introduction or exacerbation of radiological hazards (which could occur through irradiation of dispersible material, mechanical instability, inadequate shielding and/or inadequate controls for safe handling)

Interferences with other experiments ofoperations activities (which could occur through reactivity effects from more than-one source, degradation of performance ofsharcd systems - e.g., electrical, potable water, etc., physical interruption of operational activities or egress, toxic or noxious industrial hazards, unanticipated effects of pulsing. Note this evaluation should also consider potential for fire or personnel exposure to toxictnoxious material)

Determination that the proposed activity is in compliance wiLh Technical Specifications If an event or new information challenges th6 original cvaluation, the Committee shall review the experiment approval and determine if tlie original approval is still valid prior to a continuation of the experimental program. When container failure is discovered that has released material with potential la damage the reactor fuel or structure (by corrosion or other means), physical inspection shall be performed to detcrmine the coziseqtiences and need for corrective action. The results of the inspection and any corrective action taken shall be revieived by the Reactor Safeguards Committee and determined to be satisthctory before operation of the reactor is resumed.

10.4 Bibliography "250-k I TRIGA Mark)I PrulsingReacrorMechanicalMaintenanceand OperatingManual,"

Report GA-3399, General Atomic Division, General Dynamics, 1962.

"License R-88, Docket 50-188, Amended FacilityLIcensefoj the KSU TRIGA AMark) Nuclear Reactor," Nuclear Regulatory Commission, Revised 21 Dcc 1998 including Amendment 12.

"KansasState University TRIGA Mark)1ReactorFacilityOperationsManual."

"Kansas Stale University TPJGA Mark)1 Reactor FacilityAdministrative Plan."

USNRCRegulatory Guide 2.4, "Revvie ofErperimentsforResearchReactors" (July 1976)

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11. Radiation Protection and Waste Management This chapter deals with the overall radiaii6n jrbzection program -for the KSU TRIGA Mark 1I Nuclear Reactor Facility and the associated practices for management of radioactive wastes. The chapter identifies radiation sources~that may be present during normal operation ofrthe reactor and

,oni the varioiit proccdlurcs folio~ed to.rihonitor and-control these-sources: The chapter also identifies cxpeted personnel iadiition 'xploures due to normal operations. Supporting calculations aic found in the Safety Aiuilysis Rc~oft Appen~dix A.

11.i Radiation Protection The Radpati dtlProitection Program for the K-State riclear research reactorSfcility was prepared to meet the icequirements of Titli 210, Pint 20.1101, iC~dlofFcderal Rcgulations (10CFR20). The.

'Pzograin also incorporates requiremaentsof the State of Kaiisas. The Program sceek to control radiation exposures and radioactivity releases to a level that is As Low As Reasonably Achievable (ALARA) without significantly restricting oprertion of the Facility for purposes of education and research. Ihe Prograiii is executed in coordination with the Office of Radiation Safety, Department of Public Safety, for Kansas State University. TheProgram is reviewed and ppp'oved by the Reactor Safeguards Committee for hceReactorFacility..

C itain aspects of the Program detlewith radiactivi ixatcrials regulated by the StateOf Kansas (an rAgreemnt state) under licetise Crel-ea . tae eeefore, e Universit A R easdiiion Safety AComhittee (rsponaible for utdinifiscatriition of the State license) reailwid the Progras. The Radiation Paotecion Prbgra oiamivsexlodtedfollowingthe guidatnceoo f 'theOfferica n National Standasr Radcts ot Prorection dRealrch Reacior Faclities [1] and Reguliitoy Guides issued by IibeNRC [2-7j.

11.1.1 Radiation Sources Radiation sources present in the reactor facility may be gand 1 eous (airborne), liquid, solid, or form. These fonns arc treated individually L . f in successive

. e.

subsections. .

Airborne sources consist mainly of 4"Ar (1.8 h -half-life), attributable to neutron activation of nafural 40Ar in air in the reactor, bay, in the rotary specimen rack adjacent to the core, and dissolved in W6 primary coolant. The nuclide "N (7.1 i half-life) is produced'in the primary coolant as a result of the 16O(n,p)N reaction. Because of its short half-life, "N contributes negligibly 'to dff-site radiation 6xposure, but is the major souric of radiation dose to the area

above the reactor pool.

Liquid sources are principally.limited to condensate water from the facility air handling system, which occasionally contains'small concentritions of tritium: There 'are.ociasional releases of.

tritium-bearing primrariy coolant fromievel adjustments in the reactor tank or bulk-shield taik to support maintenance and operations.

Solid sources consist of rcactor fuel, a startup neutron source, and fixed radioisotope sources such.

as those used for instrumentation calibration. Solid %asteis another solid source, very linitea in volume and specific activity. Solid wastes include, ion-exchange resin usekl in' reactor-water KState Reactor .dgi . - Orignal (12/04)

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CHAPTER 11 cleanup, and contaminated tools, labware and anti-contamination clothing associated with reactor AJ experiments and surveillance or maintenancc operations. '

3. Airborne Radiation Sources U

%1 During normal operation of the reactor facility, there are three major potential airborne U sources, 3 H, "EN, and 4'Ar. Airborne tritium occurs With evaporation of primary coolant, and contributes negligibly to personnel or public radiation exposure. Assumptions and calculations used to assess production and radiological impact of "N, and 4"Ar sources l during normal operations are discussed in Chapter I1, Appendix A and summarized here.

Fuel clement failure, although not expected, could occur while the reactor is operating normally (e.g., associated with manufacturing defects or cladding corrosion), and would result in a small penetration of the cladding through-which fission products would be slowly released into the primary coolant. Some of these fission products, primarily the noble gases, would migrate from the coolant to the air of the reactor room. Although this type of failure could occur during normal operation, its occurrence is not normal and no normal operation would take place until the failed clement were located and removed from the core. Thus, failure of a single element is evaluated in Chapter 13 as an abnormal, accident situation.

Tritium in the Reactor Bay A 5-year average of tritium assay (performed monthly) indicates specific activity in the primary coolant of 228 pCi/mi for 250 kW operations. If the reactor bay atmosphere were saturated with this water at 301C, the water concentration in the air would be less than 3 x ICO5 g mL' and the activity concentration in the atmosphere would be less than 6.84E-09 pCi/ml. Based on history, tritium concentration at 500 kW would be less than 1.37 x 10 pCiIml, and tritium concentration at 1,250 kW would be less than 3A2 x 104 pCi/ml. In a21 cases, tritium concentrations are well below the I OCFR20 Appendix-B DAC of 2 x I 0"pCi mL1; and the atmospheric effluent limit of IOf pCi ml:. Even the primary coolant would meet the liquid effluent limit of IT3 pCi mL.' without further dilution.

4"Ar In the Reactor Bay Production of "Ar arises from neutron absorption in natural argon present in air spaces, notably the rotary specimen rack (RSR), and dissolved in primary coolant. Occupational exposure to 41Ar during normal operation of the KSU TRIGA reactor can therefore occur in the reactor bay. According to Appendix B of IGCFR20, the submersion DAC for occupational exposure is 3 x 106pCi mLU' and the effluent limit is I x lI'pCi mL.

Exhaust of the rotary specimen rack:' As shown in Chapter II Appendix A, the equilibrium activity of "Ar in the RSR is 0.56 Ci for sustained operation at 500 kW Exhaust of other air volumes, e.g., beam tube or pneumatic transfcr system, would release lesser quantities of 4'Ar into the reactor bay.

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RADIATION PROTECTION AND WASTE MANAGEMENT thermal power, IA Ci at 1,250 kW. If this activity were instantly dispersed into the reactor bay atmosphere, under. normal ventilation conditions, and a worker were continuously exposed thereafler, the cumulative exposure would be at 4.4x I 10' puCi h mL'I, well below the occupational exposure limit for 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months at the DAC, 3 x 10 4 pCi h maLI.

Release from primnrv coolant: As shown in Chapter *I1Appendix A; even with extremely conservative assumptions, during sustained operation at full power with ventilation, the steady-state activity concentration of 4Ar in the reactor-bay atmosphere would be 7.2 x 107 pCi mL:', less than the occupational DAC.

Offsite Impact of MIAr. As shown in Chapter 11 Appendix A, the peak off-site activity concentration during normal operations would be about 0.003903pCi mL:' at 135 In downwind under slightly unstable atmospheric conditions, occurring 0.6% of total

. time. This concentration is less than the effluent limit of 0.01 pCi xnL* 1. A full year of

. . - operation at the maximum power level maximum conccntration.would. result in an effective dose at the receptor.with the maximum concentration of only about 0.16 mirem, well within applicable limits. The highest dose to a location occurs at 2140 meters with a dose of 3.8 mrern, well below the maximum allowed 10 nirem from effluents.

'IN in the Reactor Bay

An additional source of airborne radioactivity during normal reactor operations is "N, which is generated by absorption of fast neutrons in "O present in the primary coolant within the core. Compared to molecular oxygen, dissolved oxygen oi oxygen in the RSR airspace is of negligible significance.

After "N is produced in the idre region, it rises to the tanwc surface and spreads to form a cylindrical volume source within the reactor tank, thereby leading to significant exposure rates above the tarnl A discussed in Chapter 11, Appendix A, '1N created in the coolant

is likely to remain in anionic form and in iolution, with negligible release in the gaseous

stati- Furthermore, because of the only 7.13-second hilf-life of 16N; airborne

concentrations; on-site and off-site, are of negligible significance compare to the direct dose from 36N dissolved in the primary coolant.

As shown in Chapter 11 Appendix A, conservative calculations lead to an expected exposure rate of approximately 25 nmR ig' at one meter above the center of the reactor tank during sustained operation at 500 kW thermal power, increasing to nearly 100 mR hI 1 at 1250 kW.

b. Liquid Radioactive Sources

'During normal operation of the KSU 7RIGA reactor, the only production of. liquid radioactive materials occurs through neutron activation of impurities in the primary coolant. Most of this material is captured in mechanical filtration or ion exchange in demincralizer resin; therefore, these materials are dealt with as solid waste. Non-routine liquid radioactive material could result from decontamination or maintenance activities such as resin changes. These liquids are collected in sump tanks along with condensate K-State Reactor .11-3 Original (12104)

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CHAPTER 11 from the air conditioning system. Quantities are small and these liquids are released to the sanitary sewerage system after assay and filtration. Most of these are generated from condensate drains of the air-conditioning system. The only radionuclides observed are tritium and (sporadically) trace quantities of 137CS. Typically there are three releases of liquids annually, each amounting to 2.5 m3 . Typical concentrations are 2 x 104 pCi miLs of 3H and 2 x 107 t 1iCi m1 X of 137CS. Even without dilution, all these are well below IOCFR20 Appendix-B effluent concentration limits of 1 x 10'3 pCi mIL* of 3H and I x O'PCi mL: of 137Cs, and monthly sewerage limits of 1 x 10.2 pCi mL' of 3Hand 1 x 105.,Ci rmL: of"7 Cs.

The only significant liquid radioactive source at the KSU TRIGA reactor is the primary coolant. The only measured radionuclides are 4"Ar, "'N, and 3H, and their consequences are examined in Section Chapter 11 Appendix A. Activation products such as "Mg, "AI, and 5mMn, while undoubtedly present, arc of such low concentrations and have such short half-lives that they are not detected in surveillance programs. Similarly, "N present in primary coolant as it passes through the cooling system undoubtedly contributes to ambient dose ratcs inside the reactor bay but surveillance at full power reveals no contributions of 1 mR fh'or greater.

C. Solid Radioactive Sources The solid radioactive sources associated with KSU TRIGA reactor operations are summarized in Table 11-1. Because the actual inventory of fuel and other sources continuously changes in normal operation, the information in the table is to be considered representative rather than an exact inventory.

r (-. I Solid and liquid wastes are not included in Table 11-la and ll-l .b. These sources are iddressed in Section 11.2.

11.1.2 Radiation Protection Program The Radiation Protection Program was prepared by personnel of the Kansas State University -

TRIGA Mk II Nuclear Reactor Facility in response to the requirements of Title 10, Part 20.1101, Code of Federal Regulations (I OCFR20). The goal of the Program is the limitation of radiation exposures and radioactivity releases to a level that is as low as reasonably achievable without seriously restricting operation of the Facility for purposes of education and research. The K-State Reactor 11-4 Original (12/04)

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RADIATION PROTECTION AND WASTE MANAGEMENT Program is executed in coordination with thc Office of Radiation Safety, Department of Public Safety, Kansas State University. It has been reviewed and approved by the Reactor Safeguards Committee for the Reactor Facility. Certain aspects of the Program deal with radioactive materials regulated by the State of Kansas (an Agreement state) under license Cool 1-01 and the Program has been reviewed by the University Radiation Safety Committee, which is responsible for administration of that license. The Radiation Protection Program is designed to meet requirements of IOCFR20. It has been dJeielopcd following the guidance of the American National Standard RadiationProtectionaf ReseardiReactor Facilities[1] and Regulatory Guides issued by the NRC [2-7].

a. Management and Administration Preparation, audit, and review of the Radiation Protection Program is the responsibility of the Nuclear Reactor Nuclear Reactor Facility Manager. The Reactor Safeguards Committee (chaired by the Head of the Department of Nuclear Engineering) reviews the activities of the Manager and semi-annual audits prepared by the Manager. The Reactor Safeguards Committee examines records required by the Radiation Protection Program as well as audit reportiby the Manager during their semi-annual inspections.

Training, surveillance and record keeping are the responsibility of the Reactor Supervisor who reports to the N4uclear Reactor Facility Manager. ALARA activities, for which record keeping is the particular responsibility of the Reactor Supervisor, arc incumbent upon all radiation %workersassociated with the Reactor Facility.

Substantive changes in the Radiation Protection Program require approval of the Reactor Safeguards Committee. Editorial changes, or changes to appendices, may be made on the authority of the Nuclear Reactor Facility -Manager. -Changes madc to -the Radiation Protection Program apply automatically to operating or cmcrgency procedures; corresponding Program changes may bo made without further consideratfon by the Reactor Safeguards Comnmittee. As wvith proceduresthe Reactor Supervisor or Nuclear Reactor Facility Manager may override elements of the Program on a temporary K-State Reactor 11-5 Original (1210-4)

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CHAPTER 11 emergency basis so long as the emergency changes are brought promptly to the attention of the Safeguards Commiinee.

b. Training Implementation of training for radiation protection is the responsibility of the Reactor.

Supervisor. There are two categories of radiological training at the K-State reactor.

Personnel who need access to the facility, but are not reactor staff, are either escorted by trained personnel or provided unescorted access training. Radiation training for licensed operators and staff is integrated with the training and requalification program.

The goal of unescorted access training is to provide knowledge and skills necessary to control personnel exposure to radiation associated with the operation of the KSU nuclear reactor. Specific knowledge and skills required to meet the goal have been developed as learning objectives, and training material is based on these learning objectives. Specific training requirements of IOCFR 19, 10 CFR 20, the Radiation Protection Plan, and the Emergency Plan are explicitly addressed. A facility valkthrough is incorporated.

All persons granted unescorted access to the Reactor Facility must receive the training and must complete without assistance a written examination over radiation safety and emergency preparedness. An examination score of at least 70 percent is required.

Examinations must be retained on file for audit purposes for at least three years.

The reactor staff accomplishes health physics functions at the K-State reactor following approved procedures. Therefore, procedure training for the licensed reactor staff training includes additional radiological training. Examinations for reactor staff training are prepared and implemented in accordance with the K-State reactor requalification plan.

11.1.3 ALARA Program

a. Policy and Objectives Management of the Reactor Facility is committed to keeping both occupational and public radiation exposure as low as is reasonably achievable (ALARA). The specific goal of the ALARA program is to assure that actual exposures are no greater than 10 percent of the occupational limits and 50 percent of the public limits prescribed by I0CFR20, namely, ALARA goals of:

Workers: < 500 mrem annual TEDE

< 5 rem annual dose equivalent to any organ except the lens of the eye

< 1.5 rem annual dose equivalent to the lens of the eye 5 rem annual dose equivalent to the skin 50 mrem dose equivalent to the fetus during pregnancy Public: <50 mrem annual TEDE K-State Reactor 11-6 Original (12/04)

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b. Implementation of the AkARA Program Planning and scheduling of operations and experiments, education and training, and facility design are the responsibilities of the Reactor Supervisor and/or the Nuclear ReactorFacility Manager. Any action which, in either of their opinions, might lead to as much as half the annual ALARA dose limit (Section 6.1) to any one individual in one calnidar-quartcr requires a formal ALARA review and report. Any staff member or experimenter, or any member of the Reactor Safeguards Committee may call for an ALARA review of a proposed action.

Under any of these circuinstances, it is the responsibility of the Reactor Supervisor to conduct an ALARA review and report. Only with the approval of the Reactor Supervisor and the endorsement of the Nuclear Reactor Facility Manager, may the action proceed.

c. Elements of the ALARA Review and Rcport The following topics shall be considered, if applicable. The report shall include discussion of how these topics affect personnel exposure and specific actions reconmmended, categorized by topic:

Features for External Radiation Control Shielding and construction rnaterials Radioactive material storage and disposal Monitoring systems Facility layout Control of access to high and very high radiation areas Contamination Control Ventilation and filtration

Containment of contamination Confinement of contamination spread Construction materials to facilitate decontamination Facility layout Effluent Control Gaseous effluents Liquid effluents Effluent monitoring Operations and Operations Planning Assessment of potential individual and collective exposures Application ofshielding. sutee,i diuisiance fordse reduction.

Use of ventilation and decontamination to reduce collective dose Provision of special radiac or communications instrumentation Provision of special personnel training and practice Provision of special supervision and surveillance

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CHAPTER 11 Provision of special clothing or other protective gear *1

d. Review and Audit Implementation of the ALARA Program is audited semi-annually by the Nuclear Reactor Facility Manager as part of the general audit of the Radiation Protection Program. J 11.1.4 Radiation Monitoring and Surveillance The radiation monitoring program for the KSU reactor is structured to ensure that all three categories of radiation sources-air, liquid, and solid-are detected and assessed in a timely d manner. Surveillances are

a. Surveillances .

Radiation monitoring surveillance requirements are imposed by the Reactor Safeguards Committee through the Radiation Protection Program (independent of the Emergency Plan) for:

Monthly Wipe test reactor bay and control room Quarterly Source inventory and leak test Semi-annually Environmcntal surveillance (radiation levels at full power)

b. Radiation Monitoring Equipment Radiation monitoring equipment used in the KSU reactor program is summarized in Table 11.2. Because equipment is updated and replaceid as technology and performance requires, the equipment in the table sh6uld be considered as representative rather than ciact*

c. Instrument Calibration Radiation monitoring instrumentation is calibrated according to written procedures. Whenever possible, NIST traceable sources are used for the calibration. The Nuclear Reactor Facility or the University Radiation Safety Office (Office of Public Safety) are responsible for calibration of the Table 11.2 instruments on site. Calibration records are maintained by the facility staff and audited annually by the Nuclear Reactor Facility Manager. Calibration stickers containing pertinent information are affixed to instruments.

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Table 11.2, Radiation Monitoring and Surveillance Equipment at the KSU TRIGA.

Item Location Function Continuous air monitors (2)

Effluent monitor 22-ft level Measure radioiodine, noble gases, and particulates Reactor room air monitor 12ftlevel Measure radiolodine in room air 0-ft level Measure gamma-ray exposure Area radiation G-M monitors (3) 12-ft level rates in accessible areas of the 22-ft level reactor room Measure exposure rate at pool Pool surface G-M monitor 22-it level surface (N-16 and Ar-41)

Evacuation alarm G-M monitor 22-f1 level Measure high level gamma-ray exposure rate (5 R h')

Measure exposure rate at reactor Entrance G-M monitor Control room rooni entrance Portable ion chambermeters (3) 0, 12, 22-It levels Measure gamma-ray exposure rate, sense beta particles Portable G-M survey meters (3) 0, 12, 22-ft levels Measure gamma-ray exposure rate, sense beta particles Portable neutron survey meters (2) 0-ft & 22-4 Measure ambient dose rate levels Fixed alphalbeta counter Room 11 Ward Wipe-test assay Hall Liquid scintillation spectrometer University Counts liquid and wipe-test Radiation Safety samples Office Gamma-spectroscopy systems (3) Room II Ward Gamma-ray assay Hall Direct reading pocket dosimeters Control room Personnel gammalneutron dose High volume air sampler Control room Emergency sample collection

11.1.5 Radiation Exposure Control and Dosimetry Radiation exposure control depends on many different factors including facility design features, operating procedures, training, proper equipment, etc. Training and procedures have been discussedbn Section 11.1.2. This section deals with design features such as shielding, ventilation, containment, entry control for high radiation areas, protective equipment, personnel exposure, and estimates of annual radiation exposures for specific locations within the facility. Dosimetry records and trends are also included.

a. Shielding Abe biological shielding around the KSU 7TRGA reactor is the principal design feature for control of radiation exposure during operation. The shielding is based on TRIGA shield designs used successfully at many other similaf reactors.

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CHAPTER 11 The reactor is designed so that radiation from the core area can be extracted into the 0 foot level for research, education, and training purposes. When the beam port shielding is removed, additional measures are required to control radiation exposure by either restricting access to areas of elevated radiation fields or providing a new shielding !

configuration for access. When a new shielding configuration is established (or modified), radiation surveys are required to determine access and monitoring \

requirements. When a specific configuration is used power levels higher than previous J testing, additional radiation surveys are required to ensure adequate controls at the new power level. Monitoring is not required for a well-defined shielding configuration that previously met radiological limits, but is not prohibited.

b. Personnel Exposure J Regulation 10CFR20.1502 requires monitoring of workers likely to receive, in one year from sources external to the body, a dose in excess of 10 percent of the limits prescribed in 10CFR20.1201.

The regulation also requires monitoring of any individuals entering a high or very high radiation area within which an individual could receive a dose equivalent of 0.1 rem in one hour. According to Regulatory Guide 8.7 [2], if a prospective evaluation of likely doses indicates that an individual is not likely to exceed 10 percent of any applicable limit, then there are no requirements for rccordkeeping or reporting. Likewise, -

Regulatory Guide 834 [3] indicates that, if individual monitoring results serve as confirmatory measures, but monitoring is not required by IOCFR2O.1502, then such results are not subject to the individual dose recordkeeping requirements of I OCFR20.2106(a) even though they may be used to satisfy I OCFR2O.lS0l requirements.

Table 11.3 lists results ofa 12-year survey of occupational exposures at the KSU TRIGA Reactor Facility. There have been no instances of any exposures in excess of 10 percent of the above limits. Thus, retrospectively, only confirmatory monitoring is required and 10CFR20.2106(a) recordkeeping requirements do not apply, so long as there arm no significant changes in the facility, operating procedures, or occupational expectations.

For operation at 500 kW, records of exposure are not likely to result in increased record keeping requirements. If, in the view of supervisory personnel (Reactor Supervisor, Nuclear Reactor Facility Manager, or Radiation Safety Officer), any action under consideration might lead to exposures in excess of 10 percent of any applicable limit, then the ALARA program is triggered. A consequence of ALARA program planning, which is described in Section 11.1.3, might be the imposition of federally required recordkeeping procedures.

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RADIATION PROTECTION AND WASTE MANAGEMENT Table 113. Rcpresentativc Occupational Exposurcs at 250Mk.

Numbers ofpersons in sunual-dose categories Year Immeasurable < 0.1 rem 0.1-0.5 rem > 05 rem 1992 28 0 0 0 1991 23 0 0 0 1990 20 0

0 0 1989 19 1 0 0 1988 23 3 1 0 1987 23 0 0 0 1986 26 1 0 0 1985 31 8 0 0 1984 33 1 0 0 1983 29 2 0 0 1982 26 7 0 0 1981 11 23 0 0 Monitoring of workers and members of the public for radiation exposure required by the Reactor Safcguards Committee and is described in Facility Procedure 9. Principal objectives ofProcedure 9 include:

Authorization for Access Personnel who cuter the control room or the reactor bay will either hold authorization for unescorted access, or be under direct supervision of an escort (i.e., escorted individuals can be observed by and bear instructions of the escort) who holds authorization for unescortcd access.

Access Control During Operation

%Vhenthe reactor is operating, the licensed reactor operator (or senior reactor operator) at the controls shall be responsible for controlling access to the control room and the reactor bay.

The 22-foot level access has a line of sight to the control room, and has radiation monitoring positioned directly over the pool surface and mounted on the rail surrounding the pool. The operator at the controls is responsible for appropriately controlling access to the 22-foot level based on radiological conditions.

Neutron Dosimetry If there is potential for exposure of personnel to neutrons within thi reactor bay, personnel who enter the reactorbay shall have neutron sensitive individual monitoring; this individual monitor shall be assigned to single individuals Exposure Records for Access During Operation Personnel who enter the reactor bay during reactor operation shall have a record of accumulated dose measured by a gamma sensitive individual monitoring device; at the K-State Reactor 11-11 Original (12104)

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CHAPTER 11 uP J~1 discretion of the reactor operator at the controls, a single individual monitoring device may be used for individual monitoring of no more than two people who agree to stand together in the reactor bay.

Exposure Records for Access During Non-Operating Conditions Personnel who enter the reactor bay while the reactor is secured shall have a record of accumulated dose either by measurement through individual monitoring or based on assessment of data from individual monitoring devices or surveys. J

c. Record Keeping Although the Reactor Facility is likely exempt from federally required record keeping requirements of IOCFR20.2106(a), certain records are required in confirmation that personnel exposures are less than 10 percent of applicable limits.

Records of Prior Occupational Exposures These records (NRC Form 4) are initially obtained, and then maintained permanently by the Ofmice of Radiation Safety.

Records of Occupational Personnel Monitoring The Office of Radiation Safety permanently maintains these records (NRC Form 5).

Forms in use include monthly report for the University as a whole, monthly summary report for the Nuclear Reactor Facility, and quarterly report on extremity exposures for the University as a whole.

Records of Doses to Individual Members of the Public Sclf-reading dosimeter records arc kept in a logbook maintained by the Nuclear Reactor Facility. Such records are kept permanently. Results of measurements or calculations used to assess accidental releases of radioactive effluents to the environment are tobce retained on file permanently in the Reactor Facility.

11.1.6 Contamination Control Potential contamination is controlled at the KSU TRIGA reactor by using trained personnel following written procedures controls radioactive contamination, and by operating a monitoring program designed to detect contamination in a timely manner.

There are no areas within the reactor laboratory with continuing removable contamination. More likely sites of contamination are sample ports at the rotary specimen rack (RSR) and central thimble (CI) and at a samplc-handling table for receiving irradiated samples. These sites are covered by removable absorbent paper pads with plastic backing, and are routinely monitored on a periodic basis. In some cases, an K-State Reactor 11-12 Original (12104)

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RADIATION PROTECTION AND WASTE MANAGEMENT enclosed work area is provided. If contaminated, pads are removed and treated as solid radioactive waste. While working at this or other potentially contaminated sites, workers wear protective gloves, and, if necessary, protective clothing and footwear. Workers are required to perform surveys to assure that no contamination is present on hands, clothing, shoes, etc., before leaving workstations where contamination is likely to occur. If contamination is detected, then a check of the exposed areas of the body and clothing is required, with monitoring control points established.for this purpose. Materials, tools, and equipment are monitored for contamination before removal from contaminated areas or from restricted areas likely to be contaminated.

Reactor Facility staff and visiting researchers arc trained on the risks of contamination and on techniques for avoiding, limiting, and controlling contamination.

Table 11.4 lists sample locations for routine monitoring of surface and waterborne contamination control measures. On a monthly basis, 100 cn 2 swipe tests and I mL watcr samples are analyzed for contamination.

Table 114A, Rcpresentatlvc Contamination

. Sampling Locations. -

SWIPE TESTS WVATER SAMPLES 0-toot levCl Bulk shield tank Floor Primary coolant Door handle Secondary coolant 25Cf Irradiator Source safe Source cave Sink Cleanup system 12-foot level Floor Door handle Control room 22-ft level Floor Sample table RSR loading port Cl loading port Acceptable surface contamination levels for unconditional release are given in Table.

11.5, as provided in the approved RadiationProtectionProgram. SU TR2GA Mark 11 Nuclear Reactor Facility. Limits on average contamination levels for unconditional release are calculated based on survey areas smaller than I n2 . Limits on maximum contamination levels for unconditional release arc calculated based on survey areas smallerthan l00 cm2 . * .

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CHAPTER 11 11.1.7 Environmental Monitoring Environs monitoring is required to assure compliance with 10CFR20, Subpart F and with Technical Specifications for the Facility operating license. Installed monitoring systems include area radiation monitors, airborne contamination monitors, and a radiation monitor at the pool surface. The Reactor Safeguards Committee may require additional monitoring and has (via the Radiation Protection Program) established requirements for contamination and radiation survey surveillances.

a. Area Radiation Monitors Area radiation monitors at the 22-ft level (and the 0 foot level; if beamn ports are open) are required for reactor operation. Area radiation monitor calibration .1 accomplished as required by Technical Specifications in accordance with facility procedures.

b. Airborne Contamination Monitors The facility uses two air monitoring systems, one on the 12-foot level and one in the exhaust-plume path from the reactor pool to the reactor bay exhaust system.

Airborne contamination monitor calibration is accomplished as required by Technical Specifications inlaccordance with facility procedures.

C. Pool Surface Monitor A radiation monitor is stationed directly over the pool surface. The pool surface monitor is calibrated in accordance with requirements of the Radiation protection program using facility procedures.

d. Additional Monitoring x The Reactor Safeguards Committee imposes additional requirements through the Radiation Protection Program.

Contamination Surveys Contamination monitoring requirements and surveillances.addressed in 11.1.6 prevent track-out of radioactive contamination from the reactor facilities to the environment. The K-State Division of Public Safety, University Radiation Safety Office, maintains an independent contamination-monitoring program under the Kansas State radioactive material license.

As required by 10 CFR 20.1501, contamination surveys are conducted to ensure compliance with regulations reasonable under the circumstances to evaluate the.

magnitude and extent of radiation levels; concentrations or quantities of radioactive material; and potential radiological hazards.

Guidance has been promulgated in lE Circular No. 81-07 (Control of Radioactively Contaminated Materials) for releasing materials from restricted to unrestricted areas:

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RADIATION PROTECTION AND WASTE MANAGEMENT Based on the studies. of residual radioactivity limits for decommissioning (NUREG-06132 and NVUREG-07073), it can be concluded that sufaces tmflormly contaminated at levels of 5000 dpm/lOOcm2 (bela-gamma activity from nuclearpower reactors) rouldresult in potentialdoses that total less than 5 mnremlyr. Therefore, it can be concluded that for the potentially undetected contaminationof discrete itens and materiilsat levels below 5000 d pmfl00 c2, the potentialdose to any individualwvill besignificantlylessthan 5 mremyreven if the accumulationofnumerous items contaminatedat this level Is considered.

The contamination monitoring using portable survey Instrments or laboratory measurements should be performedwith Instrumentation and techniques (survey scanningspeed, counting times, backgroundradiationlevels) necessaryto detect 5000 dpm/100 cm2 total and 100O dpm/J00 cm2 removable beta/gamma contamination. Instruments should be calibratedwith radiationsources having consistentenergy spectrum and instrumentresponsewith the radionuclidesbeing measured If alpha contamination is suspected appropriate surveeys and/or laboratorymeasurements capable of detecting 100 dpm/100 cm2 fixed and 20 dpm/100 cm2 removable ialpha activityshouldbeperformed.

Radiation Surveys Semi-annual environmental monitoring is conducted, involving measurement of both gamma-ray and neutron dose rates within the facility operations boundary and at the site boundary with the reactor at full-power operation.

Monthly surveys are conducted at the KSU TRIGA reactor for radiation levels with the reactor not in use. These are supplemented by semiannual monitoring of both neutron and gamma radiation levels during full power operations. Neutron dose rates are entirely negligible.

Gamma-ray cxposure-rate data, based on semi-annual measurcments over the ten-year period 1988-1998 (250 klW operation) is indicated in Table 11.6. Unccrtainties are I standard deviation.. Source terms are related to reactor power levels; therefore maximum radiation levels during operation at 500 kNW should not exceed twice the maximum historical values In Table 11.6.

Monitoring for Conditions Requiring Evacuation An evacuation alarm (high radiation level) is required at the 22-it level of the reactor.

Response testing of the alarm is performed in accordance with facility procedures.

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CHAPTER 11 .

Table 11.6, Radiation Survey Results for 250 kW Operation.

Leve LoctionMaximum exposure Level Location rate (rnR hIl)

Pool surface 38 i 9 22-ft 1-mabove pool 9t 5 Above bioshield I i1 Above BST 13d 10 ft Outside bioshield 2i 2 1%

2- Control room entry 0.07 :: 0.04 Site boundary 0.03 i 0.02 Outside bioshield 0.02 +/- 0.02 0-fl Source cave 3 +/-3 Cleanup/lab space 1i I 11.2 Radioactive Waste Management The KSU TRIGA reactor generates very small quantifies of radioactive waste, as indicated in Section 11.1.1. Training for waste management functions are incorporated in operator license training and requalification program.

11.2.1 Radioactive Waste Management Program Liquid wastes are released through the sanitary se verage system afer filtration and assay for beta, gamma, and alpha activity. Solid wastes generated under the Kansas State license are transferred to the University Radiation Safety Office, Division of Public Safety, where they are combined with other solid radioactive wastes from the University, allowed to decay, and disposed of under the aegis of die State of Kansas. Solid wastes generated under the reactor license are generally allowed to decay, with subsequent disposal coordinated by the University Radiation Safety Office.

11.2.2 Radioactive Waste Controls Radioactive solid waste is generally considered to be any item or substance no longer of use to the facility, which contains or is suspected of containing radioactivity above background levels.

Volumes of waste at the KSU TRIGA Reactor are small, and the nature of the waste items is limited and of known characterization, there is rarely question of what is or is not radioactive waste. Equipment and components are categorized as waste by the staff. Consumable supplies such as absorbent materials or protective clothing are declared radioactive waste if radioactivity above background is found to be present When possible, solid radioactive waste is initially segregated at the point oforigin from items that are not considered waste. Screening is based on the presence of detectable radioactivity using appropriate monitoring and detection techniques and on the future need for the items and materials involved. Kansas being an "agreement state," radioactive materials generated for K-State Reactor 11-16 Original (12104)

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RADIATION PROTECTION AND WASTE MANAGEMENT research and experiments under the federal byproduct material license of the Reactor Facility are transferred to State of Kansas license for conduct of the activities. Solid wastes resulting from experiments and activities conducted under the State of Kansas license are then physically transferred to the University Radiation Safety Office (RSO), Division of Public Safety, with representatives of the RSO and the Reactor staff ccrtifying the transfer. Solid reactor waste is stored for decay until disposal, coordinated through the RSO.

Liquid wastes in the Reactor Facility are held in temporarily in storage tanks within the facility until pumped into the sanitary sewerage system of the University. Liquid wastes are primarily condensate from the building air-conditioning system and are very-slightly radioactive because of the presence of tritium due to evaporation from the primary coolant and the bulk shield tank. To assure compliance with IOCFR20.2003, liquids wastes are assayed for alpha, beta, and gamma activity prior to release and are filtered to assure that no particulates are released along with liquids.

Although 4"Ar is released from the KSU TRIGA Reactor Facility, this release is not considered to be waste in the same sense as liquid and solid wastes. Rather, it is an effluent, which is routine part of the operation of the facility. A complete description of 4"Ar production and dispersal is provided in Chapter II Appendix A.

11.2.3 Release of Radioactive Waste The history of liquid releases since 1995 is tabulated in Table 11.7. Note the seasonal release of air-conditioning condensate.

Table 11.7, Liquld Releases.

Date Quantity Measured pCi mi L above background released (n 3 ) Alpha Beta! Gamma!

11 Aug99 . 2.5 0 38 0 6 Jul 99 2.5 0 26 0.034 26 Aug 98 2.5 0 55 0.10 27 Jul 98 25c 0 67 0 21 Jul 98 2.5 0 206 0.23 26 Jun 98 2.5 0 89 0.27 16Oct97 *2.7 0 120 0 8Aug97 23 0 250 0 27 May 97 0.7 0 0 0 5 Dec 96 3.3 0 150 0 12 Aug 96 3.0 0.004d 193 0 26Jun96 32 0 116 0-5 Sep 95 3.0 0 135 0 15 Aug 95 4.2 0 138 0 17 Jul 95 4.2 0 54 0 7 Jun 95 3.2 0 54 0

'Tritium 37 as HTO (LSC analysis) b 1 CS_'7"Ba

' Draining bulk shield tank dUnidentified anomaly K-Stale Reactor 11-17 Original (12104)

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U CHAPTER 11 11.3 Bibliography ANSI/ANS-I5.11 (FinalDrafi), RadiationProtectionatResearch Facilities,American Nuclear Society, La Grange Park, Illinois, October, 1992.

Instructionsfor RecordingandRepordingOccupationalRadiationExposureData, Regulatory Guide 8.7 (Rev. 1), U.S. Nuclear Regulatory Cornmission, Washington, D.C, 1992.

MonitoringCriteriaand Methods to Calculate OccupationalRadiationDoses, Regulatoty Guide 8.34, U.S.

ji Nuclear Regulatory Commission, NVashington, D.C., 1992.

Air Sampling in the Workpldce, Regulatory Guide 8.25 (Rev. 1), U.S. Nuclear Regulatory Commission, Washington, D.C., 1992.

PlannedSpecial Exposures, Regulatory Guide 835, US. Nuclear Regulatory Commission, Washington, D.C., 1992.

RadiationDose to the Embryo/Fetus, Regulatory Guide 8.36, US. Nuclear Regulatory Comunission, Washington, D.C., 1992.

InterpretationofBioassay Measurements, Draft Regulatory Guide 8.9 (DG-8009), US. NuclearRegulatory Commission, Washington, D.C., 1992. I-d K-Stale Reactor 11-18 Original (12104)

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SAFETY ANALYSIS REPORT This page intentionally blank, K-State Reactor

Chapter 11 Appendix A Radiological Impact of 4'Ar and '6N During Normal Operations.

di A.1 Introduction Normal operation of the KSU reactor results in two potential source terms for radioactive gaseous D effluent at significant levels, 4"Ar and "N. There arc variations in experimental configuration and jJ possible scenarios where the production of 41Ar may be different than the routine operations; these scenarios do not produce not long term, routine radioactive effluent but are assessed to determine if the amount of radioactive effluent is so high as to impact the annual exposure that might result from routine operations.

A.1. Purpose The purpose of this appendix is to show the methods and calculations used to predict the production, concentrations, and dose rates from "Ar and "N associated with normal operation of d thec KSU TRIGA MN.11 nuclear reactor. J The nuclide 4'Ar is produced by thermal neutron absorption by natural '"Ar in the atmosphere and in air dissolved in the reactor cooling water. The activation product appears in the reactor room j (bay) and is subsequently released to the atmosphere through the reactor bay ventilation exhaust stream.

The nuclide "N is produced by fast neutron interactions with oxygen. The only source of "N in the reactor that needs consideration results from interactions of neutrons with oxygen in the cooling water as it passes through the reactor core. Any interaction with oxygen in the atmosphere is relatively insignificant and is neglected in this analysis.

A portion of the "6N produced in the core is eventually released from the top of the reactor tank into the reactor bay. The half-life of "N is only 7.14 seconds, so its radiological consequences outside the reactor bay are insignificant.

Although not expected, the cladding of a fuel clement could fail during normal operations as a result of corrosion or manufacturing defect. Should a failure occur, a fraction of the fission products, essentially the noble gases and halogens, would be released to the reactor tank and, in part, ultimately become airborne and released to the atmosphere via building ventilation. This operational occurrence, taking place in air, is addressed in Chapter 13 as the design basis accident for the TRIGA reactor. -

Neutron interactions with structural and control materials, including cladding, as well as materials irradiated for experimental purposes, result in the formation of activation products. These products are in the nature of fixed sources and are mainly a sour~c of occupational radiation K-State Reactor 11.A-1 Original (12104)

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CHAPTER 11 APPENDIX A expoureAdinitraiveconrolsprelud th sinifcan foratin o aibore ativtio exposure. Administrative controls preclude the significant fornation of airbomne activation products, other than the aforcmcntioncd 4'Ar.

A.1.2 Radiological Standards The concentration to dose rate (effective dose equivalent) conversion factor for submersion in an infinite atmosphere of 4 Ar is as follows: 2.17 x 101° Sv lI' per Bq m'3, or 0.803 mrcm/h per pCilml (EPA 1993). For 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months annual occupational exposure and 50 mSv maximum permissible annual exposure, this translates to a derived air concentration of 1.15 x 105 Bq m4 or 3.1 x 104 i Ci cm- .(3 x 10 6 pCi cm3 as specified in 10CFR20). For 8766 hours0.101 days 2.435 hours 0.0145 weeks 0.00334 months annual public exposure and I mSv maximum permissible annual exposure, this translates to a derived air concentration or526 Bq n'3 or 1.4 x I0 4spCi cm' 3 (I x I 0 pCi cm 3 as specificd in I0CFR20).

