Text

Ohio State University Redacted Safety Analysis Report and Technical Specifications - December 15, 1999
	ML053400287
Person / Time
Site:	Ohio State University
Issue date:	12/15/1999
From:	Ashley D; Ohio State University
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
	Witt K, NRC/NRR/ADRA/DPR/PRT, 415-4075
References
	Download: ML053400287 (254)
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OHIO STATE UNIVERSITY RESEARCH REACTOR LICENSE NO. R-75 DOCKET NO. 50-150 REVISED SAFETY ANALYSIS REPORT AND TECHNICAL SPECIFICATIONS DATED 15 DECEMBER 1999 REDACTED VERSION*

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

Redacted text and figures blacked out or denoted by brackets

Office of the Dean 142A Hitchcock Hall College of Engineering 2070 Neil Avenue Columbus, OH 43210 Phone 614-292-2836 FAX# 614-292-3244 December 15, 1999 Document Control Desk U.S. Nuclear Regulatory Commission Washington D.C. 20555 RE: License Renewal of The Ohio State University Research Reactor (OSURR)

License No. R-75, Docket No. 50-150 This letter accompanies submission of the revised Safety Analysis Report (SAR) and Technical Specifications (TS) for the OSURR. The purpose of this submittal is to renew the OSURR license for a twenty-year period until February 3, 2020.

The SAR and TS for the 500kw, LEU fueled reactor were submitted October 7, 1987.

The current submittal updates those documents. The format is the same as the original.

A brief chapter (10) on financial qualifications has been added. Many of the changes were the result of the revisions made to IOCFR20, which eliminated the use of Maximum Permissible Concentration (MPC).

The changes to the SAR and TS were reviewed and approved by the Reactor Operations Committee in its meeting of December 8, 1999. Correspondence regarding the OSURR and this license renewal request should be sent to my attention.

Sincerely, teLe David B. Ashle Dean, College of Engineering and The John C. Geupel Chair in Civil Engineering

c. Don W. Miller, Director Nuclear Reactor Laboratory Richard D. Myser, Associate Director Nuclear Reactor Laboratory Theodore S. Michaels, U.S. Nuclear Regulatory Commission AW-GI DEC 2 3 Ing TIDL POi 05-0 o16Z)

SAFETY ANALYSIS REPORT and TECHNICAL SPECIFICATIONS for The Ohio State University Research Reactor License Number R-75 Docket Number 50-150 December, 1999 Columbus, Ohio

Table of Contents 1.0 Introduction.......................................... . .. . 1 .

1.1 Purpose ............................................. . .. . 1 .

1.2 General Facility Description ........................ . .. . 1 .

1.3 Background Information .............................. . .. . 2 .

1.4 Report Organization ................................. . .. . 3 .

2.0 Site Description and Characterization ................. . .. . 4 .

2.1 General Location ......................... ...... 4 2.2 Demographics .................. : ..... 4 2.2.1 Surrounding Population........................... .... 4 .

2.2.2 Local Activities................................. .... 4 .

2.3 Topography, Geology, and Seismology.................. . .. . .... 10 .

2.3.1 Topography....................................... .... 10 .

2.3.2 Geology ......................................... 10 .

2.3.3 Seismology.................................. . ... . .. .. .. . ... 11 2.4 Meteorology ................. ; .... . . . . . ... A.... 16 2.5 Hydrology...................................... . .. .. .. . ... 16 2.5.1 Surface Water............................... . .. .. .. . ... 16 2.5.2 Ground Water............................... . .. .. .. . ,.. 23 TVi~ 2.6 Reactor Building Description ................... .

2.6.1 Reactor Building Design and Construction

. i.

23 23 2.6.2 Layout ..................................... . ... . .. .. .. . ... 23 2.6.3 General Features............................ . ... . .. .. .. . ... 25 2.7 Chapter 2 Bibliography ......................... . . .. ... . .... ... 25 3.0 Reactor Facility Description..................... . . .. ... ... 27 3.1 Reactor Core ................................... ... . . . .... . ... 27 3.1.1 Core Structures............................. . . .. ... ... 27 3.1.1.1 Grid Plate and Support .................. . . .. ... ... 27 3.1.1.2 Fuel.................................... . . .. ... ... 27 3.1.1.3 Core Arrangements ....................... . . .. ... ... 33 3.1.1.4 Control Rod Poison Sections .......... . . .. ... ... 34 3.1.2 Associated Core Structures.................. . . .. ... ... 34 3.1.2.1 Core Reflection......................... . . .. ... ... 34 3.1.2.2 Startup Source and Source Drive ......... . . .. ... ... 34 3.1.2.3 Control Rod Housings .................... . . .. ... ... 36 i

U 3.1.3 Reactor and Shielding Pools ............................... 37 3.1.3.1 Reactor Pool .......................................... 37 3.1.3.2 Bulk Shielding Facility Pool .......................... 42 3.1.3.3 Fuel Handling Tools .................................... 42 3.1.4 In-Pool Instrumentation . . 45 3.1.4.1 Startup Channel and Drive System ...................... 45 3.1.4.2 Linear Power Monitoring Channel Detector .............. 45 3.1.4.3 LogarithifiG.Power Monitoring Channel Detector .......... 47 3.1.4.4 Power Level Safety Channel Detectors ................. 47 3.1.4.5 Temperature Sensors and Locations ..................... 47 3.1.4.6 Water Level Sensors and Locations ..................... 47 3.2 Reactor Cooling and Water Processing Systems . . ................

49 3.2.1 General Features .......................................... 49 3.2.2 Cooling System . . . 49 3.2.2.1-Primary Coolant Loop .................................. 49 3.2.2.2 Secondary Coolant Loop ................................ 52 3.2.2.3 Cooling System Instrumentation and Control Systems ..... 54 3.2.3 Regenerable Demineralizer (Makeup Water System) ........ ... 56 3.2.4 Water Processing System . . . 56 3.3 Reactor Instrumentation and Control ........................... 56 3.3.1 General Features .. 56 3.3.2 Control Room and Operator's Console . . 59 3.3.3 Nuclear Instrumentation (NI) Racks . . 61 3.3.4 Control System Inputs ..................................... 61 3.3.5 Control System Outputs .................................... 61 3.3.6 Signal Paths and Cable Runs ............................... 62 3.3.7 NI and Control System Power ............................... 62 3.3.8 Control Rod System ........................................ 63 3.3.8.1 Position Control ....................................... 63 3.3.8.2 Indicators ...... ................................ 63 3.3.8.3 Interlock Systems . .................................... 63 3.3.9 Control Rod Magnet Systems ................................ 65 3.3.9.1 Magnet Power and Actuation ........ .................... 65 3.3.9.2 Indicators ........................................... 65 3.3.9.3 Interlocks ........................................... 65 ii

3.3.10 Startup Channel Detector Positioning ................. .... 65 3.3.10.1 Position Control ........ .............................. 65 3.3.10.2 Indicators . ........................................... 67 3.3.10.3 Interlock ............................................ 67 3.3.11 Startup Source Positioning . . . . 67 3.3.11.1 Position Control ........ ............................. 67 3.3.11.2 Indicators . ........................................... 67 3.3.12 Logarithmic Power Monitoring Channel ................. .... 67 3.3.13 Linear Power Monitoring Channel . . . . 68 3.3.14 Startup Channel . . ........................................ 69 3.3.15 Period Monitoring Channel . . . . 70 3.3.16 Power Level Safety Channels .... 70 3.3.17 Period Safety Channel . . . .71 3.4 Cooling'System Controls . . ................................ .... 71 3.5 Auxiliary Controls ........................................... 72 3.6 Safety System (Reactor Protection System) . . ................... 73 3.6.1 General Features .......................................... 73 3.6.2 Types of Scrams ............................................ 73 3.6.2.1 Slow Scram (Relay Scram) ...... ........................ 73 3.6.2.2 Fast Scram (Electronic Scram) ..... .................... 73 3.6.3 Scram Functions and Setpoints .. 73 3.6.4 Scram Bypass ........................ 77 3.6.5 Alarm and Annunciator System . .77 3.6.6 Building Evacuation System (Ventilation Control) . .77 3.7 Area Radiation Monitors . . . .78 3.7.1 General Features and Purpose .. 78 3.7.2 Detectors and Location .. 78 3.7.3 Readouts, Indicators, and Alarms . .78 3.8 Experimental Facilities . . . .78 3.8.1 Central Irradiation Facility .. 78 3.8.2 Beam Ports .79 3.8.3 Rabbit Facility .. 79 3.8.4 Main Graphite Thermal Column .83 3.8.5 Bulk Shielding Facility Thermal Column .87 3.8.6 Graphite Isotope Irradiation Elements . .87 z iii

3.8.7 Movable Dry Tubes. .....

.. .. 8 7 4.0 Normal Operating Characteristics ............

4.2 Core Loading and Critical Mass ............

91 91 I

4.2 Reactivity Requirements .................. .................... 97 4.2.1 General Considerations .................................... 97 4.2.2 Operational Requirements ................................... 97 4.2.3 Experiment Requirements ................................... 98 4.2.4 Reactivity Limitations .................................... 98 4.3 Neutron and Gamma Environments ............ .. ................. 99 4.3.1 Neutron Flux Calculations ................................. 99 4.3.2 Gamma Environment Estimation ............................. 107 4.4 Reactivity Control Systems ............. .. ................... 108, 4.4.1 Control Rod System ....................................... 108 4.4.2 Cooling System Control ................................... 108 4.4.3 Control Rod Worths ....................................... 108 4.5 Temperature and Void Coefficients of Reactivity . ......... 109 4.6 Neutronics ........................ .......................... 110 4.7 Effects of Core Geometry ................ .................... 110 4.8 Heat Transfer Characteristics ............. .. ................. 112 4.8.1 Coolant Outlet Temperature ...............................

4.8.2 Fuel Plate Surface Temperature ...........................

112 114 I,

4.9 Fluid Dynamics ...................... ........................ 115 4.9.1 Detailed Analysis .......................................... 115 4.9.2 Simplified Analysis ...................................... 119 4.10 Chapter 4 Bibliography ................. .................... 121 5.0 Auxiliary Systems .......................... 127 5.1 Introduction ....................... ......................... 127 5.2 Communication ...................... .......................... 127 5.2.1 Control Room Intercom System ....................... I...... 127 5.2.2 Building Phone System ..................................... 127 5.3 Lighting Systems ...................... ........................ 127 5.3.1 Normal Interior Lighting ....... . I ......................... 127 5.3.2 Emergency Interior Lighting ........................... I.... 127 5.3.3 Exterior Lighting ................................. j.......

128 5.4 Building Services ......................................... 128 5.4.1 Electrical, Water, Air and Gas......... 128 ..........

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A n' Lvrt

".Y M%.ie ....................................................

128 5.5 Bridge crane ................................................ 129 5.6 Building Alarm System ....................................... 129 5.7 Access Control .............................................. 129 6 .0 Radioactive Waste Management.................................. 130 6.1 Source Term Estimation ...................................... 130 6.1.1 Liquid Effluents......................................... 130 6.1.2 Gaseous Effluents........................................ 130 6.1.2.1 Argon Production in Experimental Facilities. 131 6.1.2.2 Argon Production from Pool Water ..................... 137 6.1.2.3 Nitrogen-16 Production from Pool Water ............... 139 6.2 Liquid Effluent Waste Management ............................ 140 6.2.1 Pool Water Monitoring.................................... 140 6.2.2 Secondary Loop Coolant Monitoring........................ 141 6.2.3 Liquid Effluent Releases................................. 141 6.2.4 Cooling System Maintenance Operations............... 141 6.2.4.1 Draining, Blowdown, and Purging ...................... 141 6.2.4.2 Tertiary Loop Effluent Holdup ........................ 142 6.3 Gaseous Effluent.Waste Management ........................... 142 Q 6.3.1 Effluent Monitoring System............................... 142 6.3.2 Blower Effluent Monitor.................................. 142 6.3.3 Release Points........................................... 143 6.3.4 Estimated Releases in the Restricted Area ................ 143 6.3.4.1 Types of Releases ................................... 143 6.3.4.2 Puff Release from the Rabbit ......................... 143 6.3.4.3 Continuous Release from the Rabbit ................... 146 6.3.4.4 Puff Release from the Rabbit Carrier Tube ............ 148 6.3.4.5 Puff Releases from Other Experimental Facilities.. 148 6.3.4.6 Continuous Release of iAr fiom the Pool Water ........ *151 6.3.4.7 Combined Continuous Releaselfrom the Pool & Blower.... 151 6

6.3.4.8 Continuous Release of N 'from the Pool Water ......... 152 41 154 6.3.4.9 Actual Ar Releases.................................

6.3.5 Releases from the Restricted Area...... ............. 155 6.3.5.1 Dilution Factor..... .............................. 155 6.3.5.2 Puff Release from Various ,Facilities ................. 156 v

6.3.5.3 Continuous Release from the Rabbit Blower ...... 156 .

6.3.5.4 Continuous Release from the Pool Water ......... ..... 159 6.3.5.5-Combined Pool Water and Rabbit Blower Releases .... 159 41 6.3.5.6 Actual Ar Released ....................... ............ 159 6.3.6 Steps to Limit Release Levels................ ............ 159 6.3.6.1 Reducing Effective Irradiated Volumes ............ 159 6.3.6.2 Facility Purging . ............ 160 6.3.6.3 Limiting Facility Releases . ............ 160 6.3.6.4 Ventilation System Control . ............ 160 6.3.7 Estimated Doses. ............ 161 6.4 Solid Radioactive Waste Management. ............ 161 6.5 Liquid Radioactive Waste.Management. ............ 163 6.6 Byproduct Materials Management. ............ 163 6.7 Chapter 6 References . ............ 164 7.0 Radiation Protection. ............ 166 7.1 Radiation Sources During Operation. ............ 166 7.1.1 Direct Exposures......................................... 166 7.1.1.1 Gamma Dose From The Core Through Shield Walls......... 166 7.1.1.2 Gamma Dose Through an Experimental Facility........... 168 7.1.1.3 Neutron Dose Through an Experimental Facility........

7.1.1.4 Gamma Dose At The Pool Top Through Pool Water......... 171 170 0

7.1.2 Indirect Exposures.................................. 177 7.1.2.1 Nitrogen-16 At The Surface Of The Pool ............... 177 7.1.2.2 Gaseous Effluents .................................... 178 7.1.2.3 Cooling and Process System Activation Products. 178 7.1.2.4 Beam Port Plug Activation ............................ 179 7.1.2.5 Experimental Sample Activation ....................... 182 7.2 Protection Strategy ......................................... 184 7.3 Protection Methodology ...................................... 184 7.3.1 Instruction and Training................................. 184 7.3.2 Access Control........................................... 185 7.3.3 Personnel Monitoring..................................... 185 7.3.4 Area Monitoring.......................................... 186 7.3.5 Survey Instruments....................................... 186 7.3.6 Shielding ............................................... 187 vi

7.3.7 Administrative Controls........... ................ .

7.4 Chapter 7 References ............. ................ .

8.0 Safety Analysis........................ ................ .

8.1 Introduction and Overview ............ . .. .. .. .. . 189 .

8.2 Scenario Construction ................ . .. .. .. .. . 189 .

8.3 Analytical Methods and Tools ......... ....... 189 8.4 Accident Cases ....................... .... 190 .

8.4.1 Loss of Heat Sink................. .... 190 .

8.4.2 Loss of Coolant................... .... 192 .

8.4.2.1 Instantaneous Coolant Loss .... .... 194 .

8.4.2.2 Non-Instantaneous Coolant Loss .... 194 .

8.4.3 Reactivity Insertion.............. .... 204 .

8.4.3.1 Insertion Mechanisms .......... 204 .

8.4.3.2 Design Basis Accident ......... .... 205 .

8.4.3.3 Results....................... .... 210 .

8.4.3.4 Summary of Reactivity Insertion Analysis ............. 210 8.4.4 Damaged Fuel Plate....................................... 216 8.4.4.1 Design Basis Accident for Radioiodine Release......... 216 8.4.4.2 Source Term Estimation ............................... 217 8.4.4.3 Thyroid Dose Consequences ............................ 222 8.4.4.4 Design Basis Accident for Gaseous Radionuclides. 222 8.4.4.5 Whole-Body Gamma Dose Estimation ..................... 225 8.5 Natural Phenomena ........................................... 234 8.5.1 Earthquake............................................... 234 8.5.2 Severe Storms, Floods.................................... 235 8.6 Man-Made Phenomena .......................................... 235 8.6.1 Fire or Explosion........................................ 235 8.6.2 Acts of Sabotage, Civil Disobedience, Riots .............. 236 8.7 Chapter 8 References ........................................ 237 9.0 Administrative Controls....................................... 238 9.1 Organization ................................................ 238 9.1.1 Structure ............................................... 238 9.1.2 Responsibility........................................... 238 9.1.3 Support Groups........................................... 238 9.2 Training Program ........................................... 240 9.3 Recordkeeping and Reporting Requirements..................... 240 i

vii

9.4 Emergency Planning and Preparedness....... . 240 9.5 Internal Reviews and Audits ..............

9.5.1 OSURR Reactor Operations Committee.... ....

. 240

. 240 0

9.5.1.1 Responsibilities and Authority .... .... . 240 .

9.5.1.2 Committee Membership ............. . 240 .

9.5.1.3 Committee Meetings ............... .... ................ 241 9.5.1.4 Subcommittees.................... .... . 241 .

9.5.2 Experiment Approval................... .... . 241 .

9.5.3 Additional -Oversight.................. .... . 242 .

9.5.3.1 The Radiation Safety Section (RSS) .... . 242 .

9.5.3.2 Operations Manager ................ . 242 .

9.5.3.3 SRO On-Duty....................... . 242 .

9.6 Security ................................. ................ 242 9.6.1 Security Plan......................... . 242 .

9.6.2 Security Organization................. . 242 .

9-7 Quality Assurance ........................ . 243 .

9.7.1 Quality Assurance Program............. .... 243 ..........

9.7.2 Program Requirements.................. . 243 .

9.7.2.1 Program Requirements .............. . 243 .

9.7.2.2 Records and Documents ............. . 243 (W/

10.0 Financial Qualification....................... .. 244 .

10.1 Financial Ability to Operate a Non-Power React:or ............ 244 2 10.2 Financial Ability to Decommission ........... . 244 .

viii U.,

List of Figures Figure 2.1: Aerial View of the OSURR Site, West-East Perspective ..... 5 Figure 2.2: Aerial View of the OSURR Site, South-North Perspective... 6 Figure 2.3: Detailed Map of the OSURR Surrounding Areas ............. .7 Figure 2.4: Businesses and Industries in the Areas South and East of the Reactor Building and Site ........................... 8 Figure 2.5: Businesses and Industries in the Area East of the Reactor Building and Site .. 9 Figure 2.6: Seismicity of the United States: 1899-1990 ... .......... 13 Figure 2.7: Modified Mercalli Intensity Scale with Approximate Equivalent Richter Magnitude .14 Figure 2.8: Map of Historically Known Earthquakes in Ohio with Richter Magnitude 2.0 or Greater .15 Figure 2.9: Potential Regional Modified Mercalli Intensity Map Projected for a Future Possible Richter Magnitude 8.0 Earthquake in the New Madrid Area .17 Figure 2.10: Mid-Continent Regional Distribution Map of Peak Lateral Ground Acceleration (as a percent of gravitational acceleration) Having a Two Percent Probability of Being Exceeded in Fifty Years .18 Figure 2.11: Wind Speeds and Directions in the Columbus Area .19 Figure 2.12: Dimensions of the Reactor Building Shown in a Cross-Sectional View ......................................... 24 Figure 3.1: OSURR Grid Plate ........................ .............. 28 Figure 3.2: OSURR Grid Plate Frame .29 Figure 3.3: OSURR Standard Fuel Element .31 Figure 3.4: OSURR Control Rod Fuel Element ..... .................... 32 Figure 3.5: OSURR Source Drive Assembly ................ I............. 35 Figure 3.6: Overall Reactor Dimensions ....... ...................... 38 Figure 3.7: Fuel Storage Pit . ........................................ 39 Figure 3.8: Fuel Storage Pit Plugs . ................................. 40 Figure 3.9: Fuel Storage Pit Rack .................................. 41 Figure 3.10: Thermal Column Extensions ............. ................ 43 Figure 3.11: Fuel Handling Tool for Standard and Partial Fuel Elements

....................................................... 44 Figure 3.12: Fuel Handling Tool for Control Rod Fuel Elements 46 Figure 3.13: Locations of Instruments and Facilities In and Around the OSURR Core ...........-------.-------............... 48 tL ix

Figure 3.14 : Schematic Diagram of the OSURR Cooling System ... ...... 50 Figure 3.15 : Flow Diagram of the Makeup Water System .... ............ 57 Q.);

Figure 3.16 : Flow Diagram of the Water Process System ... ........... 58 Figure 3.17 : Control Room Layout ................... ................. 60 Figure 3.18 : Diagram of the OSURR Startup Channel .... .............. 66 Figure 3.19 : Simplified Diagram of the OSURR Reactor Trip System.... 74 Figure 3.20 : OSURR Beam Ports ...................................... 80 Figure 3.21 : OSURR Beam Port Plug ............. I ..................... 81 Figure 3.22 : OSURR Rabbit. System Piping Diagram .... ................. 82 Figure 3.23 : Details of the OSURR Rabbit System .................... 84 Figure 3.24 : Top Cutaway View of the Main and BSF Thermal Columns... 85 Figure 3.25 : Side Cutaway View of the Main Graphite Thermal Column.. 86 Figure 3.26 : Side Cutaway View of the BSF Thermal Column ... ........ 88 Figure 3.27 : Details of the Graphite Irradiation Elements, Tubes, &

Plugs .................................................. 89 Figure 4.1: Numbering System for Denoting Positions on the Grid Plate of the OSURR ........................................... 92 Figure 4.2: Possible OSURR Core Configurations, LEU Cores F and G... 93 Figure 4.3: Possible OSURR Core Configurations, LEU Cores J and P... 94 Figure 4.4: Reference LEU Core for the OSURR (Core I) .... .......... 95 Figure 4.5: Thermal Neutron Flux Profile for LEU Core F at 500 kW Operating Power ....................................... 100 Figure 4.6: Thermal Neutron Flux Profile for LEU Core G at 500 kW Operating Power ........................................ 101 Figure 4.7: Thermal Neutron Flux Profile for LEU Core J at 500 kW Operating Power ........................................ 102 Figure 4.8: Thermal Neutron Flux Profile for LEU Core P at 500 kW Operating Power ............................. I........... 103*

Figure 4.9: Thermal Neutron Flux Profile for LEU Core I at 500 kW Operating Power ....................................... 104 Figure 4.10: Thermal Neutron Flux Profile for LEU Core I at 500 kW Operating Power ....................................... 105 Figure 4.11: Neutron Flux Profile for One of the Two Fast Neutron Groups for LEU Core I at 500 kW Operating Power .......106 41 Figure 6.1: Ar Production Curve .................................. 147 Figure 8.1: Pool Water Temperature Rise as a Function of Shutdown Time and Power. History ..................................... 193 Figure 8.2: Power Transients From Various Reactivity Insertions Assuming a Scram Occurs ............................... 211 I;

x

Figure 8.3: Fuel Cladding Temperatures Resulting From Reactivity Insertions Assuming a Scram Occurs .... ................ 212 Figudre 8.4: Power Transients Resulting From Various Reactivity Insertions Assuming No Scram Occurs .... ............... 213 Figure 8.5: Fuel Cladding Temperature Resulting From Reactivity Insertions Assuming No Scram Occurs .... ............... 214 Figure 9.1: Administrative Organization ............................ 239 xi

List of Tables Table 2.1: Rock Types Underlying the Reactor Building Site ... ......12 Table 2.2: Temperature Summary ...................................... 20 Table 2.3: Precipitation Summary ................................... 21 Table 2.4: Other Meteorological Data ............................... 22 Table 3.1: Control Rod Positioning System Indicator Lights ... ......64 Table 3.2: Scram and Alarm Functions of the OSURR Safety System ..... 75 Table 4.1: Summary of Reactor Data ................................ 122 Table 6.1: Measured and Estimated Thermal Neutron Fluxes in Various Experimental Facilities of the OSURR .... .............. 133 Table 6.2: Calculated Air Volumes in Various Experimental Facilities of the OSURR .......................................... 135 Table 6.3: Estimated 41 Ar Source Terms for the Various Experimental Facilities of the OSURR ............................... 136 Table 6.4: Release Points of the Various Experimental Facilities of the OSURR ............................................. 144 Table 6.5: Estimated 41Ar Concentrations for Puff Releases of Saturation Activities of 4'Ar from the Various Experimental Facilities of the OSURR ............................... 149 Table 6.6: Operational Limits and Activation Time Estimates for the C)

Various Experimental Facilities of the OSURR ... ....... 150 Table 6.7: Estimated 4 1Ar Concentrations in the Lee of the Reactor Building for Puff Releases of "Ar from the Various Experimental Facilities of the OSURR .... .............. 157 Table 6.8: Operational Limits for the Various Experimental Facilities of the OSURR to Maintain Unrestricted Area Effluent Concentration Limits on the Lee Side of the Reactor Building ............................................... 158 Table 6.9: Estimated Dose Rates Resulting From 41Ar Concentrations As A Result Of Releases From Various Experimental Facilities of the OSURR ............................................. 162 Table 7.1: Geometric Parameters of Various Irradiation Facilities.. 172 Table 7.2: Measured Core-End Neutron Flux in Various Irradiation Facilities ............................................. 173 Table 7.3: Neutron Fluence to Dose Conversion Factors .... ......... 174 Table 7.4: Estimated Neutron Flux at the Exit of Various Irradiation Facilities ............................................ 175 Table 7.5: Estimated Neutron Dose at the Exit of Various Irradiation Facilities ........................................... 176 Table 7.6: Activity Estimates for Shielding Plug End Cap ... ....... 181 xii I

Table 7.7 : Dose Rate Estimates for Shielding Plug End Cap ... ...... 183 Table Fractional Heat Removal By Boiling and Steam Convection in a Partially Uncovered Fuel Channel of the OSURR .......200 Table 8.2: Reactivity Insertion Accident Analysis Input Data ....... 207 Table 8.3:: Reactivity Insertion Accident Analysis Input Data ....... 208 Table 8.4. Reactivity Insertion Accident Case Summary .... ......... 209 Table 8.5: Reactivity Insertion Accident Results Summary Assuming No Scram Occurs .......................................... 215 Table 8.6 Constants and Calculational Results For Radioiodine Production in the OSURR Core .......................... 218 Table 8.7 Number of Moles and Activities for Radioiodine Isotopes in Building Air After Release From The Pool .... .......... 221 Table 8.8 Source Terms and Other Data for Integrated Thyroid Dose Calculations .......................................... 223 Table 8.9: Thyroid Dose Estimates ................................. 224 Table 8.10 Constants and Calculational Results Used in Estimation of Submersion Cloud Whole-Body Gamma Dose .... ............ 227 Table 8.11 Integral Whole-Body Gamma Doses in The Building Assuming an Infinite Cloud and a Leakage Fraction Of 0.0042 Hr-1 (Purge Fan Off) ........................................ I ...... 228 Table 8.12 Integral Whole-Body Gamma Doses in The Building Assuming an Infinite Cloud and a Leakage Fraction Of 0.857 Hr-1 (Purge Fan On) ............................................... 229 I Table 8.13 Constants Used in Whole-Body Dose Estimates ... ........ 231 Table 8.14 Integral Whole-Body Gamma Doses in The Building Assuming a Finite Cloud and a Leakage Fraction Of 0.0042 Hr-1 (Purge Fan Off) ..................................... .......... 232 Table 8.15 Integral Whole-Body Gamma Doses in The Building Assuming a Finite Cloud and a Leakage Fraction Of 0.857 Hr'1 (Purge Fan On) ........ ....................................... 233 xiii

1.0 Introduction 1.1-Purpose This document will present a description and safety analysis for The Ohio State University Research Reactor (OSURR). This reactor, owned and operated by The Ohio State University, is located on the Columbus Campus of The Ohio State University, within the City of Columbus, in central Ohio. The descriptions and analyses presented in this report will provide sufficient information to show that the reactor can be operated with reasonable assurance that the health and safety of the public will be protected.

The description of the reactor system and its associated components are sufficiently detailed to allow an understanding of the general features, characteristics, and basic operation of the reactor. The safety analysis makes conservative assumptions to allow larger safety margins-1.2 General Facility Description The OSURR is a pool-type reactor using light water as a moderator and coolant. The core of the reactor utilizes uranium fuel, enriched to 19.5%, in uranium-silicide (U3 Si2 ) form, clad in aluminum. The fuel is in solid flat plate form, commonly called MTR-type fuel. Fuel plates are mechanically joined into fuel assemblies (also called fuel elements), which are stacked to form an approximately symmetric rectangular solid. The fuel assemblies are positioned in a grid plate forming a 5 by 6 rectangular matrix for available fuel assembly ( )

positions. The grid plate is bolted to the floor of the reactor pool.

A plutonium-beryllium (Pu-Be) startup source provides an initial population of neutrons to the core for controlled reactor startup.

Reactor control is effected by three control rods of boron-stainless steel composition (called shim safety rods), and an additional control rod composed-only of stainless steel (known as the regulating rod).

These control rods are positioned by electric motors. The three shim safety rods are held by electromagnets and can be inserted into the core under the influence of gravity by turning off the current to these electromagnets. The rods move within aluminum shrouds and extend into special control rod fuel elements. The active length of the control rods is sufficient to completely cover the active portion of the core. The control rod housings are held by brackets mounted to the sides of the reactor pool.

A number of experimental facilities converge at the reactor core. This allows simultaneous performance of a number of different experiments.

These facilities include twb beam ports, a pneumatic transfer facility (rabbit), a main graphite thermal column, a smaller graphite thermal column, a central irradiation facility (CIF) that can extend into either a water or graphite-filled flux trap, movable graphite isotope irradiation elements (GIIE), and movable dry tubes.

1

The reactor pool wall is made of barytes concrete. The interior surface of the pool is coated with a waterproof liner. The reactor pool has a capacity of 5800 gallons, with a minimum water depth of 15 feet maintained above the top of the core for shielding. Water purity is maintained by a process system composed of a circulating pump, demineralizer, and particulate filter. Makeup water is added to the reactor pool from city water supplies, after passing through a resin bed demineralizer unit.

The reactor is licensed to operate at continuously variable thermal power up to a maximum of 500 kilowatts. Operation is limited to steady-state power, with no pulsing capabilities. At maximum steady-state power, the average thermal neutron flux in the core is 4.66x1012 neutrons/cm2 /second. In cold, clean critical condition, the core contains approximately 2.6% Ak/k excess reactivity. Because of their location, geometry, and composition, the controls rods have a total worth of approximately 8.45% Ak/k. The shutdown margin is at least 1% Ak/k with the regulating rod and the highest-worth shim safety rod fully removed from the core.

The core is cooled by natural convective flow of pool water vertically through the core within the flow channels between the fuel plates.

Pool water enters the bottom of the core at an inlet temperature of approximately 20 0 C, is heated by the core, and exits the top of the core at approximately 60 0C Heated water enters an aluminum plenum, is withdrawn from the plenum and circulated through a closed-loop heat removal system. Heat rejection is achieved through a two-stage, closed-loop secondary cooling system. The primary-secondary heat exchanger removes primary coolant heat to an ethylene-glycol and water mixture, from which heat is rejected to the outside atmosphere through a fan-forced air circulation cooling unit (also referred to as a dry cooler). An additional secondary-loop heat exchanger provides further cooling of the ethylene-glycol and water coolant by using city water as its heat sink. The total heat removal capacity of the cooling system-is sufficient to remove all of the 500 kilowatts of thermal energy generated in the core, and maintain an average equilibrium bulk pool temperature of 20-25 0 C under all credible environmental conditions.

1.3 Background Information The OSURR was first operated in 1961. Its operation is regulated by the U.S. Nuclear Regulatory Commission (NRC), under facility license number R-75, docket number 50-150.

The design of the OSURR is based on the Bulk Shielding Reactor (BSR),

which was located at the Oak Ridge National Laboratory (ORNL). This reactor is in a class of reactors generally known as a Materials Testing Reactor (MTR). This class of reactors share various common features, among them are light water moderation and cooling, open 2

pools, and plate-type fuel. The reactor itself was supplied by Lockheed Nuclear Products, then a division of the Lockheed Georgia Company. Lockheed operated a reactor very similar in design to the OSURR, at a power level of 1 megawatt steady-state thermal power, in a forced convection cooling mode. When operated in the natural convection cooling mode at power levels up to 10 kilowatts, the Lockheed reactor was essentially identical in operating characteristics to the OSURR for the first 25 years of OSURR operation.

License R-75 authorized The Ohio State University to operate the OSURR at steady-state thermal power levels up to 10 kilowatts, using natural.

convection cooling. Originally, up to 8 kilograms of 93% enrichment

-U was permitted to be possessed by the university at the reactor site. This was later lowered to 4.6 kilograms upon removal of the Fission Plate from the Bulk Shielding Facility (BSF) of the OSURR, by License Amendment 7, in 1976. As of the end of 1986, nominal (i.e.,

without considering fuel burnup) 3575.81 grams of 235U was located at the reactor site in the form of fuel, and approximately 80 grams of plutonium was contained in the startup source. In 1988, LEU fuel was received to replace the HEU fuel, and the HEU fuel was shipped offsite in 1995.

The OSURR is utilized by the university for a variety of instructional, research, and service activities. Reactor use is not confined to persons employed or associated with the university. Past utilization has involved area universities and colleges, as well as local secondary and middle schools. Other individuals and groups in private industry and other state and federal governmental agencies have used the OSURR in a variety of ways. At the end of 1999, the C)

OSURR was the only operating research reactor in the State of Ohio.

1.4 Report Organization A description and characterization of the OSURR site is presented in the following chapter. Details of the facility design are discussed in Chapter 3. The operating characteristics of the OSURR under normal conditions are presented in Chapter 4. Auxiliary systems and radioactive waste management are-discussed in Chapters 5 and 6, respectively. Facility features and operational procedures for radiation protection are the subjects of Chapter 7. Chapter 8 contains the safety analysis for the OSURR. Chapter 9 presents the administrative organization and controls for the reactor facility, and Chapter 10 discusses financial qualifications. Appendix A contains the complete Technical Specifications for the OSURR.

3

2.0 Site Description and Characterization 2.1 General Location The OSU research reactor is located on property owned by The Ohio State University, west of the main campus. Aerial photographs are shown in Figures 2.1 and 2.2. A map of the surrounding territory with a 3 mile radius circle centered on the reactor building is shown in Figure 2.3.

2.2 Demographics 2.2.1 Surrounding Population The map shown in Figure 2.3 shows that the reactor building is completely surrounded by residential dwellings iwithin a three mile radius of the reactor site. Figures 2.4 and 2.5 show the locations of major industrial buildings found within a one mile radius of the site.

Most industrial and business activities are located south and east of the reactor building. To the west are primarily residential areas.

Some businesses are located to the northwest, but census data indicates that these employ relatively few people compared with those firms located to the east and south within a one mile radius of the reactor site.

Data for'both residential and industrial populations are as follows:

Type Radius Population Residential 3 Mile 141,600 Industrial 1 Mile 4,600.