A.1.3 KSU TRIGA Design Bases General System Parametcrs The calculations for 4"Ar and 16N releases during normal operations are based on the folloiving system parameters.

Table A.1, General System Parameters for Normal Operations at 500 kWt Full Power.

Parameter Symbol Value Reactor steady power P 1,250,000 V Core coolant mass flow rate It 0.150 kg Sl Core coolant density p 1.0 g cm3 Core avg. thermal neutron flux at full power (E fing)b 4X 2.05 x 10" n ca e' Core avg. fast neutron flux at full power (E ring)" r 3.00 x 1013 n Cms 4I Thermal neutron flux in RSR at full power RSR 9.00 X I 0 t2 n CM2esl Total neutron flux per watt at fast (piercing) beam port 4250 n cnes' (0.5 McV avg)

Total neutron flux per watt at tangential beam port 1400 n cm*s2 (0.1 MeV avg)

Fuel element heated length L 0381 in How cross sectional area per fuel clement' A 6.2 cm Mass flow rate per fuel element' A 108 g s7' Reactortank diameter 1.98 m Reactortank depth 6.25 m Reactor tanktvater depth above core 4.88 m (16 ft)

Coolant volume in reactor tank V. 1.92 x 107 cm 3 Air volume in reactor bay (144,000 ft3 ) V*

Y 4.078 x 109 cm2 Air volume in rotary spccimen rack Vm 3.75 x 104 cms Vcntilation rate for reactor bay (air changes hourly)d V 0.368 h'l

'See §4.6 of this report.

"See §5.8 of Operations Manual.

'See §13.2.2.2 Of this report dSee letter B.C. Ryan (KSU) to Theodore Michaels (NRC) 15 Jan 99.

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RADIOLOGICAL IMPACT OF 41AR AND 16N DURING NORMAL OPERATIONS

/ Q, Reactor Core Parameters Modeling of the reactor core for radiation transport calculations is based on the following approximations. For purposes of radiation shielding calculations the TRIGA reactor core may be approximated as a right circular cylinder 0.4572 m (18 in.) diameter (OD of F ring). The fuel region is 0.381 mn (15 in.) high. On each end axially is a graphite zone 0.0874 m (3.44 in.) high and an aluminum grid plate 0.0191 mn (0.75 in.) thick. Invcations, there aro p-e1 elements, 3 standard control rods and I transient control rod, 1 void location, I central thimble (void), I source (assume void), and I pneumatic transfer site (assume void). The fuel region may be treated as a horiigeneous zone, as may be the axial graphite zones and the grid plates.

Fuel elements are lA3-in. (3.6 cma ID and 1A7-in (3.7 cqi) OR, clad wit type 3D4 stainless steel'. Fuel density is 5996 kg rn. Fuel composition isPaniuxn,,rHu 5. The uranium is 3 U nnda 'u. Steel density is 7900 kg mn Standardcontrol rods are 0.875-in. 01), the transie rod 1.25-in. OD. Both types of rods are clad with 30-mil thick aluminum (2700 kg m 3 density). The control material may be approximated as pure graphite, with density 1700 kg mi3.

In radiation transport calculations, the core is modeled conservatively as a central homogenous fuel zone (air density neglected) bounded on either end by a homogeneous axial reflector zone, and by a 0.75-in. thick aluminum grid plate, treated as a homogeneous solid. Densities of the homogenous zones are as follow:

Fuel 3602 kg n3 Reflector 1147kgM3 Grid Plate ')2700kg m 3 Composition of the three zones, by weight fraction, are given in the following table.

Table A.2, Compositions of Homogenized Core Zones.

Element Mass Fraction Element Mass Fraction FuelZone AXalxReflector Zone C 0.0617 C 0.7920 Al 0.0010 Al 0.0033 H 0.0139 Mn 0.0041 Zr 0.7841 Cr 0.0368 Mn 0.0013 Ni 0.0164 Cr 0.0117 Fe 0.1474 Ni 0.0052 Fe 0.0469 GridPlate U 0.0741 Al 1.0000 I..

I.

I Composition, by weight, 2% Mn, 18% Cr, 8%Ni, balance Fe.

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CHAPTER 11 APPENDIX A Reactor Bay Parameters For purposes of radiation dose calculations within the reactor bay, the dimensions are approximated as follows:

The reactorbay is approximated as a right circular cylinder36 ft (10.973 mnbigh and 36.68 ft (11.18 m) radius. The reactorvessel structure is approximated as a right circular cylinder, co-axial with the bay, 22 ft (6.706 m) high and 11 ft (3.3528 m) radius. The free volume is 144,000 ftW (4078 ir'). The site boundary, at its nearest approach to the reactor bay, is about 2 m beyond the bay boundary, that is, at a radius of 13.13 m from the center of the reactor.

A.2 Radiological Assessment of 4 lAr Sources A.2.1 Production of 4 t Ar from Beams Operation with a fully open beam port is not a routine operational condition. Beam port operations normally have shielding, collimation and beam stops that prevent a full beam from penetrating the column defined by the beam port into air volume between the reactor and the reactor bay wall. Operating experience with neutron radiography performed at 10 kW involves a neutron flux of 2 x 107 cm2 el or less. We assume here that this is the flux density along the beam port, which has a cross sectional area of 324 emr (8-in diameter). In other words, in a radiography operation S - 6A8 x 109 neutrons per second enter the atmosphere essentially in a parallel beam. The nicroscopic cross section for thermal neutron absorption in 40 Ar is 0.66 barns, so the macroscopic cross section for thermal neutron absorption in Ar in air (0.0129 weight fraction) is p 1.54 x 10' cml. The maximum distance of travel of a neutron is from the reactor tank wall to the exterior wall of the reactor bay, namely, about Lb = 1020 cm. The decay constant for 4lAr is e- 0380 hl and the ventilation rate is 1 = 0368 1h. The volume of the reactor bay is V 8. 4.08 x 10 cn 3. Thus, the activity concentration of airborne "4Arafter sustained operation with an open beam port at 10 kV is given by:

C{Bq/m 3 )= S3 ,(1-e 1.27 x 10 4 Bq cn

=1) (1) or 3.42 x 10'9 pCi riL in conventional units. Operations at maximum power are not performed for radiography, and radiography is not performed long enough to achieve equilibrium 4"Ar.

Therefore, scaling the calculation for sustained operations at 1,250 kIW provides an extremely conservative bound on MAr production. Scaling the 10 kW 4Ar production value to 1,250 kV results in 428 x 10' pCi mL' which is slightly above submersion DAC for occupational exposure; however, conditions for the source term are related to a very unusual set of conditions (open beam port with no shielding) that are not continuous in two respects. Shielding for radiography external to the bema port limits the beam to less than % of the analyzed volume.

Radiography configuration is implemented only for radiography operations, a small fraction ofall operations. Typically radiography occurs less than 1 day per month. Radiography operations are inherently discontinuous as the purpose of individual operations are met when thcimage is K-State Reactor 11.A-4 Original (12/04)

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I I

J RADIOLOGICAL IMPACT OF M 1 AR AND ' 6N DURING NORMAL OPERATIONS obtained. Typically a day of radiography operations involves less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of operation at full J power. These conservatisms assure DAC and effluent limits are met with no further consideration.

A.2.2 Production of 4 tAr in Rotary Specimen Rack The air volume in the rotary specimen rack does not freely exchange with the air in the reactor bay; there is no motive force for circulation and the rotary specimen rack opening is routinely covered during operation. If the rotary specimen rack were to flood, water would force the air volume in the RSR into the reactor bay. The air volume of the rotary specimen rack (RSR). is approximately2 Vim = 3.75 x 10' cm 3 (HSR p. 28) and the thimnal neutron flux density in the RSR is . = 9.0 x 1012 cm 2 sel at 1,250 kW thermal power. Afler sustained operation at full power, the equilibrium "4Aractivity (Bq) in the RSR volume is given by 0,

V= = 5.2x0'* (2) or, 1.4 Ci in conventional units. If this activity were flushed in to the reactor-bay atmosphere as a result of a water leak into the RSR, the initial activity concentration would be AjA, = 3.5 x 104 pCi ml:'. This would instantaneously be well above even the occupational DAC for 4"Ar.

However, with radioactive decay and ventilation, the concentration would decline in time according to Att) =.A~eql+ (3)

If a worker were exposed to the full course of the decay, cumulative concentration (pCi b ml:')

in the reactor bay would be Vi.*JdI*At)= =1.6xl 04pCih ml (4)

The value 1.6 x I04 pCi s mI' 1, or44Ax 10i0 pCi h mL' , is well below the 3 x 104 pCi h ml:' annual limit of 2000 DAC hours specified in IOCFR20EPA-52011-88-020.

A.2.3 Production of 4t Ar from Coolant Water The reactor tank water surface is open to the reactor bay. Radioactive 4"Ar is circulated in the pool by convection beating, and freely exchanges with the reactor bay atmosphere during normal operation. The 4"Ar activity in the reactor tank water results from irradiation of the air dissolved in the water. The following calculations evaluate the rate at which 4lAr escapes from the water into the reactorbay. The following variables plus those in Table All are used in the calculations of 4"Ar concentrations in the core region, in the reactor tank outside the core, and in the reactor bay air.

2 Approximated as a section of a cylindrical annulus, with 28-in. OD, 24in. ID, and 14-in. height.

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CHAPTER 11 APPENDIX A Y=O,= volume of corc region (clements) = x A x L.

CQo = 40 atomic density (cfm3) in coolant a microscopic cross section (cm2 )

p density (1.0 g cm-3) v volumetric flow rate through core (cm s'l) s = residence time for coolant in core at full power (s)

T out-of-core cycle time for coolant.

Althoughplcments are assumed for thermal hydraulic analysis, the actual core water to non-water ratio is 33%; there'fore, the active region of the core contains 5160 cm3 of water. The saturated concentrations of argon in water at the coolant inlet temperature of 271C is approximately 6.1 x 10 g per cm3 of water (Dorsey 1940). If it is assumed that air is saturated with water vapor above the water tank (27 mm Hg vapor pressure at 270 C) and that the mole fraction of argon in dry air is 0.0094, the partial pressure of argon in air above the tank is 0.0094(760 - 27) - 6.9 mm Hg. By Henry's law, the concentration in water at the inlet temperature is 6.1 x 10F5 x 6.9760 =5.5 x l0 g cm3 (C4o = 83 x 1O0" atoms cm 4 ).

The number of atoms per second of "4Arproduced in the core is C4o0 x VE x 4- 2.25 x 109.

Activity is calculated as the product of the .isotope concentration and the mean lifetime. If it were assumed that all atoms escape to the containment volume, the steady-state activity concentrations in the reactor bay atmosphere would be:

C40 a.O40ah:X Bq,,,

(1 Vj~, (5)

Table A.2, Variables In A r4 Calculations Variable Units Basis C.1 0-. 83 106 cm'3. Calculated above V.c*= 5160 cma 33% of core volume a 40 := .66 104 cm'2 Cross section

.2O5105 cf 2 e1 TableA. -

xt:=.000106 a Radiological decay constant Vq -= A077625909 I00 cm 3 144,000 cu R y :-.000102 Se Bay effluent flow constant The equilibrium 4"Ar concentration during full power steady state operation at 1,250 kW in the reactor bay would be 0.072 Eq cm 4 (1.9 x 10tpCi mL') without ventilation and 0.027 Bq cm3 (7.2 x 10 7 pCi rid') with ventilation.

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RADIOLOGICAL IMPACT OF 41AR AND 18N DURING NORMAL OPERATIONS 1

Environmental Protection Agency, Federal Guidance Report No. 11 (EPA FG 11 - Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion) lists the DAC for 4"Ar as (Table L.b) 3X104 PCi/mi, or 3 pCi/ml. Therefore, equilibrium 4 Ar concentration during full power steady state operation at Q 1,250 kMV is less than DAC and there are no restrictions on activities in the reactor bay imposed by the normal Ar4 ' production mode.

Even with the extremely conservative approximation of 100%/. release of 4"Ar to the atmosphere, the estimated steady state concentration, under ventilation, is less than the DAC.

A.2.4 Maximum Impact of 4"Ar Outside the Operations Boundary Although there are three modes of 4lAr production, only the release of radioactive argon dissolved in water occurs routinely. The 4"Ar produced in the reactor bay during normal operations is released to the atmosphere via an exhaust fan at approximately height h - 11 mctcrs above grade. The flow rate is 4.17 x 105 cm3 se (884 cfm). At the steady state concentration computed in the previous section, the release rate would be Q - 129 pCi se. The maximum downwind concentration (pCi cmrr), at grade, may be computed using the Sutton formula (Slade 1968):

Coo e;ch2Q j C,() (6)

Cy in which u is the mean wind speed (m sl), e = 2.718, and C4 and a are diffusion parameters in the crosswind and vertical directions respectively. The maximum concentration downwind occurs at distance d (m) given by d= (hlC,)-, (7) in which the parameter n is associated with the wind stability condition. In this calculation, we adopt the values of n and 4 use in the McClellan AFB SAR. Mean wind speeds, by stability class are inferred from the data in Chapter 2. Calculations are shown in Table 2.3.

Table A.3, Atmospheric Dispersion Calculations.

Pasquill stability u (m sl) n C. (mu) 4 (m!) d(m) C=1 (pCi cm')

class Extremely0.53.093 unstable (A) 1.6 0.2 031 031 53 0.003903 Slightly 4.0 0.25 0.15 0.15 135 0.001560 stable (E) 3.5 0.33 4C, 0.075 393 0.000445 Extremely 0.77 0.5 84 0.035 2140 0.001013 K-State Reactor 11.A-7 Original (12104)

Safety Analysis Report

CHAPTER 11 APPENDIX A The dose conversion factor provided by EPA FG 11 for "1 Ar (Table 2.3) is 2.17 X 10.10 Sv/hr per Bqlzn3, or (using the provided conversion factor of 3.7 X 1015) 8.03 X 105 mrcni/hr per pCi/cmd, 8.03 X 10.1 mrem/hr per pCi/cm3. Using the highest maximum concentration of Table A3 (0.003903 pCi cm") at steady state full power operation for a full year (8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months ) with observed frequency of class A stability (see Appcndix 2.C) would result in a dose less than I mnrem/year. Frequency of occurrence and the concentration at the maximum dose will occur frorn class C conditions, with a maximum annual dose of 1.7 rnrcm. The maximum concentration at the highest frequency (class G) is 0.001013 pCi cm3, with a dose of3.8 rarem.

The assumed 24-7 operating history is not feasible for the KSU reactor, which has an average operating time for two decades of about 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months s/weck. Additionally, a full power, continuous operation would require a significant quantity of new fuel.

Note that over the full range of conditions examined in Table 2.3, the peak downwind concentration is substantially below the DAC of 3 pCi cK 3 established in 10CFR20 Appendix B, and less than the permissible effluent concentration of I X 104 pCikcm' for all meteorological conditions except the set of conditions with the lowest frequency of occurrence; for that stability classification, the initantaneous effluent concentration is slightly higher than the DAC.

A.2.5 Radiological Assessment of ' 6N Sources Nitrogen-16 is generated by the reaction of fast neutrons with oxygen and the only significant source results from reactions with oxygen in the liquid coolant of the reactor. The nuclide has a half-life of 7.13 s (decay constant )16 - 0.0972 e - 350 If") and enits, predominantly, 6.13-McV gamma rays. According to the McClellan AFB SAR, the cffective cross section for the "0(n,p)l6N reaction, averaged over the fast-neutron energy spectrum in the TRIGA or over the fission-neutron spectrum is afp=2.1 x l 2 'cim2 .

The atomic density C, (cmr3) of the nuclide as it leaves the reactor core is given in terms of the oxygen density in water, Co - 3.34 x 102, as CJ C~r

, *=COf *

'S16 p * (I- a * (8) where time in the core is represented by t. Fast-neutron flux varies linearly with reactor power.

Time in corm is a function of convection flow rate, a function of reactor power (see Chapter 4).

Ai power increases, the rate of production increase from increased neutron flux is mitigated by a reduced time in the core from the increase in core cooling flow rate.

As the warmed coolant leaves the core, it passes through 1.5-in diameter (A, 11= ACcm 2 )

channels in the upper grid plate, but the triflute upper end fixture of the fuel element restricts the flow. This leaves a flow area for each element of:

A0 _ AS *[l- (.) *sin3O *cos3O'] = 6.69cm2 (9)

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RADIOLOGICAL IMPACT OF 4MAR AND "N DURING NORMAL OPERATIONS Operation at power requires primary cooling; primary cooling enters the pool through a flow diverter approximately 2 feet (61 cm) above the core exit, 14 feet (427 cm) below the pool surface. Core exit is at 16 feet (488 cm) below the pool surface. The flow diverter induces mixing

'and avoids the direct rise from the core to the pool surface (which could otherwise occur through a chimney effect from core heating). A rough estimate of hydraulic diameter of the core exit (based on total flow area) is about 13 cm; calculations show the contributions to total dose rates at the pool surface are negligible at 160-200 cm below the surface of the pool, 22-25 times the hydraulic diameter of the exit into the pool. Exit flows are a small fraction of mixing flow, and under these conditions it is considered adequate to use a nuclide concentration reduced by the ratio of the total core exit surface area (approximately 555 cm2 for 83 elements) and (he pool (with a total surface area of approximately 30900 cn 2); mixing reduces the concentration of "N from the core exit by 0.018. Therefore, concentration of the radionuclide used in calculation is reduced from core exit by dilution.

Because of the short Vz life, the concentration of 16N is also reduced by decay during transit.

Since it is difficult to characterize flow velocity field from core exit to total mixing, flow rate from the core to the surface is conservatively assumed as core exit flow rate for dose rate calculations.

Dose rate calculations were modeled as a set of disk sources, each disk containing the appropriate volume source term multiplied times the difference between the disk locations. The appropriate volume source strength for each disk source calculation was modified by exponential decay of 16N, with the time element calculated from core exit surface area, flow rate, and distance form the core exit. Does rate calculations were based on the two major emissions, 6.13 MeV (69%) and 7.1 1 MeV (5%). Total dose rate at each disk (wherex is the distance form the disk to the pool surface) was therefore calculated as:

D1 2 S 2E[ c * .i (E, ,x)- (El(uxi* see D)]

Where:

k(E): MeY

cnI,' *

S =S0 (J7O. V)*o Xp A*-,- t 4 l '1(seeChapter4forcoolantflowrate)

A, Taylor buildupfactor

p, *inearattemiation coefficient, mod ifledbyTaylor buildupfactora,

. =arcla{*J K-State Reactor 1IA-9 Original (12104)

Safety Analysis Report

CHAPTER 11 APPENDIX A Parametric variation on the distance between the disk sources showed little improvement in convergence for separations smaller than 2 cm, and essentially no improvement below I cm; therefore '%cm was used for final calculations. Locations of interest for dose calculations include 30 cm (I it) abovc the pool surfacc (i.e., pool surface monitor), waist high (approximately 130 cmI51 in. above the pool, 100 cm139 in. above the bridge), and at the ceiling over the pool.(549 cm/18 t above the pool).

Table A.4, Dose Rate (mR t;W) Above Pool KCW 30 cm 130 cm 549 cm 50 0.5 0.2 0.0 100 3.5 1.1 1.3 200 15.8 4.7 0.5 300 35.7 10.5 1.2 400 60.0 17.5 1.9 500 87.2 253 2.8 750 1663 47.7 52 1000 2553 72.8 7.9 1250 347.9 98.8 10.6 Only a small proportion of the "1Natoms present near the lank surface are actually transferred to the air of the reactor bay. Upon its formation, the 16N'rccoil atom has various degrees of ionization. According to Mittd and Thcys (1961) practically all 16N combines with oxygen and hydrogen atoms in high purity water, and most combines in an anion form, which has a tendency to remain in the water. In this consideration, and in consideration of the very short half lifc of the nuclide, the occupational consequences of any airborne 1"N arc deemed negligible in comparison to consequences from the shine from the reactor tank. Similarly, off-sitc radiological consequences from airborne 1N are deemed negligible in comparison to those of41 Ar.

A.3 Bibliography EPA Federal Guidance Report 11, "Limiting Values ofRadionuclide Intake andAir Concentration, andDoseConversion FactorsforInhalation, Submersion, andIngestion,' U.S.

Environmental Protection Agency, Report EPA-520/1-88-020,1988.

KansasState University TRIGA MarkilReactorHazardsSummaryReport, by RAY. Clack, JR.

Fagan, WV.R. Kimel, and S.Z Milchail, License R-88, Docket 50-188, 1961.

"Analog is ofCertainHazardsAssociatedwith Operationof the KansasState University TMIGA Marki l Reactor at250 klYSteady State and wvith PlsedOperationto $2.00, " by R.W. Clack, et al], and the Safety Evaluation by the U.S. Atomic Energy Commission Division of Reactor Licensing, License R-88, Docket S0-188, 1968.

OperationsManial, KSU TRIGA MarklHNuclearReactorFacJlity, License R-88, Docket 50-188.

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Safety Analysis Report

'-U1 RADIOLOGICAL IMPACT OF 41AR AND '6 N DURING NORMAL OPERATIONS FacilitySafeyAnalysIs Report, Rev. 2, McClellan Nuclear Radiation Center Reactor, April 1998.

Milil, RL. and M.H. Theys, '"-16Concentrationsin EBIYR," Nucleonics, March 1961, p. 81.

Slade, D.H. (ed.), 'Meteorology and Atomic Energy," Report TID-24190, US. Atomic Energy Commission, 1968).

.1 I-11

%-S

%I-

\IS INS IS IS

'S\

IS K-State Reactor 11A-11 Original (12104) IS Safety Analysis Report NS

SAFETY ANALYSIS REPORT This page intentionally blank K-State Reactor

12. CONDUCT OF OPERATIONS This chapter describes the conduct of operations at the KSU TRIGA Mark II Nuclear Reactor Facility. The conduct of operations involves the administrative aspects of facility operations, the facility emergency plan, the physical security plan, and the requalification plan. This chapter of the Safety Analysis Report forms the basis of Section 6 of the Technical Specifications (Chapter 14).

12.1 Organization The operating license R-88, Docket 50-188, for the reactor is held by Kansas State University.

Kansas State University is a land-grant institution governed by a Board of Regents (appointed by the Governor of the State). The Chief Executive Officer of the university is the President. The organizational structure (as shown in Fig. 12.1) identifies the President (a representative of the State of Kansas via the Board of Regents) as the licensee for the KSU Nuclear Reactor Facility.

A University Provost administers academic instruction and research for the University.

Individual colleges manage these functions, with the College of Engineering responsible for the Department of Mechanical and Nuclear Engineering. The Department of Mechanical and Nuclear Engineering is directly responsible for management of the reactor.

The Department of Nuclear Engineering appoints a Nuclear Reactor Facility Manager (unclassified appointment, equivalent to a faculty appointment) for direct management of the reactor. The Nuclear Reactor Facility Manager delegates a Reactor Supervisor with the responsibility for direct supervision and coordination of daily operations (the Nuclear Reactor Facility Manager may hold these functions). The Nuclear Reactor Facility Manager and the Reactor Supervisor hold Senior Reactor Operator licenses issued by the USNRC. Additional licensed Reactor Operators (or Senior Reactor Operators) perform operations and maintenance functions under supervision of the Reactor Supervisor.

The Vice President for Administration and Finance is responsible for safety at the university.

Kansas State University provides management and independent environment, safety and health oversight functions for the University, implemented though the Division of Public Safety. Safety functions arc administered by two sections of the Division of Public Safety. the University Police Department and the Department of Environmental Health and Safety.

The University Police Department is responsible for law enforcement functions and institutional physical security. The University Police Department provides support for response to events.

University Police Department is the primary interface for external agencies during response to events at the reactor facility.

The Department of Environmental Health is responsible for compliance issues related to environment, safety and health. To meet these requirements, the Division maintains a Radiation Safety Officer, Occupational Safety Manager, Hazardous Material Manager, and (fire) Safety and Security Officer. A partial listing of germane Department of Public Safety management functions include (but are not limited to):

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CONDUCT OF OPERATIONS

Environmental rnanagement (air, water, & waste)

Ionizing & non-ioniting radiation

Sanitation and water quality

Fire prevention and emergency equipment

Accident prevention and investigation

Occupational safety and health

Industrial hygiene and toxicology

Indoor air quality

Asbestos & PCB's (TSCA)

Laboratory safety & chemical hygiene A Reactor Safeguards Committee (composed of members of the Mechanical and Nuclear Engineering Department and other }4-State faculty appointed by the President, the MNE chair, the University Radiation Safety Officer, and the Reactor Supervisor) performs the review and audit of nuclear operations for the President. The committee meets at frequencies specified in Technical Specifications. The Committee reports to the President, but also advises the Nuclear Reactor Facility Manager and the Head of the Department of Mechanical and Nuclear Engineering, Responsibility for facility operations therefore extends from the government of the State of Kansas through r the Board of Regents, the President of Kansas State University, the Provost of Kansas State University, the Dean of the college of Engineering, to the operating unit (Department of Mechanical and Nuclear Engineerin~g) and the reactor stat}, including the Nuclear Reactor Facility Manager, the Reactor Supervisor and Reactor Operators.

12.1.1 Structure As indicatedon Figure 12.3, Organizationstmc/zireforthe KSUTJGA Mark)l Nuckear.Reactor Facility, the X-State President is the licensee for the KSU Nuclear Reactor Facility. The reactor is under the direct control of the Nuclear Reactor Facility Manager, who reports through the academic administrative structure to the President.

Envimrnent, safety and health oversight and expertise is provided through the Vice President for Administration and Finance, independent of facility line management. In addition to the reactor license, Kansas State University administers a broad radioactive material license. The Radiation Safety Officer is the University broad licensee, and manages radioactive material (i.e., byproduct and non-reactor special nuclear material) inventory and the University radiation safety program for ionizing radiation.'

A University Radiation Safety Committee (reporting to the Vice President for Administration and Finance) maintains oversight and control of radiation protection functions for the University.

Radiological controls for possession and use of radioactive materials at K-State in the University Radiation Protection Program are prepared and distributed by the University Radiation Safety Committee. The University Radiation Safety Committee has authorized the Nuclear Reactor Facility Manager to possess and transfer radioactive material under the State broad license.-

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di CONDUCT OF OPERATIONS

\1

1 I.-

Figure 12.1: Organization and Management Structure for the K-State Reactor U'nivcrsity requirements and 10CFR requirements are combined in a comprehensive Reactor Radiation Protection Program. In accordance with the Reactor Radiation Protection Program, the reactor staff fil1fills most routine radiation protection functions at the XC-State reactor, with review and oversight by the Radiation Safety Officer. The Radiation Safety Officer manages the radiation worker exposure monitoring system (and distribution of related records), as wvell as radioactive material inventory control.

K-State Reactor 12-3 Original (12104) I Safety Analysis Report

CONDUCT OF OPERATIONS 12.1.2 Responsibility A description of responsibilities is provided in three categories. Responsibility forsafe operation of the reactor is described in Section 1. Independent environment, safety and health compliance and oversight is described in Section II. Advisory and oversight committees are described in Section III.

a. Reictor Operations Line Management President:

As chief executive officer for the University, the President is responsible for safe operation of the reactor, protection of the health and safety of the public, and protection of the environment. The line of authority and responsibility for reactor operations extends through thc Provost and Dean or Eneineerinf to the Head of the Department of Mechanical and Nuclear Engineering. Environment, safety and health compliance management and indepenidcnt oversight functions. are distributed through the Vice President of Administration and Finance to the Malnnager of the Division of Public Safety. Department of Environmental Snfety and Health.

Head of the Department of Meehanical and Nuclear Engineering The Dcpartment Head is the appointment authority for the Nuclear Reactor Facility Manager, Reactor Supervisor, and all Reactor Operators and Senior Reactor Operators.

The Department Head is responsible for providing resources required for safe operations of the reactor facilities. The Department Head is the Chair of the Reactor Safeguards Committee, which reports to the President, and which is responsible for approval of all plans and procedures for reactor operations and for audit of reactor operations and record keeping.

Nuclear Reactor Facility Manager The Nuclear Reactor Facility Manager, who may also serve as Reactor Supervisor, is directly responsible to the Head of the Department of Mechanical and Nuclear Engineering for all aspects of facility operation. The Manager may hold academic and research responsibilities beyond those associated with the Reactor Facility. Thc Nuclear Reactor Facility Manager is authorized to delegate responsibility for operation and use of the reactor to the Reactor Supervisor.

Reactor Supervisor The Reactor Supervisor has such duties, in regard to the operation of the reactor, as may be delegated by the Nuclear Reactor Facility Manager but whose noriiinal duties include reactor scheduling, responsibility for all irtords-with-rigard to'reactor-op'eration as 'are required by appropriate federal licenses and regulations, laws and regulations of the State of Kansas and regulations of Kansas State University including the Kansas State University TRIGA MARK 11 Reactor Operations Manual.

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CONDUCT OF OPERATIONS The Reactor Supervisor is responsible for assuring that the reactor is operated only while a properly qualified and licensed Reactor Operator is present. The Reactor Supervisor is responsible for maintenance of a Reactor Operations Manual, the manual to include prescribed operating procedures for all routine modes of operation of the reactor, procedures for loading and unloading, start-up procedures, maintenance schedule, testing procedures, operational references, and other appropriate information as determined by the Reactor Supervisor and the Nuclear Reactor Facility Manager. The Reactor Supervisor is responsible for determining that the reactor is operated in strict accordance with the Operations Manual and the Facility.License.

Reactor Operator Operators (Reactor Operators and Senior Reactor Operators) report directly to the Reactor Supervisor and/or the Nuclear Reactor Facility Manager. Operators are responsible for knowing the status and condition of the facility, and ensuring that both personnel within the facility and the general public are protected from exposure to radiation consistent with approved policies and procedures. Operators are responsible for operation of the reactor in accordance with Technical Specifications, operating procedures sand experiment procedures.

Operators are responsible for ensuring only authorized personnel (trainees for senior operator and operator positions, as well as students enrolled in academic courses making use-of the reactor, as permitted by IOCFR5S) manipulate controls under the direction of a licensed reactor operator or senior operator. Operation includes start-up, shutdown, routine instrumentation and control checkout, record keeping, routine maintenance and such other duties as may be described in the Operations Manual and/or as directed by the Reactor Supervisor.

During fuel movement, a reactor operator must be at the reactor operating console, and a senior operator inside the reactor bay directing fuel operations.

b. Environment, Safety, and Health Staff Vice President for.Administration and Finance The Vice President for Administration and Finance is responsible for safety at the university. Kansas State University provides management and independent cnvirounent, safety and health oversight functions for the University. This responsibility is implemented though the Department of Environmental Safety and Health, Division of Public Safety. Safety functions are administered by two sections of the Division of Public Safety, the University Police Department and the Department of Environmental Health and Safety.

University Police Dcpartment The -University Police Department is responsible for law enforcement functions and institutional physical security. The University Police Department is the primary interface for external agencies during response to events at the reactor facility.

K-State Reactor 12-5 Original (12/04)

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. S.

CONDUCT OF OPERATIONS Ma nager, Department of Environmental Health and Safety The Department of Environmental Health and Safety is responsible for compliance issues related to environment, safety and health. To meet these requirements, the Department has positions for a Radiation Safety Officer, Occupational Safety Manager, Hazardous Material Manager, and Safety and Security Officer. A partial listing of germane Department ofPublic Safety management functions include (but are not linited to):

Environmental management (air, water, & waste)

Ionizing & non-ionizing radiation

Sanitation and water quality

Fireprevention and emergency equipment Accident prevention and investigation Occupational safety and health

Industrial hygiene and toxicology

Indoorairquality

Asbestos & PCBs (TSCA)

Laboratory safety & chemical hygiene Radiation Safety Officer The Radiation Safety Officer reports to the Manager of the Department of Environmental Health and Safety. The Radiation Safety Officer, or an authorized representative, shall be available (upon due notice) for advice and consultation regarding radiation surveys and radiation safety in connection with isotope production and radiation streaming problems as might arise in connection with reactor operation or experimentation. 'he Radiation Safety Officer is ex oficlo a member of the Kansas State University Radiation Safety Committee. The Radiation Safety Officer serves cc officio as a member of the Reactor Safeguards Committee,-with any action (i.e., concerning potential radiation exposure or radioactive effluents) of the Committee requiring approval of the Radiation Safety Officer.

c. Principal Advisory and Oversight Committees Reactor Safeguards Committee The Reactor Safeguards Committee is composed of members appointed by the President of the university, upon the recommendation of the Chairman of the Committee, and ex-officio for specific positions. Composition and membership qualifications of the Conmnittee are explicitly stated in Technical Specifications. The Reactor Safeguards Committee is responsible for approval of all plans and procedures for reactor operations and for audit of reactor operations and record keeping.--

K-State Reactor 12-6 Original (12104)

Safety AnaWlsis Report

CONDUCT OF OPERATIONS University Radiation Safety Committee fi The University Radiation Safety Committee is an advisory committee for the Vice J President for Administration and Finance and the Radiation safety Officer. Under JJ authority of the Vice President of Administration and Finance, the Radiation Safety Committee authorizes conditions for use of radioactive material at K-State, and authorizes users (by name) to acquire and posses radioactive materials. The Reactor Radiation Protection Program incorporates requirements of the Radiation Safety J Committee.

J 12.1.3 Staffing J Whenever the reactor is not secured, the reactor shall be under the direction of a (USNRC sJ licensed) Senior Operator wvbo is designated as Reactor Supervisor. The Supervisor shall be on call, within twenty minutes travel time to the facility, and cognizant of reactor operations.

Whenever the reactor is not secured, a (USNRC licensed) Reactor Operator (or Senior Reactor Operator) who meets requirements of the Operator Requalification Program shall be at the reactor control console, and directly responsible for control manipulations. A call list of Reactor Facility Personnel, management, and radiation safety personnel shall be available in the reactor control room for use by the Reactor Operator at the controls.

During fuel movement, a reactor operator shall be at the reactor operating console, and a: senior operator inside the reactor bay directing fuel operations.

Only the Reactor Operator at the controls or personnel authorized by, and under direct supervision of, the Reactor Operator at the controls shall manipulate the controls. Whenever the reactor is not secured, operation of equipment that has the potential to affect reactivity or power level shall be manipulated only with the knowledge and consent of the Reactor Operator at the controls. The Reactor Operator at the controls may authorize persons to manipulate reactivity controls who are training either as (1) a student enrolled in acadcrnic course making use of the reactor, (2) to qualify for an operator license, or (3) in accordance the approved Reactor Operator requalification program.

12.1.4 Selection and Training of Personnel The KSU Reactor Facility maintains a training and selection program to prepare trainees for examination by the Nuclear Regulatory Commission in pursuit of operator or senior operator permits. Medical qualification for operator license program is described in Section 12.10.

Access is permitted to the reactor baj for personnel qualified for unescorted access or under the direct supervision of personnel qualified for unescorted access. Unescorted access qualification is granted after training and examination in knowledge and skills necessary to control personnel exposure to radiation associated with'th&'operation' of the KSU nuclear reactor. Training includes familiarization with salient sections of 10 CFR 19 Notices, Instructions and Reports to Workers, and Investigations, 10 CFR 20 Standards for Protection Against Radiation, the KSU Radiological Protection Program, and the KSU Reactor Emergency Plan. The KSU Radiological K-State Reactor 12-7 Original (12104).

Safety Analysis Report

CONDUCT OF OPERATIONS Protection Program requires specific instruction in the risks of occupational exposure, the nisks of prenatal exposure, provisos of IOCFRI9 and IOCFR20, and a tour of the reactor facility.

12.1.5 Radiation Safety The Radiation Protection Program for the Kansas State University TRIGA 14k 11 Nuclear Reactor Facility was prepared in response to the requirements of Title 10, Part 20.1101 (Code of Federal Regulations, IOCFR20). The Radiation Protection Program was developed following the guidance orthe American National Standard Radiafion P;otection at Research Reactor Facilities and Regulatory Guides issued by the NRC. The Program also deals with radioactive materials regulated by the State of Kansas (an Agreement state) under license 38-COI 1-01. The goal of the Program is the limitation of radiation exposures and radioactivity releases to a level that is as low as reasonably achievable without seriously restricting operation of the Facility for purposes of education and research. The Program is executed in coordination with the Kansas State University Department of Environmental Safety and Health, Division of Public Safety, Radiation Safety Office. The program was been reviewed and approved by the Reactor Safeguards Committee for the Reactor Facility; the University Radiation Safety Committee has independently reviewed the program.

The reactor staff fulfills most of the functions described in the program. The Radiation Safety Officer maintains oversight of the facility by direct observation and review of specific activities, and by acting as a member of the Reactor Safeguards Conmnittce and the University Radiation Safety Committee. Radiation Safety Officer approval is required for Reactor Safeguards Committee approval of any item under review. The Radiation Safety Officer is empowered to stop, in the interest of safety, any experiment involving radiation on the K-State campus. All campus radioactive materials users are required to eliminate any known unsafe practice or report the issue to the Radiation Safety Officer.

The details of the radiation safety program are described in Chapter I1.

12.2 Review and Audit Activities Review and audit activities are oversight actions essential to the safe operation of the facility and the protection of the health and safety of the public. The Technical Specifications, Emergency -

Plan, Radiation Protection Program and the Reactor Administrative Plan require a set of internal surveillances, reviews and audits conducted by the reactor staff and Nuclear Reactor Facility Manager, culminating in a semi-annual management audit of operations.

The Reactor Safeguards Committee holds oversight responsibility and authority. Oversight of the Manager's performance and review of managerial audits is the responsibility of the Safeguards Committee, evaluated formally at periodic intervals. In addition to periodic, scheduled reviews, the Reactor Safeguards Committee is available to conduct reviewvs on request Review and approval of administrative controls, such as programs plans and procedures, is performed by the Reactor Safeguards Committee prior to implementation.

The Radiation Safety Officer conducts periodic laboratory safety audits, and participates in radiation surveillances on a periodic basis to review facility staff conduct of radiation surveys.

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CONDUCT OF OPERATIONS 12.2.1 Reactor Safeguards Committee Composition and Qualiflc ftions VI With the exception of ea-officio members, Reactor Safeguards Committee members are appointed by the President of the University, upon the recommendation of the Chairman of the Committee.

Composition and membership qualifications of the Committee (as specified in the proposed Technical Specifications) provide expertise to evaluate reactor management, plant facilities, experimental programs, operating and experiment procedures, and radiological hazards.

The Head of the Department of Mechanical and Nuclear Engineering is the Committee Chair, and has the authority and responsibility tb allocate resources that ensure safe reactor operations. The University Radiation Safety-Officer is also an ex officio member of the Committee, with veto power over permissive Committee decisions. The Nuclear Reactor Facility Manager is also an ex officio member, non-voting, of the Committee. These ex officio Committee members (or designated alternates) are required to attend all meetings where permissive Committcc decisions are made.

At least one Committee member shall be a Mechanical and Nuclear Engineering faculty member with expertise in reactor physics, nuclear engineering or nuclear science. At least one Committee member shall have expertise in chemistry, geology, or chemical engineering. At least one-Committee member shall have expertise in the biological effects of radiation. One individual may have and represent expertise in more than one area, but the Committee shall consist of at least seven members.

a. Charter and Rules The Committee is required to meet semi-annually, as a minimum. Specific review and audit activities are prescribed for these meetings, described in Sections 1223 and 12.2A. The Chair of the Commnittee or his designee may call additional meetings. At the discretion of the Chair or his designee, the Committee may be polled in lieu of a meeting; such a poll _

shall constitute Committee action subject to the same requirements as for an actual meeting.

Any permissivc action of the Committee requires affirmative vote of the University Radiation Safety Officer as welt as a majority vote of the members present A quorum consists of not less than a majority of the full Committee, .including a offcio voting members.

Minutes of meetings of the Reactor Safeguards Committee are distributed to the Dean of Engineering, the Provost, and the President of the University. Recorded affirmative votes on proposed new or revised experiments or procedures shall indicate that the Committee determines that proposed actions do not involve unreviewed safety questions, changes in the facility as designed, or changes in Technical Specifications, and could be taken without endangering the health and safety of workers or the public or constituting a significant hazard to the integrity of the reactor core.

b. Review Function The responsibilities of the Reactor Safeguards Committee shall include but are not -

limited to the following:

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CONDUCT OF OPERATIONS Review of proposed new or revised experiments.