2.2.2 Local Activities As of Autumn Quarter, 1998, The Ohio State University is composed of approximately 65,000 students, staff, and faculty. It is a land grant institution, engaged in teaching, research and public service activities. Various community services are provided, including medical and dental services. The university is active in the performing arts, as well as in athletics, and has recently completed construction of the Wexner Center for the Arts as well as the Schottenstien Arena.

Activities in these facilities periodically draw up to 100,000 additional people to the campus area for special events such as performing arts concerts, football, basketball and NHL games, circuses, and rock concerts.

4

Fly If? ~. @o ,S vj' -

Figure 2.1: Aerial View of the OSURR Site, West-East Perspective 5

c, C Ic

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it~~~~~~~~ ,,e , 1 .4~, s ~ - ^ ,

1/4..iO IV'

.- - \ s . , . -

w - . . ..-. W . .

Figure 2.2: Aerial View of the OSURR Site, South-North Perspective 6

41*

x MI~

Figure 2.3: Detailed Map of the OSUER Surrounding Areas 7

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rApnfcATC Cur 40 _

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5 TT I A IS 3RI INGLIS __

FORTSII UELDIJG & MfG INC ISOI urnrt;---

w ORNmMrHNTAL _B CUSuTOM METAL rA!'ICATON in the Areas South and East of Figure 2.4: Businesses and Industries the Reactor Building and Site 8

r UNt; ttCTALLUkGCICL COUP D:.rJ nt %l.. t0t4 jj - L r ; FINE CA:T DIV tJAC D;...LUC O 3 IlJNCAR RD tOC I~pjj k; ;;nC:il PRECI ION TNVE.TME1JT Ftn 4f CA;Tl4GS EHP 75

-COTT IIJDEX CO I%5 /.l£:: :T rUDLISH COUNTY LAND RECORD POOl;: CMP 2,0

//ALETT HCATI1GC & COOLING 720 I1Y42 AVE

_ 4 ShEET METAL VORI; M/CCU.J EDISON :ERV/NATL ELEC COIL

_CRD // c00 KING AVE

/C'AIR ELZCTRICAL EQUIP-

_ _\/rT EnP 600 0 LrUNOX UNDUSTRI;C 1;SC EArTCRN DIV 1711 OLENTAUICY RIVER RD RC:IDCNTIAL & conmcricIAL AIR COtJDTIONIllG £1P i000 FRED D F FEllING CO 1075 STH AVE. U

/BAKCRY MACHINCRY Smr 90 INDUSTRIAL CERAMIC PRODUCT: INC p 965 5711AVE. U

_lA1 / POTTERY PIN:. RErACTORY SPECIALTIES EMP 460 DATTELLE MEMORIAL IRNTITUTE 50-5 KING AVE R & D LALS . CEIP 3000

______ /DY'ART CHEMICAL CORP Mr n144. OLENTAUGY RIVCR RD T CoHTRACT PACA;GING EMP 50 TIlE COLUMODU SIIOU CASE CO 850 SIN AVE. U

_TORE FIXTURE; EnP 230 DOtJCO INC ODA PONY ExrRc:S

_ -9095TH AVC. U PRINTING, *TYPeCSTTING.

ART, & DE IGN E11P 2S

.NCDICK DAF4ROV: cc w

AV a,".5TH AVE. U PHOTOGR4APHIC MOUN TS.

AiAT: & FOLDER: CMP 40 CHESTER A :nITH INC 1330 NORTON AVE FnuiNo IIARBLE ClIP is MURRAY CITY COAL & ICE CO

\\ TlL RLD nrc CS \13Z4 EEHILL AD EIP 20 R A0M.UATYICE & BLOCK ICC CIRCtILAR _U F:ADC PLY BOARD FABRICATO^: CO INC

\294 \DGEIIZLL RD IOutcuxEmitNsmc;C & .UPPLY Cc PALLCT:. SHIPPING CASE:.

1l41 NblcION AVE UOOD DiOXES L COMPONENTS,

mr: Ituzou: *-r 1: I RUDlCk PrODUCT: ClP 40 Emp 15 liAr :c4A4L PNRuOUCT. Zi ME CE 1S31-r. 1UL..Y AVE ; L~ --IANR t C .- OSL fr STAVflING CO AIR CONDITIONINO CL-.~NL M 'l1; 30R10 tIlULLYAVC 051 3nD AVE. U usrI;;U wu- ucr. r:rc tG3rouL. VIr: & nCTAL AIR CONDITIONING & "CAT-rX.4-. ;o ' TACtLDG. Etir A IN40GCCUIPMENT CMr 350 Figure 2.5: Businesses and Industries in the Area East of the Reactor Building and Site 9

2.3 Topography, Geology, and Seismology 2.3.1 Topography Most of Ohio includes portions of two physiographic provinces called the Appalachian Plateau and the Central Lowlands. Franklin County is divided into these two sections by a series of north-south scarps and terraces which form a gentle step-like ascent eastward to the Appalachian Plateau. The highest altitude of the county is an elevation of 1130 feet above mean sea level, located in the northeast corner of Plain Township, and the lowest is an elevation of 665 feet above mean sea level at the efflux of the Scioto River from Franklin County. In the northern part and southwestern one-third of the county, the valley floors range in altitude from 780 to 890 feet above mean sea level, hilltops range from 860 to 960 feet above mean sea level, and local relief seldom exceeds 170 feet. The range in altitude of the valley floors in the northeastern and north central parts of the county is 710 to 840 feet. The hilltops range from 900 to 1130 feet above sea level. In the south central and southeastern parts of the county the valley floors range in altitude from 670 to 760 feet and hilltops range generally from 690 to 780 feet (locally they are 840 feet) above mean sea level. Except in the extreme southeast part of Madison Township, local relief does not exceed 50 feet. Columbus is located in the center of the county with a ground elevation of about 812 feet above mean sea level. The reactor site is about 780 feet above mean sea level.

W ~ 2.3.2 Geology Columbus lies on the glaciated plains section at the eastern edge of the central lowlands physiographic province. When the Plio-Pleistocene Ice Age began two million years ago, a continental scale ice sheet originating near Hudson Bay moved southward modifying the pre-glacial landscape. The deposits from the most recent glacial advance, the Wisconsin Glaciation, lie directly on the limestone and shale bedrock underlying Columbus. The area was completely buried by glacial till, generally 10 to 30 feet thick, consisting of unsorted clay, silt, sand, pebbles, and boulders (mainly derived from the Canadian Shield) carried south at the base of the ice sheet. Large out-wash deposits in the Scioto and Olentangy Valleys resulted from the great volumes of melt water coming from the kilometer-thick ice sheet (thickening to three kilometers near its source at Hudson Bay). The deposits occur above the present drainage level as gravel terraces and serve as water recharge areas north and south of downtown Columbus, principally on the west side of the Scioto River.

The bedrock immediately underlying the area is composed of the Columbus and Delaware Limestone, and the Olentangy and Ohio Shale of Devonian Age, deposited approximately 350 million years ago when Ohio was at the eastern edge of ancestral North America and was a near-shore marine environment. The Columbus and Delaware Limestone and the 10

Olentangy Shale represent successive depositional stages from a shallow marine environment to deeper marine environment. All sedimentary rocks underlying Columbus dip gently to the southeast. The 0

pre-Devonian sedimentary rocks were derived from the Cincinnati Arch, a belt of Precambrian crystalline igneous and metamorphic rocks trending north-south along the Ohio-Indiana border. Now confined to the subsurface, this arch controls the linear pattern of Paleozoic deposition in western and central Ohio. The bedrock sediments outcrop in bands extending roughly north-south in Central Ohio. The Ohio Shale outcrops in the eastern part of the Columbus area; the Columbus and Delaware Limestone outcrop on the west side of the Scioto River. Near the close of the Paleozoic Era, the convergence of Africa and North America resulted in the Appalachian Orogeny. The western limits of this major mountain building episode are seen from northeast to south-central Ohio, where the regional southeastern dip of paleozoic strata rapidly steepens and the crustal thickness markedly begins to increase beneath the Appalachian Mountains.

Underlying Devonian strata in the Columbus area are older Paleozoic sedimentary formations. These are in turn underlain by Precambrian igneous and metamorphic basement rocks which make up the Proterozoic Craton extending northward and westward to outcrop in Canada as the Canadian Shield. These basement rocks are roughly 2 kilometers below the surface in central Ohio.

The bedrock at the reactor site is Delaware Limestone, a mixture of argillaceous cherty blue limestones and calcareous brown shales. These strata are covered by glacial drift which is predominantly gravel and clay. A boring analysis taken at a point about 500 feet southwest of the reactor site gave the information shown in Table 2.1.

2.3.3 Seismology Figure 2.6 is a map of all of the instrumentally recorded earthquakes of Richter Magnitude 4.5 or greater that have occurred in the United States from 1899 through 1990. A belt of seismic activity (the St.

Lawrence Seismic Belt) runs through Northwestern Ohio. Although the most persistently seismically active regions in the United States do not lie in Ohio, Ohio is not a-seismic contrary to popular opinion.

Figure 2.7 gives the Modified Mercalli Intensity Scale with approximate corresponding values of equivalent Richter Magnitude.

Historically, the most active seismic region in Ohio is near Anna in Shelby County, approximately 60 miles northwest of Columbus. To date this region has had at least 35 earthquakes, including the three largest seismic events instrumentally recorded in Ohio. A map of the historically known earthquakes in Ohio with Richter Magnitude 2.0 or greater is provided in Figure 2.8. Modified Mercalli Intensitiy equivalent of each event is color coded. As of 1999, the first statewide seismic network was established in Ohio became operational (Hansen, 1999). Called 'OhioSeis", this network digitally records seismic events of global origin but is intended to significantly 11

Table 2.1: Rock Types Underlying the Reactor Building Site Strata Depth (ft)

Clay 0 to 60 Slab Rock 60 to 63 Hard Clay 63 to 81 Rock 81 to 85 Hard Clay and Gravel 85 to 109 Hard Rock 109 to 115 Clay 115 to 138 Rock 138 to 142 Soft Clay 142 to 158 Limestone 158 to 190 12

Figure 2.6: Seismicity of the United States: 1899-1990 13 C C.I C

Magnitude Modified Mercalli Intensity Scale Scale I Detected only by sensitive instruments 1.5 -

Felt by few persons at rest, especially on upper 11 floors; delicately suspended objects may swing 2-Fefi noticeably indoors, but not always recog-III nized as earthquake: standing autos rock slightily. 2.5-vibrations tike passing truck Fell indoors by many, outdoors by few, at night 3-IV someawaken: dishes, windows, doorsdisturbed; standing autos rock noticeably 3.5-Fell by most people: some breakage of dishes, V windows, and plaster: disturbance of tall objects 4-Feh by all, many frightened and run outdoors; VI falling plaster and chimneys, damage small 4.5 -

Everybody runs outdoors: damage to buildings VII varies depending on quality of construction: no- 5-ticed by drivers of autos Panel walls thrown out of Irames; walls, monu- 5.5 -

VIII ments, chimneys tall: sand and mud ejected:

drivers of autos disturbed 6-Buildings shifted o foundations. cracked, thrown IX out olpiumb: ground cracked; underground pipes broken 6.5 -

Most masonry and frame structures destroyed:

X ground cracked, rails bent. landslides 7-Few structures remain standing; bridges de-Xi stroyed, fissures in ground, pipes broken, land-7.5 -

slides, rails bent Damage total: waves seen on ground surface, 8a-XII lines of sight and level distorted, objects thrown up into air Figure 2.7: Modified Mercalli Intensity Scale with Approximate Equivalent Richter Magnitude 14

I 0294; Z 197 CAR-iUT UCS '.

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2 Ne n n~mrnt l rn n O h i.Sn..i. LI ' - i. , - l, p9 ul,'a A1 7 01 2 5 5**)iw.4 0I~r.~ 19:12, 0037 195 Figure 2.8: Map of Historically Known Earthquakes in Ohio with Richter Magnitude 2.0 or Greater Notes:

1) Equivalent Modified Mercalli Intensity is color coded.

2) OhioSeis seismic network locations are shown by stars.

15

enhance the detection, location and magnitude of future seismic events p) in the mid-continent in general and Ohio in particular. In the near future, the seismic risk assessment to structures of all kinds due to earthquakes originating within and near Ohio's borders will be better understood.

Although very little historical information is available on earthquakes with epicenters in or near Columbus, or about the effects, of such earthquakes on the city, it is appropriate to assume that known mid-continent earthquakes have historically had some effects here. Figure 2.9 shows the potential Modified Mercalli Intensity distribution resulting from a Richter Magnitude 8.0 seismic event in the New Madrid area. Damage consistent with MM VIII would be likely in central Ohio.

Figure 2.10 shows the Mid-continent regional distribution of peak lateral ground acceleration (as a percentage of gravitational acceleration = 1 g) having a two percent probability of being exceeded in 50 years. The Anna, Ohio, area discussed above is clearly discernable. More detailed information about this and related seismic risk information is available at the USGS Geological Hazards Web site http://geohazards.cr.usgs.gov/eq/.

The possibility of a Richter Magnitude 7.0 or greater seismic event in the New Madrid area (southeast Missouri - northwest Arkansas - western Tennessee) affecting the Columbus Metropolitan Area 400 miles (600 kilometers) away is unfortunately real. However, The Ohio State University reactor facility is not likely to suffer sufficient damage resulting in a serious hazard. Cracking of the shield would probably be the most serious damage. Should the pool liner be ruptured, pool water would escape from the pool leaving the reactor unshielded in the vertical direction. The reactor would then be sub-critical because of the absence of the moderator.

2.4 Meteorology Figure 2.11 is a graph of the average wind speed versus-wind direction. Tables 2.2 and 2.3 contain summaries of temperature and precipitation for Columbus, Ohio. Table 2.4 contains various other relevant meteorological data. Meteorological information given in Figure 2.11 and Table 2.4 was taken from data published by the National Climatic Center in Asheville, North Carolina. The data in Tables 2.2 and 2.3 was obtained from the WWW page of the State Climate Office for Ohio at http://twister.sbs.ohio-state.edu/climoff.htm.

2.5 Hydrology 2.5.1 Surface Water Columbus is located in the center of the state and in the drainage area of the Ohio River. Four nearly parallel streams run through or 16

I 0

Figure 2.9: Potential Regional Modified Mercalli Intensity Map Projected for a Future Possible Richter Magnitude 8.0 Earthquake in the New Madrid Area Note: Refer to Figure 2.7 for interpretation of the local severity of projected structural damage in terms of local Modified Mercalli Intensity.

17

Figure 2.10: Mid-Continent Regional Distribution Map of Peak Lateral Ground Acceleration (as a percent of gravitational acceleration)

Having a Two Percent Probability of Being Exceeded in Fifty Years 18

CM¶HCOLUMB5US, OH CLASS 7 U

Figure 2.11: Wind Speeds and Directions in the Columbus Area 19 I

Table 2.2: Temperature Summary

- - - - TEMPERATURE

SUMMARY

Station: (331786) COLUMBUS WSO AIRPORT Missing Data: 0% NCDC Averages Averages: 1961-1990 Extremes: 1948-1996 #Day-Max #Day-Min Averages Daily Extremes Mean Extremes => <= <= <=

Max Min Mean High---Date Low---Date High-Yr Low-Yr 90 32 32 0 Ja 34.1 18.5 26.4 74 25/1950 -22 19/1994 39.9 50 11.4 77 0 12 26 2.2 Fe 38.0 21.2 29.6 73 25/1957 -13 02/1951 39.0 54 16.6 78 0 8.0 23 1.2 Ma 50.5 31.2 40.9 82 31/1981 -6 09/1984 50.4 73 28.4 60 0 2.3 18 0.1 Ap 62.0 40.0 51.0 88 23/1960 14 07/1982 57.4 54 45.8 50 0 0.1 6.7 0 Ma 72.3 50.1 61.2 93 30/1953 25 10/1966 70.9 91 55.4 67 0.6 0 0.5 0 Jn 80.4 58.0 69.2 101 25/1988 35 11/1972 75.0 91 63.6 72 4.3 -0 0 0 Jl 83.7 62.7 73.2 104 14/1954 43 06/1972 79.0 55 70.5 71 6.8 0 0 0 Au 82.1 60.8 71.5 101 20/1983 39 29/1965 78.4 95 68.3 67 4.8 0 0 0 Se 76.2 54.8 65.5 100 02/1953 31 21/1962 71.0 61 60.4 67 1.6 0 0.1 0 Oc 64.5 42.9 53.7 90 05/1951 17 21/1952 59.9 63 47.4 88 0 0 3.7 0 No 51.4 34.3 42.9 80 01/1950 -4 30/1958 48.2 85 33.9 76 0 1.3 14 0 De 39.2 24.6 31.9 76 03/1982 -17 22/1989 40.8 56 19.8 89 0 8.2 24 0.8 An 61.2 41.6 51.4 104 07/14/54 -22 01/19/94 55.4 91 49.7 76 18 32 116 4.4 Wi 37.1 21.4 29.3 76 12/03/82 -22 01/19/94 36.4 49 20.7 77 0 28 72 4.3 Sp 61.6 40.4 51.0 93 05/30/53 -6 03/09/84 57.0 91 46.6 84 0.6 2.4 26 0.1 Su 82.1 60.5 71.3 104 07/14/54 35 06/11/72. 75.9 91 68.6 72 16 0 0 0 Fa 64.0 44.0 54.0 100 09/02/53 -4 11/30/58 57.5 73 47.17 76 1.6 1.3 18 0

- - - - Midwestern Climate Center, Champaign IL

20

Table 2.3: Precipitation Summary

- - - - ~ PRECIPITATION

SUMMARY

Station: (331786) COLUMBUS WSOAIRPORT Missing Data: 0%

Averages: 1961-1990 Extremes: 1948-1996 Total Precipitation Snow f#Days Precip Mean High--Yr Low--Yr i-Day Max Mean High--Yr =>.01 =>.50 =>1.