Review of proposed new or revised procedures.

Review of proposed modifications of the reactor, the reactor bay, or the reactor control room.

Determination of whether items I through 3 involve unreviewed safety questions, changes in the facility as designed, or changes in Technical Specifications. The Nuclear Reactor Facility Manager may make this determination in the form of verifying an evaluation/determination.

Review ofproposed revisions to Technical Specifications Review of proposed changes to the Safcty Analysis Report Review and approval of audits of the Radiation Protection Program, the Physical Security Plan, and the Emergency Plan performed by the Nuclear Reactor Facility Manager.

Review of all operating anomalies and equipment failures.

Review of all reportable events.

Review of results of NRC inspections.

Review of critiques 6f emergency exercises.

Requalification of the Nuclear Reactor Facility Manager or Reactor Supervisor.'

c. Audit Function The Reactor Safeguards Committee shall audit reactor operations and health physics during semi-annual inspections. The inspections shall include but are not limited to the following:

Inspection of reactor operating records Inspection of maintenance activity records Inspection of health physics records.

Review of the effectiveness of trainiug and requalification activities.

Review of radiological surveillance records Inspection ofthe reactor facility.

12.3 Procedures YVritten procedures shall be prepared and approved prior to inaitiangiiiany of the activities listed in this section. The Nuclear Reactor Facility Manager and the Reactor Safeguards Committee shall t It is the responsibility of the Safeguards Committee to reach a decision on the requalification of the licensed person administering examinations to other operators and senior operators.

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CONDUCT OF OPERATIONS v approve the procedures. Under conditions specified by the Reactor Safeguards Committee, the Nuclear Reactor Facility Manager may make changes in procedures or experiments subject to validation by the Reactor Safeguards Committee. A periodic review of procedures will be performed and documented in a timely manner to assure they are current. Procedures shall be adequate to assure the safe operation of the reactor, but will not preclude the use of independent judgment and action, should the situation require. The following are actions that will typically require reviewed written procedures.

12.3.1 Reactor Operations

1. Startup, operation, and shutdown of the reactor 2.- Fuel loading, unloading, and movement within the reactor.

3. Control rod removal or replacement.

4. Routine maintenance, testing, and calibration of control rod drives and other systems that could have an effect on reactor safety.

5. Administrative controls for operations, maintenance, conduct of experiments, and conduct of tours of the Reactor Facility.

6; Implementing procedures for the Emergency Plan or Physical Security Plan.

12.3.2 Health Physics 1 * ..Testing and calibration of area radiation monitors, facility air monitors, and fixed and portable radiological surveillance instruments.

2. Conduct of radiological surveillance measurements. ..

3. Release of contaminated materials to the University Radiation Safety Office.

4. Accountability for special nuclear materials.

12.4 Required Actions Two categories of required actions are addressed, violations of facility safety limits and reportable events.

12.4.1 Violation of Facility Safety Limit In the event that a Safety Limit is not met,

a. The reactorshall be shutdown, and reactor operations secured.

b. Tbe Reactor Supervisor and Nuclear Reactor Facility Managei shall be notified.

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CONDUCT OF OPERATIONS

c. The safcty limit violation shall be reported to the Nuclear Regulatory Cormrnission within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone, confirmed via written statement by email, fax or telegraph

d. A safety limit violation report shall be prepared within 14 days of the event to describe:

(1) Applicable circumstances leading to the violation including (where known) cause and contributing factors (2) Effect of the violation on reactor facility components, systems, and structures (3) Effect of the violation on the health and safety of the personnel and the public (4) Corrective action taken to prevent recurrcncc

e. The Reactor Safety Review Committee shall review the report and any followup reports

f. The report and any followup reports shall be submitted to the NuclearRegulatory Commission.

g. Operations shall not resume until the USNRC Manager of the Division ofReactor Licensing approves resumption.

12.4.2 Occurrences Reportable to the U.S. Nuclear Regulatory Committee In the event of a reportable occurrence, as defined in the Technical Specifications, and in addition to the reporting requirements,

a. lheReactor Supervisorand NuclearReactorFacilityManagershall be notified

b. If a reactor shutdown is required, resumption of normal operations shall be authorized by the Nuclear Reactor Facility Manager

c. Thc event shall be reviewed by the Reactor Safcgunrds during a normally scheduled meeting 12.5 . Reports to the Nuclear Regulatory Commission All ivritten reports shall be sent within prescribed intervals to the United States Nuclear Regulatory Commission, Washington, D.C, 20555, Attn: Document Control Desk.

All reports shall address (to the extent known or possible) the impact of the event on safety and health of the public, workers and the facility (e g., whether or not the event resulted in property damage, personal injury or exposure). Reports (including initial reports, to the extent possible) shall describe, analyze, and evaluate safety implications, and ouitline the corrective measures taken or planned to prevent recurrence of the event.

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CONDUCT OF OPERATIONS 12.5.1 Immediate Notification A report shall be made within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of discovery of a violation of safety limit or reportable occurrence by (1) telephone and (2) fax, telegraph or electronic mail to the NRC Operation Center.

12.5.2 14-Day Notification A report shall be made within 14 days in writing to the NRC Operation Center for any violation of safety limit or reportable occurrence 12.5.3 Thirty-Day Notification A report shall be made within 30 days in writing to the Manager, Non-Power Reactors and Dccommissioning Project Manager, US. Nuclear Regulatory Commission, Washington, D.C, for a) Any permanent changes in Nuclear Reactor Facility Manager or Head of the Department of Mechanical and Nuclear Engineering b) Any significant variation ofmcasured values from a corresponding predicted or previously measured value of safcty-connected operating characteristics occurring during operation of the reactor; b) Any significant change in the transient or accident analysis as descnibed in the Safety Analysis Report. .

12.5.4 Other Reports A report is reiquired within 60 days after criticality of the reactor in writing to the NRC Operation Center resulting from a receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level or the installation of a new core, describing the measured values of the operating conditions or characteristics of the reactor under the new conditions.

A routine report is required to the US. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, within 60 days after completion of the first calendar year of operating and at intervals not to exceed 12 months, thereafter, providing the following Information:

a) A narrative summary of reactor operating experience b) The energy generated by the reactor (in megawatt-hours);

c) Unscheduled shutdowns, including corrective action taken to prevent recurrence d) Major preventative and corrective maintenance with safety significance e) Major changes in the reactor facility, including a summary of safety evaluations leading to the conclusions that no Unreviewed safety questions were involved K-State Reactor 12-13 Original (12104)

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CONDUCT OF OPERATIONS i) Major changes in procedures, including a summary of safety evaluations lading to the conclusions that no Unreviewed safety questions were involved g) New lests or experiments (or both) that are significantly different from those previously performed and are not described in the Safety Analysis Report, including a surarnary of safety evaluations leading to the conclusions that no unrcvicwvcd safety questions were involved h) A summary of the nature and amount of radioactive cffluents released or discharged to environs beyond the effcctive control of Kansas State University determined at or before the point of release or discharge.

The summary shall include, to the extent practicable, an estimate of individual radionuclides present in the effluent. If the estimated average release is less than 25% of the allowed or recommended value, a statement to this effect is sufficient.

i) A summarized result of envirorunental surveys performed outside the facility j) A summary of radiation exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed orrecommcnded.

12.6 Record Retention

'here are three categories of record retention. General operating records (as noted) are required to be kept for five years. Records related to requalification arc kept for the duration of the individuals' employment or for a complete training cycle. Records related to radiation (releases or exposure) are kept for the life of the facility.

12.6.1 Five-Year Retention Schedule In addition to the requirements of applicable Code of Federil regulations (Title I0; Parts 20 and 50), records and logs shall be prepared and retained for a period of at least 5 years for the following items as a minimum.

a) Normal plant operation, including power levels;

b) Principal maintenance activities; c) Reportable occurences; d) Equipment and component surveillance activities; e) Experiments performed with the reactor; f) All ernergency reactor scrams, including reasons for emergency shutdowns.

12.6.2 Certification Cycle Records of retraining and requalification of certified operations personnel shall be maintained at all times the individual is employed or until the certification is renewed.

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CONDUCT OF OPERATIONS 12.6.3 Life-of-the-Facility Records The following records shall be maintained for the life of the facility:

a) Gaseous and liquid radioactive effluents released to the environs b) OfMsite environmental monitoring surveys required by Technical Specifications c) Fuel inventories and transfers d) Facility radiation and contamination.surveys 1/4 e) Radiation exposures for all personnel monitored f) Corrected and as-built facility drawings 12.7 Emergency Planning An emergency plan shall be established and followed ia accordance with NRC regulations. The plan shall be reviewed and approved by the Reactor Safeguards Committee prior to its submission to the NRC. In addition, emergency procedures that have been reviewed and approved by the Reactor Safeguards Committee shall be established to coverall foreseeable emergency conditions potentially hazardous to persons within the Laboratory or to the public, including, but not limited to, those involving an uncontrolled reactor excursion or an uncontrolled release of radioactivity; 12.8 Security.Planning Administrative controls for protection of the reactor plant shall be established and followed in accordance with NRC regulations. 1/4 12.9 Operator Training and Requalification The ICSU Reactor Facility maintains a training and selection program to prepare trainees for.

examination by the Nuclear Regulatory Commission in pursuit of (senior) operator permits.

Examinations are based on those of the Nuclear Regulatory Commission and include both written and practical tests. In preparation, trainees must satisfactorily complete study for the following areas:

1. Theory and operating principles

2. Operating characteristics

3. Instrumentation and control x

4. Protection systems

5. Operating and emergency procedures

6. Radiation control and safety

7. Technical specifications

8. Title 10, Code of Federal Regulations- -

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CONDUCT OF OPERATIONS 12.9.1 Requalification Program The proposed Requalification Program follows a two-year cycle as of 1 Jan 1974. The program provides for operator medical certification, on the job training elements and proficiency, lectures, examinations,'and records. The proposed Program identifies periodic and special requirements associated with medical certification, maintaining operational proficiency, operator examinations, training lectures, and records.

a. Medical Certification The USNRC licenses operators based on physician evaluation and facility management certification that the licensee's medical condition and general health will not adversely affect the performance of assigned operator job duties or cause operational errors endangering public health and safety.

The Nuclear Reactor Facility Manager has primary responsibility to assure medically qualified personnel are on-duty. Medical qualification of the Nuclear Reactor Facility Manager, if licensed, is the responsibility of the Chair of the Reactor Safeguards Committec.

The proposed Requalification Program identifies requirements to maintain medical certification that the licensed operator is medically qualified to operate the reactor, including annual reexamination and notifications of significant changes (should they occur).

b. Proficiency During each tiwo-year cycle, each licensed operator will maintain proficiency in reactivity manipulations by performing manipulations that demonstrate skill with reactivity control systems, as specified in the Requalification Program. Changes in facility design, operating procedures, facility license, and abnormal or emergency procedures will be documented and distributed to ensure all licensed operators are cognizant of facility conditioni and requirements.

c. Examinations The proposed Requalification Program specifies two sets of annual examinations, written exams and operating exams. The program specifies that examinations should be based on a representative sample of questions covering areas in depth required to evaluate trainee understanding and capabilities. The program specifies that examinations should be based on evaluating knowledge, skills, and ability required to perform as a reactor operator/senior reactor operator, as appropriate.

Requirements for attending formal training lectures will be determined based on the results of annual written examinations administered to all licensed personnel, according to criteria specified in the Requalification Plan. TIe examinations will be prepared ahd graded by the Nuclear Reactor Facility Manager or the Reactor Supervisor.

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CONDUCT OF OPERATIONS Requirements for additional training will be determined based on the results of annual operational examinations covering normal, abnormal and emergency operating procedures (according to criteria specified in the Requalification Plan). The Nuclear Y Reactor Facility Manager or the Reactor Supervisor will prepare and administer the operational examinations, as specified in the Requalification Program.

C. Lectures The proposed Requalification Program provides guidance for preparing training material based on objectives based onoperational needs, i.e., objectives based on how the material relates to job performance.

d. Records Records demonstrating successful participation the Reunification Program will be maintained as specified in the Program, which includes operator training record folders, records of Reactor Safeguards Committee reviews, operating logs, annual training records, and retention of operator biennial requalification records for one completed cycle.

12.10 Medical Certification of Licensed Operators and Senior Operators The primary responsibility for assuring that medically qualified personnel are on-duty rests with the Nuclear Reactor Facility Manager.- Medical qualification of the Nuclear Reactor Facility Manager, if licensed, is the responsibility of the Chair of the Reactor Safeguards Committee.

Licensed personnel should be examined biennially for continued medical qualification. Medical Examination Report Forms are supplied by the Facility and must be signed by the examining physician, with the physician's license number noted on the form. .

The Nuclear Reactor Facility Manager reviews the medical report forms and makes a determination of whether the licensee (a) should be denied a license on medical grounds, (b) should have no license restrictions on medical grounds, or (c) should have certain specified restrictions. This information is entered on a Facility Medical Examiner Review Form and signed by the Nuclear Reaetor Facility Manager. The Chair of the Reactor Safeguards Committee must approve the recommendation of the Manager, with signature on the same form.

The approved recommendations of the Nuclear Reactor Facility Manager are submitted to the Nuclear Regulatoty Commission using NRC Form 396.

12.11 Bibliography ANSJ.ANS-15.1, "Development of Technical Specificationsfor.ResearchReactors, American National Standards Institute/American Nuclear Socrety, La Grange Park, Illinois, 1990.

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CONDUCT OF OPERATIONS ANSIJANS-15.1I (FinalDraft). "AmericanNationalStandardRadiationProtectionatResearch Facilities," American Nuclear Society, La Grange Park, Illinois, October, 1992.

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13. ACCIDENT ANALYSIS * $x1 This chapter provides information and analysis to demonstrate that the health and safety.of the public and workers are protected in the event of equipment malfunctions or other abnormalities in reactor behavior. The analysis demonstrates that facility design features, limiting safety system settings, and limiting conditions for operation ensure that no credible accident could lead to unacceptable radiological consequences to people or the environment.

13.1 Accident Initiating Events and Scenarios This chapter deals with analysis of abnormal operating conditions and consequent effects on safety to the reactor, the public, and operations personnel. Three conditions to be analyzed are:

Loss of coolant

Insertion of excess reactivity

Fuel encapsulation failure - the maximum hypothetical accident (MHA)

These are the three conditions considered in the initial licensing of the Reactor Facility in 1962 for 100-kW steady-state operation and in the 1968 upgrade of the license permitting 250-kW steady state operation and 250-MV pulsing operation. The analysis presented here treats the same conditions, but for steady-state operation at 1,250 kW and pulsing operation to a S3.00 reactivity insertion, estimated peak power of 1,340 Mw.

The maximum hypothetical accident for a TRIGA reactor is the failure of the encapsulation of one fuel element, in air, resulting in the release of gaseous fission products to the atmosphere.-

Failure in air could result from a fuel-handling accident or, possibly, failure in the event of a loss of reactor coolant. Failure under water, leading ultimately to atmospheric release of fission products, could possibly result from insertion of excess reactivity or operation with damaged fuel. -

This chapter addresses the several scenarios potentially leading to fuel failure, and then the potential consequences, should failure occur in air.

13.2 Accident Analysis and Determination of Consequences 13.2.1 Notation and Fuel Properties Tables 13.1-133 identify physical characteristics of the TRIGA Mark II fueL Table 13.4 identifies the assumptions and design basis values used in the accident analyses.

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CHAPTER 13 Table 13.1, Dimensions of TRIGA MkIl ZrHi,6 Fuel Elements.

Property ofIndividual Element Symbol Value Length of fuel zone i1 0381 m .

Fuel radius r 0.018161 m Clad outside radius r. 0.018669 m Fuel volume l . 0.000417 in Clad volume VI 0.0000224 m3 Fuel mass Uf 2.5014 kg Clad mass M0.1845kg Wt. Fraction U in fuel x WI Fraction ZrH1.6in fuel xw Source: Trainini ManuaL KSU TRIGANuclearReactor Facility. 1998.

Table 132. Neutronic Properties of TRIGA Mkld ZrHIA Fuel Elements.

Property Symbol Value Effective delayed neutron fractions P 0.007 Effective neutron lifetime 1 43 psec Temperature coefficient of reactivity a -0.000115 IC Source: West et al. (1967).

Table 13.3, Thermal and Mechanical Properties of TRIGA MkII ZrH,A Fuel Elements and Type 304 Stainless Steel Cladding.

Propertv Svmbol Value Temn Fuel Density py 5996 kg m7' Thermal conductivity k, 18 W m1 K' All 0

Heat capacity, Cpf 340.1 + 0.6952T( C) 340.1 Jkg"'K' 0°C Cladding Density 7900 kg mO 300 K Thermal conductivity .k, 14.9 W mn'K' 300 K 16.6 400K 19.8 600 K Heat capacity CP 477 J kg' r' 300 K 515 400K Yield strength 250 Mpa 400C Tensile strength 455 Mpa 400 C Source: fue properties from Simnad (1980); cladding properties from Incropem and DcWittj1990) and from Metals Handbook (1961).

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A ACCIDENT ANALYSIS Table 13.4, KSU TRIGA Cor-Conditions Basis for Calculations.

Steady state maximum power. P. 1.259 kW I *2)D Fuel mass per element 2.367 kg Heat capacity per element at T(Ct) 805.0 + 1.646T(J K')

Minimum number of fuel elements, N O*

Core radial peaking factor 2 Axial peaking factor nf2 Excess reactivity S4.00 (2.8% Ak/k)

Maximum pulsing reactivity insertion $3.00 (2.1% Ak/k)

Excess reactivity at 500 kW maximum power $1.16 (0.81% Ak/k)

Fuel average temperature at 500 kW maximum powers 285 'C

'Source: Data from GA Torrey Pines TRIGA reactor 13.2.2 Loss of Reactor Coolant Although total loss of reactor pool water is considered to be an extremely improbable event, calculations have been made to determine the maximum fuel temperature rise that could be expected to result from such an event taking place after long-term operation at full power of SOO kW. Limiting design basis parameters and values arc addressed by Simnad (1980) as follows:

Fuel-moderatortemperature is the basic limit of TRGA reactoroperation. This limit stems from the out-gassing of hydrogen from the ZrH. and the subsequent stress produced In ihefuel element dad materiaL The strength of the clad as arunctlon of temperature con set the upper imnit on thefiel temperature. A fud temperaturesafety limit ofJ1150Cforpulsing stainlesssteel U-ZrIfw .- fuel is used as a design value to preclude the loss of clad integritywhen the cadtemperature Ibelow500'2 i Sten clad temperaturescan equalthefuel temperature thefuel temperature limit Is 950'C There Is also a steady-state operationalfuel temperature design limit of 750 based on considerationofirradiation-andfsslon-product-Inducedfuelgrowth anddeformation_.

As this section denonstrates, even under extraordinarily conservative assumptions and approximations, the maximum fuel temperature reached in a loss of coolant accident is 2901C, well below any safety limit for TRIGA reactor fuel. Conservatism notwithstanding, the margin between computed temperature and design limits is sufficiently great to accommodate a design margin of at least a factor of twO.

a. Initial Conditions, Assumptions, and Approximations The following conditions establish an extremely conservative scenario for analysis of the loss of coolant accident.

The reactor is assumed to have been operating for infinite time at powerP* = 500 ;kW at the time coolant is lost.

Coolant loss is assumed to be instantaneous.

Reactor scram is assumed to occur simultaneously with coolant loss.

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CHAPTER 13

Decay heat is from fission product gamma and x rays, beta particles, and electrons.

Effects of delayed neutrons are neglected.

Thermal power is distributed among N = fuel elements, with a radial peak-to-average ratio of 2.0. In individual elements, thennal power is distributed axially according to a sinusoidal function.

Cladding and gap resistance are assumed to be negligible, i.e., cladding temperature is assumed to be equal to the temperature at the outside surface of the fuel matrix.

Cooling of the fuel occurs via natural convection to air at inlet temperature T1 =

3000 K Radiative cooling and conduction to the grid platekare neglected.

Heat transfer in the fuel is one dimensional, i.e., axial conduction is neglected, and fuel is assumed to be uniform in thermophysical properties.

Heat transfer in the fuel is treated on the basis of pseudo-steady-state behavior, i.e, at any one instant, heat transfer is described by steady-state conduction and convection equations.'

b. Core Geometry The following data on core geometry are derived from the K.SU TRIGA Mechanical Maintenance and Operating Manual (1962). The core contains 90 fuel positions in five circular rings (B - F), plus the central thimble (A ring). The upper grid plate is OA95 m in diameterand O.019.m thick. Holes to position the fuel are 0.03823 m diameter and the central thimble is very slightly larger in diameter, 0.0384 m.

Cooling water passes through the differential area between the triangular spacer block on the top of each fuel clement and the round holes in the upper grid plate. The nominal diametral clearance between the tips of the spacer blocks and the grid plate is approximately 0.001 m.

The lower grid plate is OA05 m diameter, with 36 holes, 0.0159 m diameter, for water

.llow. However, the bulk of the water flow is through the annular space provided between the top of the lower grid plate and the bottom of the reflector. The radial reflector is D, OA57 m inside diameter and 0.559 m height.

The effective hydraulic diameter for flow through the core, with an experiment in place in the central thimble, is given by 4*A 4 * (.4 91*:r*r)

D,= - I= 0.02127m (Equation 13.2.2.1)

P., x*Dt 91*2*z*r,.

'SeeTodreas &4Cazeri (1990) orEl-Wakil (1971)forsteady-tateconductionequations.

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ACCIDENT ANALYSIS If, in thermal-hydraulic calculations, one approximates conditions as flow through an annular section around any one fuel element, the outer radius of the annulus, say r,,, is given by D4;r(r;2-r)r Dk - or 2nr. (Equation 13.2.2.2) r= JD,r l 2 +r2= 0.02339 m, The flow area A, per fuel rod is xr(rQ2 - r,) = ;zw Dn.

/2 = 0.0006238 The total length of a fuel rod is 0.3206 m (2837 in.), of which the length of the fuel matrix, the heated length, is Lm 03810 (15 in). The lengths of upper and lower axial reflectors, LandLI are each 0.0874 m (3.44 in). Beneath the lower reflector is a bottom end fixture of length Li about 0.0824 m (3.245 in.). Above the upper reflector is a triangular spacer of length L, about 0.0191 m (0.75 in.) and an upper end fitting of length L. about 0.0634 m (2A95 in.). The zone between grid plates is L,=L,+4s+L 1 L. = 0.6382 m.

c. Decay Power The time dependence of the thermal power in the core as a function of time after shutdown is based on calculations by the CINDER code [England et al. 1976] as reported by George, LaBauve, and England [1980, 1982]. Sample results are presented in Figure 13.1 and Table 13.5 as the function R(Q defined as the ratio of the thermal power Pd(t) from gamma ray and beta particle decay at time t alter shutdown to the steady power P. prior to shutdown, based on 200 MeV energy release per fission.

For the purpose of this analysis, the time dependence function for I to 106 s may be approximated as 0.04856+0.1189x-0.0103x 2 +0.000228xX3 E tion 13 3)

= +25481x-0.19632X2 +0.05417X2 1(s) ' (3 2 1-3 in whichx is the natural log of the time after shutdown, in seconds. Time dependence of the thermal source in the worst-case element is given by pd(t) = 2P. (Equation 13.224)

N in which the worst case element generates twice the power as the core average, P. is the total-core thermal power prior to shutdown (1,250 kW), and N r83)ks the minimum number of fuel elements required for operation.

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CHAPTER 13

10. 2 ZO Ml 0 , 3 14 S 16 16 Tbe (5)

Fkjirc 13.1. Dpcav Heat Function NOW).

Table 13.5, Decay cat Funcdtion for Thermal ission of 23 Time I (s) Rt) cPd(QVPo 0 0.0526 1t(IO° 0.0486 10 (10 ) 0.0418 100 (10) 0.0282 1000(100 ) 0.0172 10,000 (104,) 0.0087 100,000 (104) 0.0044 1.000.000

_ -- - (10, _ __ 0.0025

d. Maximum Air Temperature The fundamental relationships between buoyancy- driven diffcrential pressure and pressure, losses from friction provide an independent verification for results of the previous calculations.

Buoyancy Driven Pressure Difference The total mass flow rate w (kgls) associated with the worst-case fuel element is determined by a balance between the buoyancy driven pressure difference vertically across the core and the frictional pressure loss within the core, which is discussed in the next section. The temperature rise across the core is AT. ()atft=T,()-4=,(t

= 7.(t) 2!T~,) _R(f) 2I N J (Equation 13.2.2.5) wyc-" NW)-C; K-State Reactor 13-6 Oiginal (12104)

Safety Analysis Report

ACCIDENT ANALYSIS in which the heat capacity of the air evaluated at the inlet air temperature. At time zero, for example, A?;(0)= 7(O)-2T 0.6153lw. (Equation 13.2.2.6)

Air inlet temperature Tj is assumed to remain constant at 270 C. Suppose pi and p. are respectively the densities of air at the inlet and outlet temperatures. 2 Suppose further that the effcctive chimney height is H. The chimney height is the distance between the center of the zone in which the air is heated and the center of the zone in which the air is cooled.

Evaluation of the latter is difficult to determine because of uncertainties in mixing of the air after it leaves the upper grid plate. Here we follow the lead of the UT SAR (1991) and choose 10 hydraulic diameters as the effective distance. Thus, His given by 4/2 +

L, +1ODA = 0.422 m. and the buoyancy pressure difference is given by Apb = (pi - po)*g*H (Equation 13.2.2.7) in which g is the acceleration of gravity, 9.8 m e. Since 353 *(T. - 7)

Pi -PO a 2 2T (Equation 132.2.8)

T, and T,.300 K, it follows from Eqs. (132.2-5) and (13.2.2-8) that 'S 4Ap(t) = 0.19ORQt)/ w, (Equation 13.2.2.9)

Frictional Pressure Difference In this calculation, only frictional losses within the core, computed on the basis of the equivalent annulus model, are accounted for. Based on air inlet density and an air mass flow rate per fuel rod of v,the frictional pressure difference is given by (Equation 132.2.10)

/ 2pjD4hA The laninar-flow (Moody) friction factor f for the equivalent annulus model, with r, / r, = 1.193 is given by Sparrow and Loeffler (1959) as f =lOOIRe, (Equation 13.2.2.11)

I.-

Re, the Reynolds number, is given by D&wlpA,, and j, is the dynamic viscosity of the air at the inlet temperature. 3 Equation (13.2.2-1 0) may be rewritten as 2 Density at I atm, for air as an ideal gas, is given by p ftg/n9) - 353.0/7IK). Heat capacity, from S00 to 700 'Kis 1030 J/kgK*3% (Incropera and DeWitt, 1990).

3Dynamic viscosity, over the range 250 -1000 'K is given by I0d' (Nstm2) - -106.2941 +

16.81986[710K)]m or p- 1.85 x 107r Nshn9 at 300 lC. (Incropera and DeWitt, 1990).

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CHAPTER 13

= 0J416L,pTjiv = 1780'v, (Equation 13.2.2.12)

D.,A, Equating the frictional.pressure drop with the buoyancy pressure driving force, using Equations (132.2-9) and (1322-12),

w= 0.0103Ift0, (Equation 1322.13) or T,(J)- 2 =AT(t)= 1140JEIj. (Equation 132.2.14)

Results For R(t) at time zero of 0.0526, maximum air temperature rise above 300 K is 261 K.

e. Fuel and Cladding Tcmperaturc Distribution With design power P. = 1,250 kW, a factor of two radial peak to average power, and a fuel surface area in the heated zone equal to Al = 2rwL- 0.04469 mn,the worst-case average heat flux at post-accident time I in the heated zone is:

qn"=2*1E25e6*R()/AN=3.37e5 WImn (13.22-15)

. With the conservative approximation that the axial variation of heat flux is sinusoidal, the local value of the heat flux (WVm 2 ) is given by.

q" (z) = ".. sin(= L,), (1322-16) in Which z is the distance along the fuel channel, measured from the inlet and q4.= (ir2)q", 4235 x 10'S(() W/ 2 Similarly, the local value ofthe air temperature in the coolant channel is given by

.T.. ( = 7} +0.6153R(t)I - cos(= I L,)]. (1322-17)

According to Dwyer and Berry (1970), the Nusselt number for lamrinar flow in a cooling channel is approximately Nu = 424. The corresponding beat transfer coefficient is k..Nu h=D. (1322-18)

1 By using as an approximation the air thermal conductivity 4 of 26.3 W/mK at 300%K, one computes h = 5240 Yldm 2 , and the cladding surface temperature 4For the range 200 to 1000D0 data ofincropera and DeWitt (1990) is very well fit by the fonnula k,, = -22.055+2.8057 5 inunitsoof Wn.

Orlgina! (12104)

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Safety Analysis Report

J J

ACCIDENT ANALYSIS T'ja(z) = T..(z) +q"(Z) I h. (13.2.2-19)

By using the fuel thermal conductivity ky = 18 V/rmK, and neglecting the temperature drop across the cladding, one computes the fuel centerline temperature as d(Z)= T (Z)+a -"k (13.2.2-20)

Fuel and cladding temperatures are reported in Table 13.6 and illustrated in Figure 13.2 for the case of zero time post accident. This is based on three conservative assumptions:

cquilibrium fission product buildup at full power, instantaneous loss of coolant when the reactor scram occurs, and equilibrium temperature based on initial decay heat production.

Table 13.6, Post-Accident Fucl and Cladding Temperatures.

gN (W/im2 ) T.& (°K) Tia (`K)

Tsd (0K) 0.00 0 300 300 300 0.10 17209 306 309 318 0.20 32733 325 331 348 0.30 45053 354 362 385 OAO 52963 390 400 427 0.50 55688 431 441 470 0.60 52963 471 481 508 0.70 45053 508 516 539 0.80 32733 536 542 559 0.90 17209 555 558 567 1.00 0 561 561 561 The K-State reactor does not operate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day, 7 days per week, so that the total inventory of fission products is significantly lower than the assumed value. During actual loss of coolant; overall heat transfer coefficient will be based on water rather than air.

There is at least 16 feet of water over the core that has to drain before the core is I~.-

uncovered. Experiments reported in G3A-6596 "Simulated Loss-of-Coolant Accident for TRIGA Reactors" (General Atomics, August 18, 1965) demonstrated that with a constant and continuous heat production, temperature rises to about 500ha of the equilibrium temperature in approximately 30 minutes (1800 seconds). At iOW seconds,

.R(t) is 0.0172, 33% of the heat production following shutdown. Equilibration takes a significant amount of time, while heat production is decaying.

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-CHAPTER 13 SW 5 0 --.

I-0501 C.o am 00 0.40 OM 0.60 0.70 a0 a" 100

. zA, Figurc 4.13, Axial Variation or fuel, cladding, and air temperature Immediatcly following a loss orcoolant accident, wlth equilibrium fission product heating.

f. Radiation LcvCls from the Uncovered Corc Although there is only a very remote possibility that the primary coolant and reactor shielding water will be totally lost, direct and scattered dose rates from an uncovered core following 1,250kW operations have been calculated.

This section describes calculations of on-site and off-site radiological consequences of the loss-of-coolant accident. Extremely conscrvative assumptions are made in the calculations, namely, operation at 1,250 kW for one year followed by instant and simultaneous shutdown and loss of coolant. The reactor core is homogenized and the ORIGEN-2 [CCC-3713 is used to define gamma-ray source strengths, by energy group.

Radiation transport calculations are performed using the MCNP code (Briesmeister 1997), nwith doses evaluated at the 22-ft elevation of the reactor (one meter above the operating deck), in the reactor bay at the 0-foot elevation (one meter above the floor, 30-cm from the outer wall, and at one-meter above ground level at the site boundary and at points beyond the site boundary extending to 100 m radial distance from the core.

Modeling of the reactor core for radiation transport calculations is based on the following approximations. The TRIGA reactor core is approximated as a right circular cylinder 0.4572 m diameter (OD of F ring). The fuel region is 0.381 m (I5 in.) high. On each end axially is a graphite zone 0.0874 in (3.44 in.) hih and an aluminum grid plate 0.0191lm

(0.75 in.) thick. 14&fuel locations, there arcfuel elements, 3 standard control rods and I transient conrol rod, I void location, I central thimble (void), I source (assume void), and I pneumatic transfer site (assume void). The fuel region may is treated as a homogeneous zone, as are the axial graphite zones'and the grid plates.

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ACCIDENT ANALYSIS Fuel elements arc 1A3-in. ID and 1.47-in OD, clad with type 304 stainless steels. Fuel density is 5996 kdm31Fuel composition is~3 uraniumiZrH as. The uranium iso 23'5U and W 3 1 U. Steel density is 1900 kg/n 3 . Standard control rods are 0.875-in. OD, the transient rod 1.25-in. OD. Both types of rods are clad with 30-mil thick aluminum (2700 kg/mr3 density). The control material may be approximated as pure graphite, with density 1700 kg/nl.

In radiation transport calculations, the core is modeled conservatively as a central homogenous fuel zone (air density neglected) bounded on either end by a homogeneous axial reflector zone; and by a 0.75-in. thick aluminum grid plate, treated as a homogeneous solid. Densities of the homogenous zones are as follow:

Fuel 3602 kg/rn 1 Reflector 1147 kg/rn Grid Plate 2700 kg/m' Composition of the three zones, by weight fraction, are given in Table 13.7.

Table 13.7, Compositions of Homogenized Core Zones.

Element Mass Fraction Element Mass Fraction FuelZone Axial Reflector Zone C 0.0617 C 0.7920 1=

Al 0.0010 Al 0.0033 4.-

H 0.0139 Mn 0.0041 Zr 0.7841 Cr 0.0368 Mn 0.0013 Ni 0.0164 Cr 0.0117 Fe 0.1474 Ni 0.0052 Fe 0.0469 GridPlate U 0.0741 Al 1.0000 The reactor bay is approximated as a right circular cylinder 36 ft (10.973 m) high and 36.68 ft (11.18 m) radius. The reactor vessel structure is approximated as a right circular cylinder, co-axial with the bay, 22 R (6.706 m) high and 11 ft (3.3528 m) radius. The free volume is 144,000 f' (4078 m3). The site boundary, at its nearest approach to the reactor bay, is about 2 m beyond the bay boundary, that is, at a radius of 13.13 m from the center of the reactor.

Gamma-my source strengths vs. times after shuitdown are listed in Table 13.8 for an operating time of one year and a thermal power of 1,250 kW. The source strengths are gamnm rays per second, by group, for the entire reactor core. In the MCNP calculations, the source is assumed to be uniformly distributed within the core.

The roof of the reactor bay is modeled as a concrete slab 10 cm thick and with density 2.35 g/cm'. In fact, the roof is a composite structure, so reflection from the concrete slab conservatively models gamma-ray transport and dose rates at the site boundary.

SComposition, by weight, 2% Mn, 18% Cr, 8%Ni, balance Fe.

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Safety Analysis Report

CHAPTER 13 Table 13.9 reveals substantial dose rates above the void reactor tank with long-term 500-kW operation. One day after loss of coolant, the dose rate is 17 SvIh (1700 rem/h).

However, dose rates inside the reactor bay at the 0-ft. and 12-fi. elevations are only about -

4 rem/h immediately after coolant loss, declining to about 1.5 rem/h after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 100 mrern/h after 30 days. These dose rates show that personnel could occupy the reactor bay shortly after the loss of coolant to undertakc mitigating actions without exceeding NRC occupational dose limits.

Dose rates at other receptor locations are shown in Table 13.9 as indicators of areas that should be checked following a loss of reactor pool water event. These dose rates are unrealistically conservative in view of the assumption of one-year operation at full power (not possible) and scattering from a thick concrete roo. Dose rates at the site boundary (facility fence) decline from about 3 remlh immediately after the loss of coolant to 350 mnrn/h after I day. Kansas State University has complete control over access to campus locations within the zones defined by receptor locations in the analysis.

Table 13.8, Full Core Gamma-Ray Sources Strengths (Number Pcr Second) Following Operation for One Year at 1,250 kV Thermal Power.

Time after shutdown E (MeV) 0 lb 24 b 30 days 180 days 1.OOE-02 8.72E+15 292E+1S l.OSE+15 2.43E+14 620E+13 2.50E-02 2.09E+15 7.50E+14 3.04E+14 5.97E+13 132E+13 3.75E-02 1.74E+15 7.37Ef14 3S7E+14 6.72Et13 IASE+13 5.7SE-02 1.81E+I5 550E+14 1.77E+14 4.15E+13 1.20E+13 85OE-02 1LSOE+I5 520E+14 2.33E+14 3.21E+13 8.39E+12 12SE01 1D50E+1S 7.6SE+14 4.81E+14 8.98E+13 1.31E+13 22SE-01 3.11E+1S 1.03E+15 4A.4E+14 2A2E+13 6.83E+12 3.75E-01 1.94E+1S 5.72E+14 2.20E+14 3.81E+13 3AIE+12 S.7SE-01 3.2E+15 1.71E+35 7.54E+14 I.OIE+14 1.08E+13 850E-O1 4.16E+I5 2.26E+15 894E+14 451E+14 1.lSE+14 12SE+00 2.22E+15 8.1SE+14 7.83E+13 1.62E+12 .6.01E+II 1.7SE+D0 930E+14 5.14E+14 2.23E+14 4.78E+13 7.72E+IO 22SE+00 5.00E+14 2.19E+14 6.69E+12 1.12E+12 3.72E+1 I 2.7SE+OO I.99E+14 732E+13 8.13E+12 i.22E+12 1.5SE+09 3.S0E+00 1.16E+14 1.72E+13 7.12E+IO l.SOE+10 9.37E+07 5.OOE+00 623E+13 2A2E+11 6.96E+08 4.67E+00 4.63E+00 7.OOE+00 5.OSE+II S27E-01 5.27E-01 S26E-OI S.22E-01 9.50E+00 9A9E+07 5.98E-02 5.98E-02 5.98E02 53E-02.

Total 339E+16 13SE+16 S.27E+15 120E+15 2.61E+14 McV/s lIA3E+16 6.28E+15 2.02E+15 5.77E+14 1.13E+14 K-State Reactor 13-12 Original (12104)

Safety AnalysIs Report

ACCIDENT ANALYSIS Dose rates vs. post accident times are listed in Table 13.9 for a number of receiver locations. Locations arm specified at radial distances from the center of the reactor bay.

The22-foot level provides direct access to the reactor pool, the 12-foot level access to valves controlling water flow to the reactor pool. The 0-foot level provides access to valves and pumps, which allowv water to be added to the reactor pool. The 13-meter distance marks a zone defined by the fence surrounding the reactor bay.

g- Conclusions Although a loss of pool water is considered to be an extremely improbable event, calculations show the maximum fuel temperature that could be expected to result from such an event (after long-term operation at full power of 1,250 kWV) is 2941C, well below any safety limit for TRIGA reactor fuel.

1 Maximum possible dose rates resulting from a complete loss of pool water permit mitigating actions. The area surrounding the reactor is under control of the Kansas State University, and exposures outside the reactor bay environment can be limited by controlling access appropriately. Kansas State University has complete authority to control access to campus locations.

Table 13.9, Gamma-Ray Ambient (Deep) Dose Rates (R/h) at Selected Locations for Times Following Loss of Coolant After Operation for One Year at 1,250 kWV Thermal Power.

Time post accident 0 lh 24h 30d 180d On-site (elev.)

22 ft. (center) 3.75E+06 1.38E+06 4.25E+05 1.03E+05 l.98E404 4 12 ft (boundary) 1.08E+03 4.00E+02 A.43E+02 325E+01 7.00E+O0 I..

o & (boundary 9.7SE+02 3.75E+02 1.23E+02 3.OOE+01 6.25E+O0 Off-site (radiusfromcenterofreactorbay) 13 m 6.75E+02 2.75E+02 8.7SE+01 2.15E+01 4.50E+O0 15 m 5.00E+02 2.OOE+02 6.7SE+01 1.60E+01 3.25E+00 20 m 2.MOE+02 l.00E+02 3.50E+01 8.75E+00 1.73E+00 30 m 8.25E+01 325E+O l.IOE+Ol 2.75E+00 S.50E-0I 40 m 3.75E+O1 lAOE+OI 5.2SE+00 1.18E+00 2.33E-O SO m 1.93E+01 72SE400 2.43E+00 6.25E1-01 1.23E-0l 70 m 6.SOE+00 2.75E400 9.00E-01 2.18E3-01 4.OOE-02 100 m 2.35E+O0 9.50E-01 2.75E-01 6.75E-02 1.33E-02 K-State Reactor 13-13 Original (12104)

Safety Analysis Report

CHAPTER 13 13.2.3 Insertion of Excess Reactivity Rapid compensation of a reactivity insertion is the distinguishing design feature of the TRIGA reactor. Characteristics of a slow (ramp) reactivity insertion are less severe than a rapid transient since temperature feedback will occur rapidly enough to limit the maximum power achieved during the transient. Analyses of plausible accident scenarios reveal no challenges to safety limits for the TRIGA. The fuel-integrity safety limit, according to Simnad (1980), may be stated as follows:

Fuel-moderatortemperature is the basic limit of TPJGA reactoroperation. Thfs limit stems from the out-gassing of hydrogen from the ZrH, and the subsequent stress produced in thefuel element clad material. The strengit of the clot as afinetlion of temperaturecan set the upper limit on the fuel temperature. A fuel temperaturesafely limit of l150 Cforpuing. stainlesssteel U-ZrHj/j .- firel is used as a design value to preclude the loss of clad integritywhen the cadtemperatureis below 5OO 'C When clad temperatures can equalthefuel temperature, thefuel temperaturelinit is 950'C.

Two reactivity accident scenarios are presented. The first is the insertion of 2.1% reactivity at zero power by sudden removal of a control rod.: The second is the sudden removal of the same reactivity with the core operating at the maximum power level permitted by the balance of the core excess reactivity (i.c., core excess less $3.00). Movements of control rods for the first case are controlled, in part, administratively, while movements for the second arc prevented by control circuit design.

As the analysis shows, in neither scenario does the peak fuel temperature approach the temperature limit. The nearest approach is 8691C, incurred by a pulse insertion of0.7% while the reactor is operating at a steady power of 94 kW, an action prevented both by administrative requirements and by interlocks.

a. Initial Conditions, Assumptions, and Approximations The following conditions establish an extremely conservatiVc scenario for analysis of insertion of excess reactivity.

The reactor operates with a minimum ofN=*fuel rods.

Reactor and coolant ambient (zero power) temperature is 27°C

Maximum reactivity insertion for pulsing or for the worth of experiments is set at

$3.00, &k. = 2.1% or p.., - 0.021.

Reactorpower equivalent to the core excess reactivity of Sl.00, i.e., k= Sm- Skg 0.7% (p - 0.007) isP. S.107 kW and the maximum fuel temperature at that power is T. = 1500 C. Basis: Data for the Torrey Pines TRIGA, as included in the KSU TRIGA Operations Manual.

A control rod interlock preventing pulsing operations fron power levcls greater than a maximum of 10 kW is not credited

Conservative hot channel fa&ors as'calculated in 4.5.3 arc used K-State Reactor' 13-14 Original (12104)

Safety Analysis Report

ACCIDENT ANALYSIS

b. Computational Model for Power Excursions The following relationships are for the Fuchs-Nordhcim model, modified by Scalletar, for U power excursions, as described for the TRIGA reactor by West et a]. (1967).

If the heat capacity of the fuel is given by, cor= 340.1 + 0.69527°C) (J/kg0 K), and there l are N fuel elements, each of mass mnt 2.367 kg, then the overall core heat capacity is given by K (J I K) =mNc, = C. + C;TC C) (13.23-1) in which C, 6.682 x 10' and C, 136.6. IfP. (W) is the reactor power at the initiation of the pulse power excursion, and po is the magnitude of an initiating step change in reactivity, then the maximum power increase (W) is given by

~ =(p0-i)2 C.

P.-. 2at 4 (p-/73c; 6a tl3.2.3-2 A maximum pulse of S3.00 would result in a power rise of approximately 1430 MW(t).

If T. is the average core temperature at the start of the excursion, the maximum temperature rise (0K) is given by

)TO a* a - 1) * (13.2.3-3) in which ar= (13.2.3-4) and the "9' sign applies when ar> 1. Although there are nonlinear terms in the model, calculation of temperature change as a function of temperature shows a nearly linear response. *Thereforc the major factor in determining core-avcrage peal temperature is the amount of reactivity available to pulse. To remain within 10000 C at all locations,.core average temperature cannot exceed 318°C (based on Table 13.2.1.4 peaking factors).

.The maximum worth of the pulse rod is S3.00, therefore peak temperature following a pulse was calculated based on a maximum reactivity available from the pulse rod at reference core temperature (270 C). The core-average temperature change is 229'C, with a hot spot change of 716'C (based on Table 13.2.1.4 peaking factors). Therefore, core-average peak temperature for a S3.00 pulse from 271C is 2561C, with a hot spot temperature of 7460 C.

With $3.00 reserved for a maximum pulse, the reactor has remaining reactivity $1.00 greater than critical to support operation at power. Reactivity of Sl.00 allows operation 94 kW; 94 kW operation results in an average fuel temperature of 481C, and a hot-spot K-State Reactor 13-15 Original (12104)

Safety Analysis Report .

CHAPTER 13 fuel temperature of 1500C. A $3.00 pulse increases this hot spot temperature from 1500C to 869TC.

Change in Peak Temperature Versus Pulsed Reactivity S=j E=.1 1 -- 4.1 -

I N1F 4" U

4.

to Su -

1- II I =-

IS O 2 O0 aJ 110 2to 410 410 3C Rentlvity (S)

Figure 13.3, Change in eakn Tczperaturc Versus Pulsed Reactivity lnscrtion Tahe postulated scenarios do not result in fuel damage, but physical aspects of system prevent these scenarios from occurring. It is not possible to achieve full power operation with the pulse rod fully inserted, since the pulse rod is partially withdrawn with air applied to the pulse solenoid, it physically cannot be pulsed. Although not required to ensure the safety of the reactor, an interlock prevents pulsing from power levels greater than a maximum of 10 kWV.

c. Conclusions Insertion of the maximum possible reactivity without initial temperature feedback (.e.,

fuel temperature is too low to limit core available reactivity) results in a peak hot spot fuel temperature of 7461C, well below the safety limiL Insertion of the maximum possible reactivity with initial temperature feedback (i.e., fuel temperature limits available) results in a peak hot spot fuel temperature of 8691C, well beloiv the safety limit.

13.2.4 Single Element Faildre in Air Source quantities of radioactive noble gases and iodine are computed and tabulated for a maximum hypothetical accident involving cladding failure in a single TRIGA fuel element and the escape of the radionuclides into the environment Two limiting cases of operation are considered. For short lived radionuclides, source terms are computed for element failure subsequent to eight hours full-power operation per day for five days. For long-lived radionuclides, source terms are computed for element failure subsequent to continuous operation K-State Reactor' - Original (12/14)

Safety Analyss Report

ACCIDENT ANALYSIS for 40 years at the average power experienced by the reactor over its first 33 years of operation.

Also examined are residual sources still present in fuel, but generated in reactor operations prior to local receipt of the fuel in 1973. Potential consequences of radiological releases are examined.

Even in the maximum hypothetical accident, no workers or members of the public are at risk of receiving radiation doses in excess of limits prescribed in federal regulations.

a. Assumptions and Approximations \J Following arc assumptions and approximations applied to calculations. \J I For long-lived radionuclides, calculations of radionuclide inventory in fuel are based on continuous operation prior to fuel failure for 40 years at the average thermal power experienced by the reactor during its first 33 years of operation, namely, 3.50 kW.

2, For sbort-lived radionuclides, calculations. of radionuclide inventory in fuel are based on operation at the full thermal power of 1,250 kW for eight hours per day, for five successive days prior to fuel failure, an average of 31.25 kW-hr/day.

3. Radionuclide inventory in one "worst-case" fuel element is based on 81 elements in the core for the historical period andpelements for full powver operation, which is the case for core 11-1 6,grams of "5U per element [NUREG-2382],

and a value of 2.0 as a very conservative value of the ratio of the maximum powver in the core to the average power. Thus, for the historical period, the worst case element has operated at a thermal power of (3500/8 I)x2 = 86A2 W, and for the one-week full-power operation, (500*x2 =ijkW.

4. The fraction of noble gases and iodine contained within the fuel that is actually released is 1.0 x 10'4. This is a very conservative value prescribed in NUREG 2387 [Hawley and Kathren, 1982] and may be compared to the value of 1.5 x 10 5measured at General Atomics [Simnad ct al., 19761 and used in SARs for other reactor facilities [NUREG-1390, 1990]. 1

5. The fractional release of particulates (radionuclides other than noble gases and iodine) is 1.0 x 104, a very conservative estimate used by Hawley and Kathren

[1982]*

b. Radionuclide Inventory Buildup and Decay Consider a mass of 235U yielding thermal power P (kW) due to thermal-neutron induced fission. The fission rate is related to the thermal power by the factor k 3.12 x 1013 fissions per second per kW.' Consider also a fission product radionuclide, which is produced with yield Y, and which decays with rate constant L It is easily shown that the equilibrium activity Aa (Bq) of the fission product, which exists when the rate of creation by fission is equal to the rate of loss by decay, is given by 4 = WPY. Here it should be noted that the power must be small enough or the uranium mass large enough that the 6Note that the product ofkand yield Ymay be sated as 3.12 x 1013 x YBq/MW or 843 x YCMUMY.

K-State Reactor 13-17 Orginal (12104)

Safety Analysis Report

CHAPTER 13 depletion of the 2 5U is negligible? Starting at time t = 0, the buildup of activity is given by A(t) = A. * (I - (13.2.4-l)

For times much greater than the half-life of the radionuclidc, Aa A., and for times much less than the half-life, A(t) A. ).* t. If the fission process ceases at time tz the specific activity at later time t is given by A() = A,,, *(I- e"') *e-Aia") (13A42-2)

Consider the fission product .. 1j, which has i half-life of 8.04 days (A= 0.00359 h') and a chain (cumulative) fission product yield of about 0.031. At a thermal power of I kV, the equilibrium activity is about A. 9.67 x 10" Bq (26.1 Ci). After only four hours of operation, though, the activity is only about 0.37 Ci. For equilibrium operation at 3.5 kAV, distributed over 81 fuel elements, the average activity per clement would be 26.1 x 3.5 + 81 = 1.13 Ci per fuel element. The worst case clement would contain twice this activity. With a release fraction of 1.0 x 104, the activity available for release would be about 1.13 x 2 x 1.0 x 104 2.26 x 10 Ci. This type of calculation is performed by the ORIGEN code [CCC-371] for lundreds of fission products and for arbitiry times and power levels of operation as well as arbitrary times of decay after conclusion of reactor operation. The code accounts for branched decay chains. It also may account for depletion of235U and ingrowth of"2Pu, although those features were not invoked in the calculations reported here because of minimal depletion in TRIGA fuel elements.

c. Data From Origen Calculations ORIGEN-2.1 calculation output files are included as Appendices A and B; App. A contains data for the buildup of l6ng-lived radionuclides over the 40-year entire operating history of the KSU TRIGA reactor, modeled as 86.42 W continuous thermal power. App.

B contains data for the buildup ofrelazively short-lived radionuclides during a worst-case scenario modeled as 8-hours/day operations at 12.05 };V thermal power for five consecutive days. Tabulated results for Appendices A and B are pCi activities, by nuclide, immediately after reactor shutdown, and at 1, 2, 3, 7, -and 14 days after shutdown. In Appendix A, which deals with relatively long-lived radionuclides, data arc provided only for those nuclides present at activities greater than 100 mCi in a single fuel clement at 1 day after reactor shutdown. In Appendix B, which deals with relatively short-lived radionuclides, data are provided only for those nuclides present at activities greater than 100 mCi in a single fuel element immediately after reactor shutdown. In both Appendices A and B, the activities per element are multiplied by the release fractions previously cited, thus yielding maximum activities available for release in a maximum hypothetical accident

'Negligible bumup is modeled in ORIGEN calculations by setting the fuel mass very large (I tonne) and the thermal power very low (I kV or less).

K-State Reactor 13-18 Original (12104)

Safety Analysis Report

ACCIDENT ANALYSIS

d. Reference Case Source Terms Appendices A and B data for worst case TRIGA fuel element are compared; greater 'J values for any one isotope are selected as reference case source terms for maximum hypothetical accident. Data arc presented in Table 13.10 for halogens/noble gases, and Table 13.11 for particulate radionuclides. v c- Derived Quantities J The raw data of Tables 13.10 and 13.11 are activities potentially released from a single worst-case fuel element that has experienced a cladding failure. This activity may itself be compared to the annual limit of intake (ALI) to gauge the potential risk to an individual worker. By dividing the activity by the 144,000 f 3 free volume of the reactor I bay in the Nuclear Reactor Facility, one obtains an air concentration (specific activity) l that may be compared to the derived air concentration (DAC) for occupational exposure as given IOCFR20 or in EPA federal guidance [Eckerman et al., 1988].

K-State Reactor 13-19 Original (12104)

Safety Analysis Report

CHAPTER 13 Table 13.10, Rcfercncc Case Halogen & Noble Gas Activities Potentially Rcleased In Masximum Hypothctical Accident at XSU TRIGA Mk. 1 Nnuclear Reactor. III Available activity (iiCi) at time in days after reactor shutdown Element Nuclide 0 1 2 3 7 14 28 Br

83 12110 13 0 0 0 0 0

Br 84 25388 o o o 0 0 I 131 8490 82'Is 7695 7130 5125 2810 840 I

132 23280 209M13 16930 13685 5838 1318 68 I 133 69950 335S30 15073 6773 278 0 0 1 134 192228 0 0 0 0 0 0 I 135 98750 7980 645 53 0 0 0 Kr 85 45 45. 45 45 45 45 45 Kr 87 64498 0 0 0 0 0 0 Kr 88 79048 225 0 0 o o 0 Kr 83M 9953 48 0 0 O

O 0 Kr 85M 23368 580 15 0 0 0 0 Xe 133 20363 24105 24010 22390 1411 10 5673 895 Xe 135 58955 30518 659S 1195 0 0 0 Xe 138 158548 0 0 0 0 0 0 Xe 133M 35 33 28 20 8 0 Xe 135M 205 . 15 0 0 0. 0 0 NOTE: Available activity (> 10 pCi Is from a single .worst-sefuel element as a function of time after reactor operation. Data are deried from ORIGEN 2.1 calaclations [CCC-371] as summarized In

.4ppendicesA andB. Data are raw computationalresultsand the number ofslgnlficantfiguresexceeds the A

precisionofthe calculation.

K-State Reactor 13-20 Original (12104)

Safety Analysis Report

1..1 ACCIDENT ANALYSIS.

j1-Table 13.11, Rcrcrcnce Case Particulate Activities Potentially Rcleased in a Maximum IRypothetical Accident at the KSSUTRIGA Mk. II Nuclear Reactor.

Available activity (yCi) at time in days after reactor shutdown Element Nuclide 0 1 2 3 7 14 28 BA 139 1605 0 0 0 0 0 0 BA 140 128 120 IIS 108 88 60 28 BA 141 1480 ' 0 0 0 0 0 0 BA 142 1470 0 0 0 0 0 0 CE 141 48 53 53 50 45 40 30 CE 143 5S3 340 205 125 l3 0 0 Cs 138 1713 0 0 0 0 0 0 LA 140 68 88 98 103 95 68 33 L.A 141 1140 IS 0 0 0 O 0 LA 142 1455 0 0 0 0 - 0 0 LA 143 1493 0 0 0 0 0 0 MO 99 395 303 240 I S5 68 13 0 MO 101 3273 0 0 0 0 0 0 NB 97 598 242 93 35 0 0 0 NB 98 1463 0 0 0 0 0 0 ND 147 55 50 48 45 35 23 10 ND 149 263 0 0 0 0 0 0 ND 151 1OS 0 0 0 0 0 0 I..-

PHl 151 40 23 13 8 . 0 0 0 PR 143 70 8 S 98 100 90 65 33 PR 145 638 40 3 0 0 0 0 PR 147 573 0 0 0 0 0 0 RB 88 S00 3 0 0 0 0 0 RB 89 121S 0 0 0 0 0 0 RH 105 78 65 40 2S 5 0 0 RH 107 40 0 0 0 0 0 0 RU 105 188 5 0 0 0 0 0 SE e81 53 0 0 0 0 0 0 SE 83 50 0 0 0 0 0 0 SN 128 U 3 0 0 0 0 0 0 SR 89 28 23 28 28 25 23 20 SR 91 795 13t 25 S 0 0 0 SR 92 132S 3 0 0 0 0 0 1 1c01 1273 0 0 0 0 a 0 TC 104 460 0 0 0 0 0 0 TE 129 95 3 0 0 0 0 0 TE 131 628 S 3 3 0 0 0 TE 132 250 203 16S 133 58 13. 0 TE 133 965 0 0 0 0 0 0 TE 134 1713 0 0 0 0 0 0 Y 92 W 48 38 0 0 0 0 0 Y 93 863 170 33 3 0 0 0 Y 94. 1583 0 0 0 0 0 0 YI 95 1613 0 .0 0 0 0 0 Y 912 423 88 IS 3 0 0 0 ZR 95 30 30 30 28 23 25 23 ZR 97 65S 245 93 35 0 0 0 Available activity (> 10 pCi) isfor a single worst-casefuel element as afunction of time after reactoroperation. Data arederivedfrom ORIGEN2.1 calculationsfCCC-371) assummarized in Appendices A and B. 7he table indudes onlythose nuclides with activities in excess of ]0 L Data are raw computatIonal results an) the number of sign4ficanifguresexceeds the precision of-the calculation.

K-State Reactor 13-21 Original (12104)

Safety Analysis Report

CHAPTER 13

L Comparison with the DAC and the ALI The ALI is the activity that, if ingested or inhaled, would lead to either (a) the maximum permissible committed effective dose equivalent incurred annually in the workplace,

.nominally 5 rem, or (b) the maximum permissible dose to any one organ or tissue, nominally 50 rem. The DAC is the air concentration that, if breathed by reference man for one work year (2000 h), would result in the intake of the ALl. ALI does not apply to noble-gas radionuclides.

Potential activity releases are compared to ALls, and air concentrations in the reactor bay are compared to DACs in Tables 13.12 and 13.13. Only for radioiodine does the

. available activity exceed the ALI. However, there is no credible scenario for accidental inhalation or ingestion of the undiluted radioiodine released from a fuel clement.

WVhen one compares with DACs the potential airborne concentration of radionuclides in the reactor bay, only the 1311, t3l, -and .1.I isotopes plus 'Kr and s5Kr are of potential consequence. However, annual-dose limits could be attained only with a constant air concentration over a long period of time. The 331 released in the failure of a single element, for example, would decay with a half-life of 8.04 days. Thus, even the undetected failure of a fuel clement would not be expected to lead to violations of the occupational dose limits expressed in IOCFR20 or in other federal guidance.

g. 'Comparison with the effluent concentration Effluent concentration, listed in the last columns of Tables 13.12 and 13.13, are defined in continuous exposure (8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months per year) rather than 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months per year occupational exposure. Exposure to a constant airborne concentration equal to the effluent concentration for one full year results in the annual dose limit of 100 mrem to members of the public. As is apparent from Tables 13.12 and 13.13, the reactor bay average concentrations immediately after fuel element failure exceed the efluent concentrations for several radionuclides. Thus, only for these radionuclides. is it necessary to consider radioactive decay and atmospheric dispersal after release in estimating potential risk to members of the public. For posting purposes, concentrations relative to DACs are additive. For dosimetry purposes, products of concentrations and times, relative to DAC-hours, are additive.

b. Potential downwind dose to a member of the public In this dose assessment, it is assumed that the available activity in a failed fuel clement is released instantaneously and immediately after reactor shutdown. It is further assumed that a member of the public is positioned directly downwind from the Nuclear Reactor Facility and remains in place during the entire passage of the airborne radioactivity. The very conservative approximations of Hawley and Kathren 11982] are adopted in the assessment, namely, that the atmospheric dispersion (XIQ) factor is 0.01 s/dm (2.78 x 104Y h/m 3 ) and the breathing rate Vis 1.2 tn/h. No credit is taken for partial containment, plateout, or other potential mitigating mechanisms, howeverrealistic and probable.

Let the activity of nuclide I released be A (pCi) as given in Tables 13.12 and 13.13. If one neglects radioactive decay, the activity inhaled during passage of the airborne K-State Reactor

13-22 Original (12104)

Safety Analysis Report

ACCIDENT ANALYSIS activity is A Y(x X/Q) [Faw and Shulnis, 19931. The product of the activity inhaled and the dose conversion factor 91 (mrem,/pCi) [Eckernan et al., 1988] yields Di (mrem), the more critical of the organ dose or the effective dose equivalent to the total body. Results of such calculations are presented in Table 13.14. As is apparent from the table, individual organ doses as well as the total committed effective dose equivalent are well below any regulatory limits. Entries are shown only for doscs of 0.001 mrem or greatcr.

Table 13.12, Comparison of Halogen and Noble Gas Available Activities Immediately After Reactor Shutdown wvith ALls and Reactor Bay Concentrations with DACs and Effuent Concentrations.

AV2!bllbc InheacatDC Mflewt Ratio Ratio to Nucice al-fie tiit Inaltin AC bay conc. COW to Eff E nt Half-R A(1-C il AU (pCi) (pDCifcm) Racm) WyCi/em') DAC Limit Br 83 239 h 12100 6.E+04 3.E-5 3.OOE-06 9E-08 0.1 333 Br 84 31.8 m 25375 6.E+04 2.E-5 6.2E 06 8E-08 0.3 78.1 I 131 8.04 d 8500 5.E+01 2.E-8 2.08E-06 2E-10 103.8 10375.0 I 132 2.30 h 23275 8.E+03 3.E-6 5.7SE-06 2E-08 1.9 287.5 I 133 20.8 h 69950 3.E+02 I.E-7 1.73E-0S IE-09 172.5 17250.0 I 134 52.6 m 192225 5.E+04 2.E-5 4.7SE-OS 6E-08 2.4 791.7 1 135 6.61 h 98750 2.E+03 *7.E-7 2.43E-05 6E-09 34.6 4041.7 Kr 83m 1.83 h 9950 1.E-2 2-4SE.06 SE-05 0.0 o&o Kr 85m 4.48 h 23375 2.E-5 S.75E 06 IE-07 0.3 s7.5 Kr 85 10.7 y S0 1.E-4 1.2SE-08 7E-07 0.0 0.0 Kr 87 76.3 mn 64500 5.E-6 LS8E-05 2E-08 3.2 787.5 Kr 88 2.84 h 79050 2.E-6 1.9SE.0S 9E.-09 9.8 21667 Xe 133m 2.19 d 0 1.E-4 1.00-08 6E-07 0.0 0.0 Xe 133 5.25 d 20350 1.E-4 5.00E-06 5E-07 0.1 10.0 Xe 135m 15.3 m 25 9.E-6 6.25E09 4E-08 0.0 0.2 Xe 135 9.09 h 58950 I.E-5 1.451-05 7E-08 1.5 207.1

'Bay concentration exceeds DAC.

bBay concentration exceeds effluent concentration.

K-State Reactor 13-23 Original (12104)

Safety Analysis Report

CHAPTER 13 Tablc 13.13, Comparison ofParticulatc Available Actlitics (>100 TpCI) ith ALIs and Reactor Bay Concentrations with DACs and EMuent Concentrations.

Avaibble Inhaltion DAC Reactorbay Emuent Ratio to Ratio to Eff Element Nuclide Half-life activity AJ pi pic" cone.~ co DAC Limit (CL) (jiCi) (Ci/cm ) (pc)

(n)

Ba 139 82.7 m 1600 75000 0.00001 4E-07 2E-09 4.OOE-02 2.00E+02 Ba 140 12.7 d 125 2500 6E-07 3E-8 2E-09 5.00E-02 1.50E+01 Ba 141 . 1475 175000 0.00003 3.5E207 1E-07 1.17E-02 3.50E+00

Ce 141 50 1500 2E-07 1.23E-08 8E-10 6.13E-02 1.53E+01 Ce 143 33.0 h 550 5000 7E-07 1.35E-07 2E-09 1.93E-01 6.75E201 Cs 138 32.2 m 1700 150000 0.00002 425E-07 8E-08 2.13E-02 5.31E+00 La 140 75 2500 5E-07 1.85E-08 2E-09 3.70E-02 925E+00 La 141 .3.93 h 1150 22500 0.000004 2.75E-07 1E-08 6.88E-02 2.75E+01 La 142 92.5 m 1450 50000 0.000009 3.5E-07 3E-08 3.89E-02 1.17E+01 La 143 1500 225000 0.00004 3.75E-07 1E-07 9.38E-03 3.75E+00 Mo 99 66.0 h 400 2500 6E-07 9.75E-08 2E-09 1.63E-01 4.88E+01 Nb 98 1475 125000 0.00002 35E-07 7E-08 1J5E-02 5.00E+00 Nd 147 1.73 h

50 2000 4E-07 1.23E-08 1E-09 3.06E-02 123E+01 Pmn 151 . 50 7500 D.000001 1.23E-08 4E-09 1.23E-02 3.06E+D0 Pr 143 75 1750 3E-07 1.BE5-8 9E-10 6.17E-02 2.06E201 Pr 145 5.98 h 650 20000 0.000003 1.6E-07 1E-08 5.33E-02 1.602+01 Rb 88-17.8 m 800-150000-0.00003 -1.5-07E79E-08-6.50E-3 -217E+00 Rb 89 1225 250000 O.006 3E2-7 2E-07 5.OOE-03 150E+00 Ru 105 4.44 h 200. 25000 000005 sE-08 2E-08 1.00E-02 2.50E+00 Se 81 - 18.5 m 50 S00000 0.00009 123E-08 3E-07 1A6E-04 4.08E-02 Sn 128 59.1 m 75 75000 0.00001 1.85E-08 4E-08 1.85E-03 4.63E-01 Sr 89 25 250 6E-08 625E-09 2E-10 1.04E-01 3.13E+01 Sr 91 9.5 h 800 10000 0.000001 1.95E-07 SE-09 1.95E2-1 3.90E+01 Sr 92 2.71 h 1325 17500 0.000003 325E-07 9E209 1.08E-01 3.61E+01 Te 129 69.6 m i10 150000 0.00003 2.45E-08 9E-08 8.17E-04 2.72E-01 Te 131 25.0m 625 .12500 0.000002 1.53E-07 1E-09 7.63E-02 1.53E202 Te 132 78.2 h 250 500 9E-08 625E-08 E2-10 6.94E-01 6.94E+01 Te 133 12.5 m- 975 50000 0.000009 2AE-07 8E-08 2.67E-02 3.00E+00 Te 134 41.8 m 1725 50000 0.00001 425£-07 7E-08 425E-02 6.07E+00 Y 91m 49.7 m 425 500000 0.00007 1.05E-07 2E-07 1.5E-03 5.25E-01 Sr . 91 9.5 h 800 10000 0.000001 1.95E-07 5E-09 1.95E01 3.90E401 Sr . 92 2.71 h 1325 17500 0.000003 32sE-07 9E-09 1.08E-01 3.61E+01 Te 129 69.6 m 100 150000 0.00003 2.45E-08 9E-08 8.17E-04 2.72E-01 Te 131 25.0 m 625 12500 0.000002 1.53E-07 1E09 7.63E-02 1.53E+02 Te 132 782 h 250 500 9E-08 625E-08 9E-10 6.94E201 6 94E+01 Te 133 12.5rm 975 50000 0.000009 2AE-07 BE-08 2.67E-02 3.00E+00 K-State Reactor 13-24 Original (12104)

Safety Analysis Report

i ACCIDENT ANALYSIS Table 13.13, Comparison ofParticulatoAvailableActivitics (>100 pqCi) vithALls and Reactor Bay Concentrations with DACs and EMuent Concentratlons.

Available InIbalation DAC Reactor bay Effuent Ratio to Ratio to Eff Element Nuclide Half-life activity AIpQ ommi cone. conc. DA Lit (pCi) A, (jCi) ( 3 ) nc

((CCtc icm, RDACt Limitc Te 134 41.8 m 1725 50000 0.00001. 4.25E-07 7E.08 4.25E-02 6.07E+00 Y 91m 49.7 mr 425 500000 0.00007 1.05E-07 2E-07 1.50E-03 5.25E-01 Y 92 3.54 h 850 20000 0.000003 2.08E-07 1E-08 6.92E-02 2.08E+01 Y 93 10.1 h 850 5000 0.000001 2.08E-07 3E-09 2.08E-01 6.92EI01 Zr .95 25 250 5E-O8 625E-09 4E-10 1.25E-01 1.56E+01 Zr 97 16.9 h 650 2500 5E-07 1.6E-07 2E-09 3.20E-01 B.OOE+01

i. Residual Activity from Fuel Utilization Prior to Receipt 1

All but a few instrumented Mark-lI fuel elements in the original 1962 core loading were replaced by Mark-Ill elements on July 10, 1973. The replacement elements had seen considerable use prior to their installation at Kansas State University. The two most heavily used elements, with serial numbers 4078 and 4079, had experienced, respectively, consumption of 11.27 and 10.33 g of 235U. Even after about 25 years of subsequent use, considerable '37Cs, "Sr. and '5Kr remain from fission during the pre-1973 use. However, the 8Kr atmospheric concentration inside thc reactor bay immediately after release would be orders of magnitude lower than the DAC.

Therceore only 37Cs and "0Sr offer a potential for occupational or public risk. In the absence of knowledge about the pattern of early fuel utilization, it is assumed that all the generation of fission products took place in 1973 and that fission product decay took place over the period of 28 years from 1973 until 2001.

If Yis the fission yield, Xis the decay constant (el), and N. is Avogadro's number, the activity A (Bq) of any one radionuclide immediately after fissioning ofmncssm (g) of 2 3U is

.A= N- mYA. (13.2.4-3) 235 Activity calculations using this formula and consequences, as computed in 13.14 are reported in Table 13.15.

K-State Reactor 13-25 Original (12104)

Safety Analysis Report

CHAPTER 13 Table 13.15. Worst Case Source Terms and Consequence Calculations for a Single TRIGA Fucl Eilemcnt Experiencing 11.27 g of 35U Consumption 28 Years Prior to Element Failure.

'RAT)TnN11f3mF FACTOR iUSr tics Half life (y) 29.12 30.00 Decay constant A.(s) 7.54 x lo"' 7.32x 10.10 Fission yield Y 0.0577 0.0615 Release fraction l.OOx 10e6 l.00x 10e Initial Bqlg contained in element

1.12x I0o" 1.15x 10' Initial PCi available for release 34.0 35.2 pCi available for release in 28 y 17.4 18.4 AU (PCi) 4 200 Reactor bay concentration (pCi/cm3 ) 4.3x 10e 4.5x 10°9 DAC (pCi/cm') 2.x 10e 6.x IO0s Tissue at risk Bone surface Total body Dose conversion factor (mrem/pCi) 2690 32 Maximum downwind dose (inrem) 0.16 0.0020 Whereas the -Sinactivity availabli for release would exceed the occupational AI and, if dispersed within the reactor bay, would have a concentration in excess of the DAC, credible mechanisms for ingestion or inhalation of the full available activity or even its full dispersion are not apparent. Thus, neither the 90Sr nor the 1'Cs would pose a significant occupational threat. Even if the total available activity were somehow dispersed to the free atmosphere, no person downwind of the accidental release would receive doses even approaching regulatory limits.

3. Conclusions Fission product inventories in TRIGA fuel elements were calculated with the ORIGEN code, using very conservative approximations. Then, potential radionuclide releases from worst-case fuel elements were computed, again using very conservative approximations. Even if it were assumed that releases took place immediately after reactor operation, and that radionuclides were immediately dispersed inside the reactor bay workplace, few radionuclide concentrations would be in excess of occupational derived air concentrations, and then only for a matter of hours or days. Only for certain nuclides of iodine would the potential release be in excess of the annual limit of intake.

However, there is no credible scenario for accidental inhalation or ingestion of the undiluted radioiodinc that might be released from a damaged fuel element.

For the residual "Sr and 37Cs remaining in fuel elements from consurmption of 235U prior to receipt of the fuel at Kansas State University, bnly the former would pose any conceivable occupational threat. However, the total "0Sr activity available for release is K-State Reactor 13-26 Original (12/04)

Safety Analysis Report

ACCIDENT ANALYSIS estimated to be at most about 4 times the ALI and there is no credible scenario for its consumption by a worker.

As far as potential consequences to the general public are concerned, only for the few radionuclides listed in Tablc 13.14, are maximum concentrations inside the reactor facility in excess of effluent concentrations listed in IOCFR20 and potential doses 0.001*

mrem or greater. However, even in the extremely unlikely event that radionuclides released from a damaged fuel element were immediately released to the outside atmosphere, very conservative calculations reveal that radionuclides inhaled by persons downwind from the release would lead to organ doses or effective doses very far below*

re gulatory limits. As is shown in Table 13.15, the same is true for residual "0Sr and 131Cs remaining in fuel elements from early operations.

13.3 Bibliography ANSIYANS-5.) -1994, "American NationalSfandardforDecayHeatPower in Light lWater Reactors, "American Nuclear Society, 1994.

ReportLA-12625-M. Version 4B, "MCNP- A GeneralMonte Carlo N-ParticleTransportCode.'

Los Alamos National Laboratory, Los Alamos, NM (1997). Briesmeister, J.F. (ed),

CCC-3 71, "ORIGEN 2.) Isotope GenerationandDepletion Code: Matrix ExponentialMethod,"

Radiation Shiclding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tcnnessce, 1991_. *.-

_____- KansasStale University-TRIGA-MkHl-ReactorHazardsSummaiy-Report,-License-R-88,-Docket 50-188 (1961) Clack, RW., .R.Fagan, W.R. Kimel, and SZ Mikhail,.

"Laminar-FlowHeatTransferforIn-LineFlowThrough UnbaifiedRodBundes,1NucL Sci.

Engg. 42, 81-88 (1970), Dwyer, O.E. and H.C. Berry.

"Nuclear Heat Transport, " International Textbook Company, Scranton, 1971, El-Walil, M.M.,.

Federal Guidance Report No. I1, Report EPA-5201/1-88020, U.S. Environmental Protection Agency, "Limiting Values ofRadionuclideIntake andAir Concentrationand Dose Conversion FactorsforInhalation,Submersion, andIngestion"Washington, DC, (1988), Eckernan, KF.,

A.B. Wolbarst, and A.C.B. Richardson.

Report LA-5885-MS `CINDER-7: An Interim Reportfor Users,,, England, T.R, et aL., Los Alamos Scientific Laboratory, Los Alamos, NM, 1976.

"RadiologlcalAssessmenl,'Prentice Hall, Englewood Cliffs, NJ., 1993, Faw, R.E., and J.K.

Shultis.

Report LA-9362, Application ofAdjusted DataIn CalculatingFissionProduceDecay Energies and Spectra," Los Alamos Scientific Laboratory, 1982.George, D.C., RJ. LaBauve, and T.RI England.

NUREG/CR-2387 (PNL-4028), 'CredibleAccident Analysesfor TRIGA and TPJGA-Fueled Reactors,I Report Pacific Northwest Laboratory, Richland Washington, 1982., Hawley, S.C., and R.L. Kathren, K-State Reactor 13-27 Original (12104)

Safely Analysis Report

CHAPTER 13 "FurndamentalsofHea andMass Transfer,' 3d ed., Wiley, New York, 1990, Incropera, F.P. and D.P. DeWitt, GA-3399.KSU TPJGA ReactorMechanicalMaIntenance and OperatingManual, General Atomics Report, 1962.

"Fission Product Analytical Source Functions," Nuclear Tcchnology 56,332-339 (1982).LaBauve, R.J., T.R. England, and D.C. George, See also Reports LA-9090-MS (1981) and LA-UR-80-3305 (1980), Los Alamos Scientific Laboratory, Los Alamos, NM.

Metals Handbook; e ed., Vol. 1,American Society for Metals, Metals Park, Ohio, 1961.

NUREG-1282, 'Safety EvaluationReport on High-Uranium Content, Low-EnrichedUranium.

Zirconium HydrideFuelsforTRIGA Reactors," Office of Nuclear Reactor Regulation, US.

Nuclear Regulatory Commission, 1987.

IVUREG-1390, "&SetyyEvaluation Report Relating o theRenewval of the OperatingLicensefor the TWGA TrainingandResearchReactoratthe University of.Arizona, 'Report NUREG-1390, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1990.

NUREG-1537, -#GuidelinesforPreparing andRevievinjApplicationsforthe Licensing ofNon-PowerReactors, Formatand Content," Report NUREG-1537 Part 1, Office of Nuclear Reactor Regulation, US. Nuclear Regulatory Commission, 1996.

"The U-Zr-HxA lloy: Its Propertlesand Use In TRIGA Fuel" Report E-l 17-833, Simnad, M.T, General Atomics Corp, 1980.

"Fvel ElementsforPulsedTPJGAResearchReactors," Nuclear Technology 28,31:56 (1976)

Simnad, M.T., F.C. Fausbec,and GB. West "LongitudinalLaminarFlov.Betveen CylindersArranged in a RegularArray,"AIChE-JournaI 5,325 (1959), Sparrow, E.M., and A.L Loeffler, and H.A. Hubbard.

NuclearSystems J: ThermalHydraulicFundamentals,Todrcas, N.E. and MS. Kazimi, Hemisphere, New York, 1990.

SafetyAnalysis Report, TGA ReactorFacility,NuclearEngineeringTeaching Laboratory, University ofTexas atAustin. Revision 1.01, Docket 50-602, May, 1991.

"Kinetic Behaviorof TFIGA Reactors, "Report GA-7882, W~est, G.B., W.L Whiltemore, J.R Shoptaugh, Jr., J.B. Dee, and C.O. Coffezr, General Atomics Corp., 1967.

K-State Reactor - 13-28 Original (12104)

Safety Analysis Report

dii APPENDICES TO CHAPTER 13

' I A Origen 2.1 input file for235 U fission at 1 W thennal power for40 years. .-

B Origen 2.1 input file for 235U fission at I W thermal power 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per day for 5 days.

C Origen 2.1 output file extracts for " 5 U fission at I W thermal power for 40 years.

D Origen 2.1 output file extracts for 23sU fission at I W thermal power 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per day for 5 d~l Days.

E Maximum activity available for release from a single TRIGA fuel clement as a function of

'J time after shutdown for a 235U-fuelcd thermal reactor operating at 3.5 kW thermal power for, 40 years, based on one element of 81 at 86.42 W. J1 F Maximum activity available for release from a single TRIGA fuel element as a function of J1 time after shutdown for a 23SU-fuclcd thermal reactor operating at 500 kW kW thermal power J1 for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per day for S days, based on one element of 83 at 12.05 kW 1-1-

K-State Reactor Original (10104)

Safety Analysis Report

CHAPTER 13 APPENDIX A ORIGEN Input file for 1 tonne U-235 at 1 watt for 40 years

-1

-1 RDA ORIGEN2, VERSION 2.1 (8-1-91) TRIGA REACTOR REFERENCE PROBLEM RDA UPDATED BY: Richard E. Faw, Kansas State University BAS ONE TONNE OF U-235 RDA Continuous operation for 40 years at I watt RDA WARNING: VECTORS ARE OFTEN CHANGED WITH RESPECT TO THEIR CONTENT.

RDA THESE CHANGES WILL BE NOTED ON RDA CARDS.

CUT -1 RDA LIBRARY PRINT (1-PRINT,0-DON'T PRINT)

LIP DOD

. RDA DECAY LIBRARY CHOICES (0-PRINT: 1 2 3 DECAY LIBRARIES; 601 ...

RDA CROSS SECTIONS; ETC, SEE P'. 47)

LIB 0 1 2 3 201 202 203 9 3 0 1 38 PHO 101 102 103 10 <<< PHOTON LIBRARIES, P. 47 TIT INITIAL COMPOSITIONS OF UNIT AMOUNTS OF FUEL AND STRUCT MAT'LS RDA READ FUEL COMPOSITION INCLUDING IMPURITIES (1 G)

INP -1 1 -1 -1 1 1 TIT IRRADIATION OF ONE TONNE U-235 MOV -1 1 0 1.0 HED 1 CHARGE BUP IRP S .000001 1 2 5 2 1 W/Tonne FOR 5 YEARS IRP 10 .000001 2 2 5 0 1 W/Tonne FOR 5 YEARS IRP 15 .000001 1 2 5 0 1 W/Tonne FOR 5 YEARS IRP 20 .000001 2 2 5 0 1 W/Tonne FOR 5 YEARS IRP 25 .000001 1 2 5 0 I1 W/Tonne FOR 5 YEARS IRP 30 .000001 2 1 5 0 W/Tonne FOR 5 YEARS IRP 35 .000001 1 2 5 0 .1 W/Tonne FOR -5 YEARS IRP 40 .000001 2 1 5 0 I W/Tbnne FOR 5 YEARS .