Ja 2.18 8.29 50 0.65 61 4.79 21/1959 9.2 34.4 78 13.7 1.8 0.4 Fe 2.24 5.15 90 0.31 78 2.15 23/1975 7.0 16.4 79 11.6 1.3 0.2 Ma 3.27 9.60 64 1.01 79 3.40 9/1964 4.4 13.5 62 13.8 1.8 0.3 Ap 3.21 6.39 96 0.67 71 2.03 30/1983 1.1 12.6 87 13.2 2.1 0.6 Ma 3.93 9.11 68 0.95 77 2.12 29/1982 0.0 0.8 89 12.5 2.7 0.7 Jn 4.04 9.75 58 0.71 84 2.55 13/1981 0.0 0.0 49 10.7 2.9 1.0 Jil 4.31 12.36 92 0.99 51 5.13 13/1992 0.0 0.0 49 10.8 3.0 1.1 Au 3.72 8.63 79 0.58 51 3.17 5/1995 0.0 0.0 49 9.4 2.3 0.8 Se 2.96 6.76 79 0.51 63 2.66 14/1979 0.0 0.0 49 8.5 1.9 0.5 Oc 2.15 5.24 54 0.11 63 1.69 3/1986 0.1 4.6 93 9.0 1.1 0.3 No 3.22 10.67 85 0.60 76 2.38 10/1985 1.9 15.2 50 11.6 2.0 0.5 De 2.86 6.99 90 0.46 55 1.74 8/1978 5.3 17.3 60 13.2 1.6 0.3 An 38.09 53.18 90 24.51 63 5.13 7/13/92 29.0 47.5 78 138.8 24.6 6.7 Wi 7.28 14.39 50 3.52 77 4.79 1/21/59 21.9 46.4 78 38.5 4.6 0.8 Sp 10.41 17.91 64 5.02 76 3.40 3/ 9/64 5.6 18.5. 87 39.5 6.6 1.6 Su 12,07 22.02 58 6.00 51 5.13 7/13/92 0.0 0.0 49 30.9 8.2 2.9 Fa 8.33 13.78 85 1.42 63 2.66 9/14/79 2.0 15.2 50 29.6 5.2 1.3

Midwestern Climate Center, Champaign IL

21 C)(N C,

Table 2.4: other Meteorological Data 633666I0S. 016 &HHUAL.

664100 @6aCOS 60, 66366643 INTL. so 29233 Gas,.,39630*1:1 706fl TABLE 10. CEILING. VISIBILITY. AND WEATHER BY WIND DIRECTION (PERCENT FREQUENCY OF OBSERVATIONS) a]uLIND 10,95 VI90310UT7 56653625 WEATHER, WIDdo oM low mm 301 m OVRo, 104 to2 3 o TO 3 3 ow T To To To T To 710 mm lo To T. OC 0 SW o o dew~ mm WIG5 31 4 2313 6 s 101 6 .3 .7 . .1 .6 .3 .7 I.4, .6 .0 .1 .6 .3 .6 3.7 .3I 16 . 6 . 1 .A .2 .3 . .3 .6 .8 .3 .7 2. .3 .6 .31 . 7 .

62 6 . 3 . .2 . 6 . .7 a0 .6 .6 .1 . . 7 . 1 . 0 34 .

SE 6 . 3 .4 .8 . 3 . .3 0. . . . . A.? 3. .9 .6 . 1.8 .6 a. .

6 . 3 .I .4 .0 . . . 67 . 6 3.023 7. 2.3 6 . . 3 32 .

S 6 . 3 . 6 .

.6 8 . . .6 .0 2.3 4. .3 .4 .02 .6 .I 3.1 .34 0S . .0 .0 . 3. 3. 3. .4 26 . .6 . 2. 49 .6 . 0. .6 . 3. .2 .

.J.3 .2 .3

3

.0.4 37 .0 . .2 .6.2 0 .3.6 .4 .4 .3 CALM . .6 . .6 .6 .6 . 2 . .3 2 . 6 . .9 Z.1 .2 .6 8 .6 .

Toy .3 '.I .9 3.7 4.3 2.3 9.2 &1.3 9.7 63.2I .3 .3 .6 6. 8 1.6 69.9 9.0 .3 4.1 11.6 .2 1 I . .1 . .6 5P* 013PILLITI006PAC622L66T 430 NAL3m§.140 TABLE IL. WINO DIRECTION VS. WINOSPEED (PERCENT FREQUENCY OF OBSERVATION5)

A. ALL WEATHER L IFR

.1 ".00,00Vet0 0660T7 00600 -6D 06907 GIN 0060 63 46 1.56 50.0617.21v.27V 4

3. 0.8 "Y TOT f AIN of .7 . 12.9 .4 .6 9.4 7.3 6 . .3.6.3 .3 6.*

6 .6 3.6 S.0 .4 .N. .

ISE .6 :'. . .

as .3 2:.3 2. I 9 . 6 . .

I4 . .8 S324 6 .83 2 . ..4 L 3 6 .3 7.

NW . 1. 31:26 .. 293.3 5.3 I

,,SW .1 .43. .3 .3 3 .63. 13.4 lwo . :,L .2 .2 . .6.13 0.

"NW .33. .6 3. .2 6 .? 6.3 CAL 3.2 1.3 ToT 6.6 23.9 24. 23.7 3.0 .3 .3 20.

lov6 6.21 TOT .61 3.4 3.0 3.3 .3 .6 .6 to.. 6.3 ALL2 WEATHER: ALL2.W050 O62164VATIONS 0W6. 6302.36CC534 F? 0600506VISIBILIT Cl36ANSLT Z 34 FT MO0 Z 113W.

TABLE 12 WEATHER CONDITION BYHOUR 9MEAN NO. OF DAYS) 0003333379 3M*T6I6 COND.700S2 asI6 I S I7to I is I I ' I kg9 3l 6*59*3M ORIN ONI3Z U 32.3 I 3.7 3 . 3 . 33.7 3 6. 3. 6 SNOW10 103051 ICE FILLET 36.3 3$.6 31.3 30.7 33.7 37.4 13.9 30.5 PRECIPITATION1 44.7 s0.. 10.3 26.0 49.0 32.4 46.3 40.

FOG 3 6 6 3.6 363.2 3.17 35.3 3.9l 3I. 3:2, .

sw066*AND/OR06436 39.4 .6.6 336.6 39 .91 I 2.7 011.6 3 6.4 *21i CALMI 32.2 it.. 31.21 4.3 1.a 8 33 3.

- - 9. 37 .9 .6 .8 2.3 L . . 37 a _ 3 2 3 .1 .4 .3 .

6- 663. .64316 .2.

3 8- AS Ua . 40 .4 . . S.3 2 .

6 2 a. 66 is6 . 7 . 3 4. 6 4 9. 66. 6 4. 0 6

OVERa04.0 302'.6 034.6 33 .6 6 :.2 i91s T . 00.

ZER6O03.0O6n.6L.3 2.. 3. .3 .2 .4 .2 66 4 06 0 69 6 6. 63 o7.7 40.1 0 0.3 i 3.

137 .7 1 43.61 3 2 7.9 3 6 3.21 "I, .7 9 3. 2 9. 2.

0 0 U7 1.4 6 6 6 1 6 3 3 6.2 3 38. 6 I 2. 9, 3 43. 3 l 6. ,

3- 0 .4 T.4 .6 WAR W60 4163.316 ARE 6055006 TO 66*61ST T10676.IOT NOT JUDAISM TO 6A667906063PAU EXACTLY 6033*3TO 093U3580OAVOWTOTAL&

75563 VAOWOSA*1066 3(0 0003-3 0UOLY 0620 V4r730602 22

adjacent to the city. The Scioto River, located to the west of the city and of the reactor site, is the principal stream and flows from the northwest into the center of the city and then flows straight south toward the Ohio River. The Olentangy River, which is located to the east of the reactor site, runs almost due south and empties into the Scioto just west of the business district. Two minor streams, Alum Creek and Big Walnut Creek, run through portions of Columbus and skirt the eastern and southern fringes of the area. Alum Creek empties into the Big Walnut southeast of the city and the Big Walnut empties into the Scioto a few miles downstream. The Scioto and Olentangy feature gorge-like formations with very little flood plain and the two creeks have only a little more flood plain or bottomland.

2.5.2 Ground Water Infiltration from the reactor site enters the groundwater table which is approximately 45 to 50 feet below the surface. The groundwater flows toward the Olentangy River which is 1.1 miles to the east. The Olentangy River joins the Scioto River at a point about 2.51miles south of the site, and flows past the city of Columbus in a- southern direction. The Dublin Road Water Treatment Plant for the city of Columbus is located on the Scioto River 2.2 miles south by southeast of the reactor site. This location is upstream of the confluence of the Scioto and Olentangy Rivers and uses water collected from the Scioto River basin. Thus, contamination of the city of Columbus' water supply by infiltration of the reactor pool water is virtually impossible. The next major town using water from the Scioto River is Circleville, located 30 miles south of Columbus.

2.6 Reactor Building Description 2.6.1 Reactor Building Design and Construction The reactor building is a steel framed structure with insulated metal wall panels and built up roof. The ground floor is a concrete slab on grade. The second floor slabs are concrete supported on steel beams.

Elevated platforms are checkered plate supported on steel beams.

Interior partitions are plasterboard. Floor drains provide drainage.

2.6.2 Layout The overall exterior ground floor dimensions of the reactor building are 62 feet by 48 feet. The reactor building is divided into three 48 foot sections, extending from west to east (see Figure 2.12). The western section extends 16 feet into the building. It is a single level section with an internal floor to ceiling elevation of 11 feet.

The center section, known as the bay area, houses the pools, reactor, and reactor facilities. It stretches 29 feet 3 inches from the west section with a ceiling height of 35 feet. The eastern section of the building is a two level section with an overall floor to ceiling height of 22 feet. On both the first and second levels there is a walkway extending a distance of.3 feet 9 inches east from the bay area 23

I 5,,__

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=

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^

- _ L

- -= :_ -- - -: :-

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_1 , 9'3" I-Figure 2.12: Dimensions of the Reactor Building Shown in a Cross-Sectional View 24

to an interior wall. From here, the building continues a distance of 12 feet to the eastern exterior wall of the Building.

2.6.3 General Features The reactor building area is serviced electrically by the Columbus and Southern Ohio Electric Company via 13,800 volt-ampere lines and has an electric failure probability of 0.2 failures per year. The in-house electric service is 120/240 volts, three wire, single phase; 240 volts, three phase; and 120/208 volts, four wire, solid neutral for lighting and power.

Water is supplied by the city of Columbus via piping distribution from its treatment and pumping plant located 2.2 miles south by southeast of the reactor site.

Natural gas service to the reactor building is provided by the Columbia Gas System of Ohio. The building also contains service air at a pressure of about 70 pounds per square inch. A forced warm air system, using gas-fired furnaces, heats the building. The building air conditioning system is tied into the ductwork for the furnace system heating the east side of the building. Window air conditioner units are used in certain rooms for additional cooling capacity (e.g.,

control room).

2.7 Chapter 2 Bibliography Section 2.2.1: Dan Redman, Ohio Emergency Management Agency, Personal Communication, 1999.

Section 2.2.2 Dr. H. C. Noltimier, Department of Geology &

Mineralogy, The Ohio State University, Personal Communication, 1999.

Section 2.3.1: C. R. Stauffer et al, 1911. The Geology of the Columbus Quadrangle, Ohio Geological Survey Bulletin 14, 133 pp.

J. J. Schmidt, 1958. The Groundwater Resources of Franklin County.

Ohio Division of Water, Bulletin No. 30, pp. 5, 18-21, 51.

Section 2.3.2: R. Melvin and G. D. McKenzie, 1992. Guide to the Building Stones of Downtown Columbus: A Walking Tour, Guidebook No. 6, Geol. Soc. America Annual Meeting, Cincinnati, OH, 33 pp.

Dr. H. C. Noltimier, Department of Geology &

Mineralogy, The Ohio State University, Personal Communication, 1999.

25

[ pi Section 2.3.3: ibid., Dr. H. C. Noltimier M. C. Hansen, 1999. "Earthquakes in Ohio". Educational Leaflet No. 9, ODNR, Division of Geological Survey.

Section 2.5.1: Dr. Robert Stiefel, Water Resources Center, Personal Communication, 1987 26

3.0 Reactor Facility Description 3.1 Reactor Core 3.1.1 Core Structures 3.1.1.1 Grid Plate and Support The Reactor Core Grid Plate is located at the west end of the reactor pool. The grid plate, shown in Figure 3.1, is a 15.66 inch by 18.70 inch by 5.03 inch A356-T6 aluminum alloy plate, cast with a 5 by 6 array of 2.48 inch diameter through holes. The grid plate holds the core elements in place by fitting the lower end boxes of the elements into the through holes. A 0.25 inch diameter dowel pin fixes the rotational position about the vertical axis of each element. The grid plate is positioned in the pool so that the east and west sides have five holes, and the north and south have six.

The grid plate is attached to a frame by eight 0.5 inch hex head screws. The grid plate frame, shown in Figure 3.2, is made of aluminum extrusion having four feet resting on pedestals mounted to the reactor pool floor with four threaded studs that penetrate the pedestals and frame feet. Leveling shims are located between the frame feet and pedestals. The grid plate and frame assemblies are positioned so that the distance from the reactor pool floor and the center of the core is 3 feet 6 inches. The.core is centered between the north and south walls, which are 1 foot 10.25 inches from the center of the core. The j inner sides of the east and west walls are a distance of 8 feet 9 inches and 1 foot 10 inches respectively, from the center of the core.

3.1.1.2 Fuel The OSURR is fueled with uranium enriched to 19.5%, in the form of uranium-silicide (U3 Si2) and aluminum. This fuel type was developed as part of the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program of the Argonne National Laboratory (ANL). It is the fuel -type of choice by the Department of Energy, in consultation with the National Organization of Test, Research, and Training Reactor Operators (TRTR), in which the OSURR staff is active.

The fuel is supplied by the Babcock & Wilcox Company under the terms of the DOE Fuel Assistance Program, managed by EG&G Idaho, Inc.

The uranium-silicide and aluminum fuel exhibits superior properties under a variety of conditions, including high burnup and elevated temperatures. The fuel shows a strong resistance to mechanical degradation and breakaway swelling for burnup densities considerably in excess of the burnup expected in the OSURR fuel cycle length. It is expected that the fuel lifetime in the OSURR core will be reactivity limited, rather than burnup limited. Blister tests conducted in the RERTR program on high burnup fuel have provided data which can be used to show that no fission product releases will occur from the fuel in 27

Figure 3.1: OSURR Grid Plate 28 C.' C' (C

A; Figure 3.2: OSURR Grid Plate Frame Note: Startup source is at thet_]

29

the OSURR core even under maximum credible accident conditions that do not involve direct mechanical damage to the fuel plates. These accident scenarios are discussed in more detail in Chapter 8 of this report.

The active portion of the fuel, commonly referred to as the "meat", is contained within flat aluminum plates 0.050 inches thick. The meat thickness is 0.020 inches. Fuel plates are joined to aluminum side plates to form either standard",partial, or control rod fuel elements (also called fuel assemblies). Each fuel assembly contains both fueled and unfueled (or "dummy") plates, the unfueled plates being made of pure aluminum with no U3Si2 content, but otherwise identical to fueled plates. Each fuel assembly has an upper and lower end box. The upper end box consists of a bracket and handle assembly to allow handling of the fuel element with a special handling tool. The lower end box is formed from a tapered aluminum cylinder which inserts into the sockets in the grid plate. A positioning dowel pin is located near the top of the tapered cylindricaltsection. This pin inserts into a corresponding hole in the grid plate to allow reproducible positioning of the fuel assemblies on the grid plate.

Standard fuel assemblies have a total of 18 plates per fuel element.

Of these 18 plates, 16 contain uranium fuel, and two are unfueled, pure aluminum, or dummy plates. The dummy plates are the two outer plates of each fuel element. Use of two dummy fuel plates allows a lower loading of uranium in each standard fuel assembly, permitting the core of the OSURR to be physically larger than it would be if all plates of a standard fuel element contained uranium fuel. If dummy plates were niot used, either a smaller core would result for a given excess reactivity, or a larger core could be loaded with a higher excess reactivity. In either case, undesirable operational and/or safety consequences would result, given h design and operational characteristics of the OSURR. A total of; % tandard fuel elements are available for the OSURR. Given the reactivity lii OSURR, a cold, clean critical core sizewaspredicted using t-ndard fuel assemblies, along with 4 control rod fuel elements. Details of a standard fuel assembly are shown in Figure 3.3. Chapter 4 provides information on predicted core size.

The control rod fuel elements are similar to the standard elements in that they utilize the standard fuel plates and a similar side plate.

Several of the inner fuel plates have been completely removed to allow room for the control rod to pass through the element. On either side of this gap for the control rod are pure aluminum guide plates. The outer two fuel plates in each control rod fuel assembly contain uranium fuel. The upper end box of the control rod fuel elements have a modified grappling section to allow handling with the special handling tool used for manipulation of control rod fuel elements. An aluminum guide plate for the control rod is also mounted at the top of the assembly. Figure 3.4 shows a control rod fuel element in more detail.

30

.. .7.62 cm Aluminum Plate 0.48 cm

_~~

I-=] -

0 H2O~ Alj

. -, Al 3 2 I USi Aluminum Plate I Figure 3.3: OSURR Standard Fuel Element 31

r z

,T

- U Al~i Gul-

-~C Pae l

-L U

I0 Q.0.,

mr U

I L-

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-4 5j Figure 3.4: OSURR Control Rod Fuel Element 32

The partial fuel elements are physically identical to standard elements with the exception of their uranium loading. They utilize standard fuel plates, but some of the fuel plates have been removed and replaced with pure aluminum dummy plates. Partial elements are available with 25, 37.5, 50, and 62.5 percent of the nominal uranium loading of a standard element. Use of partial fuel elements allows precise adjustment of the excess reactivity of the OSURR core.

All standard, partial, and control rod fuel assemblies are stamped on both side plates with an identifying alphanumeric symbol unique to the assembly. This identifier, formed of 2-inch high letters and numbers, can be read underwater at a depth of at least 3 feet. These identifiers are used in verifying core loading and in material accounting procedures.

When loaded in the core, fuel elements are positioned on the grid plate. When not in use in the core, or when being stored as spare or spent fuel assemblies, fuel elements are kept in the fuel storage pit at the east end of the reactor pool. A single fuel assembly may be in transit at any one time between the core grid plate and the fuel storage pit. Exceptions to these storage procedures require approval by the Reactor Operations Committee and the Nuclear Regulatory Commission.

3.1.1.3 Core Arrangements Up to 30 positions on the core grid plate are available for use as fuel element positions. Control rod fuel elements occupy 4 of these positions, and one is reserved for the Central Irradiation Facility (CIF) flux trap (water or graphite-filled). Also, a total of 5 graphite isotope irradiation elements (GIIE) are available for positioning on the grid plate, although not all GIIE assemblies need be mounted. Specially fabricated hollow, water-filled, plug-type elements of pure aluminum will also be made available for use in the grid plate. Allowing for the 4 control rod fuel elements and one CTF flux trap position, a total of 25 positions are available for use as fuel or GIIE assembly positions. More information on both the GIIEs and CIF is given in the following section concerning experimental facilities.

Maintaining a uniform, well-behaved, and predictable flow of coolant through the core is achieved by assuring that all grid plate positions are occupied by some type of assembly. In this way, no holes are left in the grid plate which could result in anomalous flow patterns through the core. The type of assembly occupying a given grid plate position will depend in part on experimental requirements, fuel burnup, and reactivity limitations. In general,, however, as symmetric a core as achievable, within these limits, is desirable. Control rod elements should be surrounded with fuel, in as much as possible given reactivity and burnup constraints, so as not to reduce the effective worth of the control rod. A regular pattern of control rod element 33

positions relative to the overall core geometry is desirable to W maintain predictable control rod worth.

A cold, clean critical core size is predicted using 'tandard fuel elements, 4 control rod fuel elements, and a sing e, central position for the CIF. This leaves 2 or 3 positions to be occupied by other kinds of assemblies, either GIIE elements or plugs. Thermal hydraulic and reactivity requirements and/or limitations will determine which will be used and their locations on the grid plate.

3.1.1.4 Control Rod Poison Sections Those sections of the control rods which actually effect control of the neutron population in the core, known as poison sections, penetrate into the core boundary through gaps in the control rod fuel elements. A total of 4 control rod poison sections penetrate the active boundary of the core. Three of the rods are shim safety rods and one-is-vrsregulating rod. The poison sections of the shim safety rods are 26 inches long, while that of the regulating rod is 24 inches long. They travel through a maximum stroke of 61 centimeters (24 inches). More detail on the control rod system is presented in section 3.3.8 of this chapter.

3.1.2 Associated Core Structures 3.1.2.1 Core Reflection The core of the OSURR is reflected by reactor-grade graphite and

- high-purity water. Solid graphite reflectors, each encased in its own aluminum shell, are mounted to the reactor pool walls and reflect the west and south sides of the core. The GIIE assemblies, if used, provide reflection at the locations where they are placed on the grid plate. A detailed description of the solid graphite reflectors, which also form the extensions for experimental thermal column facilities, is presented in Section 3.1.3.1, while the GIIE assemblies are described in Section 3.8.6 of this chapter.

Where graphite is not present, most reflection is achieved by the high-purity, light water filling the reactor pool. Water provides the primary neutron reflection effects at the top and bottom and along the north face of the core. Some of the reflecting power of the water at these positions is lost by the presence of other structures and facilities, such as the grid plate below the core, and the experimental facilities located along the north face of the core (beam ports, rabbit).

3.1.2.2 Startup Source and Source Drive A plutonium-beryllium (PuBe) source provides an initial source of neutrons to the OSURR core controlled startup of the reactor. The initial source activity was Llcuries, which provides a neutron emission rate of 8.95x106 neutrons/second. A minimum of 106 34

'N Al e-ob l1 C60PLe rt-IN- ,C-I'f

'/4 J41 PCOD C,4DMA Figure 3.5: OSURR Source Drive Assembly 35

neutrons/second is required. The source activity has slowly increased as a result of buildup of alpha-emitting decay products of plutonium.

Total source activity at the end of 1986 was such that the neutron emission rate was slightly above lxlO neutronIZ4 dThensource is

.1a2 *&

n a fixed housing assembly, located

} lightly offset from the vertical center e core.

e ot rm of the housing is mounted to a positioning bracket attached to the grid plate pedestal, while the top of the drive housing mounts to brackets attached to the pool walls. The lower section of the source housing is wrapped with a cadmium sheet of sufficient thickness to be "black" to thermal neutrons.

Within the source housing is a movable platform on which the source rests. The position of this platform can be raised and lowered as selected from a switch in the control room. The source drives through a maximum stroke of 95 centimeters. The source is normally in its raised position for reactor startup, in which position the neutron emitted by the source are available to induce fissions in the fuel.'

When raised, the startup source is-located between the planes defined by the grid plate and the tops of the fuel elements. After the reactor has achieved a critical or supercritical condition, the source is typically lowered into its shielded position, wherein neither the core nor the startup source is exposed to thermal neutrons from each other.

Source shielding consists of cadmium sheet wrapped around the lower section of the source housing. When the reactor is shutdown, the source may be in either its raised or lowered position.

The electromechanical source drive system is operated from the control console. The electric drive motor controller and position encoder is located on the south edge of the top of the reactor pool wall. The source drive motor is mounted on the beam supporting the source drive housing tube, extending outward from the edge of the pool wall, above the surface of the pool. The shroud forming the source drive housing tube, which runs down the length of the reactor pool to the vicinity of the core, is made of aluminum. More detail on the operation of the ____

source drive system, including its safety system interlock features, is contained in a following section on reactor instrumentation and control.

3.1.2.3 Control Rod Housings The control rods move within aluminum housings. These housings join to the tops of the four control rod fuel elements. The control rod poison sections move upward into these housings above the core when being withdrawn from the core. The aluminum shrouds forming the rod housings are made from 2.75 inch outer-diameter tubing, with 0.125 inch wall thickness. The housings are supported at the top of the reactor pool with brackets attached to the walls of the pool. Holes are cut in the lower sections of- the aluminum housings to allow coolant flow and water ejection during sudden rod insertion movements (reactor SCRAM, or rod drop experiments). More information on the control rod system, 36

including information on the rod housings and associated apparatus, is contained in Section 3.3.8 of this chapter.

3.1.3 Reactor and Shielding Pools 3.1.3.1 Reactor Pool The reactor pool is built on a 15 inch thick regular concrete foundation. The pool has inner dimensions of 3 feet .8.5 inches by 10 feet 7 inches by 20 feet deep, with a capacity of 5800 gallons of water. The walls are constructed of barytes concrete to an elevation of 13 feet 6 inches above the foundation. The remaining 6 feet 6 inches of the walls are made with regular concrete. A scupper is located along the top edge of the east wall. Figure 3.6 shows the layout and dimensions for the reactor pool with scupper. The pool is completely lined with a layer of fiberglass-reinforced epoxy paint coating to prevent leakage, leaching of the materials in the concrete by the pool water, and to facilitate decontamination and repair of the walls.

Constructed at the east end of the reactor pool is the Fuel Storage Pit (refer to Figure 3.7). The sides of the pit are made with 7 inch thick regular concrete, the lid section of which is coated with a 0.25 inch layer of the same sealant used for the reactor pool. The dimensions of the lid section are 16.75 inches by 3 feet 5.75 inches with a depth of 8 inches. The pit dimensions are 12.75 inches by 3 feet 3 inches with a depth of 4 feet 1.25 inches extending from the bottom of the lid section. There are two Fuel Storage Pit Plugs that cover the pit. These plugs are identical except for the positions of the eyebolts used to lift them out of the well. The plugs are made of type 3003-H112 0.25 inch aluminum welded at the joints, filled with lead brick. The dimensions for the plugs are shown in Figure 3.8.

A Fuel Element Storage Rack, capable of storing a complete core loading in a 10 by 3 array exists n t shows detail on the storage rack. The rack is made with 0.25 inch 3003-HI12 aluminum sides and 0.25 inch boral spacers that span the width of the rack and act as a safeguard against inadvertent criticality. The effective multiplication of the fully loaded Fuel Storage Pit, when filled with water, is 0.68 or less. The floor is a 0.75 inch 1100 aluminum plate, which rests on four 0.5 inch 1100 aluminum bars that are welded to the sides of the rack. The bottom side of the rack floor is elevated 9 inches above the Fuel Storage Pit floor. The plate forming the rack floor has 0.25 inch deep slots into which the boral spacers fit, as well as tapered holes at each fuel element position. The tapered holes act as sockets into which insert the lower end boxes of the fuel elements. The tops of the boral spacers are supported by two 0.25 inch by 0.75 inch 3003-H112 aluminum bars that extend lengthwise across the top of the rack. These bars are welded to the two sides of the rack and have 0.25 inch-width slots for the boral spacers. The secondary purpose of the bars is to hold the top ends of the stored fuel elements in place.

37

C: 1-' 47 Figure 3.6: Ovi rall Reactor Dimensions 38

5ANDELAST & PAINT VWITH ONE COAT OF ZINC CHROMATE INTEIO2 OF WELL & rOP FLG.

OF ANGLtE A ,FUEL ELEMEWT S;TORAGE WELL PLUG A I I

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Figure 3.7: Fuel Storage Pit 39

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_1 I u Figure 3.9: Fuel Storage Pit Rack 41

The reactor core is surrounded on the south and west sides by Thermal Column Extensions. These extensions are made of 0.25 inch 3003-H112 aluminum plates welded at the joints and filled with reactor-grade 0

graphite. They mount to their respective pool walls via slotted plates that are welded to the two sides of the extensions and fit over studs in the walls. Figure 3.10 shows more detail on these extensions.

Behind both of the thermal column extensions are 0.125 inch thick stainless steel plates which lead to the Thermal Columns. The Thermal Columns are graphite filled cavities in the reactor pool walls. These will be discussed further in Section 3.8. In addition to the Thermal Columns, other penetrations of the pool walls include the beam ports and rabbit facility, all of which are located on the north side of the reactor pool. These will be discussed in more detail in Section 3.8.

Two movable, 400 Watt pool lights are suspended in the reactor pool.

These lights can be positioned so that they illuminate any desired portion of the pool. The power ON/OFF switch for the lights is located on the control panel in the control room.

3.1.3.2 Bulk Shielding Facility Pool To the south of the reactor pool lies the Bulk Shielding Facility (BSF) Pool, which is built on the same 15 inch thick Portland concrete foundation that the reactor pool is built on. The inner dimensions of the BSF pool are 8 feet by 12 feet by 15 feet 8 inches deep. It holds 11200 gallons of water. The walls of the pool are made of regular concrete with the exception of the reactor pool side, which is made of barytes concrete from ground level to elevation 13 feet 6 inches. The BSF pool is completely lined with a layer of the same fiberglass reinforced epoxy based paint used on the reactor pool.

As of the end of 1999, the BSF pool contained a 60Co source, licensed separately from the OSURR, and (currently) unused equipment storage racks.

42

7Su R P THERMAL COL.

I B. _.F 10 _ I I.

All REACTOR

'9 L'to

_ 7=

-I II PI THERMAL COLUMI'

-1 to TIIE mezO/JTIPJG LA.

CLrE LOCATION FOR e THE MAWN THLEMAL COLVIMU exTrEPSICJ is iei

-arLD5 M4ooAerTidV We EchoesA ftL*TE LAC.47710J

-4 FCT1Z4 I5SF71IGV i*f S j-* S.S. CINCH ANCHOI FURNISHED BY POOL CONTRACTOR 'I A COATING L -

-CiNC-H ANCHOR TO FDN (INSTALL ANCHOR BEFORE COATING)

Figure 3.10: Thermal Column Extensions 43

44 3.1.4 In-Pool Instrumentation 3.1.4.1 Startup Channel and Drive System A fission chamber used in the Startup Channel is located in the reactor pool above the northwest corner of the reactor core. A protective guide tube is mounted to the ionization chamber bracket, which in turn is mounted to the west wall of the reactor pool. The tube has a guide running down the inside of the tube which fits into a slot at the top of the fission chamber canister. This allows the canister to move up and down within the guide tube without rotating.

The Fission Chamber Drive Motor is attached to the top of the west reactor pool wall and uses a teleflex cable to position the fission chamber. Section 3.3.14 has a more detailed discussion of the Startup Channel.

3.1.4.2 Linear Power Monitoring Channel Detector The Linear Power Monitoring Channel Detector is a compensated ion chamber (CIC) that measures the neutron flux in the reactor over the full operating range from startup to full power. The chamber 45

46 compensates for gamma interaction to produce a signal that is proportional to neutron interaction. It is located above the west thermal column extension, and is supported laterally by a rack mounted to the pool wall above the extension. The chamber is supported vertically by a holder mounted to the top of the reactor pool wall.

The cable pipe that extends from the detector canister to the top of the pool fits through the holder and is adjusted so that the detector sits in the desired vertical position. A clamp is attached to the pipe to keep it from sliding through the holder. Figure 3.13 shows the location of the Linear Power Monitoring Channel Detector more precisely. A more detailed discussion is found in Section 3.3.13.

3.1.4.3 Logarithmic Power Monitoring Channel Detector The Logarithmic Power Monitoring Channel Detector is a CIC identical to the Linear Level Power Monitoring Detector described above, except that the instrumentation is such that the output is in logarithmic rather than linear form. The detector is located next to the Linear Power Monitoring Channel Detector, and is held in place in the same manner. Figure 3.13 shows the precise location. Section 3.3.12 contains a more detailed description of the Logarithmic Power Monitoring Detector.

3.1.4.4 Power Level Safety Channel Detectors There are two uncompensated ion chambers (UICs), known as the Power Level Safety Channel Detectors, that are located next to the compensated ion chambers as shown on Figure 3.13. The uncompensated ion chambers do not compensate for gamma interactions. Therefore the signals produced by the UICs have components resulting from both gamma and neutron interaction. The UIC detectors are held in place by the same system that positions the CICs. Further discussion of the Power Level Safety Channel Detectors is located in Section 3.3.16.

3.1.4.5 Temperature Sensors and Locations Temperature of the pool water at, various locations throughout the volume of the reactor pool is measured by thermocouples. These thermocouples indicate water temperatures at the inlet (bottom) and outlet (top) of the core. Additionally, thermocouples monitor the temperature at the inlet to and outlet from the cooling system. A thermocouple is also used to monitor the temperature of the secondary coolant. Digital panel meters provide temperature information display.

3.1.4.6 Water Level Sensors and Locations The water level sensing float switch is used with the reactor safety system. This sensor initiates a reactor trip if the water level in the reactor pool should fall to the lip of the scupper at the east end of the pool. This provides assurance that the requirements 47

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£ 04 Y7rj~COvewrr ZXAFN/r and Facilities In and Around the Figure 3.13: Locations of Instruments OSURR Core 48 QSjV

for minimum water depth in the reactor pool, as specified in the OSURR Technical Specifications, are met.

3.2 Reactor Cooling and Water Processing Systems 3.2.1 General Features Light water in the reactor pool serves as the primary coolant. This water passes through both a heat removal system and a water processing system. These systems are independently controlled and operated. The heat removal system serves to remove reactor-generated heat from the pool water and transfer it to a secondary heat sink (either the outside atmosphere or city water). A schematic diagram of the cooling system is shown in Figure 3.14. The water processing system removes impurities from the pool water to meet required limits on water purity. Both the cooling system and water processing system are designed to meet their required operational parameters with an additional margin of safety.

3.2.2 Cooling System 3.2.2.1 Primary Coolant Loop Natural convective cooling is the primary means of heat removal from the OSURR core. Water enters the core at the bottom, flows upward between the fuel plates through the flow channels in the fuel elements, and is heated by the warm surfaces of the fuel plates. The heated water rises as result of buoyancy forces, and enters a plenum at the top of the core. This plenum, made of 6061-T6 aluminum plate, is essentially a box surrounding the core, with a sloped top (or cap) above the core, in the shape of a truncated pyramid. The sides of the plenum surrounding the core serve to limit bypass coolant flow between the fuel assemblies, while the plenum cap helps to confine the heated water (which also contains some 1(N isotope) to the area immediately above the core, from which it is withdrawn by the suction from the cooling system. A hole at the top of the plenum cap allows natural convection flow of the water in the event that the cooling system pump is turned off or fails. The size of the plenum cap is such that natural convective flow through the core is not significantly disturbed. The plenum can be manipulated from the top of the reactor pool. It is mounted in a way that helps preclude the possibility of flow blockage due to plenum misalignment, and the chance of damaging the tops of the fuel assemblies by mechanical impact.

Water is withdrawn from the plenum cap by the cooling system.

Withdrawing water near the top of the core allows introduction of relatively warm water into the cooling system heat exchanger, and subsequently maximizes the temperature difference between the primary and secondary coolant. Flow rate-is adjusted so that the natural convection cooling flow through the core is not significantly disturbed by the suction from the cooling system. Suction is taken 49

I REACTOR BUILDING INTERIOR

.r Outlet Primary Coolant Loop t floI. news { e-.sZCAir Intake l- I ACity Water Inlet Figure 3.14.: Schematic Diagram of the OSURR Cooling System 50 C C C-

from two sides of the plenum to help equalize the water temperature distribution above the top of the core.