BUP OPTL 24*8 ACTIVATION PRODUCT OUTPUT OPTS P. 56 OPTA 24*8 ACTINIDE OUTPUT OPTIONS P. 59 OPTF 6*8 5 17*8 FISSION PRODUCT OUTPUT OPTIONS P. 59 RDA DECAY TO 28 DAYS DEC 1 1 2 4I 2 DEC 2 2 3 4I 0 DEC 3 3 4 4I 0 DEC 7 4 5 4I 0 DEC 14 5 6 4I 0 DEC 28 6 7 4I 0 OUT -7 1 -1 0 OUT 7 1 -1 0 END 2 922340 0.0 922350 l.E06 922380 0. 0 0.0 PURE U-235 0

K-State Reactor 13A-1 Original (10104)

Safety Analysis Report

CHAPTER 13 APPENDIX B ORIGEN Input File for 1 tonne U-235 at 1 watt 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per day for 5 days

.I

-1

-1 RDA ORIGEN2, VERSION 2.1 (8-1-91) TRIGA REACTOR REFERENCE PROBLEM RDA UPDATED BY: Richard E. Faw, Kansas State University BAS One tonne U-235 RDA Fuel composition CUT -1 RDA LIBRARY PRINT (l-PRINT,0-DON'T PRINT)

LIP 0 0 0 RDA DECAY LIBRARY CHOICES (0-PRINT; 1 2 3 DECAY LIBRARIES; 601 RDA CROSS SECTIONS; ETC, SEE P. 47) 1*

LIB 0 1 2 3 201 202 '203 9 3 0 1 38 -

PHO 101 102 103 10 <<c PHOTON LIBRARIES, P. 47 TIT INITIAL COMPOSITIONS OF UNIT AMOUNTS OF FUEL AND STRUCT MAT'LS RDA READ FUEL COMPOSITION INCLUDING IMPURITIES INP -1 1 -1 -1 1 1 TIT One tonne U-235 8 hId for 5 days at 1 kWt MOV -1 1 0 1.0 HED I CHARGE BUP IRP 8.0 0. .001 1 2 3 2 OPERATE FOR 8 UR AT 1 kW DEC 24.0 2 3 3 0 COOL FOR 16 HOURS IRP 32.0 0..001 3 4 3 0 OPERATE FOR 8 HR DEC 48.0 4 5 3 0 COOL FOR 16 HOURS IRP 56.0 0..001 5 6 3 0 OPERATE FOR 8 HR DEC 72.0 6 7 3 0 COOL FOR 16 HOURS IRP 80.0 0..001 7 a 3 a OPERATE FOR 8 HR DEC 96.0 8 9 3 0 COOL FOR 16 HOURS IRP 104.0 0..001 9 10 3 0 OPERATE FOR 8 HR OPTL 24*8 ACTIVATION PRODUCT OUTPUT OPTS P. 56 OPTA 24*8 ACTINIDE OUTPUT OPTIONS P. 59 OPTF 6*8 5 17*8 FISSION PRODUCT OUTPUT OPTIONS P. 59

RDA MOVE COMPOSITION VECTOR FROM 10 TO 1 MOV 10

1 0 1.0 RDA DECAY TO 0.1 UNITS (2-MINUTES) FROM COMP VEC 1 TO VEC 3 DEC 1 1 2 4 2 DEC 2 2 3 4 0 DEC 3 3 4 4 0 DEC 7 4 5 4 0 DEC 14 5 6 4 0 DEC 28 6 7 4 0 OUT -7 1 -1 0 OUT 7 1 -1 0 END 2 922340 0.0 922350 1.E6 922380 0.00 0 0.0 1 g U-235 0

K-State Reactor 13.B-1 Original (10104)

Safety Analysis Report

CHAPTER 13 APPENDX C ORIGEN Output File Extracts for 1 tonne U-235 at I watt for 40 Years NUCLIDE TABLE: RADIOACTIVITY, CURIES Time post discharge 0 1.0 D 2.0 D 3.0 D 7.0 D 14.0 D 28.0 D AG 111 1.65E-04 1.51E-04 1.37E-04 1.25E-04 8.62E-OS 4.50E-05 1.22E-05 BA 140 5.23E-02 4.96E-02 4.70E-02 4.45E-02 3.58E-02 2.45E-02 1.25E-02 BA 137 2.93E-02 2.93E-02 2.93E-02 2.93E-02 2.93E-02 2.93E-02 2.92E-02 CE 141 4.93E-02 4.85E-02 4.75E-02 4.65E-02 4.27E-02 3.68E-02 2.73E-02 CE 143 4.98E-02 3.03E-02 1.83E-02 1.21E-02 1.47E-03 4.32E-05 3.72E-08 CE 144 4.56E-02 4.55E-02 4.54E-02 4.53E-02 4.48E-02 4.41E-02 4.26E-02 CS 137 3.10E-02 3.10E-02 3.09E-02 3.09E-02 3.09E-02 3.09E-02 3.09E-02 EU 155 2.75E-04 2.75E-04 2.75E-04 2.75E-04 2.75E-04 2.74t-04 2.72E-04

£U 156 1.13E-04 1.l0E-04 1.06E-04 1.01E-04 8.43E-05 6.13E-05 3.23E-05 I 131 2.37E-02 2.20E-02 2.03E-02 1.87E-02 1.33E-02 7.28E-03 2.18E-03 I 132 3.56E-02 2.95E-02 2.39E-02 1.93E-02 8.23E-03 1.86E-03 9.45E-05 I 133 5.68E-02 2.62E-02 1.18E-02 5.29E-03 2.16E-04 8.00E-07 1.10E-11 I 135 5.31E-02 4.29E-03 3.46E-04 2.80E-05 1.19E-09 2.66E-17 1.34E-32 KR 85 2.10E-03 2.10E-03 2.10E-03 2.10E-03 2.10E-03 2.10E-03 2.09E-03 KR 85H 1.07E-02 2.64E-04 6.44E-06 1.57E-07 5.58E-14 2.88E-25 0.00E+00 LA 140 5.24E-02 5.19E-02 5.06E-02 4.99E-02 4.08E-02 2.82E-02 1.32E-02 LA 141 4.93E-02 7.77£-04 1.13£-05 1.64E-07 7.27E-15 9.89E-28 0.00E+00 MO 99 5.06E-02 3.94E-02 3.06E-02 2.38E-02 8.68E-03 1.49E-03 4.36E-05 Im 95 5.38E-02 5.38E-02 5.39E-02 5.37E-02 5.35E-D2 5.28E-02 5.04E-02 N3 97. 4.93E-02 1.85E-02 6.90E-03 2.58E-03 5.03E-05 5.14E-08 5.70E-14 NB 95M 3.78E-04 3.77E-04 3.76E-04 3.74E-04 3.64E-04 .3.41E-04 2.95E-04 NB 97H 4.66E-02 1.74E-02 6.51E-03 2. 43E-03 4.74E-05 4.84E-08 5.01E-14 UD 147 1.90E-02 1.79E-02 1.68E-02 1.58E-02 1.23E-02 7.91E-03 3.29E-03 PH 147 1.91E-02 1.91E-02 1.91£-02 1.91E-02 1.91E-02 .1.90E-02 1.89E-02 PH 149 9.11E-03 6.89E-03 5.04E-03 3.68E-03 1.05E-03 1.17E-04 1.46E-06 PH 151 3.52E-03 1.97E-03 1.10E-03 6.11E-04 5.86E-05 9.70E-07 2.65E-10 PR 143 4.98E-02 4.93E-02 4.80E-02 4.64E-02 3.86E-02 2.71E-02 1.33E-02 PR 144 4.56E-02 4.55E-02 4.54E-02 4.53E-02 4.48E-02 4.41E-02 4.26E-02 PR 145 3.29E-02 2.06E-03 1.27£-04 7.89E-06 1.16E-10 4.06E-19 5.14E-36 PR 144 5.4BE-04 5.46E-04 5.45E-04 5.43E-04 5.38E-04 5.29E-04 .5.11£-04 RH 105 8.53E-03 6.09E-03 3.83E-03 2.39E-03 3.64E-04 1.35E-05 1.86E-08 RH 106 3.27E-03 3.26E-03 3.26E-03 3.25E-03 3.23£-03 3.18E-03 3.10E-03 RH 103 2.37E-02 2.33E-02 2.29E-02 2.25E-02 2.09E-02 1.85E-02 1.45E-02 RU 103 2.63E-02 2.58E-02 2.54E-02 .2.49E-02 2.32E-02 2.05E-02 1.60E-02 RU 105 8.53E-03 2.08E-04 4.90E-06 1.16E-07 3.57E-14 1.44E-25 0.00E+00 RU 106 3.27E-03 3.26E-03 3.26E-03 3.25E-03 3.236-03 3.18E-03 3.10E-03 SB 125 2.49E-04 2.49E-04 2.49E-04 2.49E-04 2.48E-04 2.47E-04 2.45E-04 SB 127 1.10E-03 9.28E-D4 7.75£-04 6.47E-04 3.15E-04 8.93E-05 7.18E-06 SB 129 5.33E-03 1.15E-04 2.45E-06 5.20E-08 1.06E-14 2.08E-26 O.ooE+00 SM 251 9.34E-04 9.34E-04 9.34E-04 9.346-04 9.34E-04 9.33E-04 9.33E-04 SH 153 1.36E-03 9.55E-04 6.69E-04 4.68E-04 1.13E-04 9.30E-06 6.34E-08 SN 125 1.13E-04 1.05E-04 9.75E-05 9.07E-05 6.80E-05 4.116-05 1.50E-05 SR 89 4.05E-02 3.99E-02 3.94E-02 3.89E-02 3.68E-b2 3.34E-02 2.76E-02 SR 90 2.97£-02 2.97£-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 SR 91 4.94E-02 8.59£-03 1.49E-03 2.59E-04 2.35E-07 1.12E-12 2.51E-23 SR 92 5.04E-02 1.09E-04 2.35E-07 5.07E-10 1.10E-20 0.00E+00 0.00E+00 TC 99M 4.43E-02 3.76E-02 2.95E-02 2.29E-02 E.36E-03 1.43E-03 4.20E-OS SE 127 1.09E-03 1.02E-03 8.88E-04 7.68E-04 4.49E-04 2.27E-04 1.37E-04 SE 129 5.25E-03 6.43E-04 4.99E-04 4.96E-04 4.48E-04 3.88E-04 2.90E-04 SE 131 2.13£-02 3.95E-04 2.27E-04 1.30E-04 1.426-05 2.92E-07 1.24E-10 SE 132 3.54E-02 2.86E-02 2.32E-02 1.87E-02 7.99E-03 1.80E-03 9.17E-05 SE 127 1.53E-04 1.53E-04 1.52E-04 1.52E-04 1.50E-04 1.44E-04 1.33E-04 SE 129 7.91E-04 7.78E-04 7.63E-04 7.47E-04 6.86E-04 5.95E-04 4.46E-04 SE 131 3.04E-03 1.75E-03 1.01E-03 5.79E-04 6.30E-05 1.30E-06 5.52E-10 XE 133 5.68E-02 5.47E-02 5.03E-02 4.52E-02 2.74E-02 ).09E-02 1.72E-03 XE 135 5.52E-02 2.02E-02. 4.15E-03 7.39E-04 5.35E-07 1.47E-12 1.09E-23 XE 131 2.63E-04 2.62E-04 2.616-04 2.58E-04 2.41E-04 1.97E-04 1.11E-04 XE 133 1.65E-03 1.51E-03 1.24E-03 9.65E-04 2.96E-04 3.27E-05 3.96E-07 XE 135 9.52£-03 6.976-04 5.55E-05 .4.48E 1.91E-20 4.27E-18 2.14E-33 Y 90 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 2.97E-02 Y 91 4.94£-02 4.91E-02 4.86E-02 4.0E6-02 4.SE-02 4.21E-02 3.57E-02 Y 92 5.05E-02 1.60E-03 1.70E-05 1.60E-07 1.l0E-15 5.65E-30 0.00E+00 Y 93 5.44E-02 1.06£-02 2.04E-03 3.93E-04 5.41E-07 5.32E-12 5.15E-22 Y 91M 2.86E-02 5.46£-03 9.47£-04 1.64E-04 1.49E-07 7.09E-13 1.60E-23 ZR 95 5.38E-02 5.32E-02 5.26E-02' 5.21E-02 4.99E-02 4.62E-02 3.97E-02 ZR 97- 4.92E-02 1.84E-02' 6.87E-03 2.57E-03 5.00E-05 5.1E-08 5.29E-14 K-State Reactor 13.C-1 Odginal (10104)

Safety Analyss Report

d.l CHAPTER 13 APPENDIX D ORIGEN Output File Extracts for 1 tonne U-235 at 1 H for 8 hid, 5 Days NUCLIDE TABLE: RADIOACTIVITY, CURIES Time post discharge 0 1.0 D 2.0 D 3.0 D 7.0 D 14.0 D 28.0 D AS 78 1.24E-01 2.17E-05 5.66E-10 1.16E-14 1.11E-33 0.00E+00 0.00E+00 EA 139 5.33E+01 3.47E-04 1.99E-09 1.14E-14 1.28E-35 O.OOE+00 O.OOE+00 BA 140 4.22E+00 4.00e+00 3.79E+00 3.59E+00 2.89E400 1.98E+00 9.27E-01 BA 141 4.91E+01 O.OOE+00 0.00E+00 0.00E+00 0.00E400 O.OOE+00 O.OOE+00 BA 142 4.88E+01 O.OOE+00 0.00E+00 O.OOE+00 0.00E400 O.OOE+00 0.00E+00 8R 83 4.02E+00 4.13E-03 3.92E-06 3.72E-09 3.01E-21 0.00E+00 0.00+00 BR 84 8.43E+00 2.19E-13 5.13E-27 O.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 CE 141 1.57E+00 1.74E+00 1.71E+00 1.67E400 1.54E+00 1.321+00 9.81E-01 CE 143 1.84E+01 1.13E+01 6.83E+00 4.13E+00 5.49E-01 1.61E-02 1.39E-os CE 144 1.85E-01 1.84E-01 1.84E-01 2.83E-01 1.82E-01 1.79E-0O 1.73E-01 Cs 138 5.68E+0O 3.40E-12 1.17E-25 0.00E+00 0.00E+00 O.OOE+00 O.OOE+00 I 131 2.82E+00 2.74E+00 2.55E+00 2.37E+00 1.70E+00 9.33E-01 2.79E-01 I 132 7.73E+00 6.95E400 5.62E+00 4.54E100 1.94E+00 4.371E-01 2.22E-02 I 133 2.32E+01 1.11E+01 5.00C+00 2.25E+00 9.17E-02 3.40E-04 4.66E-09 I 134 6.38E+01 1.62E-06 9.36E-15 5.37E-23 0.00+00 0.00+00 O.OOE+00 I 135 3.28E+01 2.65E+00 2.14E-01 1.73E-02 7.35E-07 1.65E-14 8.26E-30 KR 87 2.14E+01 4.51E-05 9.39E-11 1.96E-16 0.00E+00 0.001+00 0.00E+00 KR 88 2.62E+01 7.49E-02 2.14E-04 6.09E-07 4.03E-17 6.19E-35 0.00E+00 KR 83M 3.30E+00 1.59E-02 1.65E-05 1.58E-08 1.29E-20 0.00+00 0.00E+00 KR 85M 7.76E+00 1.93E-01 4.70E-03 1.15E-04 4.07E-ll 2.10E-22 0.00E+00 LA 140 2.25E+00 2.88E+00O 3.22E+00 3.38E+00 3.19E+00 2.27E+00 1.07E+00 LA 141 3.79E+01 6.111-01 8.86E-03 1.29E-04 5.72E-12 7.77E-25 0.001+00 LA 142 4.83E+01 1.15E-03 2.43E-08 3.12E-13 1.01E-31 0.00+00 0.00E+00 LN 143 4.96E+01 O.OOE+00 0.00E+00 0.001+00 0.00+00 O.OOE+00 0.00E+00 O 99 1.31E+01 1.02E+01 7.93E+00 6.17E+00 2.25+00 3.85E-01 1.13E-02 HO 101 4.23E+01 O.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 O.OO+00 0.00E400 NB 97 1.98E+01 8.19E+00 3.06E+00 1.14E+00 2.23E-02 2.28E-05 2.53E-11 NB 98 4.86E+01 0.00E+00 O.OOE+00 O.OOE+00 0.00+00 0.00+00 0.00E+00 No 147 1.79E+00 1.69E+00 1.59E+00 1.49E+00 1.26E+00 7.50E-01 3.12E-01 ND 149 8.74E+00 5.96E-04 3.98E-08 2.65E-12 5.23t-29 O.OO+00 0.00E400 ND 151 3.50E+00 O.00E+00 O.OOE400 O.OOE+00 O.OOE400 0.001+00 0.00E+00 PD 109 1.19E-01 3.50E-02 1.02E-02 2.95E-03 2.11E-05 3.69E-09 1.13E-16 PH 151 1.36E+00 7.72E-01 4.30E-01 2.39E-01 2.30E-02 3.80E-04 1.04E-07 PR 143 2.31E+00 2.92E+00 3.22E+00 3.33E+00 3.03E+00 2.16E+00 1.061E00 PR 144 1.86E-01 1.84E-01 1.84E-01 1.83E-01 1.82E-01 1.79£-01 1.73E-01 PR 145 2.12E+01 1.33E+00 8.25E-02 5.l1E-03 7.52E-08 2.63E-16 3.22E-33 PR 147 1.90E+01 O.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 RB 88 2.65E+01 8.37E-02 2.39E-04 6.80E-07 4.50E-17 7.40E-35 O.OOE+00 RH 105 2.57E+00 2.17E+00 1.37E+00 8.55E-01 1.30E-01 4.84£-03 6.68E-06 RH 107 1.37E+00 1.84E-20 0.001E00 O.OOE+00 0.00E+00 0.00E+0O 0.00E+00 RU 103 7.45E-01 7.33E-01 7.20E-01 7.08E-01 6.59E-01 5.83E-0l 4.55E-01 RU 105 6.23E+00 1.54E-01 3.62E-03 8.54E-05 2.64E-11 1.07E-22 0.00+00 SB 127 2.19E-01 1.96E-01 1.63E-01 1.36E-01 6.64E-02 1.88E-02 1.51E-03 SE 81 1.76E+00 2.49E-09 6.77E-17 1.84E-24 O.OOE+00 O.OOE+00 0.00E+00 SE 83 1.64E+00 8.98E-20 O.OOE+00 0.00E+00 0.00+00 0.00+00 O.OOE+00 SM 153 4.27E-01 3.01E-01 2.11E-01 1.49E-01 3.55E-02 2.93E-03 2.00E-OS SM 155 2.76E-01 8.53E-21 0.00E+00 O.OOE+00 O.OOE00 0.00t+00 0.00E+00 SN 127 6.52E-0O 2.37£-04 8.59E-08 3.12E-11 5.40E-25 0.00+00 0.00E+00 SN 128 2.78E+00 1.25-07 5.62E-15 2.53E-22 O.OOE00 O.OOEt00 0.001E00 SR 89 9.30E-01 9.27E-01 9.15E-01 9.021-01 8.54E-01 7.76E-01 6.40E-01 SR 91 2.64£+01 4.60E+00 7.99E-01 1.39E-01 1.26E-04 5.98E-10 1.35E-20 SR 92 4.40E+01 9.50E-02 2.05E-04 4.42E-07 9.59E-18 1.621-36 0.00E+00 TC 101 4.23E+01 0.00E+00 0.00E+00 O.OOE+00 0.00E+00 O.OOE+00 0.00E+00 TC 104 1.53E+01 0.00E+00 0.001E00 0.001E+0 0.00E+00 O.OOE+00 0 .00E+00 TE 127 1.61E-01 1.77E-01 1.55-01 1.31E-01 6.49E-02 1.94E-02 2.81E-03 TE 129 3.11E+00 1.18E-01 1.87E-02 1.62E-02 1.49E-02 1.29E-02 9.65E-03 TE 131 2.09E+01 1.51£-01 8.68E-02 4.99E-02 5.43E-03 1.12E-04 4 .76E-08 TE 132 8.32E+00 6.74E100 5.45E+00 4.411+00 1.88E+00 4.241-01 2.16E-02 TE 133 3.20E+01 6.42E-08 9.61E-16 1.44E-23 O.OOE+00 O.OOE+00 O.OOE+bO TS 134 5.69E+01 2.42E-09 1.03E-19 4.401-30 0.00E+00 0.O00E+0 0.00E+00 xe 133 6.76E+00 8.00£00 7.97Et00 7.43E+00 4.68E+00 1.88E+00 2.97E-01 XE 135 1.96E+01 1.01E+01 2.19E+00 3.97E-01 2.91E-04 7.99E-10 5.94E-21 XE 138 5.26E+01 0.00+00 O.OOE00 0.00+00 O.OOE+00 0.00E+00 0.00E+00 y 91 7.71E-01 9.16E-01 9.32E-01 9.25E-01 8.84E-01 8.13E-01 6.89E-01 Y 92 2.81E+01 1.25E+00 1.35E-02 1.28E-04 8.81E-13 4.51E-27 0.00t00 Y 93 2.86E+01 5.64E+00 1.09E+00 2.09E-01 2.88E-04 2.83E-09 2.74E-19 Y 94 5.26E+01 0.001E00 0.001400 O.OOE+00 O.OOE+00 0.00f00 0.00E+00 Y 95 5.35E+01 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 K-Stale Reactor 13.D-1 Original (10104)

Safety Analysis Report

CHAPTER 13 -. A.

y 92H 1.42E+D2 2.93E+DO 5.VBE-O1 8.S1E-02 8.00E-D5 3.80E-10 8.56E-21 ZR 95 9.72E-O1 9.68E-01 9.58E-01 9.47E-01 9.07C-01 8.41E-01 7.23E-02 ZR 97 2.28E+O2 8.25E+OO 3.05E+DO 1.14E+00 2.22E-02 2.27E-05 2.34E-21 K-State Reactor 13.D-2 Original (101/04)

Safety Analysis Report

ACCIDENT ANALYSIS CHAPTER 13 APPENDIX E Maximum Activity Available for Release One SRIGA Element at 86.42 W for 40 Years (Release fractions: 1E-04 for halogens and noble gases, 1E-06 for particulates)

Potential Activity Release Ipci) Inhalation Initial Bay Sime post discharge (days) ALI DAC Concentration 2

0 1 2 3 7 14 28 pci pCi/cm pci/=m AG 111 0.0 0.0 0.0 0.0 0.0 0.0 0.0 9.E+02 4.E-07 3.5E-12 BA 140 4.5 4.3 4.1 3.8 3.1 2.1 1.0 1.E+03 6.E-07 1.1E-09 BA 137 2.5 2.5 2.5 2.5 2.5 2.5 2.5 na na 6.2E-10 CE 141 4.3 4.2 4.1 4.0 3.7 3.2 2.4 6.E+02 2.E-07 1.0E-09 CE 143 4.3 2.6 1.6 1.0 0.1 0.0 0.0 2.E+03 7.E-07 1.1E-09 CE 144 3.9 3.9 3.9 3.9 3.9 3.8 3.7 l.E+01 6.E-09 9.7E-10 Cs 137 2.7 2.7 2.7 2.7 2.7 2.7 2.7 2.E+02 6.E-08 6.6E-10 EUi 155 0.0 0.0 0.0 0.0 0.0 0.0 0.0 9.E+01 4.E-08 5.8E-12 EU 156 0.0 0.0 0.0 0.0 0.0 0.0 0.0 S.E+02 2.E-07 2.4E-12 I 131 204.6 190.1 175.3 161.4 114.9 62.9 18.s S.E+01 2.E-08 5.OE-08 I 132 307.3 255.3 206.4 166.8 71.2 16.0 0.8 8.E+03 3.E-06 7.5E-08 I 133 490.5 226.5 101.8 45.7 1.9 0.0 0.0 3.E+02 1.E-07 1.2E-07 I 135 459.6 37.0 3.0 0.2 0.0 0.0 0.0 2.E+03 7.E-07 1.lE-07 KR 85 18.1 18.1 18.1 18.1 18.1 18.1 18.1 1.E-04 4.5E-09 KR 85H 92.2 2.3 0.1 0.0, 0.0 0.0 0.0 2.E-05 . 2.3E-08 LA 140 4.5 4.5 4.4 4.2 3.5 2.4 1.1 l.Et03 5.E-07 1.1E-09 LA 141 4.3 0.1 0.0 0.0 0.0 0.0 0.0 9.E403 4.E-06 1.OE-09 HO 99 4.4 3.4 2.6 2.1 0.7 0.1 0.0 1.E+03 6.E-07 1.1E-09 NB 95 4.6 4.6 4.6 4.6 4.6 4.6 4.4 1.E+03 S.E-07 1.1E-09 NB 97 4.3 1.6 0.6 0.2 0.0 0.0 0.0 7.E+04 3.E-05 1.OE-09 NB 95H 0.0 0.0 0.0 0.0 0.0 0.0 0.0 2.E+03 9.E-07 8.0E-12 ND 97H 4.0 1.5 0.6 0.2 0.0 0.0 0.0 na na 9.9E-10 ND 147 1.6 1.5 1.4 1.4 1.1 0.7 0.3 8.E+02 4.E-07 4.0E-10 PM 147 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.E102 5.C-08 4.0E-10 PM 149 0.8 0.6 0.4 0.3 0.1 0.0 0.0 2.E+03 8.E-07 1.9E-10 PM 151 0.3 0.2 0.1 0.1 0.0 0.0 0.0 3.E+03 1.E-06 7.5E-11 PR 143 4.3 4.3 4.2 4.0 3.3 2.3 1.1 7.E+02 3.E-07 1.lE-09 PR 144 3.9 3.9 3.9 3.9 3.9 3.8 3.7 1.E+05 S.E-05 9.7E-10 PR 145 2.8 0.2 0.0 0.0 0.0 0.0 0.0 8.1+03 3.E-06 7.0E-10 PR 144 0.0 0.0 0.0 0.0 0.0 0.0 0.0 na na 1.2E-11 RH 105 0.7 0.5 0.3 0.2 0.0 0.0 0.0 6.E+03 2.E-06 1.8E-10.

RH - 106 0.3 0.3 0.3 0.3 0.3 0.3 0.3 ma na 6.9E-l2 RH 103 2.0 2.0 2.0 1.9 1.8 1.6 1.2 1.Et06 5.E-04 5.OE-10 RU 103 2.3 2.2

2.2 2.2 2.0 1.8 1.4 6.Et02 3.E-07 5.6E-10 PM 105 0.1 0.0 0.0 0.0 0.0 0.0 0.0 1.E+04 5.E-06 1.SE-10 BU 106 0.3 0.3 0.3 0.3 0.3 0.3 0.3 l.E+Ol 5.E-09 6.9E-11 SB 125 0.0 0.0 0.0 0.0 0.0 0.0 0.0 5.E+02 2.E-07 5.3E-12 sa 121 0.1 0.1 0.1 0.1 0.0 0.0 0.0 9.E+02 4.E-07 2.3E-11 SB 129 0.5 0.0 0.0 0.0 0.0 0.0 0.0 9.E+03 4.E-06 1.1E-10 SM 151 0.1 0.1 0.1 0.1 0.1 0.1 0.1 1.E+02 4.E-08 2.OE-11 SM 153 0.1 0.1 0.1 0.0 0.0 0.0 0.0 3.E103 1.E-06 2.9E-11 SN 125 0.0 0.0 0.0 0.0 0.0 0.0 0.0 4.E402 1.E-07 2.4E-12 SR 89 3.5 3.5 3.4 3.4 3.2 2.9 2.4 1.E102 6.E-08 8.6E-10 SR 90 2.6 2.6 2.6 2.6 2.6 2.6 2.6 4.E+00 2.E-09 6.3E-10 SR 91 4.3 0.7 0.1 0.0 0.0 0.0 0.0 4.E+03 1.E-06 1.OE-09 SR 92 4.4 0.0 0.0 0.0 0.0 0.0 0.0 7.E103 3.E-06 1.1E-09 TC 99M 3.8 3.3 2.5 2.0 0.7 0.1 0.0 2.E105 6.E-OS 9.4E-10 TE 127 0.1 0.1 0.1 0.1 0.0 0.0 0.0 2.E+04 7.E-06 2.3E-11 SE 129 0.5 0.1 0.0 0.0 0.0 0.0 0.0 6.E404 3.E-05 1.1E-10 TE 131 1.8 0.0 0.0 0.0 0.0 0.0 0.0 5.E103 2.E-06 4.5E-10 TE 132 3.1 2.5 2.0 1.6 0.7 0.2 0.0 2.E+02 9.E-08 7.5E-10 SE 127 0.0 0.0 0.0 0.0 0.0 0.0 0.0 3.E+02 1.E-07 3.2E-12 1.*

SE 129 0.1 0.1 0.1 0.1 0.1 0.1 0.0 2.E102 1.E-07 1.7E-11 SE 131 0.3 0.2 0.1 0.1 0.0 0.0 0.0 4.E+02 2.E-07 6.4E-l1 XE 133 490.7 473.1 434.5 390.3 236.8 94.5 14.9 1.E-04 1.2E-07 Xz 135 477.0 114.2 35.8 6.4 0.0 0.0 0 .0 1.E 1.2E-07 XE 131 2.3 2.3 2.3 2.2 2.1 1.7 1.0 4.E-04 5.6E-10 XE 133 14.3 13.1 10.7 8.3 2.6 0.3 0.0 1.E-04 3.5E-09 XE 135 82.3 5.9 0.5 0.0 0.0 0.0 0.0 9.E-06 2.OE-08 Y 90 2.6 2.6 2.6 2.6 2.6 2.6 2.6 1.E+02 5.E-08 6.3E-10 Y 91 4.3 4.2 4.2 4.1 4.0 3.6 3.1 8.E+03 3.E-06 1.0E-09 Y 92 4.4 0.1 0.0 0.0 0.0 0.0 0.0 2.E+03 1.E-06 1.1E-09 Y 93 4.7 0.9 0.2 0.0 0.0 0.0 0.0 2.E+03 1-.E-06 1.2E-09 Y 91M 2.5 0.5 0.1 0.0 0.0 0.0 0.0 2.E+05 7.E-05 6.1E-10 ZR. 95 4.6 4.6 4.5 4.5 4.3 4.0 3.4 I.E+02 5.E-0S 1.1E-09 ZR 97 4.3 1.6 0.6 0.2 0.0 0.0 0.0 1.E+03 5.E-07 1.0E-09 K-Stale Reactor 13-3 Original (12104)

Safety Analysis Report

CHAPTER 13 APPENDIX F Maximum Activity Available for Release One TRIGA Element at 31.125 kW, 8 h/d. 5 Days (Release fractions: 2E-04 for halogens and noble gases, 2E-06 for particalates) i '.

AS 78

3.75 0 0 0 0 0 0 2.E+04 9.E-06 9.25E-10 NA BA 139 1604.3 0 0 0 0 0 3.E+04 1.E-05 4.OOE-07 NA BA 140 127.25 120.5 114.3 108.3 87.25 59.5 28 1.E+03 6.E-07 3.00E-08 NA BA 141 1480.3 0 0 0 0 0 0 7.E+04 3.E-05 3.75E-07 NA BA 142 1470.5 0 0 0 0 0 0 1.E+05 6.E-OS 3.50E-07 NA BR 83 12110 12.5 0 0 0 0 0 6.E+04 3.E-OS 3.00E-06 NA BR 84 25386 0 0 0 0 0 0 6.E+04 2.E-OS 6.25E-06 NA CE 141 47.25 52.5 51.5 50.25 46.25 39.75 29.5 6.E+02 2.E-07 1.15E-08 NA CE 143 553.5 340.75 205.8 124.3 16.5 0.5 0 2.E+03 7.E-07 1.35E-07 NA CE 144 5.5 5.5 5.5 5.5 5.5 5.5 5.25 2.E+01 6.E-09 1.38E-09 NA CS 138 1711.9 0 0 0 0 0 0 6.E+04 2.E-05 4.25E-07 NA 1 131 8489.3 8245.3 7694 7131 5124 2810 840.5 S.E+01 2.E-08 2.08E-06 103.8 1 132 23281 20943 16930 13686 5838 2317 67 8.E+03 3.E-06 5.75E-06 1.9 I 133 69950 33529 15072 6772 276.3 1 0 3.E+02 1.E-07 1.73E-05 172.5 1 134 192228 0 0 0 0 0 0 5.E+04 2.E-05 4.75E-05 2.4 1 135 98750 7980 644.3 52 0 0 0 2.E+03 7.t-07 2.43E-05 34.6 KR 87 64498 0.25 0 0 0 0 0 na 5.E-06 1.58E-05 3.2 RR 88 79048 225.75 0.75 0 0 0 0 na 2.E-06 1.95E-05 9.8 KR 83H 9953.3 47.75 0 0 0 0 0 na 1.E-02 2.45E-06 NA KR 85H 23368 580.25 14.25 0.25 0 0 0 na 2.E-05 5.75E-06 NA LA 140 67.75 86.75 97 101.8 96 68.5 32.25 1.E+03 5.E-07 1.68E-08 NA LA 141 1140.3 18.5 0.25 0 0 0 0 9.E+03 4.E-06 2.75E-07 NA LA 142 1454.5 0 0 0 0 0 0 2.E+04 9.E-06 3.50E-07 NA LA 143 1493.3 0 0 0 0 0 0 9.2+04 4.E-05 3.75E-07 NA H0 99 395.5 307.5 239 185.8 67.75 11.5 0.25 1.E+03 6.E-07 9.75E-08 NA Ho 101 1272.8 0 0 0 0 0 0 1.E+05 6.E-05 3.00E-07 NA NB 97 596.75 246.75 92.25 34.5 0.75 0 0 7.E+04 3.E-05 1.48E-07 NA NS 98 1463.5 0 0 0 0 0 0 5.E+04 2.E-05 3.50E-07 NA ND 147 53.75 51 48 45 35 22.5 9.5 8.E+02 4.E-07 1.33E-08 NA ND 149 263.25 0 0 0 0 0 0 2.E204 1.E-05 6.50E-08 NA ND 151 105.5 0 0 0 0 0 0 2.F+05 8.E-05 2.50E-08 NA PD 109 3.5 . 1 0.25 0 0 0 0 5.E+03 2.E-06 8.75E-10 NA PM 151 41 23.25 13 7.25 0.75 0 0 3.E+03 1.£-06 1.0oE-08 NA PR 143 69.5 a8 97 100.3 91.25 65 31.75 7.E+02 3.E-07 1.70E-08 NA PR 144 5.5 5.5 5.5 5.5 5.5 5.5 5.25 I.E+05 5.E-05 1.38E-09 NA PR 145 638 40.25 2.5 0.25 0 0 0 8.E+03 3.E-06 1.58E-07 NA PR 147 572 0 0 0 0 0 0 2.2+05 8.E-05 1.40E-07 NA RD 88 799 2.5 0 0 0 0 0 6.E+04 3.E-05 1.95E-07 NA RH 105 77.5 65.25 41.25 25.75 4 0.25 0 6.E+03 2.E-06 1.90E-08 NA RH 107 41 0 0 0 0 0 0 2.E+05 1.E-04 l.00E-08 NA RU 103 22.5 22 21.75 21.25 19.75 17.5 13.75 6.E+02 3.E-07 5.50E-09 NA RU 105 187.5 4.75 0 0 0 0 0 1.2+04 S.E-06 4.50E-08 NA SB 127 6.5 6 5 4 2 0.5 0 9.E+02 4.E-07 1.63E-09 NA SE Si 53 0 0 0 0 0 0 2.E+05 1.E-04 1.30E-08 NA SE 83 49.5 0 0 0 0 0 0 I.E+05 5.E-05 1.23E-08 NA SH 153 12.75 9 6.25 4.5 1 0 0 3.E+03 1.E-06 3.25E-09 NA SM 155 8.25 0 0 0 0 0 0 2.E+05 9.E-05 2.05E-09 NA SN 127 19.75 0 0 0 0 0 0 2.2+04 8.E-06 4.75E-09 NA SN 128 83.75 0 0 0 0 0 0 3.E+04 1.E-05 2.05E-08 NA SR 89 28 28 27.5 27.25 25.75 23.25 19.25 1.E+02 6.£-08 6.75E-09 NA SR 91 796.25 138.75 24 4.25 0 0 0 4.E+03 1.E-06 1.95E-07 NA SR 92 1325.8 2.75 0 0 0 0 0 7.E+03 3.E-06 3.25E-07 NA TC 101 1273 0 0 0 0 0 0 3.E+05 1.E-04 3.00E-07 NA TC 104 459.75 0 0 0 0 0 0 7.t+04 3.E-05 1.13E-07 NA TE 127 4.75 5.25 4.75 4 2 0.5 0 2.E+04 7.E-06 1.182E-09 NA TE 129 93.75 3.5 0.5 0.5 0.5 0.5 0.25 6.E+04 3.E-05 2.30E-08 UA SE 131 629.75 .4.5 2.5 1.5 0.25 0 0 5.E+03 2.E-06 1.55E-07 NA K-State Reactor 13.F-1 Original (10104)

Safety Analysis Report

I dl di CHAPTER 13

-NM TE 132 250.75 203 164.3 132.8 56.75 12.75 0.75 2.E+02 9.E-08 6.25E-08 NA TE 133 964.25 0 0 0 0 0 0 2.E404 9.E-06 2.38E-07 NA TE 134 1712.5 0 0 0 0 0 0 2.E+04 I.E-05 4.25E-07 NA XE 133 20362 24106 24010 22389 14111 5673 894.75 na 1.E-04 S.OOE-06 NA XE 135 58955 30517 6594 1195 1 0 0 na I.E-OS 1.45E-05 1.5 XE 138 158548 0 0 0 0 0 0 na 4.E-06 4.00E-OS 10.0 Y 91 23.25 27.5 28 28 26.5 24.5 20.75 1.E+02 5.E-08 5.75E-09 NA Y 92 946.?5 37.75 0.5 0 0 0 0 8.E+03 3.E-06 2.08E-07 NA Y 93 861.25 169.75 32.75 6.25 0 0 0 2.E+03 1.E-06 2.10E-07 NA Y 94 1583.8 0 0 0 0 0 0 8.E+04 3.E-OS 4.OOE-07 NA Y 95 1613 0 0 0 0 0 0 l.E+0S 6.E-05 4.OOE-07 NA Y 91M 423.5 88 15.25 2.75 0 0 0 2.E+05 7.E-05 1.05E-07 NA ZR 9S 29.25 29.25 28.75 28.5 27.25 25.25 21.75 1.E+02 S.E-08 7.25E-09 NA ZR 97 657 245.5 91.75 34.25 0.75 0 0 1.E+03 S.E-07 1.60E-07 NA

-.11

.1-1 1

K-State Reactor 13.E-2 Original (10104)

Safety AnalosIs Report

SAFETY ANALYSIS REPORT This page intentionally blank K-State Reactor

CHAPTER 14 TECHNICAL SPECIFICATIONS SEPARATELY PUBLISHED AS REPORT KSUNE-00-05 DEPARTMENT OF MECHANICAL AND NUCLEAR ENGINEERING, KANSAS STATE UNIVERSITY

SAFETY ANALYSIS REPORT This page intentionally blank K-State Reactor

15. Financial Qualifications 15.1 Financial Ability to Operate the Reactor The KSU TRIGA Reactor has been in continuous operation since 1962. From 1962 until 1997, the Department of Nuclear Engineering operated the Reactor at the University. In 1997, the Departments of Mechanical and Nuclear Engineering merged to become a single department.

The budget for the reactor is integrated. into the department budget. The department has substantial resources, in that it supports a student body of 424 undergraduate students and 61 graduate students, a full-time faculty of 24, and $2.5 million of externalfunding for research support Appendix A contains information related to demonstrating financial ability of Kansas State University to operate the K-State reactor.

Fuel Cycle Assistance Contract (including compliance with the Nuclear Waste Policy Act of 1982) and Amendments thereto

Support and Responsibility for Decommissioning

Nuclear Liability Insurance Indemnity Agreement and Amendments thereto 15.2 Financial Ability to Decommission the Facility As indicated in Appendix C, the university administration intends to support extension of the operating license of the TRIGA reactor. Whenever a decision is reached to decommission the reactor, the university will request legislative appropriation of fiuids, or otherwise provide funds sufficiently in advance of decommissioning to prevent delay ofrequired activities.

153 Bibliography NUREG/CR-1756 "Technology, Safety, and Costs of Decommissioning Reference Nuclear Research and Test Reactors, " US. Nuclear Regulatory Commission,, March 1982; Addendum, July 1983.

K-State Reactor 15-1 Original (9102)

Safety Analysis Report

-- FINANCIAL QUALIFICATIONS Appendix 15.A, Financial Statements I

K-State Reactor 15AA- Original (9/02)

Safety Analysis Report

FINANCIAL QUALFICATIONS j

_ tV 1196 32:M" AiR I rvRnY li n P."