Primary coolant then passes into a decay, or holdup tank, vertically mounted in the reactor pool, near the southeast corner. The delay time for the water in this tank is equal to about 11 half-lives of 26N, which significantly reduces 16N concentration in the primary coolant as it passes into the downstream components of the primary coolant loop.

The decay tank is made of type 304 stainless steel, and has an overall length of 20 feet 11 inches. It has an outer diameter of 10.75 inches, and a wall thickness of 0.109 inches. The tank has a baseplate made of type 304 stainless steel which allows it to rest on the bottom of the reactor pool. Piping runs from the bottom of the decay tank to the core plenum, and out of the top of the decay tank. All piping and decay tank surfaces exposed to the bulk pool water are insulated to prevent heat loss into the bulk pool water. Pipe runs do not penetrate the reactor pool wall, so its shielding capabilities are not compromised. Where possible, cooling system piping runs parallel with existing water processing system piping.

Outlet from the decay tank passes to the primary coolant pump. The primary coolant pump is a centrifugal pump, with the pump made of PVDF. A siphon breaker protects against accidental siphoning of the pool water through the inlet leg. The primary pump does not have to re-prime on each startup, since it is self-priming as long as the pump head is immersed, which prevents air from being introduced into the primary heat exchanger (which could reduce its effectiveness). The pump is protected from running dry in that if the water level drops low enough to expose the pump head, the reactor will have already tripped, blocking operation of the primary pump (if the reactor is tripped, the cooling system need not be operating, since the thermal capacity of the reactor pool is sufficient to remove core decay heat).

The primary coolant pump is driven by a 3 HP electric motor. A modulating valve is available in the primary coolant leg to adjust flow rates. This feature allows adjustment of the primary coolant flow rate to match that required to maintain natural convective flow through the core. The primary coolant pump provides the driving head for coolant flow through the entire primary coolant loop.

Water is pumped into the primary heat exchanger from the primary coolant pump. This heat exchanger removes heat contained in the primary coolant (pool water) to a secondary cooling fluid, which is a mixture of water and ethylene glycol. The heat exchanger is a plate-type design (sometimes called a plate-and-frame type), with stainless steel suracsw whrcotct with the primary~ coolant occursAW The heat exchanger outlet passes through a flow sensing device to coolant flow conditions. A siphon breaker is added to the of the primary coolant loop. This siphon breaker is a small o an revents a siphon problem in the return leg in the event of a 51

significant leak in the primary coolant loop. Since the volume flow rate of water through the siphon breaker tube will be small during operation, it will not significantly disturb the overall flow characteristics of the primary loop.

The cooled primary water is returned to the pool at a point above the core outlet plenum. The return flow is directed in a manner that provides dispersion of any core outlet flow that was not drawn off by the primary coolant loop suction, while not causing a disturbance of the convective flow pattern through the reactor core. This dispersion technique commonly used in TRIGA-type reactor designs, prevents buildup of 16N-rich water at or near the surface of the reactor pool.

3.2.2.2 Secondary Coolant Loop The secondary coolant loop removes heat transferred from the primary coolant to the secondary coolant. This coolant, a mixture of ethylene glycol and water, passes through two separate heat exchangers, which removes heat in the secondary coolant to an ultimate heat sink (the outside atmosphere, or, under certain environmental conditions, a supply of city water).

The secondary coolant mixture is prepared in a way that minimizes the chance of buildup of activation products in the pool water should a leak occur from the secondary loop into the primary coolant.

Commercially available ethylene glycol solutions generally contain corrosion inhibitors that may contain activatable elements. Use of a coolant mixture making use of buffers such as lithium tetraborate helps reduce activation product production. Borate compounds assure that any possible secondary coolant leakage into the primary coolant results in a negative reactivity insertion.

Primary to secondary coolant heat exchange occurs in the primary heat exchanger. This unit was-desertibed in-the-pierious section. Secondary coolant outlet from the primary heat exchanger then flows into a bypass control valve. This valve is an electrically actuated three-way device and provides a means to control cooling system heat removal capacity. A valve position signal is generated from a secondary coolant temperature sensor located at the inlet to the secondary cooling loop pump. The bypass control valve has 2 inch diameter inlet-outlet dimensions, and has an ANSI class VI leakage rating.

Outlet from the primary heat exchanger can be directed into either the outdoor, fan-forced cooling unit or the bypass leg. The amount of fluid flowing through each of these legs is a function of the hydraulic resistance of the fan-forced cooling unit and the position of the bypass control valve. CoQlant directed through the outside cooling unit is cooled by fan-forced air flowing through the device,i while that flowing through the bypass leg is not cooled. This arrangement allows for adjustment of total system heat removal 52

capacity in the event of low outside air temperatures, and it limits the possibility of overcooling the primary coolant.

Fluid flow directed through the bypass leg is monitored by a remote-indicating flow meter. Flow indications are used to provide information on total secondary coolant flow rates through both the bypass and outside cooling unit flow paths.

The outside cooling unit is a fan-forced drycooler, sized to provide a total heat rejection capacity sufficient to remove all heat generated by 500 kilowatt operation of the core, if the outside air temperature is 78 'F or less. A total of eight electrically operated fans provide air flow through the unit to cool the secondary coolant. These fans, banked in groups of four, are remotely actuated by two separate (one for each bank) 24 volt starting relays. The unit operates on 230 VAC 3-phase power, and is mounted on an 8 foot by 14 foot concrete pad located on the east side of the reactor building. This unit does not have a "fan cycling" option. Signals to indicate fan activation and flow are provided, with readouts in the control room.

'A common flow point is provided for bypass leg outlet and the outlet from the outdoor fan-forced air cooling unit. This common flow point is also a common point for inlet to a surge tank and the return from a secondary pump bypass leg.

The surge tank accommodates volumetric changes in the secondary coolant resulting from temperature changes and leakage. The tank also

serves as a charging port for the secondary coolant system. It is placed at an elevated location relative to the remainder of the system, and can serve as an inlet for makeup coolant and/or flushing fluid for secondary system blowdown procedures.

The secondary coolant pump is a positive displacement device with variable flow capability. The variable flow capability provides flexibility in adjusting secondary heat removal capacity and enhances system efficiency both for daily operations and in the initial system startup and optimization. Use of a positive displacement pump in this loop reduces some of the concerns associated with pumps (such as centrifugal pumps) which are sensitive to system head and coolant viscosity. Positive displacement pumps provide an inherently stable flow rate. The secondary pump is driven by a 10 HP variable speed electric motor.

The small bypass flow leg around the secondary coolant pump has a filter which removes accumulated corrosion products. This filter is a stainless steel filter sump with replaceable filter cartridges. Flow through the filter is controlled by a small throttling valve. A flow meter is provided to monitor filter performance.

Outlet from the secondary coolant pump passes through a remotely indicating flow meter to provide an indication of total secondary 53

system coolant flow rate. The coolant then enters a heat exchanger which uses city water as its heat sink. This unit provides additional cooling capacity for the secondary coolant in the event that the outside fan-forced air cooling unit does not provide a sufficiently low secondary coolant temperature. This heat exchanger, like the primary heat exchanger, is of the plate-and-frame design and is located on the main floor of the reactor building. Outlet from this heat exchanger passes back into the secondary side of the primary heat exchanger, which completes the secondary coolant system loop.

City water flow through the secondary side of the city water heat exchanger (sometimes called the tertiary loop) provides additional cooling capacity for the secondary coolant in the event that its temperature is not lowered sufficiently by the fan-forced outdoor air cooling unit. This unit is sized to provide necessary heat removal capacity while not exceeding local limits on discharge water temperature.

City water enters the tertiary loop and passes through a backflow preventer. This device, which operates on a reduced pressure principle, prevents introduction of non-potable water into the city water supply main. It is installed in Room 102 of the reactor building, near the service water line penetration. In the event of control valve closure, the pipe volume between the city water flow rate control valve and the backflow preventer is restricted. A pressure relief valve is installed in this length of pipe to limit pressure buildup and thereby protect the valve seats.

Flow rate of city water through the tertiary loop is regulated by a remotely controlled 2 inch, two-way electrically actuated valve. This device, similar to the bypass leg flow control valve, operates on 110 VAC power, and has an ANSI class VI leakage rating. City water flow rate is determined by secondary coolant temperature and the position of the bypass control valve.

Valve outlet flows through the city water heat exchanger on the secondary side. This heat exchanger was described previously. City water flow rate is monitored by a remotely-indicating flow meter.

Discharge from the tertiary loop is to a floor drain on the main floor of the reactor building. Driving head for tertiary loop circulation is provided by the normal city water service pressure, thus eliminating the need for a circulation pump.

3.2.2.3 Cooling System Instrumentation and Control Systems The reactor cooling system-has a variety of sensing devices to indicate the condition and performance of the system. These sensors include temperature, pressure, and flow measurement devices, as well as "flow" or "no-flow", pump and fan actuation, and valve positioning indicator lights.

54

The following temperature indications are provided:

Primary Loop Inlet Temperature Primary Loop Outlet Temperature Secondary Loop Coolant Temperature City Water Inlet Temperature City Water Outlet Temperature Outdoor Air Temperature The following pressure indications are provided:

Primary Loop Inlet-Outlet Pressure Drop Secondary Loop Pressure Drop Across Heat Exchanger The following flow indications are provided:

Primary Loop Flow Secondary Loop Bypass Leg Flow Rate Secondary Pump Bypass Flow Rate Secondary Pump Outlet Flow Rate City Water Flow Rate In The Tertiary Loop The following valve position indications are provided:

Bypass Leg Control Valve City Water Flow Control Valve The following actuation indications are provided:

Primary Coolant Pump Off-On Secondary Coolant Pump Off-On Air Cooling Unit Fan Bank Off-On (Two Banks)

Air Cooling Unit Fans Off-On (Eight Fans)

The information provided by the cooling system instrumentation system is sufficient to indicate system conditions and performance under normal operating conditions. Sensor readouts are available in the control room or at the location of the sensor.

The following cooling system controls are available:

Primary Pump Start/Stop Secondary Pump Start/Stop Pump Actuation Control (Ganged or Separate)

Air Cooling Unit Fan Bank 1 Start/Stop Air Cooling Unit Fan Bank 2 Start/Stop Air Cooling Unit Fan Actuation (Ganged or Separate)

Bypass Control Valve Control Mode (Auto/Manual)

Bypass Control Valve Manual Positioning (if in Manual)

City Water Control Valve Mode (Auto/Manual)

City Water Control Valve Manual Positioning (in Manual) 55

These control systems allow flexible operation and control of the cooling system, while not imposing an inordinate amount of control requirements on the operator.

3.2.3 Regenerable Demineralizer (Makeup Water System)

To assure availability of deionized water for the reactor pool, makeup water is taken from city water supply lines and passed through a series of filters, then a carbon filter, and finally through two mixed-bed ion exchange cartridges. A valve can be opened to allow the makeup water to be added to the reactor pool as needed.

3.2.4 Water Processing System A water processing system provides for the water purification of reactor and bulk shielding facilities. Incorporated in the system are particulate and ion-exchange filters which serve to retard fuel cladding corrosion and capture associated corrosion products. A flow diagram of the assembly and associated components is provided in Figures 3.15 and 3.16.

Water from the reactor and bulk shielding pools recirculates through separate particulate and ion-exchange filters of the process assembly, on an approximate 5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per day duty cycle. Monitoring instruments, which are a part of the process assembly, are utilized to assess whether operating values remain within their specified ranges, as required by OSURR Technical Specifications. These monitors provide (

temperature and conductivity indications at various points throughout the system.

3.3 Reactor Instrumentation and Control 3.3.1 General Features The reactor instrumentation and control systems provide a means to monitor the condition of the reactor and control its operation. The monitoring instrumentation provides readings on a variety of process variables important to safe and reliable reactor operation such as reactor power level, rate of power change (reactor period),, water temperatures, and others. The control system allows for flexible and reliable control of reactor power and for safe reactor shutdown under a variety of conditions.

The basic design of the OSURR instrumentation is similar to that of the Bulk Shielding Reactor and Tower Shielding Reactor, both of which were located at Oak Ridge National Laboratory, and the Breazeale Reactor at the Pennsylvania State University. All of these reactors have operated safely for many years.

Similarly, the essential features of the OSURR control system resemble those of the control system used in similar US research reactors.

56

CARBON C1 IY FLTERS FILTER I NLET I TITY EWATER EOUTLET F I LTERS DEM IN DEM IN IDEZDNIZED

,WATER bOUTLET.

-D< ~DEIONIZED WATER TO REACTOR POOL TO BSF POOL OUTLETS PROCESS SYSTEM PROCESS SYSTEM Figure 3.15: Flow Diagram of the Makeup Water System 57

FROM REACTOR F

FILTERS POOL :-:

DEMIN I I

FROM

BSF POOL

-o

- I I . I I e r e ace:_

LlrTC i urVIjL I j TO REACTOR POOL I ETO BSF NOTE I1 POOL THE BSF POOL PROCESS SYSTEM ENCLOSED FROM IS IDENTICAL TO THE REACTOR POOL MAKEUP_ PROCESS SYSTEM ENCLOSED WATER SYSTEM Figure 3.16: Flow Diagram of the Water Process System 58 C C), C71

Again, these reactors have operated for many years without major difficulties.

3.3.2 Control Room and Operator's Console All instrumentation and control systems for the OSURR have readouts and control points in the control room (Room 205 in the reactor building). This arrangement allows reactor monitoring and control from a central point. The control room also houses various interlock and permit switches which control operation of various experimental facilities located throughout the reactor building. Building communications systems (intercom, telephone dialcom, and public address) allow contact between the control room and all points within the reactor building.

Figure 3.17 shows a floor plan for the control room. The control room is heated and cooled by the reactor building HVAC system. Additional cooling and ventilation is available from a window air conditioner and vent fan. Reactor instrumentation and control system cables penetrate the floor of the control room below the control console. The door to the control room can be locked to limit access to the room. Windows along the west wall of the control room permit visual observation of activities in most of the central room (reactor room) of the building.

A telephone is located in the control room for offsite communications.

The control console has three main panels mounted in 50 inch-height, 19 inch-width cabinets. The cabinets are mounted to the floor of the control room. The central panel contains controls and readouts for positioning the four control rods, as well as the control rod withdrawal interlock system gang-lowering controls. Positioning readouts and controls for the startup source and startup channel fission ion chamber are also located on the central panel. To the left of the central panel is the reactor safety system annunciator panel. A i- series of indicator lights and pushbutton controls for safety system readout and operation are the main features of this panel. To the right of the central control panel are the readouts and controls for E auxiliary reactor systems such as permits and interlocks for experimental facilities, water processing system pump actuation, and pool lighting switches. The linear power monitoring channel range and local readout is also located in this panel. A desk-type shelf is provided at the console at which the reactor operator sits during reactor operations. The three control panels at the console slope back and away from the operator to allow convenient access to the controls and comfortable viewing of the indicators.

Above the central panel of the control console is a 19 inch-width, 24 inch-height vertically mounted rack containing additional controls and readouts. Located here are remote readouts (analog meters) for the logarithmic power monitoring channel, linear power monitoring channel, and reactor period. These meters provide the operator with convenient readouts of these parameters in a location easily readable when seated at the central console.

59

(QI Window Figure 3.17: Control Room Layout 60 C

3.3.3 Nuclear Instrumentation (NI) Racks To the left of the control console are the four NI racks housing the majority of the NI system components. These racks are 72 inches in height, and 19 inches in width. They are mounted to the floor of the control room, parallel to the east wall of the room. Access to the rear and interior of each rack is available.

Along the top of each rack are mounted the strip-chart recorders for various instrumentation systems. The remainder of each rack contains specific system components. These systems and the details of their components and operation will be discussed in following sections.

3.3.4 Control System Inputs Control system signal inputs include detector outputs, switch contacts (e.g., from manual scram switches), control rod position indications, startup source and startup channel fission chamber position signals, and safety system inputs related to instrumentation and control system performance. Auxiliary systems also provide informational displays in the control room, including conditions of experimental facilities (e.g., beam port open/closed indications), area radiation monitor outputs, and stack and effluent monitoring system readouts.

Neutron detectors used in the power level safety system, the linear power monitoring channel, and the logarithmic power monitoring channel (from which period information is derived), provide a DC signal as input to their associated signal processing electronics. The magnitude of this DC signal and its time behavior contain information necessary for these systems. The fission ion chamber (FIC) used in the startup channel provides a pulse train (one pulse for each detected neutron event in the chamber) whose pulse output rate and time rate of change of the pulse rate are used to derive reactor power and backup period indications.

Position signals for the control rods, startup source, and startup channel fission ion chamber position are generated by servo positioning transmitters. These signals are used to align position pointers (dials) on the central panel of the control console.

3.3.5 Control System Outputs The reactor control system output signals control a variety of systems located in the main reactor room. In general, the primary control system outputs concern the .position of the four control rods.

Positioning from the control room involves actuation of electric drive motors mounted at the top of the control rod drive housings. The control system also provides magnet current to hold the three shim safety rod magnet armatures (one for each rod) to the electromagnet attached to the drive rods. This magnet current can be turned off by 61

the control (safety) system to cause a sudden reactor shutdown (scram) by allowing the shim safety rods to drop into the core.

Control system outputs also include positioning commands for the startup source and startup channel fission ion chamber location. A teleflex cable driven by an electric motor provides positioning capability for the fission chamber, while a linear electric motor (electric cylinder) positions the startup source.