July 6, 1988 I Mr. Varren Strauss. Comptroller Kansas Statea iniersity M4asatt, KaIsus 6650C TAS ORDER 10. I UNDER SWCONTAX T 11%. CSS-1a86 (DE-ACG7.760082) WtlH THE KANSAS STATE UIIYUMM FOR REACTOR NEL ASSISISCI -*M-6.9-*

Dear tr. Strauss:

I

1. This Task Order 4*.1, effective July 1, 2988. is In accordance NMtb:

A. The torus of Subcontract go. C8-l101BS6.

B. The Subcontractor's letter to continue the Reactor Fuel U AsSistae Progsra.

. As aresult of this 7ask reder. the Subaatractowill provide for l utilization of Ue reactor owned by the Subcontractor In a progm of education and training of studets In sucltar science *nd The contract engineering, and for Faculty and student resexrch.DE-owned pro"ides for the continued possession and use of nucleiar materials. Includla: enriched uraslu In reactor fuael vitout charge of use. burn-up, er rsprocessing while used for research, education and traiilag purposes. A report shall be subeitted eanually eanth use ef tie leased material during the past year.

3. IT=G Idaho Administration.

A. Contractual ruponlbillities uner this Task Wraer shall be

administered by AJm Pydalch. I A. All work to be performnd under tOis Task Order shall be endir the technical Jurisdiction and direction of Keith Irow.

4. Subcontractor Adainistratien. I A. The Subcontractor's contract responsibilities under this Tafs Order shell be administered by Warree Strauss.

.... O.RjtI IDa 15 Idm.

K-State Reactor 1 i Original (9102)

Safety Analysis Report

FINANCIAL&WAUFICATIONS

- J11 1% 321W~n I, Ar WNMMzr W=CR 1tIL *.

P.4/9 x,

er. Warran Strauss July 6. Is" AN-69-83 Page 2 D. Technical imis titration: All twrk to be performed Ynder this Task Order shall be under the technical Jurisdiction and direction of the Director of tIe Reactor fuel Assistanca higi.

S. Subject to the terms and conditions ef.Sudc*3tract No. M8-101856, the Subcontractor shall uste all reasonable efforts to complete the Scope of Work by tober 30, 1993.

6. The total estimated cost end ceiling mount for this Task Order Is Q:. Fwnds may bi provided at aI tin under this agreeent, 1i. wd s meeded.

The mount co..ently obligtted for performance of this Tast Order Is the maxiz mount obligatsd and mst mot be exceeded without the rIor aprovnl of the Subconttct ladinletrator designated herein.

If stto thae thractbcr tciler has nen to believe that the total cost to the Ccntractor will be Peater thaneh obligatttd nvoun'.

And/or cellig avount etbew on tkis Task Order, the Subcnntractor shall notify the eontractor In writing to that effect, 91vir the revised estimate or souc total cost for the performace of te Task Order.

7. Title to all speclal muclear material loaned to the Subcontractor uander this Task Order shall at all times be and reralr In the United States overunt. .l

8. The Contractor will not charge the Sctcontractor for such amunt of material as Is (3) consmed is the eperation of the facility until expiraticn of this last Order and 2) mat recovered In reprecessilg subsequent to the situate eturn the special meclear material.

will walve all useschaes on the material antil expiration of this Task Cresrg and vili, at the Contractor's option, eithr reimburse the Subcontractor for cost of shiping spent fuel for clinical reprocessing or accept rtetur of e material t the Saubcaetractor's facility. Ike Subcontractor's obligation Is to otn material In the form defined above. as affected by use for the contract purposes, and, therefore the Subcontractor bas no responsibility for reprocessing of sech material.

Except as otherwise provided hertin, the Subcontractor is responsible for and will pay Ue Contractor any thargas imposed by the Contractor for material delivered to the Suhcuntracter and mot Ultimately returned to the Comtrctor.

K-State Reactor *r15Aiii .iginal (9102)

Safety Analysis Report

FINANCIAL QUALIFICATIONS

, 11 9G6 12204PH ATR L UtrM lY 1ZX=F nL P.5" Br. Warren Strauss

,%o '2Ialy S. 1588 AR.59.83 Page 3

9. Upon 9xiration of Use tsrcs of this Tast Crder, the Subcotractor ma. continum to use thb material under Its license aud In such veitt, the Contractor his the option to require the Subcontractor to he TBsponsible for all charges requred by the Contractor relating to the material from the date of expitatlec. including charges for use and constaption.

10. Notwithstanding any other provision of this contract or Task order.

the Covenmnt sba r not be responsible for or havs any oblcation to the Subcontractor for decostaxmtat on or deca missianlvg (0 ) of any of the Subcontractor's facilities.

L1. The Sutcontractor Is responsible for verification and accuntsbility of any DOE-Owned nuclear materials la Its possession. A report shat be submitted annually an the use of the loaned material duing the ast year. In the event the ters and conditions or this contract or ask Order art not In agreement with MMC rules and reulations, the PRC requiranats will take precedence.

12. Invoices for seruices provided under this Task Order Ihere applicable, may be accmanled by a certifled statem it of costs in tie foreat set forth In Appendix A; however, must accOpany the tisal Invoice.

The orginal A two copies of this Tesk Order are forwarded. Please excrte the original amd cet copy d return them to £EG6 Idaho, Inc. The rmaining copy, executed by £"3 Idho, Inc., may be retained for your records.

£16 IwDoI, INC. XS STATE INIVERSM BI y _ _f_ _B' Ann tydalda Wzmun Stas .12-Title Su=contret Admfoistrater Title 2mnw ' '

CIW ce: Ka1th Eran I?,

K-State Reactor 15.Aiv Qriglnal (9102)

Safety Analysis Report

W!i'

!!:,j.-

FINANCIAL QUAUFICATIONS November 29, 1993 Mr. Richard E. Faw Kansas State University Manhattan, KS 66506 MODIFICATION 1N. 1 TD TASK ORDER NO. I UNDER SUBCONTRACT NO. C88-101856 (DE-ACO7-76ERO~4f6) W1TH KANSAS STATE UNIVERSITY FOR REACTOR FUEL ASSISTANCE -

KF-201.93 ;Aoj2 od%

Dear Mr. Fal:

Ihis IModfication Ho. I to Task Order No. I under Subcontract No. CSt-101856, effective lovember 1, 1993.is to:

1. Expand the scope of work to include additional compliance requirements,

2. Modify the EGIG Idaho Administration,

3. Extend the term five years.

NOV, THEREFORE, the parties mutually agree to the following:

1. Item 2 - is expanded to include the Attachment A - UNIVERSITY REACTOR FUEL ASSISTANCE PROGRAM - Subcontract Materials Management Requirements.

2. Item 3 - E&AG Idaho AdminIstration Ismodified to change the contractual representative and the technical representative as fol ows:

A. Contractual Responsibilities under this task order modification shall be administered by Ken Feliciano.

E. All work to be performed under this task order modification shall be under the technical jurisdiction and direction of Anthony Vifnola.

3. Item 5 Ismodified to extend the term five years. Therefore, Item 5 is modified to read as follows:

Subject to the'terms and conditions of Subcontract No. C8SOE1856, the RECEIVED subcontractor shall use all reasonable efforts to complete the Scope of Work by October 30, 1998.

DEC 07 P.O. Bax 1= Idho ft ID M5 .. 3j. rui EprlmnI SbUm K-State Reactor 15Av Original (9102)

Safety Analysis Report

'ji I,

FINANCIAL QUAUFICATIONS 'ii JI Mr. Richard E. Faw Novncber 29, 1993 KF-201-93 Page 2 Except only as changed by this Modification No. 1. all of the terms and conditions of Task Order No. I Under Subcontract No. C8B-101856 shall remain infull force and effect.

Two copies of this ModifIcation No. I .to Task Order tto. 1, executed by EGcG Idaho, Inc., are forwarded. Please execute one copy and return Itto ELG Idaho, Inc. One copy my be retained for your records.

KAXSAS STATE UNIVERSITY [GIG IDAHO I, J1~

By f,ttJr aShanteau Xej felJzTAi0o Title 'Title Subcontract adidt__

Znterim Assoc. Vice Date v Pr.n"vqt fFmr tp.rpl,, Data o - 'J.

IN.

K-State Reactor 15A-vi Original (9102)

Safety Analysis Report

FINNCIAL QUAIRCAVONS

[F J Office of the Provost ZA2 AS AndeMronHall Muahamv AnsKana kSC 13-2=424 Ouly 17, 1990 United States Nuclear Regulatory Commission, Region 1V 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76012 PR: License 9-98 Docket 50-186

Dear Sirs:

This concerns the ultimate decommissioning of the Kansas State University TRICA Nuclear Reactor Facility, currently licensed for operation by the University until August 15, 2001. Pursuant to Federal Regulations, Title 10, Part 50, this Is to assure that the University, an entity of the Stite of Xansas, will obtain the funds for decommissioning when necessary.

It Is our Intention, at the appropriate time, to request an extension of the Reactor Facility operating license beyond the August 2001 termination of the current license. Never-theleas, whenever a decision In reached to decommission the facility, the University will request legislative eppropri-ation of funds, or otherwise provide funds sufficiently in advance of decommissioning to prevent delay of required activities.

Enclosed Is a preliminary cost estimate for decommissioning of the Reactor Facility. The cost estimate was prepared on the basis of the methods described in the May 1989t Nuclear Regulatory Commission Draft Regulatory Guide on Assuring the Availability of Funds for Decommissioning luclear Reac-tors.' Cost escalation factors other than those prescribed in publications cited by the Guide are based on U.S.

Department of Labor statistics.

With assurance that I have the authority to sign this statement of Intent, I am Yoqrs truly,

^4 ac .Cffmpl Encl.

K-State Reactor 15A-vii Orginal (9102)

Safety Analysis Report

FINANCIAL QUAFICATIONS tNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTOR D.C. 20555

-*.,, NHOV 27 t896 Docket Ila. 50-188 Dr. Richard E. Faw, Director Nuclear Reactor Facility Department of Nuclear Engineering Yard Hall '

Kansas State University Manhattan, Kansas 66506

Dear Dr. Faw:

SUWECT: KANSAS STATE UNIVERSITY - AIENDIIUT TO INDEMNITY AGREEMENT Y~eare enclosing herewith an asmeneent to your Indemnity agreement reflecting the changes to 10 CFR Part 14D, "FInancial Protection Requirements and Indemnity Agreements,- effective July 1, 1989. The amendments to Part 140 reflect the increase from $160 million to $200 million in the primary layer of nuclear energy liability insurance provided by American Nuclear Insurers and Mutual Atomic Energy Liability Underwriters. The amendment also conforms to changes made to the Price-Anderson Act by The Price-Anderson Aaendents Act 1 of 1958 which was enacted on August 20, 1988.

Please signify your acceptance of the amendment to your Indemnity agreement in the space provided and return one signed copy to Ira DInitz, Senior Insurancel Indemnity Specialist, U. S. Nuclear Regulatory Comaission, Fall Stop 12E4.

Washington, D.C. 20555. If you have any questions about the foregoing, please contact Mr. Dinita at 301-492-1289.

Sincerely, Theodore S. Iiichaels, Project Manager Non-Pcfwer Reactor, Decoumissioning and Environmental Project Directorate Division of Reactor Projects - Ill, IT, Mand Special Projects Office of Nuclear Reactor Regulation

Enclosure:

Amendment to Indemity Agreement cc w/enclosure:

See next page K-State Reactor 15.Aviii Orginal (9)02)

Safety Analysis Report

.1 e.s A i -C.::

FINANCIAL QUALIFICATIONS I 11 ;

Kansas State University tocket 1o. 50-188 cc: Office of the Governor State of Kansas Topeka, Kansas 66612 Mayor of Manhattan P. 0. Eox 748 Manhattan, Kansas 66502 1

K-State Reactor 15A-i OrIgInal (9102)

Safety Analysis Report

FINANCIAL QUAUFICATIONS UCA °,, rUNITEDSTATS 1.

5NUCLEARREGULATORY COMMISSION

lI"Hinarot 0O.C.0553 Docket tNo.50-18e Amendment to Indemnity Aj:eemt Ila. [-1 3nient Ho. j3 Effective July l, 2919, ladtmnity Agreement la. £-l, between Kansas State

-I University of Argriculture and Applied Science, and the Atomic Energy Commission, dated J7ure 18, 1962, as amended, is hereby further amended as follows:

The amount *S160,00,000' isdeleted wherever it appears and the amount $200,000,000' Is substituted therefor.

7he amount S3124,000,000' isdeleted wherever it appears and the amount 'S155,000,000' is substituted therefor.

The amount *S36,000,000 is deleted wherever It appears and the amount 'S45,000,000' is substituted therefor.

Paragraph 1, Article I is modified to read as follows:

1. "Nuclear reactor,* byproduct material,' 'person.' 'source .aterial,'

special nuclear material,' and precautionary evacuation shall have the meanings given then In the Atomic Energy Act of 2954, as amended, m"d the regulations issued by the Commission; The definition of 'public litability in paragraph 7, Article I is deleted, and the following Is substituted therefor:

"Public liability' means any legal liability arising out of or resulting from a nuclear incident or precautionary evacuation (including all reasonable additional costs incurred by a State or a political subdivision of a State, in the course or responding to a nuclear incident or precautionary evacuation),

except (1) claims under State or Federal Vorlmen's Compensation Acts of exployees of persons indemnified who are employed (a) at the location or, it the nuc ear Incident occurs in the course of transportation of the radioactive material, on the transporting vehicle, and (b) In connection with the licensee's possession, use or transfer of the radioactive material; (2) claims arising out of an act of war; and (3) claims for loss of, or damage to, or loss of use of (a) property which is located at the location and used in connection with the licensee's possession, use, or transfer of the radioactive material, and (b) If the nuclear incident occurs in the course of transportation of the radioactive material, the transporting vehicle, containers used In such transportation, and the radioactive material. e cs Paragraph 4(c), Article ll Is revised to read as follows:

(c) Any Issue or defense based on any statute of limitations If suit Is Instituted within three years from the date on which the claimant ft nrt knew, or reasonably could have known, if his InJury or damage and the cause thereof.

K-State Reactor 15 Anx Original (9102)

Safety Analysis Report ¶

FINANCIAL QUALIFICATIONS

.1 II 2

Paragraph 1. Article IV Isrevised to read as follows:

1. When the Comaission determines that the United States will probably be required to make indemnity payments under the provisions of this agreement, the Comoisslon shall have the right to collaborate with the licensee and other persdns indemnified 1n the settlement and defense of any claim (including such tgal costs of the licensee as are approved by the Cormission) and shall have the right (a) to require the prior approval of the Commission for the settlement or payment of any claim or action asserted against the licensee or other person indemnified for public liability or damage to property of persons legally lIable for the nuclear incfdent which claim or action the licensee or the Comeission may be required to indemnify under this agreement; and (b) to appear through the Attorney General of the United States on behalf of the licensee or other person indemnified, tate charge of such action and settle or defend any such action. If the settlement or defense of any such action or claim is undertaken by the Commission, the licensee shall furnish all reasonable assistance in effectIng a settlement or asserting a defense.

In paragraph 1, Article VIII, the aMount *SS,Oo,000o Is deleted and the amount *S63,0OD,000- is substituted therefor.

fOR THE U.S. NUCLEAR REGULATORY COMMISSION Policy Develo nt and Technical Support Branch Program Kanagement Policy Development and Anas1s Staff OffIce Nuclear Reactor Regulation Accepted /

-7: 19f/

It a ve LA0 UgriculState undve d St c et tfgiculture and Applied Sc knce K-State Reactor 15Axi Original (9102)

Safety Analysis Report

.. I

FINANCIAL QUALIFICAIONS_

ci UNITEDSTATES I

NUCLEAR REGULATORY COMMISSION JJ

A ,,OOH0C.O5 Docket Ro. 50-188 b Ametdzent to Tndemntty A!ement Ito. E-1 ,

zmrament No. 13a-Effective July 1, 1989, Indemnity Alreesect No. E-1, between Kansas State UnIvers'ty of Argriculture and Appi ed Science, and the Atomic Energy Comoission, dated June I8, 1962, as amended, is hereby further.amended as follows:

The amount W$160,Coo,000' Is deleted wherever it appears and the amunt S20,O00,000V Is substituted therefor.

The amount S5124,coo0000C Is deleted wherever it appears and the amount $155,000,0004 Is substituted therefor. C The amount *S36,00D,00 Is deleted wherever it appears and the amount s45sooe000' Is sub;tituted therefor.

Paragraph 1. Article I Is modified to read as follows:

1. "Huclear reactor, "byproduet materfal," person, "source naterial,"

special nuclear material, and 'precautionary evacuation' shall have the meanings given them in the Atomic Energy Act of 1954, as amended, and the regulations issued by the Comnission.

The definition of public liability' In paragraph 7, Article I Is deleted, and the following 1s substituted therefor:

oPublic liablity" esans any legal IMability arisig out of or resulting from a nuclear Incident or precautionary evacuation (Including all reasonable additional costs Incurred by a State or a polItical subdivision of a State, In the course or responding to a nuclear incient or precautionary evacuation),

except (11 claims uander State or Federal Ioren's Compensation Acts of employees of persons indemnified who are employed (a) at the location or, If the nuclear Incident occurs In the course of transportat1nn of the rad1oactive material on the transporting vehicle, and (b) In connection with the licensee's possession, use or transfer of the radioactive material; (2) claims arising out of an act of war; and (3) claims for loss of, or damage to, or loss of use of (a) property which Is located at the location and used in connection with the licansee's possession, use, or transfer of the radioactive material, and (b) if the nuclear Incident occurs In the course bf transportation of the radioactive material, the transporting vehicle, containers used in such transportation, and the radioactive material.

Paragraph 4(c), Article It isrevised to read as follows:

(c) Any issue or defense based on any statute of limitations Ifsuit is instituted within three years from the date on which the claimant first knew, or reasonably could have known, of his Injury or damage and the cause thereof.

K-State Reactor 15Axii Original (9102)

Safety Analysis Report

to:*

SAFOTO ANALOSIS ROQORT This page intentionally blank K-State Reactor

FINANCIAL QUAUIPCATIONS I 1 :

2 Paragraph 1, Article IV is revised to read as follows: 1

1. Uhen the Coculssion determines that the United States will probably be required to make indemnity payMents under the provisions of this agreement the Commission shall have the right to collaborate with the ltcensee and o her persons Indemnified inthe settlement and defense of any claim (including such legal costs of the licensee as are approved by the Commission) and shall have J the right (a) to require the prior approval of the Commission for the settlement or payment of any claim or action asserted against the licensee or other person aenwiffied for public liability or damage to property of persons legally liable for the nuclear incident which claim or action the licensee or the Commission may be required to indemnify under this agreement; and (bJ to appear through the Attorney General of the United States on behalf of the licensee or other person indemnified take charge of such action and settle or defend any such action. If the settlement or defense of any such action or claim Is undertaken by the Commission. the licensee shall furnish all reasonable assistance in IJn effecting a settlement or asserting a defense. U In paragraph 1. Article VI!t, the aiount *5.OC00,000 Is deleted and tke amount S63,0C0,000- Is substItuted therefor.

FOR 7HE U.S. NUCLEAR REGULATORY COl1ISSlOII U CeCllI0. 1hoas, LhBUT Policy Development and Technical Support Branch Program Management Policy Development and Analysis Staff Office Nuclear Reactor Regulation ..

Accepted _ 7 /.. , 19/

a By .

as tat u er ot Ariculture and Applied Science 9

. 4.

K-State Reactor 15Axiii Original (9102)

Safety Analysis Report

Safety Analysis Report:

Add:itional Information Kansas State University TRIGA Mark II Nuclear Reactor Facility License R-88 Docket 50-188 21 December2004 Department of Mechanical and Nuclear Engineering K-State Nuclear Reactor Facility Kansas State University 110 WardHall 302 Rathbone Hall Manhattan, KS 66506 Manhattan, KS 66506

Q'tz53-.'-..-':, .ii.;.,?<TAB..J., *aTAUATION:OFCHANGES-TODRA.TKSU..TRIGA;ISAFlTYANALYSTSREPORT-,Y.;i.. '

w t<* l6l;Tt~. @t"*1!8s t.l w-,2" , .":i ^ , -Revised; l t v. ., LYT . ..v !.X ' ':;'

.. R.1..;: ;.,'. *."", * ¢; . *.;.^....; £ .. :. ;.****^;.. **;5.*^-* :.> ,;..

.'..* .* .* .. *. ..-... ... 1*-'

. aThis report is bssed on the Xarnis Slale Unhcnlt,y TRIGA2 Mm*ar 11 Jthznrdr Snrmmaty R'tport (1961) fior the initial oexration of the reactor at 100 kzW thermal powter, the 1968 lids reportis basedontheJ X umwstateUn hnIO Safety Analysis Report and Safety Evaluation Report fior TRYGA Af rkllthransmSumm ryReporl(i961) fiorthe license amendment to allow .2S0 kW steady state thermal initial operation of the reactor alt 1;00kW theunni power, power (250 MW pulsing capability), and subsequent analyses 1968forSafiety steady state operations at S3.00 1.250reactivity kW (pulsing to a CIarifiato the licenseAnalysis Report to and allowSafiety Evaluatlon t- t. 11supporting

.. Repott amendment 250kW Xsteady . 11 nominal S3.00 reactivity insertion. A Insertion Ca c o state thermal power (250 MW pulsing caability), and Isexpected to resull in a peak thermal power orapproximately subsequent analyses supporting steady state operations at 1,340 MWV). Basedf on proposed reactivity limits, tho TKSU

$00 kW(pulsingto anominal S3.00 reactivity insertion. reactor will only be able to achieve about %A the proposedf

. ~maximum power level fior steady state operation; therefiore

. ltherma1 hydraulicF and source term calculations are

-. conservative bv a factor of 2 in analysis.5_______ p This report addresses safiety issues associated with operation of the reactor at 1,250 kW, including Increased pulsing This report addresme sarety Issues associatedt with capabilities. The maximum excess reactivity permitted by operalion of tho reactor at 500 kW. Including Increased Technical Specifications cannot achieve a continuous steady pulising capabilities. This report reflects the assbuilt state power level greater than about 500 kW; therefore analysis condition of the facility, and includes experience with the performed fior steady state operations at 1,250kIW Isextremely operation and performance of the reacior, radlation conservative in evaluating consequences and characteristics of surveys, and personnel exposure hstsories related to normal and accident scenarios. This report reflects the as built operation of the reactor at 250 kW steady-state po°wer condition of the facility, and Includes experience with the Where appropiate,radiological characteristics have been operation and performance of the reactor, radliation surveys, extrapolated to reflect operation at 500 lCW. The and personnel exposure histories relatedto op~eration of the Power level ror bounding 1111 consequence of routine genertion of radioactive cimuent 11 .1 reactor 3t 250 kW steady state powver. WVhere appropriate, analysis and other waste products from steady state operation at radiological characteristics have been extrapolated to reflect 500 kW is addressed in Chapter 11. Radiation workerand operation at 1,250 KW. The consequtence of routine generation public doses from radiation associated with routine of radioactive effuent and other waste products from steady

.operations are we~llwithin the limits ofTitole0, Codeoor state operation at 1,250 kWV Is addressed in Chapter 11.

Federal Regulations, even under unrealistically Radiation worker and public doses from radiation associated conservative scenarios. The consequence of accident with routine opeatlions are well within the limits of Title 10, acenarios frona operation at 500 KW steadystate power Code of Federasl Regulations, even under unrealistically and pulsing isptresented InChapter 13. conservative scenarios. The consequence of accident scenarios from operation at 1,250KlW steady-state;power and pulsing Is

- - Th.ecito fteratrcr n hra yrui prestnetl inChapter 13.

Aercto dsniilo cocorte ant tema hytuli _ _ The description of the reactor core and thermal hydtraulic analysis presented in Chapter 4 the Secondary Cooling: 1.2 11 analysis presented inChapter 4, the Secondary Cooling Power level for bounding System InChapter 5 and the Reactor Control System in I.Systemn in ChapterS, and the ReaCtor Control System in analysis

_____ __ _Cihapter 7 are based on S00 KCEoperations. Chapter 7 are based on I.2S0 KW operatlons.___________

-. . ~~TAULiATION:OFCHANGES -TO DRAYRKSU TRIGA:II SAFETYANLSSROT.' Z:.:J2 I C*.

Pagovarag~i: ~-Texft.! Z.. .. ". IC. *-Snipii. .Tex t$ :~v As of'July, 1999, there were over 66TRG atn -- TRIGA reactors in AofJl,19,tccwrovr7 uco use or under construction at universities, government and Andsonsulnict999 athr unieversi 0Ties, g rmenctornd indusetria Industrial laboratories, and medical centers in 24 labdroratoruieond meial centvers ini24 gouemntriesd Industruialy countries. lHistorically, analysis and testing or TRIGA lanoalysies and mestingocTIaul inshateemounstrates. thatofueall fuel has demonstrated that ftc! cladding integrity is not clyssaddin itesigriyI norTG challne as leongsstraedsothatfel challenged as long as stress on the cladding remainscldigntrtysnochleedalngsstssnth within yield strength at cladding temperature Elevated cladding remains within yield strength ror the cladding 1-21.21TIGAfue teperturs eolv hyrogn rornthe 1-2 1.2 temperature. Elevated TRIGA fuel temperatures cvolve Powcr level for bounding 1z12 rconifuel empeat ureswit evolvemhydrgtnpfromute bui2dup2 hydrogen from the zirconium matrix, with concomitant analysis zircoadnium mThrix, w M thcocmtanth pressue buladu asa pressure buildup Inthe cladding. Therefore, the strength or the unthcldiong Thernefotre, testrengis ostheperlimd as a ue clad as a function or temperature establishes the upper limit on untooftemperature Fe tesprtuablsess theaupelimitiongfauel fuel temperatre Fuel temperatre less than limiting values temperature. Fcla ntemprtaturevaluaess thn limingvalues wilt ensure clad integrity (as cvaluated in NUREG 1282) and aill hensure claontainraioatyi(a evmatedrinl proued1282 therefore contain radioactive materials produced by fission in

_ __ fission In the reactor core the _ __ __ __ __ ___ctor_ __ __ __ __ __ __

1-2 1 1000

,21 1.2 lf 1,250 Power level for bounding Para _ _ _ _ _ _ _ _ _ _ _ _ _ Pa na 2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ analysis Consequently, maximunt possible power using .... Consequently, maximumn possible power using TRIGA 1-2 1.2. TRTGA fuel Iscontrolled by the amount of iuel loading. Muel Is controlled by limnitin the amnount of fuel loading. A Power level for bounding pamn 3 A minimum of eighty-three elements is required to allow minfimum of elighty-threo elements Is required to allow analysis oneration at 500 kV. .. ___ operation at 1.250 kW...___________

As indicated in Chapter II. radiation sources are discharged As Indicated In Chapter I 1. radiation sources are from the reactor Facility In gaseous (airborne), liquid or solid discharged from the reactor fhclllty In gaseous (airborne), form. These forms are treated individually Insubsections of 1.2 1.2.2 liquid or solid form. Those forms are treated inividually 1-2 1.2.2 Chapter I 1. Airborne radiation sources consist mainly of Clarification Insubsections of Chapter I 1. Airborne radiation sources Argon-Il, Nitrogen.16 and Tritiunm. with Argon 41 the major consist mainly of Afgon-4l, Nitrogen-16 and Tritium. contributor to off site dose. Limits on Argon-41 and Tritium Argon 41 is the major contributor to off site dose. are tabulated below, with Cesium 137, the othier significant 1-3 1.2.2 NA 1-3 1.2.2 isotope orinterest for the KSU reactor.

Added chart of IOCFR2O App Bvalues &related Informnation Clatriication

'I

. (- (- r t-' (- C (- (- (- (- (-'- ( (- C--C--r- C--L-'- r- (-- (- (- 6- C- C--C-.C-.G-. (-. (-. 6 6. 6. (-=, (.- (-=, -= -= (=-= (=-=

'. ' *s *. %~ '. '- ' .. I . S. % %. .5 S. . . % I I % .

I I .. 5, I - I I .

Argon 41 Is the major contributor to radiation exposure Incident to the operation or the K-Stale reactor. Argon 41 Is attributed to neutron activation or natural argon (In air) In the reactor bay atmosphere. rotary specimen rack adjacent to the core, and dissolved in primary coolant. Argon 41 has 1.8 h Argon 41 is the major contributor to radiation exposure hairflire. Calculations based on 1,250 kW steady state incident to the operation orthe K-State reactor. Argon 41 continuous operations shov that doses In the reactor bay Is atnributed to neutron activation ofnatusal argon (in air) remain below Inhalation DAC. Using extremely conservative In the metor bay atmosphere rotmry specimen rack assumptions of operational conditions In concert with the adjacent to the core and dissolved in primny coolant. worst-case wind stability class, the oftsite dose from Argon 41 Argon 41 has 1.8 h hal lifc. Even under unrealistically Is slightly less than 10N or the 10 mreinlyear limit. A conservative assumptions of operational conditions, the summation orall relative Frequencies for winds under Pasquill off site dose from Argon 41 Is well within limits and stability category A (Table B-3) indicates frequency tess than 1-3 doses In the reactor bay are below the levels requiring 1.3 1.2.2 0.6%. I.e., the contribution to off site doses from Argon 41 controls of an aliborne radioactivity area. Chapter II produced during a year of full power, steady state operations Appendix A shows peak off-site actIvity concentration accounts for less than 0.6% of the total dose. All other during normal operations would be about 6 x 1O4 pCimL atmospheric dispersion calculations show that the off site dose at 135 m downwind under slightly unstable atmospheric from Argon 41 Is well within limits, and doses in the reactor conditions, less than the effluent limit of 0.01 pCIImL A bay ar below the levels requiring controls of-an aitbome WHIt year exposure at the maximtnn concentration would radioactivity area Chapter II Appendix A shows peak off lead to an effective dose or only about 3 mrem, well site activity concentration during normal operations would be within applicable limits about 4.5 x t0' pCVmL at 53 m downwind under extremely unstable atmospheric conditions, less than the emuent limit of 0.01 pCimL. A full year exposure lo equilibrium argon concentration for 1.250 kW operations under normal atmospheric conditions would lead to nn effective dose of less than 7 mrnem. well within armlicable limits.

. :.'..;h.sZ%.~TBJAI~O GESTO RATSU.TRIA*1SAFTYNAYSISEPR~

CRANX..:tl,.TS~ D ah. ': ':8';::.a or "fiidl Evsd § : ,,A - i 1 Ž Reason!:.,,-

_ i-;- :Pwa tt :; . *h: ; !'t; 4Text:.. .. Par! i ; *:" i . v.- .':';.

Nitrogen 16 is the major contributor to radiation fields directly over the reactor pool during operation. Nitrogen 16 is produced by a fast neutron reaction with oxygen (as a natural component orwater Inthe core). Nitrogen 16 has a 7.1 second half lifec and consequently does not remain at concentrations capable ofcontributing significantly to off-site dose. Chapter Nitrogen 16 Isthe major contributor to radiation fields II shows very conservative calculations load to an expected directly over the reactor pool during operation. Nitrogen exposure rate or slightly less than 100 mnremlhr at one meter 16 Is produced by afast neutron reaction with oxygen (as above the center or the reaclbr tank during sustained operation a natural component or water inthe core). Nitrogen 16 at 1,250 kW thermal power. The 22.foot level has radiation has a 7.1 second hairflire, and consequently does not monitors directly above the pool and at thc rail surrounding contribute to off-site dose. Chapter II shows veiy access to the pool. Measured exposure rates directly above the 1-3 1.2.2 conservative calculations lead to an expected exposure 1-3 1.2.2 pool surface are about 20-30 mR/h at 250 kW operations, and Poweraevel fior bounding rate of4Oml/h at onemeterabove the center ofthe measurements at the rail approach 2 mR/hr. During normal, nalysis reactor tank during sustained operation at 500 kW thermal steady state S00 kW operations dose rate can be expected to power. Measured exposure rates are about 20 rl/h at achieve 40-60 mrem/hr, and during steady state operations at 250 kW operation. Therefore radiation dose rates directly 12 0 kW the ara directly above the pool surface may become above the reactor pool during operation are within a high-radiation area. Therore, radiation dose rates directly required levels for a radiation area . above the reactor pool during expected operations at levels up to 500 kW are within required levels for a radiation a as derined In IOCFiR20, and additional administrative controls for access to the area directly above the reactor pool 500 kW to the maximum license power level of 1,250 kW may be required. Installed monitoring systems provide informnntion necessary to Identify appropriate access controls.

Tritium Isgenerated by sequential activation of hydrogen Tritium Is generated by sequential activation or hydrogen (in (inwater) in thecore arca Mcasured tritiuin specific water) in the core are Measured tritium specific activity in activity inprimary coolant is less than 10'3pCg. If the primnary oolant Is less than 5x 10D pCVg. Ifthe reactor bay reactor bay atmosphere wcr saturated with this water at atmosphere were saturated with this water at 30°C the water 30C, the water concentration In the air would be less tOam concentration in the air would be less than 3 x I04W/mL and 3 I *g an *- * *

the activity concentration in the atmosphere 1.5 x 10' 1rMIr Power level for bounding 1.3 12.2 atm loshewll blhe Iv OCFtRt ndtAC 1-3 122 well below the DAC limit and well below the atniosphenr analysis adteatmosphere, b~elow cl th imit2 AEvendinder DACeffluent limit with the dilution factor of 200 for discharge fromn urand sthe assmopheioneftlutet cmlm t ory te trtumIvenue the top of the reactor bay. Even under the extremely unrhemaliticoruptoionethasetothe mlt ritiurnbnvat conservative assumption that the complete tritiumi inventory of otrtshereator poo itiumleasedninto theorector bay the reactor pool Isreleased into the reactor bay atmosphere, the atmospher, thlC tritum ctritium concentration would be within limits for an unrestricted limits rot an wunrstricted area.

_ _ _ __ _ _ _ _ _ _ _ _area._ ___

( ( L--(__ (--- (__ (_ L--(_ L. (-' (_ (_ (_ C_ r- (_ " C_ C_ (_ (__ r- C_ G__ L= (__ C__ C__ (__ C__ Q' C_. C_. 6-a C_. 6_. (.-.(6. "_ "_ G-W

. . % I v. . s v I I. Iv S v S -

v I.

CHANGES-TO'DRA~r.KSU.TRIGXH'SAFLY';.-N:X *YSIS

^p .^ ! _______

, ;:t____________________________

l: h i l-^TS>~~~~" '

F4 .d.-*.s;** abn -

J .8tv 'Pmt';, I<,W*lii6¢S ;l p^ .st; rs h

Although total loss orreactorpool vater is considered to Although total los orreactor pool water is considered to be an be an extremely improbable event, calculations have been extremely improbable event, calculations have been made to 1.3b thatmade to determine the maximum fiuel temperature rise determine the maximum fuel temperature rise that could be could be expected to result from such on e~vent taking expected to ratilt fromn such an event taking place after long-I4 123b placeslierlong-termoperatlonat 1 under extraordinanly onservative MI powveri500kw 1-6 123.bp term opertlon at full power or 12o kw. Even under Power level Iorbounding assumptions and 1 extraordinarily conservative assumplions and approximations, approximations, the maximum fuel temperature reached the maximum fuel temperature reached in a loss or coolant in a loss oreoolant accident is less than 290C, well accident Isless than 300eC, well below any safety limit for below any safety limit forTRIGA reactor fuel. TRIGA reactor fuel.

Radiation doses ftom loss of coolant accident under Radiation doses from loss ofcoolant accident under extremely extremely conservative assumptions are computed and conservative assumptions are computed and have been have been tabulated In Chapter 13. WaterInjection Into tabulated in Chapter 13. Waterinjection into the reactor pool I.23b the reactor poolis accomplished by operating valves and 123.b Is accomplished by operating valves and pumps on the 12.foot Power level ror bounding p2 pumps on the 12-root and O-foot levels. Radiation levels p2 and 0Mroot levels. Radiation levels calculated under these analysis calculated under these assumptions are calculated to be assumptlions are high, but valve operation can be accomplished less than 500 mRJhr on the 0 and 12 foot levels at the time In a time period which will ensure doses do not exceed ofwaer loss.. IOCFR20 limits Two reactivity accident scenarios are presented. Thefirst is the Insertion of 2.1X (S3.00) reactivity at zero power by sudden removal ora control rod. The second Isthe sudden removl or TwVo reactivity accident scenarios are pesented. The first the same reactivity vlth the core operatIng at a power level is theinsertion of2.1% (S3.00) reactivity at zero power equivalent to the remainder or the core excess reactivity.

by sudden removal ofa control rod.The second Isthe Analysis shows that peak fuel temperatures in the first case sudden removal orthe nine reactivitywith thecore does not reach fuel temperature limits, with a maximum operating at a power level equivalent to the balance of the temperature less than 750C at the peak in the hot channel for core excess reactivity. Movements of control rods for the conditions whereInitial steady state power level Is regulated

-5 1.23.e first case are controlled,In part, administratively, %vhile 1-5 1.23c only by he balanc of core excess racIvity, hil cladding Claicaion movementsfor the secondare prevented by control circuit temperature remains below SOOtC. In the second case design. Analysis shows that(in neither scenario) peak fuel maximum fue temperature iscalculated at anmaxtmum of less temperatures reach limits, with a maximum of869C for than 870C al thepea. inthe hotchannela again with cladding conditions where Initial steady state power level Is thanpem le tha In te Ahoug tha e wo scnadiom regulated only by the balance ofcore excess reactivity (a conditionprevented tertperatureless S00CeuhanAlthough the two scnarios meetl by interlocks) criteria required to ensue cods for the first case aretbel integrity,(in controlled movements ofcontrol part) administratively, while movements for the second case are prevented by control circuit design.

I" 1 t.'-;1.X>~~..- ,..,, Ih ., -.

t.l2 C.-.-,- *j....

The KSU TRIGA reactor is a water-moderated, water-cooled thermal reactor operated in an open pool. The reactor is fueled with heterogeneous clemts clad with stainless steel, consisting or nominally t ,crcent enriched uranium In a zirconium hydride matrix. In 1968. the KSUTAII was licensed to operate at a steady-state thermal power of 250 kW with a pulsing thermal power limit or2so MW. Application is made concurrently with license renewal to operate up to a maximum steady state power level of 1,250 kW steady-state thermal powers and pulsing lo S3.00 (nominal 1,340 MW peak The KSU TRIGA reactor Isa water-moderated, water- power). Reactor cooling Is by natural convection. The 250.

cooled thermal reactor operated In an open pool. The kir core consists typically aO fuel elements (a minimum of reactor is fueled with hctrogeneous eljspts clad with elanned fior the 1250kW core), each containing as much as stainless steel, consisting of nominall(ercent ". The ractdor core Is In the rorm or a right o rams enriched uranium in a 2irconium hydrt7 auix tn 1968, circular cylinder about 9 in. (23 cm) radius and 15 in. (38 em) Clarify Power level for 1-6 1.3.3 the KSUTMII was licensed to operate at a steady-state 1-6 133 depth, positioned with axis vertical near the base of a bounding analysis thermal power of250 kW with a pulsing thermal power cylindrical water tank 1.98 m (65 It) diameter and 6.25 m (16 limit of 2SO MW. Application Is made concurrently with Q) depth. Criticality Is controlled and shutdown margin license renewal to operate at 500 kW steady-state thermal assured by control rods in the form of alumintun or stainless.

power and nominal 1000 MW pulsing maximum power. steel clad boron catbide or brated graphite. The 250 kW core originally used three control rods, the 1,250.kW core will be controlled by four. To reactor tank Is surrounded on the side and at the base by a biological shield of reinforced concrete at least 8.2 fn (2.5 m) thick. The tank and shield are In a 4078 m' (144,000 t') dynamic confinement building made of reinforced concrete and structural steel, with composite sheathing and aluminum siding. Sectional views orthe reactor are shown in Figures 1.1 and 1.2. with a floor layout in Figure 13 showing the 0-foot, 12-foot and 22-foot levels or the aefllihv.

1.6 13.4-

- I Rcficrcnce to 500 kW T11TJA Changed to 1,20 kW nlyi level for bounding

-I Power C ( (* C, ( ( (( -( -( - -( -( -(~( ~C- -L (-C .C-( - - - - -

I '% I . .

%TABULAUTION.U OF.C-NGES.T D AF.T U.RGA-,SAYSIS- EPORT; ..:t.:,;./-;;; .. :rn,;

_ n.T e..- '.s... *; i '... '.' ' .-

The reactor control system includes the mechanical and electrical systems fot control rod drives, and infsniments The reactor control system Includes the mechanical and that monitor control rod position. Each control rod can be electrical systems for control rod drives, and instruments that independently manipulated by pushbutton console monitor control rod position. Each control rod can be controls COne control rod can be operated In an automatic independently manipulated by pushbutton console controls.

mode to regulate reactor power according to amanual One control rod can be operated In an automatic mode to setpoint indicatedpowreronthelinearpowerlevel regulate reactor power according to a manual setpolnt, monitoring channel and a wide range power level indicated power on the linear power level monitoring channel 1.7 13.S monitoring channel (period) feedback.The autotnatic 18 13.5 and a wide range power level monItoring channel (period) Editorial portion or the reactor control system isInterconnected feedback. The wide range power level monitoring channel or thorough the automatic mode. The wide range power the reactor protection system provides Interlock signals and level moniltoring channel of the reactor protection system actions to the reactor control system. The reactor control provides Interlock signals and actions to the reactor system Is also Interconnected to the reactor protection system control system The reactor control ytem Isalso through a manual scram bar above the control rod drive Interconnected to the reactor protection system through a switches (allowing the reaetor protection system to be actua ted manual siram bar above the control rod drive switches, manually) and the automatic mode control (as described allowing the reactor protection system to be actuated above).

manually. -

Primaywater temperaturIs measured biaresstanee Primary water temperature is measured In the water box and dtcr RT)Inmhewatere box rssance Pmrywertemperature displayed on the console. A manometer Indicates flow rate 1-8 13.5.b displayed on the console. A manometer indicates flow 1.9 1.3.5.b through the cleanup loop locally. Equipment modifications rate through the cleanup loop locally.

-

_ Changed 'Imvl' to "levels:'

The reactorprotection system Isdesigned to ensure The reactor protection system Isdesigned to ensure reactor and reactor and personnel safeiy by imiting parameters to personnel surety by limiting parameters to operation within 1-8 l.53.c opemihoiitwithin analyzedopeatingrange Parameters 1.9 I.S3.c analyzed openting ranges. Process parameters that can Editorial that can automatically Initiate reactor protection system -utomatically initiate reactor protection system actions include mnclude msmronev rale orriso (perlod)J and fuel tempc tron l neutron level, rate orfise (period) and fMel temperature.

1-8 13.C 14 L53.C"...Aditmonliy."1.9

"- AddiTonally, 1. 1-9 iS3.cc Deleted "Additionally."

location ofthe temperatureintroduced "although' to identiryEdtra scram switch _Editoial Radiation monitors are Installed to monitor radiological conditions at the facility. One monitor is stationed on the top orthe reactor, with a local, high range indicator and alarm (at 5 1.8 13.d A system of fixed and moveable radiation monitors are Rllhr) to initiate evacualton of the reactor bay. One monitorl1 Modiflcation inpro installed to monitor... stationed at the control room door to the reactor bay, with a 2.5.mrem/hr-atarn setpoint. Electrical connections are Installed near each beam port, permitting control room and local Indication orradiation levels near an open beam port.

1.9 135.e .. subcriical. secured..." andreferencetobackuppower .. ..13.e Removed secured: reserved term: evacuation alarm does not Editori I

____1_____ for evacuation alarn have backup power suppVly at this time dto a

...'Potentlal concerns over these systems are address3ed 1.10 . 13.6 Inthis chapter of the Sarety Analysis 1-11 13.6 Removed sentence: ypo, should be and vice an Editorial

_________ Report .......eaipment an operton..." - -

U.. ITA.BUIATION. OPCHANG-ESTO DRAITKSUTRA:R SA.ANASIS EPART;:

-'-i- ." .-.. ;

. ' .7*.* 'Re.

1-tl 1.3.6hb Watcrretumns to an open surge tankby gravity. 1-12 1.3.6.c Water returns to an open surge tank (located in the reactor bay) by gravity. Clarification 1-11 1.3.6.d Reference louse orwaler transfer system to recirculate Deleted System changes Liquid sources ar limited principally to rnium-icaring- ng condensate water from the facility air handling system Liquod souates ar f imthe faciltity air handlin system. andi and occasional releases of tritiumbaring pimary coolanl occasional releases or t filufbeating primany coolant from from level adjustaents In the reactor tank or bulk-shield lcvel ajsensn or lank orbulk-shilank All tank. All reaclorbay loordrains and the HVAC level adjustments In the reactor tank or bulk-shicld tank. All condensate drains discharge to a .rb reactor bay floor drains and the hIVAC condcnsatc drains Il .- 1.3.7.

_for Discharges the r.actorbay sump are sampled .* and 11 13.7.b discharge to a reactor bay sump. Contents or the reactor bay Clarification and system assayed to assure limits for discharge am rnct prior to sump are sampled and assayed to assure limits for discharge changes discharge. A recirculation system filters sump water to arm met prior to discharge. Sump effluent Is filtered prior to allow discharge to campus sewerage (when radiological dischargo to meet NPDES requirements for discharge to requirementsar met). Liquid wastes a released through capus sewerage. Liquid wastes are released through the the sanitary sewerage system after filtration and assay for sanitary seweage system after filtration and assay for beta, beta.rram and alpha activity, gamma, and alpha activity.

1.12 13.8 Left ustified 1-12 1.3.8 Full iustification Editorial

...Although the shield water may be removed to allow ...Although the shield water may be removed to allow extraction of a vertical thermal-ncutron and ganuSaoray extraction of a vertical thermal-neutron and gamma-ray beam 1.12 13.&a beam (not done at the KSU facility), four 0.2S-in (63. 1-13 1.3.8a (not currently done at the KSU Facility at the time this report Prevent inhibiting the mm) holes are located In the tube at the top of the core to was completed), four 025-in (63-rmm) holes are located in the experiment prevent expulsion or water from the section or the tube tube at the top or the coro to prevent expulsion of water from within the reactor core the section ofthe tube within the reactor core.

1-12 13.8.b A rotary 40-position rotary specimen rack (RSR) Is 1-13 1.3. b A 40-position rotary specimen rack (RSR) is located in a well Editorial 13.8. located in a well In the top ofthe raphite radial reflector. In the top orthe graphite radial reflector. __E ________

The design of the Muel for the KSU TRIGA is similar to that for Fuels used in 70 reactors in 24 nations (General Atomics te fuel te K<SU. TWGA. is Thee designn of of tho Nt fior the Is similar

,_ to July 1999 data).

in operation Of total or under number ofwith construction trcactors, 45 are 40 rated forcurrently steady-that for fuels used in 70 reactors in 24 nations (General state thermal powers of 250 kW or greater, 22 at 500 kW or Atomics July 1999 data) Of total number of reactors, 45 greater, and 20 at I MW or greaer. Nine of the larger power are currently in operation or under construction with 63 reactors are TRIGA Mark 11. The TRIGA Mark It design is a rated for steady-state thermal powers or 250 kW or substantial fraction of the 70 reactors using TRIGA fuel world-greater, 36 at 500 kW or greater. There have been 35wieClrfclo.orctnfr 1-13 IS TRIGA reactors In the US., with 21 currently in 1-14 1 wuds. Claificionf corrcction for operation. As Indicated inTable 1.1,11 United States In the United States, there have been 26 TRIGA reactors built, useoftabies TRIGA Installations operate at power levels of at least with 19 currently in operation (S TRIGA facilities and 3 non-500kW. .... TRIGA reactors converted to operate with TRIGA fucl at power levels greater than 1,000 kW, as indicated in Table Principal decsign paramtectrs for the ICSU TRiGA aro1) given in Table 1.3. 1.1)

Major design parameters for the KSU TRIGA arc given In Table 1.3.

1-14 Table Tablenunbcr & placcrnc 1-15 Table nbcr placent Clarification, correction for

_ _ I__ 1.1 I _ _ _ _ _ __ _ _ _ _ _ __ _ _ _1___ _ _ _ .2 Ta lIub r l c m n use o ftables

- CLL LLLLaLLCcLC-C(--C-XC-a-L- LLL-LL-L-L-L-L-L-L-L-L-L-L-6666(-=

. %. *. .. S. . S. S. I % S *. I -. - - I ..

- kAB1LA0MONKFCHANGESrO:DRAFriKSU TRIGAB -SAFETY'ANALYSIS REPORT-..;i:_.?1~'"~-?-~i Table Table Table 1.2, U.S. MARK It TRIGA REACTORS. <Insetted Clarificati on, correction iro 1-14 . Table Id1, US.MARKItIITRIGA REACTORS. f-IS 1.2 bottom lineIn table> use ortabies. editorlal I-s Table Table 1.2, US. TRIGA REACTORS AT 500 kV OR 1-14 Table Table 1.1, US. TRIGA REACTORLS AT 500 kW OR Clariflcation, correction for 1-6 Table Table 1.3, KCSU TRIGA Reactor PrIncipal Design Table Deleted footnotes added line at bottom of table, reformatted . &

1.3 rarameters fit 500 kW Stesft-State Power - 1.3 table___________

1.16 1.6 List ofroutside users 1.17 16 Added University or Chicago & University or Nebraska at Upa userbase

________Lincoln 1-7 Table 1998 value for MWhi ofthermal enerff. 154 1-17 Table 1993 value for MwhT orthermal energy: 26 Correction

____ _ 1.4 1.IA________________________ ___________

1.8 Table Table 1- 9 1. Year.: 2002 Scheduled activities 1.17 I RevIsed to reflect accom plishment, &current estimates Overtaken by events Prodction or hydrogen fromn a high temperature zirconium.

water reaction Is a well-known phenomenon. Zirconium 3-12 3.5.1 OrigInal printing showed "I?' as or superimposed on usW 3.12 3.5.1 hydride does not exhibit the same chemical reactivity as Editorial zirconium, and tests demonstrate this reaction Isnot an issue

_______rot TRIGA fuel.

Fuel growth and deformation can occur during normal operatilons, as described In General Atomics technical report 13.1174933. Damage mechanisms Include fission recoils and fission gasecs, strongly Influenced by thermal gradients.

3-13 3.5.1 No co ~rrepodingtext 3.13 3.5.1 Operating with maximum long-term. steady state fuel Add inrormation temperature or 7S0'C does not have significant time- and temperature-dependent Mbel growth. Since the KSU reactor wHIlnot be operated In the regime vulnerable to this

___________________________________degradation, the damage mechanism Is not applicable.

Teeretme uead erisratvycrTherefore, temperatures and chemical reactivity orTRG TRJGA fuel matrix ensure that a zirconium water reactor Teeoe eprtrsadceia eciiyo RG 3.13 3.5.1.a will not occuir at magnitudes that could causeliarrd to 3-14 3.5.La Muel matrix ensure that a zirconium water reaction -will not Typographical error the reactor, occur at magnitudes that could cause haard to the reactor.

I . W

.., 11-1_,-.__'21__--_; _-_'

PtI.I.

'. ';, -', " - .1""'- -z, 7 4* _1 4.1 -IList of e erm na faiiis anet l(te al4-1 1 4.1 -Tangentlal (thermnal neutron) (Jr" Typographical error I ~~~neutron) (2rWI

.TAIJIATIN:O~CRN(ETO~lAT*I!~t TR(AA.~A~1 ANT.VJ~1~fllT:~ .. *~ . Y-

_____-____h

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to

Teo reactor was licensed in 1962 to operate ata steady-state The reactor was licensed in 1962 to operate at a steady. thermal power of 100 kilowatts (k). The reactor has been statethmlpowera oflOkilowatts (kW). The reactor licensed since 1968 to operate at a steady-state thermal power has been licensed since 1968 to operate aat Teaclor or 2S0 kW and a pulsing maximum thermal power of 250 thermal power of250 kW and a pulsing maximurm MW. Application is made concurrently with license renewal thertnal power of'250 MW. Application Is made to operate at a maximum of 1,250 kW, with fuel loading to onu tl ihlicense rene op o bmt W support 500 kW steady state thcrmal power with pulsing to concurrntly withlicenserenwal toopSt500kW S3.00 reactivity Insertion. All cooling is by natural tol statekW theronaitspower coteady. felwitmhnpulsing rr bondin 4-1 4.1 reactivity insertion. All cooling is by natural convetion. 4-1 4.1 fcionctcnt Tho(acoe.cnsiss.o.81f.clel.ent Powcrlevelforbounding The SO-W Icree)onsstsof lemntstypically (yialat leas'na~e for the 1,25-kW core). each analysis t he 550.kWcronsistsof8l Ibcl elem entstypicaly containing as mucor2rams or23U. The reactor core is in (at lcas nned for t2sThe S o ore)eachtcontaiinn r t form or a tight circular cylinder about 23 cm oasmucsf ramsoclU. Thereactor core ius anth (approximnaely 9 in.) radius and 38 cm (14.96 in.) depth, form eta oiztu eireular cywidcr about 23cm radiwuand positioned with axis vertical near the base of a cylindrical 3rm depth, positioned with auis vertical nearthebase watcr tank 1.98 m (6.5 QL)diaxnctcr and 6.25 m (20.5 ft) ota cylindrical.watertank 198m iameter and6.2S depth. Criticality is controlled and shutdown margin assured depth..... by control rods in the form of aluminum or stainless-steel clad boron carbide orborated graphite

....Criticality is controlled and shutdown margin assured by control rods in the form of aluminum or stainless-steel clad boron carbide or berated graphite. . Reactivity requirements

.... Critieality is controlled and shutdown margin assured (i.e., minimum shutdown margin with the most reactive rod 4-1 4.1 by three control rods in the form ofaluminum or 4-1 4.1 fully withdrawn and maximum excess reactivity) can be met Clarfication stainless-steel clad boron carbide or borated graphite. ror250 kW with three control rods, but reactivity required to compensate for fuel temperature and fission products ror operations at power levels of 500 kW requires four eontrol rods to meet reactivity requirements.

4-1 4.1 The tank and shield ar in a 4078-m3-contaianent 4-2 4.1 The tank and shield ar in a 4C78re (144,000 00) terminology building.... Colicicrtbfctdn 4-2 4.1 In 1968 pulsing... .2 4.1 In 1968, pulsing... Editorial it is more convenient to set a power level limit that is It is convenient to set a power level limit that is based on based on temperature The design bases vglysis indicates temperature. The design bas alysis indicates that operation thatoperationatupto t900kW(with I an ent ore at up to 1900 kW (with an element core and 120@F inlet and 120°F Inlet water temperature) with natural water temperature) with natufrconvective flow will not allow convective flow will not allow film boiling therefore high 4-4 film boiling thereiore high fuel and clad tipcramt coapable 42 the! and clad temperatures wh ceould cause loss ofclad of causing loss of clad Integrily cannot occur. Ancemcnt Editonal integrity could not occur. Aricement core distributes core distributes the power over n largeryjume of heat the power over a larger volinn'oheatgenerating generating clement$, and therefore usingUelements in elements, and therefore results Ina more favorable, analysis results In a less favorablc moreon ative thermal more conservative thermal hydraulic response. hydraulic response.

4_ __ 4.2.1.a Various 4-5 42.L.a Added Sl orSAB units, as approptiate Add Inroatiaion 4 4.2.1.b I J. ________________

Reernco tonroblems olfdcnsilt chances 4-5 1 4.2.1.b Added tin earlier dbsians." Clarification

( C fCL L(_ L((_c_(_cC_ CG- C-a-G-XL--XC-a-a-a-G-SCG-S-A

. .i'~TABUTiATION.OF'CHAGES .TO.DRAFT-KSU:-TRIGA-ITSAF TNANALYSTS'REPflRT'.Y~", ';."-f " -

.0 ' :e :t.

! e';t! . ! , :, f i z 1 '.' {,:;-,9, "~. I,:. *.." *v ' !\i:* ;2 :;-' ':4i l:* R-:;s A typical layout for a KSU TRIGA 11 250.kW core (Core I1-The layout otr the 2SO-kW core IsIllustrated in Figure 18) Is illustrated in Figure 4.4. The layout for the I.2SO-kW 4.7 42.1.c 4.4.' e layout for theSO-kW cor Is expected lobe 42.1.c core is expected to be similar, exept that the graphite Correct error vety similar, with about 4 graphite elements replacing elements will be replaced by fuel elements one additional three fuel elements and one control rod. controt rod will be added, and control rod positions will be adiusteld.

The additional fuel elements are required to compensate for higher operating temperatures from the higher maximum steady state power level. The additional control rod Isrequired to meet reactivity control requirements at higher core reactivity associated with the additional fuel. The control rod positions 4-7 4.2l.c No corresponding text 4.7 4.2.1.c will be difetrent to allow a higher worth pulse rod (the 2S0 kV Clarification puise dreactivity worth Is S2.00, the 1,250 kW core pulse rod reactivity worth is S3.00), balancing the remaining control rods worths to meet minimum shutdown margin requirements, and meeting physical constraints Imposed by the dimensions of the pool bridge The pulse rodis1.25 In.diameter. Other rods are 78 In.

4.7 4.22

-borsted diameter.gtsphite, Control cladrods area 20 In.long boron eatbide with 30-mill alumninumt sheath or . 4.8 42.2 With exception ofinitiat paragraph, extensiveremwrite.

_ were RA1t2 While three control rods adequate to meet Technical Specification requirements for reactivity, operation at 500kW requires control byfour control rods (three standard and one transientlpulsing control rod)... Interlocks ensure operation of the control rods remains withinnnalyzed conditions for v oreactivity Re4ctivi.yaord control or limit potential for accident scenarios.

hecoSUr eatoraIs ient(pulsing) c ontroll our while scrams operate at limiting safcty system settings. A Correct error & clarify use of 4.7 4.22Zn standard rods plus one translent(pulsing) ontrol 4-10 4.2.2.a detailed eontroldescription of the control-rod system is provided in interlocks rod Chapter 7; a sumanry orinterlocks and scrams is provided below In Table 4.2 and 4.3. Note that (1) the highflel temperature and period scrams are not required. (2) the fuel temperature scram limiting selpoint depends on core location for the sensor, and (3)the period scram can be prevenled by an Installed bypass switch.

4-8 4.2-a No conresponding text 4-11 4Z2.a Table 42 &43 added 4-8 Add information 42.b Reference to 500kV operation 4.12 42-2b Added approximately toquallty 500kW Clarification Nominal speed orthe slansie Nominal speed of the standard control rods Is about 12In.

control ods 22Isabout in. (305 cm) per minute (viththe stepper motor specifically 4-8 422mb witha otl travelsaboutSin. M aimumratof 4-1Z 4.2b adjusted to this value), or the transient rod ISabout 24 In.(61Addinonnaion reactivity changefor standard control rods IsspecilledIn Maximum rate orreactivity change forstIndatd control rods Is Technical Specifications. cper inuTehwita totaltalao u I .

-: ;TABULATION OF.CANGESTO'DRAF.KSUTRIGAHSAFETVY.ANALYSISREPORT.?t-'i' qi-- -

Pag g;1it,.:

.. . Pa;rp .;s . 4.2'i .' -*e;ts;> t.Y ; - - *j .... *...*Rt n Hydrogen in the Zr-H fuel serves as a neutron moderator.

Dcmincralized light water in the reactor pool also provides neutron moderation (serving also to remove hcat from operation or the reactor and as a radiation shicid). Water 4.2.3 No corresponding text 4-12 4.3 occupies approximately 3S% or the core volume. A graphite Add information consistent rcflector surrounds the core, except for a cutout containing the with rcvicw standard rotary specimen rack (described In Chapter 10). Each fuel element contains graphite plugs above and below fbel approximately 3A in. in lcngth, acting as top and bottom reflectorsd 4-8 4.14 St and SAE units used 4-12 42.A Added SI orSAE units parenthctically Add information The fMel elements are spaced and supported by two 0.7S- The fuel elements are spaced and supported by two 0.75-in.

In. thick aluminum grid plates. The grid plates have a (1.9 cm) thick aluminum grid plates. The grid plates have a total of91 spaces, up toSS ofwhlch ar Filled with fuel- total of 91 spaces, up to 85 or which are filled with fuel-moderator elements and dummy elements, and t moderator clcmcnts and dummy clements, and the remaining remaining spaes with control rods, the central thimble, spaces with control rods, the central thimble, the pneumatic 4.9 42.5 the pneumatic transfer tube, the neutron source holder, 4.13 42.5 transfer tube, the neutron source holder, and one or more Editorial and one -ormore voids. The bottom grid plate, which ids Thebottomgddplatewhichsupports the weight orthe supports the weight of the fuel elements, has holes for fiul element has holes for rceiving the lower end fxtures.

receiving the lower end fixtures. Space Is provided for Spacc b provicd for the passageo r cooling water around the th passage of cooling water around the sides of the sides of the bottom grid plate and through 36 cxpcimen bottom grid plate and through 36 special holes init. The penctrations; The l.S-in. (3.p la) diameter..

-.5-in. diameter.. pn___

__ os__'Me_________3 ___cm)_diametr..

I

- (_ L_ C_ (_ L_ C_ (_ ('_ (_ (_ (_ (_ (_ (_ C_ cm L_ L_ L_ L_ (__ L_ L_ C. C__ L. C. C. (_ L. C. (_ (_ C. (_ C. I-. (_ (_ (__ (_ L.

The KSU TRIOA reactor core support structure rests on the base ofa continuous, cylindrical aluminum tank surrounded by a reinforced, standard concrete structure (with a minimum concrete thickness or approximately 249 cm. or 8 ft 2 in), as Illustraled in Figures 4.1 and 42. The tank is a welded aluminum structure with 0.63S cm. (114-in.) thick walls. The tank Is approximately 198 cm (63-fl) In diameter and approximately 62S cm (203-fl) In depth. The exterior of the tank was coaled with bituminous material prior to pouring concrete to retard corrosion. Each experiment Facility penetration In the tank wall (described below) has a water collection plenum at the penetration. All collection plenums are connected to a leak-offr volume through Individual lines with isolotion valves, with the leak-orr volumes monitored by a pressure gauge. The bulk shield tank wall Is known to have a small leak into the concrete at the thermalizing column plenum. therefore a separate Individual leak-off volume (and The KSU TRIGA reactor core support structure rests on uesisure gauge) Is Installed for the bulk shield tank; all other plenums drain to a common volume. In the event of a leak the base of a continuous, cylindrical aluminum tank from the pool through an experiment faeillty, pressure in the surrounded by a reinforced standard concrete structure volume will increase. Isolating Individual lines allows (with a minimum thickness ors It 2 In.). as llustrated In Add information consistent 4.9 43 4-13 43 Identificatlon orthe specific facdity with the leak. with review standard Figures 4.1 and 4.2. The tank Isa welded aluminum structure with 1/4-1n. thick Walls. The tank 1s 6s-nt In A bridge or steel plates mounted on two ratils or structural steel diameter and 205-fl Indepth. The minimum thickness or provides support for control rod drives, central thimble. the concrete shielding at the core level is 8 R.2 In. rotary specimen rack, and instrumentation. The bridge Is mounted directly over the core area, and spans the lank.

Aluminum grating with clear plastic attached to the bottom Is installed that can be lowered over the pool. The grating normally remains up to reduce humidity at electro-mechanical components or the control rod drive system and to prevenl the buildup of radioactive gasses at the pool surface during opations. The grating can be lowered during activities that could cause objects or matcrat to fall Into the reactor pool.

Four beam tubes extend ftm the reactor wall to the outside or the concrete biological shield in the outward direction. Tubes welded to the Inside otthoe wall extend toward the reactor core.

Three of the tubes (NW, SW, and SE) end at the radial reflector. The NE beam tube penetrates the radial reflector, extending to the outside or the core. TWo penetrations In the tank allow neutron extraction Into a thermal column and a l _ . thermalizing column (described inChapter I10).__

49 __ l 4A4 Sland SAE units used j4-13 j4.4 _ Added SI or SAE units parenthetially Add inrormation

CRNGET'nnlrrT S TR-a; IRA AAT.:: 4fl1TiV.r'flPnAIONTI flf:iVC.

,.- P *',- , - - -, ,.;: :; .;

...For fuels with j cenrichment, the value Is . ...For lbels witl U,c]enrichmcnt, the value isnearly 4-9 4.5 ncarlyconstanaxpol er IC, and varics only 4.14 4.5 constant at 0.01% /k per eC, only wcakly dependent on Editorial weakly dependent on geometry and temperature. geometry and temperature

...The design bases analysis Ini~i tes that operation atThe desIgn bases ainaly j indicates that operation at I,250kW 500 kW thermal power with znF*Icmcnt across a broad thna oe iha leIntarsabodrngofce range of core and coolant Inlet temperatures with natural hrapoewihn lmntcosabodrngofoe convective flow will not allow film boiling leading to and coolant inlet temperattures with natur convective flow 4410 4.5 high fuel and clad temperatures that could cause loss or 4.14 4.5 wilt not allow film boiling tat could lead to high fuel and clad Powcr level for bounding clad integrity. caingrttemperatures that could cause loss of clad Integrity. analysis Increase Innmaximum thermal power from 250 to 50 kW Increase In maximum thermal power rrom 250 to 1,250 kW kW does not affect An_____ does not affect flndamental...

The limiting core configuration for this analysis is a The limiting core configuration for this analysis is a compact compact core defined by theTRIGA Mk 11 grid plates core defined by the TRIGA Mk It grid plates (Section 4.2.S).

(Section 42.5), namely.. Me grid plates have a total of91 spaces, up to BSorwhich are filled with fuel-modcrator elements and graphite dummy Thgrid plateslhaved total a of91spacesustoand elements, and the remaining spaces with control rods, the 4-10 4.1 which are filled wilh Fuclmoderthor graphile dummy elmcents, and theremainin lementsspaeeswith sad 4-15 43.1I central holdcr thimble, and one theor pneumatic more voidstranser In the tube, E or the neutron source F (outermost two Clarificalion control rots, the central thimble, the pnnunatic transfcr rings) as required to support experiment operations or limit The bottohngrid platowhichoslrandponor Thet boettom grid polaeso t orevoids.

wh cich ngrthe lweigto ten orts excess reactivity. hle bottom grid plate, which suppofts the weight or the fule elements, has holes for receiving the lower fiue lemenis, has holes for reiving the lowe end end fixtures.

4.10 Table Position change 4-S 4Ta2bl Position change Editorial 4.12 45.3.a No corresponding text 4-16 4.53.a Substantial rewrite, incorporaling steady state calculations Add infiormation 4.12 43.3.b Substantial re-write 4-20 453.b Substantial re-write Power level for bounding

___ _________ ___analysis-,

___ ____ RAT # 3 & 21 4-14 45.3.c Substantial ro-write 4-23 4.53.c Substantial re-write Power level for bounding analysis 4-17 Table 4.6 Upperpowcr lcvel S00kW 4-26 4.11 Table extended to hlghcrmaxlmumn power Power level for bounding

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _analysis I t,,

(

C -( ( C-( -( -( -( -( -C -L-(-L -(-L -(-(-L - - - - - - -

I % % %. k. % % I a... 'I . % . *. ' . .- %. .- S. *- % % '

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% I . . - I B As described In 3.5.1 (Fuel System) and NUREG 1282. fuel As described In3.5.1 (Fuel System) and NUREG 1282 tlemperature limits both steady-state and pulse-mode operation.

fuel temperature limits both steady-state and pulse-mode The Iuel temperature limit stems from potential hydrogen operation. The fuel temperature limit stems rmnt oulgeassing from the fuel and the subsequent stress produced In potential hydrogen outgassIng from the fuel and the the fuel element clad material by heated hydrogen gas. Yield 4-17 4.7 subsequent stress produced Inthe fuel element clad strength orcladding material decreases at a temperature of RAT #4 & 5, Clatifleation material. The maximum temperature limits oft 15050C 4.0OC; consequently limits on fuel temperature change for (with clad c 500°C) and 950iC (with clad > 500C) for cladding temperatures greater than SOOC. Amaximum U-Zr!! (H/Zri. 6,) have been set to limit internal fuel temperature of I IS30C (with clad 'c500C) and 9501C (with cladding stresses that mihtl lead to clad Integrity clad > 500°C) for U-ZrI (lH/Zrl m)will limit internal ruel (NUREG 1282)... cladding stresses that might lead to clad integrity (NUREG 1282) challenges 4-18 4.8.2 Substantial re-write 4.26 4.8.2 Substantial we-rite RAI f 6.10 4.18 4.83 Substantial re-write 4-27 4.8.3 Substantilal re-write RAI# 6.7 4-18 Table Deleted 4-27 NA Removed table RAT ff 8 This cooling system combination provides enough heat removal ior continous full-power operation. Inaddition This cooling system combination provides enough heat to the cooling system, the reador Isprovided with a bulk- removal for continuous full-power operation. In addition to shielding tank. This 650-gallon (25 l.) tank contains the cooling syslem, the reactor Is provided with a bulk-distilled water, and can be used to supplementn mkeup shielding tank. This 6500-gallon (25 kL) tank contains 52 5.1 water for thepri arytank or provide temporasy fuel 5-2 5.1 distilled water, and can be used to supplement make-up water Clarification storage. Makeup water for boh systems Isiprovided by a for the primary tank (using a makeup water system steam-powered still.... Although the cooling system Isnot Independent of the above drawing) or provide temporary fuel required to be operating during reactr operation, there is storage. Makeup water for both systems Is provided by a an administrative requirement that the system be capable steam-powered still ....ln...

ofoveratilon (operable). In normal...

Principal functional requiremenis of the primar coolant rienipal functional requirements of the primaty coolant system are to (1) transfer heat rom the reactorcore to the system are to (I) Iransrer heat from the reactor core to the secondary cooling system, and (2) provide radiation seconday cooling system, and (2) prvide radiation shielding Co tion

-3 52 shielding directly above the reactor core. Although S-3 S2 directly above the reactor core. Although natural convection orree natural convection cools the reactor core proposed cols the reactor core, primary bulk waler temperture should Technical Specifications require primaly bulk water be kept below 130'F (48.9 ?C).

- - temperamture be kept below 130FP (48.9 C).--

A plate-type compact heat exchanger Is used to remove A plate-lypo compact heat exchanger is used to remove heat heat rrom the primary coolant (see Flgur 5.2). The heat r the primary coolant (see Figure 5.2). The heat exchanger 52 54 exchanger consists orsandwiched alternatelycarrying stainless steel primary and secondary plates cooling water. 53 S.2 consists canying orsandwiched stainless cooling primary and secondary steel plates alternately water. The heat Modification TheMheat exchanger has a transfer capacity or (2,327,080. exChangerhas a transfer capacity or 682 kV (1,709,000. BUT BUT h') (500 kW) under normal conditions. h') under normal condilTons. _

Iramajor loss of coolant were to occur, there are two Ifa major loss ofcoolant were to occur, thee ar three level levet sensors that would Illuminate a light on the control sensors that would illuminate lights on the control panel. Two 5-4 5.2 panel. One sensor is located inthe reactor bay sump, 5-4 52 sensors are localed Inthe reactor bay sump. activating when activaling when the surnp Id flill. Since all floor drains In the sump level Ishigh. Since ill floor drains In the reactor bay Equipment modification the reactor bay connect to the sump, any leaks would connect to the sump any leaks would accumulate there. A

- - accumulate there. A second sensor Islocated at the top of third sensor is located at the top of the tank....

__________ ~the tank.... _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

54 53 5-4 53towercapacty Cooling Coolin tower capacity wasupgradd was upgraded 5-4 SJ 33 Thi secondary operalion cooling system at approximately 723iskW.

designed for continuous Carfcto The water from the surge tank is then drawn into the The water from the surge tank is then drawn into the system by system by a dircct-coupled, sclf-priming, centrifugal a direct-coupled, selr-priming, centrifugal pump (see Figure pump (see Figure 53). When the controls are Ina normal 53). When the controls are in a normal configuration, the 5.6 53.1 configuration, the purnp Is energized (with the primary 5-6 5.3.1 pump is energized (with the primary pump and cooling tower Equipment modification pump and cooling tower fans) from a backlit push button fans) from a backlit push button switch in the control room.

switch in the control room. Normal flow rate through the Nomnal flow rate through the system is 250 gallons per minute system is 176 gallons per minute (11.1ILs'). (IS.8 lest ).

5-6 Ftg 53 New picture 5.6 Fig 5.3 New picture Equipment modification To detect possible leaks In the heat exchanger, the To delect possible leaks In the heat exchanger, the secondary secondary water is tested monthly for radioactivity. Since water is tested monthly for radioactivity. Since the primary the primary tank has 22-feet (6.7 n) orstatic had, as tank has 22-fect (6.7 m) ofstatic head, as opposed to 3 to 9-opposed to 3to 9-eet (0.9 to 2.7 rn) in tho secondary feet (0.9 to 2.7 an) in the secondary (depends on surge tank (depends on surge tank level), a breach In the heat level), a breach in the heat exchanger would result in flow exchanger would result in flow from the primary tofom _ the is os s e secondary cooling system when the cooling system Is fromin sythe pm tsseconed.Lary wilcooling h system whn sh secured. Leaks as a result ofrsystem operating pressur coolraingsytmunisey s herdLeaks whileather cooinprsysuetemtis 5-8 5.33 ar unlikely test pressure was 150 psig while since the 548 53.3 peratoig aea nlikietheyd ha t escune w prese tes Clatrfication maximum pressure differential across the heat exchanger het exchsg.andgter (rmaximpump presunnng diserentilacros thep Claicaio (primary pump running, secondary pump or!) is30 psig. hs3 sdAsaleatrahi exchangerh mig e ondr wpiaypm ump ofe A small breach In the het exchanger would be evidenced wide3 p Arstill y breacinathe heat ech ar woul r. A by tritlium contamination or the secondary water. A larger largernndicad breach would be by loss of prinary coolant brcach would be indicated by loss orptimary coolant from the reactor tank. Primary water typically remains within from the reactor tank. Regardless, primary water ICFR20 limits for release to sanitary sewes and is not a typically remains; within IOCFR2O limits for release to hazard even if a leak were to occur.

sanitary sewers, and is not seen as an eminent hazard.

The cooling systems return line to tho reactor pool cnters the pool through a diMiser. The diffuser is constructed to induce a helical flow pattern in the reactor tank. This extends transport The cooling systems return line to the reactor pool cntets time or the convection flow of water from the core to allow the pool through a diffuser. The diffuser is constructed to much or the nitrogen-16 generated during operation to decay induce a helical flow pattern in the reactor tank. This before reaching the pool surface. A radiation monitor directly extends transport time of the convection flow orwater above the pool surface provides the control room operator with from the core to allow much of the nitroSen-16 generated information to prompt exposure controls (generally energizing during operation to decay before reaching the pool the primary cooling pump to initiate th heliical flow for deay 5.10 5.6 surface. A radiation monitor directly above the pool or limiting access to the arma directly over the pool). Pool 5r10 Ssuface provides the control room operator with S9 S.6 surface monitor radiation measurements at 250 kWV directly Clarifcation informalion to prompt exposure controls (generally above the pool surface are typically 30 to 40 MR.h 1 from all energizing the primary cooling pump to initiate the helical sources, and Is expected to be 60.80 mttht at 500 kW flow for decay, or limiting access to the area directly over operation.

the pool). Waist-level radiation measurements at rull licensed power directly above the pool surface am A radiation monitor at the rail around the pool provides the typically 5 to 10 mR-h l (10 pGy-h') form all sources. control room operator with information to prompt exposure controls for personnel on the 22-foot level but not directly over the reactor pool. At 250 kW, radiation levels at the rail are less than 2 msiRhr.

% I...;

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-(_ L L L (_ L L (_ L (_ C (_ (_ (_ C. r _ C__ (__ (, C_- L- L_ L C__ L_ L C__ L_ L_ L_ C_- C_- C___ C.. L_- C_- r- r-- (__ (_- (- (_

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The control console and display Instruments are primarily he control console and display Instnnnents are primarily housed In i control console, with auxiliary instruments located housed In a control console, with auxiliary Instruments in arack next to the console. At the console, the reactor located Ina rack next to the console. At the console, the operator has direct control over mode of operatlon. control rod reactor operator has direct control over mode or drive positions, cooling system operation, opening ofreactor operation, control rod drive positions, cooling system bay doors, and manual scram orthe reactor. Display operation, opening orresctor bay doors, and manual Instruments located in the control console provide scram orthe reactor. Display Instruments located In the measurements orrectlor power, control rod positions, primary 7-2 7.1 control console provide measurements of reactor power, 7.2 7.1 water temperature, and fuel temperature Indicators in the Equipment modifications control rod positions, primary water temperature, and fuel conole display scram Infornmation. low air pressure, low lemperature. Indicators in the console display scram primary water level, high reactor sumip water level, sump high itrormatlon, low ait pressure, low primary water level, water level, sump overflow water leveij secondary surge tank high reactor sump water level, sump full, secondary surge level low, source Interlock status, reactor bay upper door open.

tank level low, source interlock status, reactor bay upper reactor bay lower door open, thermal column door open, door open, reactor bay lower door open, thermal column person on stairway, and rod drive status. Secondafy surge tank door open, person on stairway, and rod drive status. makeup Is controlled with a backlit pushbutton that indicates

- - surge tank low level and surge tank makeup valve operation.

The primary function of the RCS is to goven the manner The primary firrction orthe RCS Is to govern [be manner In Inwhich reactivity Isvatied in the reactor core. The RCSThpimrfnconoteRSstogentemaern sysemhoul preavyIsvarent thei react prtor cm. uii RC which reactivity Isvaried Inthe reactor corm The RCS system system Ins thnleargor oterator eutm enshoall should prevent the reactor operator front unintentionally unintentionally inserting large amounts ortreadivity, Inscrting taw amouts orreactivity, through vartous intaiocyc through various Interlock systems. The operator shouldIneinlagamutoratitytruhvaosItroc 7-3 7.2 only be abe to remove one rod at a time fon the reactor 73 7.2.2 systems. oeoperator should only be able to remove one rod 7 nl prevabenting v oe o at a73 time fmmtemlr 7Z talm rom the reactor core, preventing large insertion rates. Correct wording core, prevenling large Insertion rates. The pulse rod must The pulse rod must not be able to be rapidly ejected fromn the not be able to be rapidly ejected from the core while In cote while In steady-state operation. Furthnermore, the pulse steady-state operalion. Furthermore, the pulse rod should rod should be the only rod that can be Withdrawn fi pulse be the only rod that can be can be moved Inpulse mode. mode. preventing supercriticacpulses.

Preventing supercritical pulses.

Tlie primary function of the RPS is to automatically Insert the control rods into the reactor core when certain The primary function ofthe RPS is to automatically insert the parameters deviate from limited safety system settings. control rods Into the reactor core when certain parameters Several scrams involve the neutronle channels In the devieae fom limited satety system settings. Several scrams 7-4 72.2 RCS. Irf 101h rated power level Isexceeded In steady 7I 7.22 Involve the neutronic channels in the RCS. 111107%rated SCRAM removed form state mode, one oftwo trip-poInts will scram the reactor. power level is exceeded in steady state mode, one or two trip- Technical Specifcalions Failure orthe high voltage power supplies ror operaling pointswill seram the reaclor. Failure orthc high voltage neutronlo channels will also caU a scram. For pulsing power supplies for operating neutronic channels will also operstlons a scram will be actuated when the fuel cause a scram.

temperature IsInexcess of 450C.

I _'_e overall system layout Isdepicted In Figure 7.2. lbe The overall system layout Is depicted InFigure 7.. 'he majority orthe RCS Is housed in 4 General Atomics (GA) majority of the RCS Is housed Ina General Atomics (CA) console originally manufactured for the USGS reactor, which 7-S 7.23 console originally manutaetured for the USGS reactor, 7-5 7.2.3 Is shown with modifications In Figure 7.2. A detailed Clarification which Is shown with modifications In Figure 72. A description of this figure is provided in Table 7.11 Figure 73 detailed description or this figur Isprovided shows arepresentative layout or the auxiliay instirumntation

_ rack 7-6 Fi 7.3 Labels changed 78 Fi- 73 Timer& Pool Light controls removed Equipment modifications Table Scram Status, Source Interlock, Low Air Pressure, Person Table Scram Status, Sour Interlock, Low Air Pressure,Hi & HI-Hi 7-7 7.l,No. on Stairway, Upper& LowerDoors, and Cooling System 7-7 7.1, No. sump level, surge tank level & makeup, Upper& Lower Equipment modifications 13 Powerr - 13 Doors, and Cooling System Power

,-.Y-.Y - - -- - r The remainder of the channel circuitry is located in the The remainder otthe channel circuitry is located in the NLW-NLV- 1000 unit in the central eonsole. The NLW-l000 1000 unit in the central console. The NLW-I000 unit supplics unit supplies thc high voltage for the detector and power the high voltage for the detector and power for the for the prearrpliier. The instrument switches from pulse preamplificr. The Instrument switches from pulse mode mode operation to current mode as reactor power operation to current mode as reactor power increases out of the Increases out of the source range allowing the instrument source range, allowing the instrument to measure reactor to measure reactor power In the upper ranges. Three power in the upper ranges. Three displays indicate reactor displays indicate reactor power, high voltage, and reactor power, high voltage, and reactor period. Tho power signal is period. The power signal Is permanently recorded via an pennanently recorded via an opto-isolated output to a strip-7.9 7.3.1 opto-isolated output to a strip-chart recorder located In the 7-8 7.3.1 chart recorder located In the instrumentation rack. The period Clarification instrumentation rack. The period meter has a scramn at 3 meter has a scram at 3 see and there Is a high voltage scram, see and thee Is a high voltage seram, both of which are both of which arm bypassed in pulse mode. This channel also bypassed In pulse mode. This channel also provides a provides a protective interlock which prevents rod withdrawal protective interlock which prevents rod withdrawal when when indicated neutron nlux is < 2 cps, which is also activated indicated neutron flux Is < 2 cps, which Is also activated In pulse mode to prevent removal orthe shim, safcty and In pulse mode to prevent removal of the shim, safety and regulating rods. Another interlock prevents pulsing when regulating rods. Another Interlock prevents pulsing when reactor power is abovo 10 kW (normally set at I kW). The reactor power Is above I kV. unit has two calibration checks in pulse mode, two In current mode, and checks for the neriod and high voltage scrams.