Bias voltages for the various NI channel detectors are provided by the instrumentation and control system. DC power for the startup channel fission ion chamber preamplifier is also provided from the control room (the preamplifier is located in the area of the pool top).

3.3.6 Signal Paths and Cable Runs Signals from the NI channels are passed by coaxial,' or, in some cases, triaxial cables. For neutron sensors, the sensor itself is electrically isolated from ground. Sensor output is connected to the central conductor of the coaxial or triaxial cable. NI channel ground is established at the NI racks, which serve as a single-point, local ground in the control room. Low current, low bias voltage transmissions are made using typical signal cable, such as RG-58/U.

Higher bias voltages are passed by appropriately rated cables such as RG-59/U. In general, signal cables use BNC-type connectors, while high voltage connections are made with MHV or SHV connectors.

Other signals such as temperature indications, servo motor position transmitter signals, switch contacts, etc., are passed by multiconductor cables or twisted wire pairs.

Many of the NI and control cables feed into a cable tray running the length of the-reactor pool top, down the southeast corner of the reactor shielding pool, and overhead on the main floor of the reactor building. The cable tray enters room 103A and runs along the ceiling to a point below the control console. The cables then run through a ceiling penetration into the cabinets of the control console located in the room above. Cable routing to the NI racks is achieved through penetrations between the console cabinets and the NI racks.

Where cable trays are not used for cable traverses, wires and cables are shielded by conduit. System interconnections are made, where necessary, within appropriately labeled terminal boxes. These boxes, located at various points in the reactor building, are securely mounted to the walls of the building or pool.

3.3.7 NI and Control System Power Building electrical service provides the initial source of electrical power for the NI and control systems. Separate constant voltage isolation transformers provide electrical service to the control console and instrument systems. The console transformer is located on 62

the main floor, while the instrument power transformer is housed in the control room.

Within the NI and control systems, circuitry is provided to supply the necessary voltages and currents. The various subsystems in the NI and control systems require voltages ranging from indicator and control voltage levels on the order of volts up to high voltage detector biasing and B+ bias levels for vacuum tube-based amplifiers. Circuit capacity is appropriate for the system load under normal conditions.

Circuit breakers and fuses are provided for safety and protection of equipment.

No auxiliary backup power system is provided. Loss of offsite power results in deactivation of the NI and control systems, which includes the electromagnet power supplies. Loss of magnet power results in a reactor shutdown from the dropped shim safety control rods. In this sense, the control system is failsafe with respect to a loss of building power event. OSURR operating procedures dictate a minimum warmup time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for all vacuum tube-based systems.

3.3.8 Control Rod System 3.3.8.1 Position Control Control rods are positioned in the core by electromechanical (electromagnet-motor-lead screw) devices controlled from the main console in the control room. A toggle switch (raise-neutral-lower) is t provided for each of the four control rods. The toggle switch is spring loaded to return to the neutral position upon release. In addition, prior to the rod being raised (except for the regulating rod) its associated "raise" switch must be depressed. The rod drive speed is the same in the raise or lower direction.

3.3.8.2 Indicators Several lamp indicators and two dial indicators on the lower part of the main console give control rod position in the core. A description of the indicator lamps and their function is shown in Table 3.1.

Two dial position indicators are provided on each control rod instrument module. One provides fine position readout (1 centimeter travel in 0.02 centimeter increments) and the other indicates coarse position of the rod from 0 centimeters' to its upper limit of rod travel in 1 centimeter increments.

3.3.8.3 Interlock Systems The main console has eight interlocking pushbutton switches in the switch plate. The left bank of four black pushbutton switches initiates downward motion of the control rods. Each switch can be pushed independently or in conjunction with any of the other pushbutton switches. Once depressed, the control rod specified on the 63

Table 3.1: Control Rod Positioning System Indicator Lights Jam Indicates control rod malfunction (lamp color:red).

Up Indicates control rod lead screw at upper limit of travel (lamp color:orange).

Intermediate Indicated control rod lead screw is near the middle range of travel (lamp color:yellow).

Down Indicates control rod lead screw at lower limit of travel (lamp color:green).

Bottom Indicates control rod at bottom of the core (lamp color:green). This light is not on the regulating rod.

Engage Indicates control rod engaged to the electromagnet assembly (lamp color:white).

This light is not on the regulating rod.

Servo Permit Indicates power within 5% of servo setpoint, as set on the Linear Level Recorder (lamp color:red). This light on regulating rod only.

Servo On Indicates servo activated and controlling regulating rod to maintain reactor power within 2% of setpoint (lamp color:blue).

This light on regulating rod only.

64

Q Lface of the switch drives is driven into the core. A blue light adjacent to the switch informs the operator that the switch has been depressed. The right bank of four red pushbutton switches underneath the column marked "raise", controls movement of the three shim safety rods and the startup channel fission chamber. A mechanical interlock prevents more than one switch from being depressed at a time.

Therefore, only one shim safety rod or the fission chamber can be moved away from the core. This prevents excessive reactivity additions to the core or loss of startup channel information.

3.3.9 Control Rod Magnet Systems 3.3.9.1 Magnet Power and Actuation Magnet power is supplied to each shim safety control rod by a magnet current amplifier module. Current is adjusted from a potentiometer located on the front of each amplifier. The current is normally set for 150% of the minimum holding current or 10 milliamperes greater than the minimum rod pickup current (as determined during quarterly rod parameter testing).

3.3.9.2 Indicators H Magnet currents for all three shim safety rods are displayed on the magnet current indicating module. Also, lights on the control rod instrument panels give indication of shim safety rod engaged or disengaged.

3.3.9.3 Interlocks Interlocks are provided to prevent application of magnet current to the electromagnets until certain conditions have been met. The safety system can turn off relay coil current and initiate a slow scram under a variety of conditions, as described in Section 3.6 later in this chapter. Other interlocks include the magnet power key switch, magnet current on/off switches on the magnet control modules, and the magnet power permit pushbutton switch on the scram activation and annunciator panel.

3.3.10 Startup Channel Detector Positioning A schematic diagram of the startup channel is shown in Figure 3.18.

The remainder of this section will discuss the startup channel positioning system in more detail.

3.3.10.1 Position Control The fission chamber of the startup channel is moved away from the reactor core as power increases. This prevents saturation of the channel at high power levels. The fission chamber position is controlled from the fission chamber instrument module. An electric 65

IDLER WHEEL

-T ELF-FLEXO DP-%VE CABLE 0 dQ DRIVE M OTOR.

TOOTHED DRIVE Wi&EEL.

P051T1ON TRANSMITTER LOWER LIMIT 5W.

LUPPER LIMIT SW.

5TART UP CHANNEL DRIVE A5M.

,5- PROTEC.TIVE SHROUD

.s-FIS5ION CAIr N ER Figure 3.18: Diagram of the OSURR Startup Channel 66

motor mounted at the top of the reactor pool drives a gear which is linked to a teleflex cable. A position transmitter indicates chamber position.

3.3.10.2 Indicators Indicators are provided on the instrument panel to indicate the position of the fission chamber relative to a position above the core.

A dial type indicator, with 1 centimeter increments, gives overall position (0-160 centimeters), and two lights are provided to give indication of reaching the lower limit (green) and upper limit (125 centimeters - blue) of travel.

3.3.10.3 Interlock An interlock is provided on the master switch plate to prevent outward motion of the fission chamber while any shim safety rod is being withdrawn, or when any of the other rods are being lowered. This prevents adding reactivity while at the same time reducing power level the detector would see. The purpose is to provide reliable information concerning power indication while pulling rods in the core. No interlock prevents moving the fission chamber inward while simultaneously withdrawing any of the control rods.

3.3.11 Startup Source Positioning 3.3.11.1 Position Control The startup source is positioned adjacent to the reactor prior to startup to provide an initial source of neutrons. As the reactor creates its own neutrons, the source is driven back into a protective cadmium shield to prevent burnout. The source is positioned with a toggle switch on the source drive instrument panel. No interlocks are provided.

3.3.11.2 Indicators Indicators are provided on the instrument panel to indicate position of the source relative to its cadmium storage cask. Two lights are provided to give indication of lower limit (in the cask) and upper limit (95 centimeters) of travel.

3.3.12 Logarithmic Power Monitoring Channel The Logarithmic Power Monitoring Channel (Log-N) provides continuous indication of reactor power covering 8 decades from 1012 to 10-amperes. The Log-N uses a compensated ion chamber located above the graphite reflector at the west end of the reactor pool.

The Log-N channel consists of a detector, high voltage and compensating voltage power supply, logarithmic and period amplifier, local and remote meters, and a recorder.

67

The detector is a compensated ion chamber manufactured by Reuter-Stokes, model number RSN-15A. Operating voltages are provided by a power supply module. Positive high voltage and negative compensating voltage are displayed on a digital panel meter located on the front of the power supply module. Compensating voltage is adjusted by a 10-turn potentiometer located next to the meter on the front panel.

The Log-N amplifier, a Keithley model 25012A, receives its input signal from the detector and provides outputs to local and remote current meters, period safety amplifier, and the Log-N recorder.

Additionally, an internal signal is provided for the period amplifier portion of the Log-N amplifier (see Section 3.3.15).

Local and remote analog panel meters are provided for monitoring of the reactor power level. The remote meter is located on the center of

-the main console with various other meters so that a centralized meter location is available for monitoring reactor power. A logarithmic recorder gives a continual, permanent record of the power history of the reactor. Its range is 8 decades, covering power levels from startup to the full operating power of the OSURR.

Although no reactor trip signals are generated directly by the power level signal from the Log-N channel, a continuity circuit is provided such that the front panel switches must be in their "normal" positions in order for the reactor to be operated. As noted above, however, one output of the Log-N amplifier is used as an input to the period safety amplifier system, which can generate a reactor trip. Also, the internal connection to the period amplifier section of the Log-N amplifier provides a signal from which the reactor period signal is U

derived, and from which an independent reactor trip function may be generated. The Log-N channel stripchart recorder also has a trip function associated with its "on-off" switch, to provide a trip if the device is turned off.

3.3.13 Linear Power Monitoring Channel The linear level power monitoring channel provides indication of reactor power in discrete decades of current/power covering a range from 10-11 amperes through 10-03 amperes full-scale deflection. This channel uses a compensated ion chamber located adjacent to the reactor, above the graphite reflector at the west end of the pool (see Figure 13).

The linear level power monitoring channel consists of a detector, power supply, amplifier, local and remote meters, and a recorder. The detector is an RSN-15A compensated ion chamber manufactured by Reuter-Stokes. Positive high voltage and negative compensating voltage are provided by a power supply module. Compensating voltage is adjusted by a potentiometer located on the front panel of the power supply. Both voltages are monitored locally by 31 digit panel meters.

68

The linear current amplifier receives an input signal from the RSN-15A detector and amplifies it according to the feedback resistance selected by the selector switch located on the right panel of the main console. The amplifier provides outputs to local and remote meters, as well as a chart recorder. All meters and the recorder read 0-150;,

with 100% corresponding to full range current of the decade selected.

Also, the recorder has an adjustable setpoint for the Servo System.

The linear level system provides three "slow" scram signals. The positive high voltage power supply has a low voltage trip, the recorder has a trip at 120% indicated power (according to the range selected), and the recorder also has a trip for the "on-off" switch in the "off" position.

3.3.14 Startup Channel The startup channel provides information about reactor power at initial and low levels of power operation. It also provides information at high power levels, but the detector must be repositioned away from the reactor core to prevent saturation of the channel sensor.

The startup channel uses a fission chamber, Reuter-Stokes model number RSN-1OA, as the detector. This detector uses a 235U coated tube to capture neutrons. The detector is positioned above the reactor core on the NNW side. The detector travels through a stroke of 154 centimeters.

Q l The detector position is controlled by a toggle switch on the fission chamber instrument control module in the main console. An interlock is provided so that the shim safety rods cannot be withdrawn while the fission chamber is moved out, and vice-versa.

The positioning system consists of a synchronous motor connected to a position indicator transmitter and a toothed wheel (see Figure 18). By moving the "in - none - out" toggle switch to the "in" or 'out" position, the motor rotates causing the toothed wheel to turn in a clockwise or counterclockwise direction. A teleflex cable attached to the top of the detector housing rides on this wheel and moves the fission chamber up or down (i.e., away from or towards the reactor cQre).

Two lamps are provided to indicate upper and lower limit, blue and green, respectively, of travel of the fission chamber. Micro switches are placed at the top and bottom of the support stanchion and are activated when the chamber is positioned at the top or bottom of its travel.

Additional startup channel instrumentation consists of a high voltage power supply, linear amplifier, discriminator, digital timer/counter, linear-log ratemeter, and a recorder.

69

The high voltage power supply provides a positive voltage to the fission chamber. An analog display on the front panel of the power supply indicates the magnitude of detector voltage (0-2.0 KV).

K Two slow scrams are associated with the startup channel. The stripchart recorder ion-off" switch must be in the "on" position, and the recorder must indicate a count rate of greater than 2 counts per second. The latter requirement insures a detectable, measurable population of neutrons is present in and near the reactor core, which provides a reliable startup.

Periodically, it is necessary (e.g., during core loading experiments) to withdraw control rods when counts (as indicated on the SU recorder) are < 2 counts per second. A bypass key, locked in the key locker, is provided to allow the slow scram alarm protection function to be bypassed. However, the alarm lamp annunciator still lights to indicate a scram condition has occurred. Additionally, a bypass lamp, labeled "rsource", is illuminated to give the operator indication that the SU system is in a bypassed condition.

3.3.15 Period Monitoring Channel The signal for the period monitoring channel recorder comes from the period amplifier section of the Log-N/Period amplifier (see Section 3.3.12). The output of the logarithmic amplifier is applied to a differentiating circuit and amplified.

The system is composed of the Keithley model 25012A Log-N/Period amplifier, Leeds and Northrup "Speedomax" type G recorder, and remote )

indicator. Analog meters are provided for local and remote indications of reactor period. A remote meter is located on the main console instrument panel. Both local and remote meters are analog devices covering a range of period from -30 seconds to +3 seconds.

The recorder provides period indications and a slow scram function.

The recorder is set up so that at a period of 5 seconds causes a slow scram reactor trip. This scram function prevents generating reactor power transients which could be difficult to conveniently control.

3.3.16 Power Level Safety Channels Two uncompensated ionization chambers (UICs), Reuter-Stokes model number RSN-36A, are positioned at the west end of the reactor pool above the graphite reflector. These detectors, along with a high voltage power supply, amplifier, and output meter, comprise the power level safety channels. The channels are independent and redundant. A signal from a UIC is applied to an amplifier which provides an output voltage proportional to input currents between 0 and 2xl00°5 amperes, corresponding to 0 to 200% reactor power. A 31 digit panel meter, Datel DM-500, is provided to monitor reactor power. Limitations of the data display device restrict actual reactor power readout on the local meter to 0.001 to 200 percent reactor power.

70

Outputs are provided to the slow scram and fast scram reactor protection systems when alarm conditions occur. Additionally, LEDs give local status to all alarm conditions of the Safety amplifier.

A system continuity circuit is installed in each level safety channel to provide a signal to the Slow Scram system on loss of the +12 volt supply voltage, or when the instrument is removed from its NIM bin. A condition of low output voltage from the high voltage power supply (high voltage to bias the detector) trips a slow scram relay causing a reactor scram. Also, the ground connector to the detector is monitored to detect when the UIC is not connected properly or removed from the system, and trips a relay in the slow scram system.

A "Fast Scram" occurs when reactor power reaches 150% of full power.

Section 3.6.2 covers the scram mechanisms in more detail.

3.3.17 Period Safety Channel The input signal to the Period Safety system amplifier comes from the Log-N amplifier, model 25012A, "Log Analog Out" voltage jack. The signal is differentiated in the period safety amplifier and the outputs sent to a comparator amplifier and external meters and recorders.

The period is displayed locally by an LED display indicating reactor periods from X to 1 second. Also, LEDs are provided to give the status of alarm conditions occurring within the period safety amplifier.

The period safety channel has a system continuity circuit installed that trips a slow scram relay upon loss of the +12 volt supply voltage, or removal of the amplifier from the NIM bin.

A fast scram output occurs when reactor period reaches 1 second. This output is applied to the gate of the current-controlling field effect transistor (FET) of the Magnet Current amplifier (see Section 3.6.2, and the basic schematic diagram of the slow and fast scram systems).

3.4 Cooling System Controls The control panel for the heat removal system is located above the control console, above the panel containing the safety system annunciators. Controls and indicators are mounted in a vertical rack of 19 inches width. Indicators can be viewed from a seated or standing position.

Section 3.2.2.3 discussed the various monitoring instruments and control available for the cooling system. System operation is controlled from the cooling system panel rack. Various pushbutton and/or toggle switches allow actuation of the various system components such as pumps, valves, and fans. The system is designed so 71

that various subsystems (e.g., primary water circulation loop) can be controlled independently or in conjunction with associated systems.

Cooling system indicators include displays of temperature pressure, flow rates, valve positions, component actuation and operation, and yes/no indications of flow. Numerical data is displayed by either digital or analog meters, while simple on/off indicators are provided by LED displays.

The cooling system instrumentation and controls are interfaced with the reactor safety system. Cooling system operation must be effective under specified reactor power conditions. Primary and secondary cooling loop pumps must be on and flow indications received from both primary and secondary loops If the reactor is operated above a specified power level to avoid a scram. Additionally, loss of cooling system capacity (as reflected in elevated outlet water temperatures) results in a reactor trip condition. Sudden changes in outlet temperature occurring over short times results in a warning annunciator indication to alert the operator to anticipate a reactivity effect.