The original insttumnents that the N-1000 series units replaced are still housed in the control console for backup use. These older analog devices have all or the same measurement and RPS features, except that they lack opto-isolatcd outputs for comnputer acquisition or reactor data. Thoy also require more manual Input as the linear 7-11 73.1 channel 'does not possess outomranging fcatures These 7.9 7.3.1 Deleted Not needed instrutnents were used for many years at USGS and for one year at K-State until the N-1000 units arrived, and provide adequate backup for an Interim time while the N-1000 series tuits are serviced. Wiring diagrams and calibration procedures for these Instruments are located in the maintenance manual for the USGS console.

Temperature Indications for the pinmaty water and specific 13-Ring fuel elements ae provided on the front Temperature indications for the primary watcrand specific B-section of the control panel and in the instrumentation Ring fuel elements aro provided on the fiont section of the rack. The instnunented fuel elements have three chromel- control panel and In the instrumentation rack. The alumel thermocouples in the fuel element that are used for instrumented fuel elements have three chromel-alumel temperatur indication on the console or In the thermocouples in the fuel element that are used for instmentation rack. Since all three thermocouples are temperature indication on the console or in the instrumentation Clanfiation located 0.76 cmlO34in below the fuel surface and with rack. The thermnoouples ar located 0.76 em (0.3-in) below two spaced only 2.5 cm rom the third at the midpoint of the fuel surfacelspaced at the midpoint orthe element and at the eeirment, an averaged value from all three +, 2.5 cm from the midpoint; an averaged value from all three thermocouples fora single element Is typically used for thermocouples is typically used for instnument readings.

instrument readings. -

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v. % % . S.

v x. v % . a. ** I S. % I. I I - . - .

Four control rods aire reuired roFour reactor ~i . sa eco operations

~~~~~~ Four~ocontrol rods~ are required~ fOr reactor

~ operations at 1,250 si ronlro aosacrq~e 500 kW: a shim rod, a regulating rod, a transient rod, and el tio meenracivt consientro, and a s: a shirod rod, a a sarety rod. The shinu regulating and safety rods share regulating rod safansient rodsand enta oynrod. The shim Identical and circuity and provide coarse and fine power (Fgularten and sprely rods share identieal control circuitry control. Two orthe rod drives are original analog (Figur 7.7) and provide coarse and Gne powerconlrol. Two 7.12 7.3.3 systems. Oneoftheroddrivesusesnastepermotor. 7.11 7.3A otriherddrvesnreoriginralana'ogsystems. Oneofiherod RAr#I 12 Theut"rod s tht sI dsiged canbe aily eectd didves uses a stepper motor. Drive position IsdetermIned by Tothe pulse tod is dresigedts thagt toiiticate Tlowthcie orest penionseiaghst to initiate l arapl ejctrped reator puIsn. voltage rod control dropdrive across a potentiometer Is moved. that Is The position adjustedfor indicator as the the toevdser it still All ctons as a normal control rod In analog molors is attached to a shall coupled to the drive motor steady state mode. Al1 rods eanbo individuallyrscramned shart with a setscrew, while the stepper motor Is connected to writhout shutting down the reactor, the position indicator with achain drive _ _ _ _

The analog rod drive motor Is dynamically braked and held by an electrically locked motor. In the static condition, both Thc rod drive molor Isdynamically braked and hld by an vindings are energized with the same phase (see Figure 7.7),

electrically locked motor. In the static condition, both electrcally locking The mtor. Cloc s (p or tera windings are energized with the same phase (seeFgre clockwise (down) rotation Isenabled by sli fling the phase 7-12 73.n 7.7) electrically locking the motor. Clockwise (up) or 7-11 73.4.a between the windings with a I-pF capacitor; motor control RAT# 12. 13 counter-cl ockwise (down) rotation Isenabled by shilling switches allow the appropriate phase shift. The stepper motor the pha.ebetween the windings with a .I _pcolmit.o operates using phase switched direct current power. The mtorcontrolpswhtcase alow the mnotoreontrol switches allowv theapopriate ase Imotorshaft advances 200 steps per revolution (1.% degrees per appropriate phase shilt. step). Sinece urrenl Is maintained on the motor winding when the motor Isnot being stepped, high holding torque Is

- maintained. A translatorrmodule drives the steppingmotor.

The pulse rod is the only control rod that can be moved 4. The pulse rod Is the only control rod that can be withdran" 7.17 73.4.c.4 when the reactor IsIn the PULSE mode (this does not 73A.oc when the reactor is in the PULSE mode (this does not prevent Correct wording prevent the scramming ofany control rod). the scramming orany control rod).

Additionally, there Is an Interlock that prevents reactor Additionally, ther Isan Interlock that prevents reactor pulses pulses from being fired isthe reactor power Isabove I from being tired Itthe reactor power is above 10 kW 7-17 7.3.4 kW.There is also akey switch for bypassing the source 7.16 73.4 (nomally set at I k J There Isalso a key switch for Clarficalion bypassing the source interlock during Nbet loading operations Carfato interlock during Nuel loading operations to check for to check for criticality.

criticality.

in pulse mode, the mode selector switch Is set to the Ill In pulse mode, the mode selector switch is set to the HI PULSE position, Interrupting detector signal to the linear PULSE position. Interrupting deteclor signal to the linear 7.18 7.4 channel. When the pulse Interlock Is activated (to Initiate 7.17 74 channel. mhen the pulseinterlock is activated (toInitiate the the source Interlock) to prevent movement otlhe shim, source Interlock) to prevent withdrawal of the shim. safely and Correct wording safety and regulating rods.the detector signal to the regulating rods, the detector signal to the logarithmic wide logarithmic wide range detector is interrupted. raon detector IsInterrupted.

Ther ar several additional pieces orcquipment in the There ar several additional pieces of equipment In the control control room. Directly behind the operator are the circuit room. Directly behind the operator are the circuit breakers to breakers to Interrupt power to electrical devicesIn the interrupt power to electrical devices In the control room and control room and reactor bay (see Figure 7.13). A halon reactor bay (see Figure 7.131). A hlaon fire extinguisher Is fire extinguisher Is located next to the breakers for useIn located next to the breakers for useIn fighting electrical fires 7.18 7.6 fighting electrical fir. Currentcore and ichility 7.17 7.6 Current core and facility configuration is shown In a display ClarIfication configuration s shown In a display cabineL A wall- cabinet (Figure 7.13.b). A wall-mounled box In the control mounted box In the controlroom hasIlluminated switches room has illuminated switches to Indicate personnel In the to indicate personnel In the reactorbay. Aradiation area reactor bay. A local radiation area monitor (including monitor is located above the doorIn the control room to Indicator and alarm) Is located above the door Inthe control the reactor bay. room to the reactor bay.

The pneumatic transfer system or rabbit Isused to rapidly 'De pneumatic transer systern or rabbit is used to rapidly transport samples between an in-core location and the transport samples between an in-core location and the Neutmn Neutron Activation Analysis Laboratory. From Activation Analysis Laboratory. From commercial cylinders.

commercial cylinders, compressed helium OIlfs small compressed helium fills small tanks at either end orthe 9.6 9.7.3d tanks at either end orthe system. Pressure Is limited by 73.d systen Pressure Is limited by release valves at 275 kPa (40 Clarification release valves at 275 kPa (40 psi). The system is opratled psQ). The system Isoperated from the instrumentation rack from the control room where the reactor operator sets the (Fig 7.3) In the control room, where the reactor operator sets direction ofmotion by positioning vent valves and applies the direction otmollon by positioning vent valves and applies the helium by another valve. Indicator lights show the helium by another valve. Indicator lights show position of

_ _position ofvalves. valves.

~

The Theinterface of experimental facilities (beam pninstfhoe The orexperimental column and thermallning facilities column) (beam and the ports, reactor thermal pool liner thermal columnn and thermalizing column)1 and the reactor contains an open p)lenumn with piping cormected to a leak oft pool liner contains an open plenum with piping connected volume. The leak detetiton piping is connected to a single to aleak'orrvolume. The leak detection piping Is vlm.Tela eeto iigi once oasnl connected to asingle volume, except that a separate teak volume except that a separate leak ofrvolume and pressure 10.1 10.13.a offvolume and pressure gauge has been Installed for the 10-i 10.13.a gauge has been Installed rorthe thermalizing cotlumn. Itbe Add information theprimalzng column.theraliingcolun.

rth pol leks ntothePool Irthe pool lerks into the leaksthe overflow Intoplenum the experiment and fill thefacilities, the water wvill leak ofrvolume. Pressure expeniment faclities, the water will overflow ithe plenum monitors Inthe leak offvolume indicate when the volume is cnd 11 the leak atevoine. threvolue isonitors intholeak partially or rally filled. The leak oft volume and the pressure oRfdvolum indicae when tvolume partily or fully monitors are located on the north wall of the biological shieldingncar the northwest (radial) beam port. _

Various Various Included both Si and SAE units Various Various Included both SI and SAE units Add informalton 11.1.1, ..i Because ofits short halr-kION 'lcontriutes 11n1.1 . ause otofi shorthadlf-if,"N contributesngligibly to li-I 2 ngiil oorst aito xoue It.! oft-siate P2 negl~gibly to ofrBsito radialton exposure. dose to radiation exposure, the area above but Isthe the reactor major pool. - source ofrradiation Clarification AS-year average of tritium assay (performed monthly) indicates specific activity in the primary coolant of 228 pCi/ml for2SO kWopcralIons. irthe reactorbayatmosphere were 1l.l.l.a. t l.l.t.a saturated with this water at 300C,the water concentration in Trintum Measured tritilum specific activity in primary coolant Is Tritium the air would be less than 3 x 10-5g mt;' and the activity 11-2 in the less than I 0' Ci g*' I t-2 Inthe concentration Inthe atmosphere would be less than 6.84B.09 New informailon Reactor Reactor pCi/ml. Based on history, tritium concentration at 500 kW Day Bay would be less than 1.37 x 104pCi/ml, and tritium concentration at 1,250 kW would be less than 3A2 x 10' U-C/ml. In all cases, tritium concentrations are 11.2 11.1.1.a During norial operation of the reactor facility. there are 1 1 1...a During normal operation ofthe reactor Facility there ate three Clarification 1- 1111a three lshome sources, 'H. 'O. and 41ArI*3 .1o major potential sirbome sources. 'it 'IN and i Ar. Calclo Exbaust ot the rrtary mneIlmen rack, As shown in Exhanst ot the rntarv meriaermn rack-' As shown In Chapter Chapter II Appendix A. the equilibrium activity or "Ar It Appendix A. the equilibrium activit or "Ar In the RSR Is In the RSR is 0.56 Cl for sustained operation at 500 kW 06 Ci for sustained operation at 501 kW thermal power, 1A thermal power. IfthIs activity were instantly dispersed Ci at 1.250 kW. Ithis activity were Instantly dispersed Into RAT N 14 & Power level for 11.2 I l.l.n.A Into the reactorbayatmosphere, under normal ventilation 11.2 1l.l.l.o the reactorbay atmosphere, under normal ventilation bounding analysis conditions. and a worker were continuously exposed conditions,and a worker were continuously exposed theraiter. thecumulativeexpose vouldbe 1.8 x ItT thereafler.thecumulativeexposure wouldbeat4Ax I 0T pCI pCi hmL', well below 6 x lI OpCi h mL4 for2000 h mt; ,well below the occupational exposure limit ror 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months occupational exposure at the DAC- hours at the DAC, 3 x I04O'uCt h ml: .

Radiation monitoring systems are employed throughout the Radiation monitoring systems are employed throughout reactor facillty G-M detectors at the reactor pool surface and 7.19 77 the reactor facility G-M detectors at the reactor pool 718 7 cleanup loop, 7remote area monitor channels (3 general area Modification in progress

. surface and cleanurp loop S e mote area monitor channels or process monitors. 4 ehtnncls for beamn ports - I beam port (3 permanent, I mobile), a.... channel iscurrently Instrumented, with the remainder I scheduled for instrumentation near term), a..

Fig 7-20 Fig 7.13 Changed figures 7-20 7.13a & Split into 2 separate pictures Configuration change The remote aa monitors utilize G-M detectors located throughout the reactor bay (typical unit illustrated in throughouthe reonactorbayu(tical uMnit Iluectors located Figure 7.14). Permanent locations arn: at the top of the 7ho4hu teractor b typca t ilusrated tangr reactor tank, above the bulk shield lank, and near the ion 7.14). Penmanenllocations arc: at the top of the reactor tankt exchanger in the primy coolant system and directly above the bul hield tan, and near the ion exchangier in the 7-21 7-7 over the primary water tank. A movable detector can be 7.20 7-7 prImary coolant systemhn dinelanyd power lines p o suwport Modification in progress located at any beamrport or the thermal column. Meak ahba othssgnladpwrlnst upr ldetectors ndated aoally feature wiat an astoalr irheatorlin both analog readout o utecnr m control Installing bothmonitor.

a beaminport analog readout the controlThe roomdetectors feature and locally with anvsa room and locally with visual indicators for normal, alcm indicators for nornal, alert, and alarm conditions. he control and alarm conditions. The control rorn0s11has a room alanm has an audible signal as well.

________ _ audible signal as welL - -

7-22 Fig 7.15 Changed figures 7.21 Fig 7.15 ChangedFigures New location for air

____ ______ ___ ___m onitoring syste mn c_ ~

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-; s*-,^ V ; .. -~ .:. -;' ;4vi~ ;6a- '; ;:.3'::L.- ' t. ?:;1,::

Backup batteries supply the evacuation alamn; a 12 V deep-cycle battery for the air hom and a 6 V battery for the detector. These batteries are tested annually and a 8.3 82 charger is permancnily incorporated into the system. As 8-2 9.2 Deleted Configuraiton change mentioned InChapter7, the evacuation alamm is localtd at the upper level of the reactor and signals a need to evacuate the reactor bay.

A polar crane Inthe reactor bay Isused lo manipulate loads or A polar crane in the reactor bay isused to manipulate upto 3630 kg (8000 lb). Various lifing straps and loads of up to 3630 kg (8000 lb). Various liting straps attachments are available for handling varied loads. A breaker and attacs uandling ments am available for varie 10ads Inthe reactor control room supplies power to the crane. A A breaker inthe reactor control room supplies power to s eurigp erto the rane. a o tthe crane 9.5 9.7.2 the crane.: A basket is attached to the outside edge of the 9-4 9.7.2 serming pOersolto trcing the crane a local and positive Equipment modification crane for changing light bulbs in the ceiling of the reactor control overspower to the crane Abaskleis altached Sthe o bay. Duet! safpety concerns lighting t was installed outside edge of the crane for changing light bulbs in the around the peripheay of the reator bay eliminating the ceiling of the reactor bay. Due to safcly concerns, lighting need forthe overhead lightsand thebasket. was installed around the periphery or the reactorbay, reducing

. . the need for the overhead lights and the basket.

/ 'e' /' f' I el / t ,r e, r (' e, I 1, I' (' f, (' f (' r ( 1, . (' . (-- (-' (- (-' (- (- (- r- r- C ( (-- C--(-- r - (-- (--

Releasefromprimsaryoalant: AsshowninChapterti Release from primnry coolant: As shown in Appendix A, evcn with extremely conservalivo Chapter 11 Appendix A, evcn with extremely conservative 11-3 II l assumptions, during sustained operation at full power 11-3 111 .la assumnplions, during sustained operation at b11 powert with Power levcl for bounding with ventilation, the steady-state activity concentration or ventilation, the steady-state activity concentration of ' Ar in analysis "Ar in the reactor-bay atmosphere would be 3 x 10 ,uCi the reactor bay atmosphere would be 7.2 x 10t'pCi mL;', less ml:,, less than the occupational DAC. than the occupational DAC.

OMfsile Impact of Uar,_As shown in Chaptcr II Appendix A, OApsene Isthe ImxA, t efpa f-As shown in C eaptir 11 peak would operations off-site beactivity concentration about 0.003903 duringat normal pCi mL:' 13S m Apprndinra,theperatin wourlt e activity6co InO'utio downwind under slightly unstable atmospheric conditions, during normal operaions would be aboult 6x 10 pCi occurring 0.6% or total time. This concentration is lass than mL at 13Smidownwind under slighly unstable the emuent limit of 0.01 pci mL'. Afull year of operation at Power level for bounding 11.3 l ~ltl.a atmospheric conditions. Thisconentraion islessthan 11o3 1.l.1.a the maximum power level maximum concentration would analysis the e limt mauent of 0.01u pCI rnLt. Afull year o a result in an effective dose at the receptor with the maximum exposure at the maximum ontration woll t oin concentration of only about 0.16 intern, well within applicable efaectivcse ofonly about 3 rnrcm, well wihin limits. The highest dose to a location occurs at 2140 meters applicable limits, with a dose or3.8 inrem, well below the maximum allowed 10 mrcr from effluents.

As shown in chapter II appendix A,very conservative As shown in Chaptcr 11 Appendix A, conservative calculations lead to an expected exposures rate of40 mR calcushtowns ldtn Chp erxIApentdexpAonsu e rvatie o 11-3 1I h.L fat one meter above the center ofrthe reactor tarlc 11.3 1l.l.L~a alcroxlatioslea toIf atne mexeted aboposuhe enter ofth Power level ror bounding 11-3 ll.l.a durng sustained operations are500kW thermal power. approxim telyd25imng h' at onmerte rea abov te centerof th alysis Measured exposure rates are about 20 mR h' at 250 kW rectr, tnk during tained opertn t at 1,5 kW t Orations.

. .powe increasing to nearly 100 m_ h at 1,250 k 11_4 Table Gamma Cell location as room 3 115 Table Gamma Cel location as room 14 Change in location I 1.l.5S, b, The 22-foot level access has a line of sight to the control room, Access Al.l. and has radiation monitoring positioned directly ovcr the pool l-l Control ecs o corresponding text Il-Il Control sutrface and mounted on the rail surrounding the pool. The Power level for bounding 11Cnt NDuring operator at the controls Is responsible for appropriately analysis controlling access to the 22-fool level based on radiological Ops -Ops conditions.

Acceptable surface contamination levels for unconditional Acceptable surface contamination levels for unconditional Aeleptabeam giveninc Table I tI.Slimits [onaverageurelease are given in Table. I 1.5, as provided Inthe approved release are given in Table. 11.5, Limits on average Radilation Poleculon Progamn, KSU TRIGA tark!!l Nuclear contamination levels ror unconditional releaseam Rareatorotcion Lmits o acg conlark n 1Nulear 11-13 11.1.6 calculated based on survey aras smaller than I rn. 11-13 11.1.6 Reactor FacII16. Limits on average contamination levels for RaI 15 Lmtonm immcontamination levels (b unconditional release are calculated based on survey areas Limits on mraximumor calculate on survey smaller than I in. Limits on maximum contamination levels unconlilional release are calculated based on survey aras fior unconditional release arm calculated based on survey areas smaller than 100em. _smaller than 100 cm'

-cLL Lc(Cc c- C GCLL-C- - - L- c- -L- L-L-L-L-C-L-L- - -.- (-Q-(-L-Q

. % .I % S. *%

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As required by 10 CFR 20.1501, contamination surveys are conducted to ensure compliance with regulations reasonable under the circumstances to evaluate the magnitude and extent of radiation levels; concentrations or quantities of radioactive material; and potential radiological hazards.

Guidance has been promulgated in IE Circular No. 91.07 (Control of Radioactively Contaminated Materials) for releasing materials rom restricted to unrestricted areas:

Dived on the stuidies of restdital racdoactivity limits for decommisloning (NtUREG-06132 mndNUREEG07073). It can be conci'nled that strfaces tmiformnly confaminatedat levels of 5000 dpr41OOcm2 (befft-amma activity from nucdear power reactors) word result Inpotential doses that total less than 5 mremnr. 7eefore. It can he concluded that for the 11-14 11.t.7.d No correspondingtext 11-14 11.1.t7. potentally undetected contamination of discrete Items and RAI# 15 material at leels byelow 5000 dpm/lOOcmZ thepotential dose to nT Indihal will be signifcanftly less than Smm 'r even Ifthe amrnialion of numerous Items contaminated at this level Is considered.

The contamination monitoring using portable survey Instruments or laboratoty measurements should be perforned with Instnimentatlon and techniques (survey scanning speed, counting times, background radiation levels) necessary to detect S0O dpmII0O cmt2 total and 1000 dpmilOO cm2 removable belaigamma contamination. Instruments should be calibrated with radiation sources having consistent energy spectrun and instrument response with the radionuclides being measured. It alpha contamination Is suspected appropriate surveys and/or laboratoty measurements capable of detecting 100 dpm/100 cm2 Fixed and 20 dpmWI00 cm2 removable alpha activity should be perfonned.

Normal operation of the KSU reactor results In two potential source terms for radioactive gaseous effluent, 4 Ar and 'Wt.

There ae variations in experimental configuration and possible scenarios where the production of "Armaybe I I.AI A.1 No corresponding text diiterent than the routine operations; these scenarios do not Clarification produce not long term, routine radioactive effluent but need to be assed to determine Irthe amount orradioactive effluent Is so high a to impact the annual exposure that might result from routine opetiLons. _

Tbe concentration to dose rate (effective dose equivalent) The concentration to dose rate (effective dose equivalent)

I l.A-1 A.I.2 conversion tactor or submersion In an Infinite conversion factor for submersion in an Infinite atmosphere or Power level for bounding atmosphere of"ArIsas follows: 23 x IO"OSvh ' per ll.A-2 A.Is2 arollows: 2.17 x 10 '0Svh ' pr Dq ms orQ0.03 atArls analysis Bq m; (EPA 1993). mrem/h per pCilml (EPA 1993).

I Il-A-2 Table Ad1 Prmeters for 500 kW I I-A-2 Table A.1 Change1 parameter orolXO kWV analysislevel for bounding Power

Various Various In some cases, Si or SAE units used; in some cases. 500 Vaous Various Included both sets of units for fundamental measurements; Power level for bounding

-M kused -- used I.250kW tor miaximum power analysis & Add inromralton Operation with a fully open beam port is not a routine operational condition. Beam port operations nonmally have I 1.A4 A2l2 No corresponding lext I I.A.4 A22 shielding, eollimation and beam stops that prevent a full beam Clarification from penetrating the column defined by tbe beam port into air

___ volume between the reactor and the reactor bav wall.

or 3A2 x 10' pCi mL' In conventional units. Operations at maximum power are not performed for radiography, and radiography is not performed long enough to achieve equilibrium 4"Ar. Therefore, sealing the calculation for or 3A22x 0I' pCI mL' In conventional units. Operations sustained operations at 1,250 kW provides an extremely at maximum power are not performed forradiography, conservative bound on "Ar production. Scaling the 10 kW and ramdionrahy is not performed long enough to achieve "4Arproduction value to 1,250 kW results In 4.28 x l0 'pCi equilibrium Ar. Therefore, scaling the calculation for mL:' which is slightly above submersion DAC for sustained operations at 500 kW provides an extremely occupational exposure however, conditions for the source conservative bound on 't Ar production. Fiky times the term arm related to a very unusual set of conditions (open beam Power level for bounding llJ.A4 A2.1 10 kW "Ar production value results in 1.7 x l07 pCi mL I I.A-4 A22 port with no shielding) that nre not continuous in two respects. analysis

'which meets the submersion DAC for occupational Shielding for radiography external to the bena port limits the i exposure with no further consideration. This value is beam to less than %4of the analyzed volume. Radiography slightly higher than the emuent limit for continuous conTguradon is implemented only for radiography operations release; meeting the effluent release limit Id assured at K- a small fraction or all operations. Typically radiography State through conscrvaitsms In the calculations and occurs less than I day per month. Radiography operations are because the reactor is not operated continuously. inherently discontinuous as the purposes or Individual operations are met when the Image is obtained. Typically a day or radiography operations involves less than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of operation at full power. These conservatisuns assurc DAC and

- - emuent limits are met with no further consideration.

The air volume in the rotary specimen rack does not freely exchange with the air in the reactor bay;-therc is no motive I I.A-4 A2.2 I l..4 No orreponing

.2.2 NoIext correspondr~ing extll.A4 I I.A41 A2.2 All force for circulation routinely and the rotary specimen covcrcd duringope'ration. rackspecimen iftherotay opening is Claiiictlion rack were to flood, water would force the air volume in the

-RSR into the reactor bay.

ThsIswllblo he6X 03pih l'annual limit of The value 1.6 x I0' PCI s ml;, or 4.x I " PCI ht ml:', is I l.A5 Al This 2000 isDAC welhours belowspcsified thci6ix in 10 IOCFR2O.

pCI hCmL. A.2 A22 well below the 3 x 107p pCi hltWnl annual limitof2000 hours specified In I0CFR20IEPA-52011-8O. 020.

DAC RAI # 16 The reactor tank water sureace Is open to the reactor bay.

I I.A5 orresondin A.23 No textI LA- A2.3 Radioactive 't Ar Is circulated in the pool by convectionClrfcto 11 A-5 A.2.3 No corresponding text I l.A.S A2.3 heating, and freely exchanges with the reactor bay atmosphere Clarification

_ durina normal operation.

NA NA NA 1.A46 Table Added Clarification Edat - LCtUC L L CC tCL -a-L- C L-L-L-L-L-L-L-C- C- (- (-. -C-.,G (-.1G G.-1-;-I-=, -

\ I. I \. , 1% \ %. I %. I \1. 11 %. II, I I The equilibrium 4'Ar concentration during fMll power steady state operation at 1,250 kW in the reactor bay would be 0.072 Dq cm71 (1.9 x 10'e pCi ml:') without ventilation and 0.021 Bq cma3 (7.2 x I0O pCi ml;') with ventilation.

Environmental Protection Agency, Federal Guidance Report Power level rorbounding llA.6 A2.3 See previous version I I.A-8 A23 No. II (IA FG 11 - Limiting Values of Radionuclide Intake analysis and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion) lists the DAC for "Ar as (Table l.b) 3X104 pCitml or 3 pCVmI. Thererore, equilibrium 't Ar concentration during full power steady state operation at 1,250 kWV Is less than DAC and there are no restrictions on activities In the reactor bay imposed by the

-1ms A.41 ntA61 -11 Although there are three modes of" Ar production, only the I I.A-6 A2.4 No corresponding text I I.A-6 A2.4 release orradloactive argon dissolved Inwater occurs Clarification routinely.

IA6 Table Cocnrto ausfr50k I-'Table Cnetainvlsro125kWPower level for bounding 11 A.6 A3 Coneetration u rOr 5 kW 1A7 A.3 Conetitrionvelua for 1,250kW analysis The dose conversion factorprovided by EPA FO II ror "'Ar (Table 2.3) is 2.17 X 10' Sv/hr pr Bqt@m, or susing the provided conversion fictor of3.7 X I0") 8.03 X 10 mrem/hr per pClerm, 8.03 X 10`' mrem/hr per pCecm'. Using the hihest maximurn concentration or Table A3 (0.003903 pCi cm ') at steady state full power operation ror a Mll year (8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months ) with observed firequency or class A stability (see Appendix 24C) would result in a dose less than I mrern/year.

Frequency or occurrence and the concentration at the maximum dose will occur from class C conditions, with a maximum annual dose of 1.7 mresn. The maximum concentration at the highest Frequency (class G) Is 0.001013 pCicm, with adose or3.i mern.

I I.A-6 A.7-4 No corresponding text A.24 I I.A-7 The assumed 24-7 operating history Isnot feasible Clarification, information add for the ISU reator, which has an average operatling time for two decades orabout 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months smheek. Additionally, a full power, continuous operation would tequire a signirtcant quantity of new Fuel.

Note that over the Murange orconditions examined InTable 23, the peak downwind concentration Is substantially below the DAC or3 pCI em' established in IOCFR20 Appendix B.

and less than the permissible eluent concentration of t X 10' pCi/cm' for all meteorological conditions except the set of conditions with the lowest frquency or occurrence fior that stability classilication, the instantaneous efluent concentration Is slightly higher than the DAC.

I I.A-7 A.2.5 ISubstantial rewrite I I.A8 I A2.5 I Substantial rewritl RAI # 17, Clarification, add I e.inronnation

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e Radialion Safety Oiccr reports to the Manager of The Radiation Safety Officer reports to the Manager of the the Department of Environmental Hcalth and Safety. The Department of Environmental Hcalth and Safety. ohe Radiation Safety O Mcer, oran authorized representative, Radiation SafetyO Mcer, oran authorized reprcscntalivc, shall shall be availabic (upon due notice) for advice and be availablc (upon due notice) for advice and consultation consultation regarding radiation surveys and radiation regarding radiation surveys and radiation safety in connection safety In connection with isotope production and radiation with isotope production and radiation staming problems as 12.6 12.l2b streamingproblem asmightariseinconnectionwith 125 12.1b might arise in connection with reactor operation or RAI# IS reactor operation orexperimentation. The Radiation experimentation. The Radiation Safiety OMcer is5x officio a Safcty Offcer issex officio a member ofthe Kansas State member of th Kansas State University Radiation Safcty University Radiation Sarety Committee. The Radiation Commitce. The Radiation Safety OMcer serves cx offlcio as Safety OM=cserves cr officlo as ammenber ofthe a member or the Reactor Safcguards Committee, with any Reactor Safeguards Committee, with any action of the action (ie, concerning potential radiation exposure or Committee requiring approval of the Radiation Safety radioactive emuentis) or the Committee requiring approval of Olficer. the Radiation Safety OMcer.

Whenever the reactor is not secured the reactor shall be Whenever the reactor is not secured, tho reactor shall be under under the direction ora (USNRC licensed) Senior the direction ofa (USNRC licensed) Senior Operator who Is 12-7 12.1.3 Opcrator who is designated as Reactor Supervisor. The 12-7 12.13 designated as Reactor Supervisor. The Supervisor shall be on Change Supervisor shall be on call, within ten minutes travel time call, within twenty minutes travel time to the facility, and to the facility, and cognizant of reactor operations- cognizant orreactor operations.

A report shall be made within 10 days In writing to the A report shall be made within 14 days in writing to the NRC 12-13 125.2 NRC Operation Center for any violation of safety limit or 12-7 123.2 OperatIon Center for any violation of safety limit or reportable RAI # 19 renortable occurrence oc-urrence 12-15 12.63 Noeorresponding text 12-15 12.6.3.f Corrected and as-built facility drawings RAI #20 A physical security plan for protection of reactor plant Administrative controls for protection of the reactor plant shall 12-IS 12.8 shall be established and followed in accordane with NRC 12-1S 12.8 beestablishled and followed in accordance with NRC Correction regulations, regulations.

These arm the t conditions considered In the initial Thcse are the three conditions considered in the initial licensing of the Reactor Facility In 1962 for m0iakW licensing or the Reactor Facility in 1962 for 100-kW steady-steady-state operation and in the 1968 upgrade of the state operation and in the 1968 upgrade of the license 13-1 13.1 license permitting 250-kW steady state operation and 13.1 13.1 permiUing 250-kW steady state operation and 250-MW Power level for bounding 250-MW pulsing operation. The analysis presented here pulsing operation. The analysis presented hero treats the same analysis treats the same conditions. but for steady-state operation conditlons but for steady-state operation at 1,2S0 kW and trcats tho samocond ituli onsgstead apulsing operation e operation to a S3.00 reactivity insertion, estimated at 500 kW and pulsing operation at 1,000 MW. - - pcakpowerof 1.340 MW.

various various 500 kW various various 1,250 kW Powerlevel for boundiny Fuel and cladding temperatures are reported in Table 13.2 Fuel and cladding temperatures are reported in Table 13.6 and 13-8 132.2e and illustrated in Figure 13.6 for theease orzacro time 13-8 132.1e Illustrated in Figure 13.2 for the ease ofzero time post Correctreferences post accidenL accident.

13-8 Table Values ror500kW 13.8 Table Values for 1,250 kW 13 S13.6 1 _ _ _ _ __ _ _ _ _ __ _ _ _ _ __ _ _ _ _ _13.6 __ _ _ _ _ _ _ __ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _

Power analysislevel for bounding

.LCLLLLLLLL-CL-CLCC-C-C-L-L-L-L-C-C-L-L-L-L- .L-Q C- -.1 Q Q Q (-.' Q L-C-(-.. L-(-. L-

\. k \ , 1% \. It.. I \ ,,- 'k. ". \. ,, . , \ I.. %, "... 1% %. % % %. 1. I.. % .. %

Although a toss orpooi wafer is considered to be an Although a00 of pooit vater Is consIdered to be an extremely extremely Improbable event, calculations show the I 13 2 2. rnium fUe* Iempeatre titat could** ected to temperature event, ituprobeble calCUlatiOns show the maximiumsuchfued Correction t3-13 m Mulfrom such ain event (slner long-term operation at 13-13 13.2.2ug that could event (afler long-term operation at to be expectaed fullresult of 1P250an powerfrom kW) Is fiull power of 500 kW) Is 261C, well below any safety 294ceC, well below any safety limit forTRiGA reactor fMel.

limit for TRIGA reactor fuel.

13-13 T3b9 4 table values In error 13.13 Table Rhr values, I,2S0kWV Correction Rapid compensation or a reactivity insertion is the distinguishing design feature or the TRIGA reactor.

13.14 132.3 Rapid compensation of a reactivity insertion is the 13 14 1323 Characterislics or a slow (ramp) reactivity Insertion arm less RAT U24 distinguishing design feature orthe TRIGA reactor. . severe than a rapid transient since temperature feedback will occur rapidly enough to limit the maximum power achieved

. during the transient....

A control rod Interlock preventing pulsing operlions from power levels greater lhan a maximum of 10 kV Is not 13-15 1323 Nto corresponding text 13-t1 13.23 credited Clarification

Conservalive hot channel factors as calculated In 4.5.3 are

. - -used A maximum pulse of S3.00 would result in a power rise of 13-15 1323.b TfIsthenser oretempe. tureatthestutofIhe 13-15 13.b approxlmately 1430 MW(t). IrT. Is the average core Power level for bounding excursion, the mnaximunn temperature rise (OM)is given by temperature at the start of the excursion, the maximum analysis temperatus rise (MK) Is given by 13-15 13.23.b Extensive rewrite 1315 1323.b Extensive rewrite Clariflcatlion CASE 1: Analysis oF a 2.1% (S3.00) Reactivity Insertion at Zero Power For this ese, Eqs. (1323.2) and (13.2.3-3) yield a power rise of 1430 MW and a coremavelage Mel temperature rise of temperature or229MK Peak temperature rise would be the core average multiplied by the overall peak-to average ratio orf n resulting in a hot spot temperature of'27 + 719 Insertion or the maximum possible reactivity without initial

- 7460a Therefore, insertion orthe maxImum possible temnpeture feedback (I.e.. flel temperature Istoo low to limit reactivity without initial temperature feedback(l.e hihel core available reactivity), results In a peak hot spot fuel temperature is too low to limit core available reactfvfiy) temperature of746'C, well below the sarety limit.

13-16 132.3.c results InMuel temperatures well below the solely limit. 13-16 13.2.3.c RAT # 22,23 of a 0.7% ($1.00) eetivly CASEIfAnalysis Insertion or the maximum possible reactivity with initial ICASE 1i: 9t at Power * (temperature feedback (lme fuel temperature limits available) nsertion at 94 kW Power results In a peak hot spot fuel temperature or s69'C, well For this ease, Eqs. (1323.2) and (13W.3) also yield a below the safety limit.

powver rise of 1430 MW and a core-average fuel temperature rise of tempeature of 229°K In this case.

the hot spot temperature Is 150 + 719 - 869'C.

Therefo, insertion or the maximum possible reactivity

,with Initial temperature feedback (Le.. fuel temperature limits available) results In fuel lemperatures well below

.the saety limt. -

For shoit-lived radionuclides, calculations of radlonuclide For short-livcd radionuclldes, calculations of radionuclide 13.24.a.2 13-17 invcntory In fuel are based on operation at the Mil t hcma 13-17 13.24L2 Inventory In fuel are based on operatlion at the full thcrmal Powcr level ror bounding powerof500 kW foreighlhours perday. for five power of 1,250 kWV for eight hours per day, for five successive analysis successive days prior to tuel failur days prior ro rucl failure. an average of 31.25 kW-hrlday.

For times much greater than the half-life of the For tIms much greater than the half-life orlhc radionuclide, Cotlon/typographical 13-17 132.4.b radionucllde, A mA4, and for times much less than the 13-17 132.4.b A R 4, and for times much less than thz half-lirc, A() -

halflife. A(l) - A.* ?* t A. *

L crror 131 Table Values for 500 kW 13.20 Power lcvel for bounding

-13.Vauesrot3,200 analysis 13-20 Table Values forSOO kW 13-21 Table Values for 1,20 kW Power level for bounding

______13.11 13.11 _________________________ analysis 13-22 13.24.h Reference to table 12.14 incorrect 13-22 13.24.h Changed to 13.22 Incorrect tabIl no.

13-22 Table values for 500 kW 13-23 Table Values for 1,250 kW Power level for bounding

______ 13.12 13.12 aayi 13-23 Table Values for 500 kW 13-24 Table Values for 1,250 kW Power level for bounding I_____ 13.13 ______________________ - 13.13 _________________________ analysis App 13.F App Values for500 kW App App1 Values for 1,250 kW Power level for bounding BY_____ BY BYF ______

analysis

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KSUResearch,Reactor DocketNo.50-18S

REFERENCE:

TACNO. MBR7966 I

Pagel of 3

KSU Research ReactorDocket No. 50-188

REFERENCE:

TACNO. MB7966 I

I I

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BB .1 KSU Researchs ReactorDocket No. 50-188

REFERENCE:

TACNO. MB7966

'.-.: toR d~rffoA~dItI I.lnfd s.:;:~s'u:.<>,':8Ro~sc on2,. ^.-

18 18 12.6 12-6 Section The RSO Committee. hasIS veto power there in theestablished a process Reactor Safeguards to override the Revised to indicate veto poxvcr ovcr iaonctrlsse I2.1.2.b

____veto?

vo radiation control issues 19 12-13 Scction 12.5.2

________days What is the reason to restrict the 14 day report to within IO Corrected 20 12-15' Section 12.6.3 Please include the update, corrected, and as-built d\facility Included 20 1-5

_______ Seon1.. drawings in this section as Indicated In TS Section 6 1.b nl)6

___d Section Please discuss the calculation of the peak to average core 21 13-16 13.2.3(3) Case tmpeteratio equal tororprovide a reference. n Use of the factor discussed in rewritten 1 and Table appears to be too high a value for this parameter for your material of Chapter 4 & referenced 13.4 reactor.

Please discuss the limitation of the initial power for Case ll Section 13.2.c to 94kW. Please discuss the possibility ofan experiment 22 13-16 Case .I reactivity change while at power greater than that analyzed. Rewritten Please correct the inconsistency between this analysis and TS Section 3.1.5 with regard to the Initial power.

1 Section Please discuss the conclusion that the core power rise will be R 1-16 J13.2.3(3) Case the same for 2.1% and 0.7% insertion orreactivity. Rcrten Justify not analyzing a ramp accident and using the results as 24 13-16 Also TS bases for the LSSS and the ractivity change rate limits fr Discussd ramp as being boundcd by pulse Section 5.3 moveable experiments and control rod motion. How are the

_ consequences of such accidents limited?

Pnge 3 o3

ML052580517

Text

Dear tr. Strauss:

Dear Mr. Fal:

Dear Sirs:

Dear Dr. Faw:

Enclosure:

REFERENCE:

REFERENCE:

REFERENCE:

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