3.5 Auxiliary Controls Operation of the circulating pumps for the water processing system is controlled from the control room. Local pump control is available at the pumps themselves through actuation of pump motor contactors and a timer manual override .switch. Manual pump control allows pump startup at any time, while automatic control allows the timer to activate the pumps. All valve positions are manually set at the valve location, with the exception of the makeup water inlet valve, which is a switch-activated solenoid valve.

Ultimate control of certain experimental facilities is maintained in the control room in the form of permit switches (interlocks).

Operation of systems with control room permit interlocks is not possible until the system interlock switch is activated. Permit switches are provided for rabbit system operation and use of the removable reactivity oscillator.

Those experimental facilities not directly controlled from the control room generally have remotely-reading indications of their condition.

For example, opening the gamma shield shutters in the beam ports, or moving the main thermal column shield door away from its fully closed position causes a lamp to be lit on the auxiliary systems control panel.

72

3.6 Safety System (Reactor Protection'System) 3.6.1 General Features The safety system protects the reactor in the event of an occurrence that could result in operation outside the allowable power ranges. The Reactor Protection System (RPS) shuts the reactor down by allowing the shim safety rods to be inserted into the reactor core under the force of gravity. The system is designed to release the shim safety rods within 50 milliseconds of a scram signal input to the fast or slow scram systems. Additionally, shim safety'rods are designed to be inserted into the reactor core within 600 milliseconds upon receipt of this signal.

Alarms are also provided to alert the operator to the system or systems that have sent an output scram signal to the fast or slow scram systems. In addition, provisions are provided such that expected alarms can be pre-acknowledged and the "Low Source" alarm can be bypassed.

3.6.2 Types of Scrams 3.6.2.1 Slow Scram (Relay Scram)

The slow scram system initiates a reactor scram by turning off the current to the electromagnets holding the shim safety control rods to the rod drive assemblies. A set of relay contacts are opened to turn i j off the power supply generating the current through the electromagnets, as shown in Figure 3.19. With the electromagnet current turned off, the control rods drop into the reactor core.

3.6.2.2 Fast Scram (Electronic Scram)

The fast scram system turns off shim safety rod electromagnet current by biasing a current controlling element in the magnet control modules which reduces the current flow through the electromagnets. When current is reduced below the minimum holding current, the control rod drops into the reactor core.

The fast scram system is somewhat quicker than the slow scram system since there is no lag time associated with the opening of a relay.

Also, both systems are fail-safe (initiate a reactor scram) with respect to loss of voltage to the system or removal of a safety device from the instrumentation system.

3.6.3 Scram Functions and Setpoints The safety system functions and their associated trip setpoints are shown in Table 3.2.

73

POWER SUPPL IES LEVEL SAFETY AMPLIFIERS 3 MAGNET COILS SLOW SCRAM

) CONTACTS 3 I PERIOD SAFETY AMPLIFIER 3 CURRENT CONTROL F.E.T.a Figure 3.19: Simplified Diagram of the OSURR Reactor Trip System ws 74 C

CA')

C

C C 2 Table 3.2: Scram and Alarm Functions of the OSURR Safety System System Signal Setpoint System Bypass Conditions Origin Ranges Action Provisions Low Count Rate Startup Channel > 2 counts/sec slow scram & bypass only if Fission Chamber alarm Keff < 0.9 Fast Reactor Log-N/Period CIC >.5 seconds slow scram & alarm None Period (recorder switch)

Fast Reactor Log-N/Period CIC > 1 second fast scram & None Period alarm (sigma bus amplifiers)

Excessive Servo Linear Level CIC < + 7% servo inhibit servo None Demand demand operation Reactor Overpower UIC in Level < 120% full power fast scram & None Safety Channel alarm 1, 2 Reactor Power Linear Level CIC > 10% full scale inhibit servo None Above Setpoint of linear level operation recorder Reactor Overpower Linear Level CIC < 80% full scale on slow scram & alarm None linear level (recorder switch) recorder Manual Scram Switch at Switch Contacts Manual Actuation slow scram & alarm None console, pool top, BSF area, thermal column, or beam ports Magnet Power Key Not In Switch Contacts Manual Actuation slow scram None The "On" Position 75

Table 3.2: Scram and Alarm Functions of the OSURR Safety System (continued)

System Signal Setpoint System Bypass Conditions Origin Ranges Action Provisions Startup Channel LCRM Switch Contacts Manual Actuation slow scram & alarm None Not In "Use" Function Log-N/Period Amp. Switch Contacts Manual Actuation slow scram & alarm None operate switch not in "Operate" Position Indicated Dose Rate ARM System < 10 iar/hr local & remote alarm Temporary Above Setpoint Detectors Replacement By Equivalent Portable Monitor High Core Inlet Temperature < 40 deg. C. slow scram & alarm None Temperature Monitor Below The Core Reactor Power Above Log-N/Period CIC < 100 KW slow scram & alarm None Setpoint With No and Flow/No-Flow Primary Coolant Flow Monitor Core Inlet Temp. Below Temperature > 20 deg. C. alarm None Setpoint Monitor Below The Core Safety System Safety System Manual Actuation slow scram & alarm None Continuity Continuity Bus High Voltage Failure On Monitor Circuit Variable slow scram & alarm None CICs or UICs In Bias Supplies 76 C C( C,

3.6.4 Scram Bypass Provisions are made to bypass the reactor trip initiated by low counts (less than 2 counts per second) on the source range channel. This allows shim safety control rod withdrawal daxiy rzr iai&3tions. The bypass is controlled by a[_

_ of the slow scram panel. A lamp pro n icati sTVgy feature has been overridden. Activation of this bypass also defeats the scram function associated with turning off the stripchart recorder in the startup channel monitoring system.

The procedure for implementing the bypass are contained in the OSURR operating procedures. This procedure is used only when the effective neutron multiplication factor of the core is not greater than 0.9, source counts are less than 2 counts per second, and shim safety control rod withdrawal is necessary.

3.6.5 Alarm and Annunciator System Closely integrated with the slow scram system are the alarm and annunciator system. The alarm system provides a visual alarm prior to slow scram or fast scram actuation. The annunciator system alerts the operator to an alarm condition, slow scram occurrence, or a fast scram trip by emitting an audible signal.

An annunciator lamp switch flashes to indicate the presence of a slow or fast scram condition (a separate switch indicates an alarm condition), and is cleared by depressing the switch. Additionally, a time delay relay has been provided such that in the event the alarm or scram condition is not acknowledged within 15 seconds, the building evacuation alarm will sound.

3.6.6 Building Evacuation System (Ventilation Control)

The building evacuation system is activated by two switches. The first switch, located on the slow scram console in the control room, causes a building evacuate klaxon horn to sound. At this signal, all personnel present in the reactor building leave the premises. This action removes people to a safe, alternate location away from the building so that appropriate equipment and personnel can have unimpeded access to the area. The other switch, located underneath the wall-mounted telephone in the control room, turns off all ventilation fans eXhausting to the outside of the building. Effluent discharge to the outside environment is thus prevented.

As a precaution, the reactor operator will scram the reactor prior to activating either the building evacuation alarm or ventilation system cutoff, and will announce over the building PA system what is about to occur. Applicable emergency procedures as outlined in the NRL Operating Procedures Handbook are followed to insure safety of all personnel and protection of property.

77

3.7 Area Radiation Monitors 3.7.1 General Features and Purpose The Area Radiation Monitors (ARM) provide information on radiation levels around the reactor building. Their purpose is to alert the reactor operator and people in the surrounding areas if radiation dose rates above a specified setpoint exist. These conditions, if 'not an anticipated part of an experiment, can then be corrected to minimize possible radiation doses to personnel.

3.7.2 Detectors and Location The ARM system uses four Geiger-Mueller detectors. They are located above the reactor pool, opposite the thermal column and beam ports, near the primary coolant loop heat exchanger, and next to the water processing system.

3.7.3 Readouts, Indicators, and Alarms Duplicate instrumentation is provided at both local (near the detector position) and remote (control room) locations. Analog meters indicate radiation levels from 0.1 to 1,000 millirem per hour at the local position, while digital ratemeters provide control room readouts.

Additionally, a test button is provided, simulating high radiation levels, to test the instrumentation. Amber and red colored lamps are also provided to indicate high background and alarm conditions respectively. An alarm annunciator is located in each of the local ARM meters, and provides a local warning to personnel in the area.

The setpoints for the background and alarm levels are adjusted on the instrument modules (remote) located in the control room. No provisions.

are made for silencing an alarm except adjusting the setpoint at a higher level (nominally set at 10-30 millirem per hour).

3.8 Experimental Facilities 3.8.1 Central Irradiation Facility The Central Irradiation Facility (CIF) consists of a 1.5 inch outer-diameter 6061-T6 aluminum tube that extends from the top of the reactor pool down into the central array position of the core. The tube is 270 inches long, with a 0.06-inch wall thickness. An aluminum plug is welded to the bottom. Additionally, a lockable cast aluminum cap is attached to the top. The core end of the CIF tube is inserted into a liner which has an end box attached to it and fits into the grid plate in a similar fashion to the fuel elements. The liner most commonly used is a water filled, 3 inch x 3 inch x 27.30 inch box, made of 0.125 inch aluminum plate, welded at the edges. The end box is welded to the bottom ends of the aluminum plates. The CIF tube is laterally supported by a square collar that fits into the top of the liner. Vertical support is accomplished by a clamp that is attached at 78

the I-beams supporting the control rod drive structure at the top of the reactor pool.

3.8.2 Beam Ports Two portholes, known as beam ports, penetrate the north wall of the reactor pool. A drawing of one of the two beam ports is shown in Figure 3.20. Each beam port consists of a shutter assembly, beam port tube, and beam port cap. The shutter mechanisms are housed in 0.1875 inch carbon steel boxes. Any voids are filled with barytes concrete.

These shutter assembly housings fit into liners in the pool wall. The outside ends of the shutter assemblies are 7.125 inch inner-diameter, type 304 stainless steel tubes, with 0.1875 inch thick walls. These tubes have beveled edges that lead to the actual shutter mechanisms, which decrease the inner diameters of the beam ports to 6,125 inches.

The shutter mechanisms are lead-filled cylinders, positioned horizontally across the path of the beam port .with their axes perpendicular to the path. These cylinders have 6.125 inch inner-diameter through holes, which line up with the beam port tube in the open position. With the cylinder rotated 90 degrees and the opening is vertically oriented, the shutter is closed. A 6.125 inch inner-diameter, type 3003 H14 aluminum alloy tube with welded flange is screwed to the shutter assembly liner. A 0.062 inch sheet lead gasket is located between the liner and flange. The tubes penetrate the inner pool wall, and are welded to a 0.25 inch aluminum support plate which is embedded in the wall. A 6061-T6 aluminum alloy cap with flange is bolted to each tube. Between the caps and tubes are 0.062 inch natural rubber gaskets.

3.8.3 Rabbit Facility The rabbit facility makes use of a vacuum system for insertion and extraction of samples to be irradiated. The vacuum system is used instead of a pressure system because the evacuated air used for operation can be expelled away from the place of operation rather than at the place of operation. This is desirable since a small amount of 41Ar is produced in the rabbit facility.

A pipe line diagram is provided in Figure 3.22. When sending a sample to the reactor core end of the facility the vacuum causes air to flow into the system through an air filter into the manifold. The manifold has a series of four 2-position valves, operated by a solenoid.

Between each pair of valves is an airway whose flow is controlled by the corresponding pair of valves on either side of the airway. The valves are opened and closed so as to allow air to flow down the rabbit tube toward the reactor core. The rabbit tube is a 2.25 inch diameter 2S aluminum tube, perforated at the end. A 3 inch diameter 2S aluminum tube sheathes the rabbit tube. Thus, the air directed down the rabbit tube flows through the perforations and into the sheathing tube. The sheathing tube is completely sealed around the rabbit tube except for an outlet tube running back to the manifold, where the valves on either side of the tube force the air into the tube that 79

or POOL WALL Figure 3.20: OSURR Beam Ports 80 IC, CI C'

4 3/8 in. Aluminum Plate 20 in. Reactor Grade Graphite, Expansion in the Axial Direction Beam Port Plug Body / Assembly:

1. 6 in. O.D. Aluminum, 1/4 in. wall turned down to 5.975 -1/- 0.005

2. Tension spring, 3 in. da., 2.5 in.

extended, stainless steeL

3. Beam Plug Filler Materials trimmed to fit, or poured to fit 16 3/4 in. Reactor Experiments Boronated Concrete, No. 244 3 in. Boronated Polethylene 3/8 in. Aluirnum Plate 4 1.69 un Spring Assembly 3/8 in. Aluminun Plate Figure 3.21: OSURR Beam Port Plug 81

4LOWE'le .1Y wlarcq 5A~

fZovoir I.t0e, ez- icaeve WVrsor ae*Mutzor ro Wor f#95 mCO(Cr.

Figure 3.22: OSURR Rabbit System Piping Diagram 82 C- Cl C'

leads to the vacuum. Air expelled by the vacuum is piped to the vicinity of the building exhaust fan where it is expelled from the system and removed from the building.

When extracting a sample from the reactor core, the valves on the manifold change state to alter the airflow. Air flows into the manifold through the filter, but is forced into the tube that leads to the sheath around the rabbit tube. Once inside the sheath, airflow is directed through the perforations in the rabbit tube and down the tube toward the manifold. The air is once again forced into the tube that leads to the vacuum and is exhausted.

The solenoid that controls the valves on the manifold can be controlled by either a manual push button or an automatic mechanism actuated by a timer. In the event of an error in timer setting or unintentional release of a sample, the manual control can be used to override the automatic timer control. A permit switch in the control room provides overall control of the rabbit system from the operator's console. Figure 3.23 shows additional details of the rabbit system.

3.8.4 Main Graphite Thermal Column A cavity exists in the reactor pool wall behind each of the thermal column extensions-. The main thermal column, which is the cavity in the west wall of the reactor pool, is stepped twice to help prevent radiation streaming. Figures 3.24 (top cutaway) and 3.25 (side cutaway) show the details of thermal column configuration. The thermal column has a support liner made from 3/8 inch steel plate, which is supported from the foundation by an iron angle structure imbedded in the reactor pool with concrete. An 1/8 inch stainless steel plate is welded to the reactor core end of the liner to provide a waterproof seal. A layer of lead brick is stacked against the inside of this plate. The remaining cavity is filled with 4 inch x 4 inch graphite bars, thirteen of which are removable for foil or sample insertion.

The front face of the thermal column is sealed with a removable 4 foot 11 inch x 4 foot 11 inch, 1i inch 3003-H112 aluminum face plate. Two 0.25 inch thick boral plates are riveted, side by side, to the inside of the face plate for shielding purposes. A 2-inch-width, 0.1875 inch-thick rubber gasket is cemented around the outer edge of the aluminum plate to form an airtight seal around the thermal column opening.

Additional shielding is provided by the main thermal column shielding door. Made of barytes concrete and attached to a "railroad car" type chassis, this door moves on two rails set in the floor of the building. It is supported by an angle iron frame around the edges, and evenly-spaced vertical and horizontal reinforcing rods welded to this frame. The overall dimensions of the shielding door are 96 inches by 31 inches, with a height of 78 inches. The door is moved along the rails by a hand-operated crank mounted at the southwest corner of the door. The hand crank drives a compound sprocket-and-chain system that directly turns the wheels at the south edge of the shielding door.

83

fReael, core.

3 In. dls.. 29 AlnumTu lng Conconlrie slangd e

0 0 0 0 0 3I Aluminum Plus I, .

-1/4 In. di, 29S Aluminum Tubing-/

.I It Core Terminal ion Plan fall polyelheloef f ,ri- -\ - D

-- - r1 'I-A-/ - i.

.Rabbit' Figure 3.23: Details of the OSURR Rabbit System 84 c C1 C

MLEAD

~ OSARYTES CONCRETE 9 GRAPHITE P L

.4 PLATE OORAL REACTOR :

Fu REACTORaer Figure 3.24: Top Cutaway View of the Main arnd BSF Thermal Columns 85

LEAD BARYTES CONCRETE M MI GRAPHITE Thermal Column Figure 3.25: Side Cutaway View of the Main Graphite 86 C, C C

3.8.5 Bulk Shielding Facility Thermal Column The Bulk Shield Facility (BSF) thermal column is located between the reactor and BSF pools, behind the thermal column extension on the south wall of the reactor pool. The BSF thermal column, shown in a side cutaway view in Figure 3.26, is an unstepped cavity, lined on the top, bottom, and two sides with 0.375 inch steel plate. The reactor pool end is capped with a 0.125 inch stainless steel plate. A layer of lead brick is stacked along the inner edge of this plate. The remainder of the cavity is filled with removable blocks of graphite. A shutter assembly exists between the graphite thermal column and the BSF pool. The shutter is made up of two h inch plates, one made of boral and the other of 6061-T6 aluminum, that are set side by side on a slide rail. Only one plate is partitioned behind the thermal column at a time. On the BSF side of the shutter, the hole in the BSF wall is capped with a support plate upon which is mounted an end plate. At one time, the support plate housed a heater coil and fission plate, both of which have since been removed.

3.8.6 Graphite Isotope Irradiation Elements Graphite isotope irradiation elements (GIIEs) can be mounted in any grid plate position. Each of them consists of a graphite element with a one inch diameter hole that runs vertically through the center, and a graphite reflector plug that fits into the hole. The outer dimensions of the graphite elements are identical to those of the fuel elements. Attached diagonally across the tops of the elements are lifting bails which can be engaged by the standard fuel element handling tool. With this tool, the elements can be moved to any desired position in the core.

When an element is not being used for sample irradiation, the graphite plug-is inserted into the vertical hole. The plugs have handles on the tops of them that can also be engaged by the standard fuel element handling tool. For sample insertion, the graphite plug is removed and an isotope production tube with sample holder is inserted in its place. Features of the GIIEs, tubes, and plugs are shown in detail in Figure 3.27.

3.8.7 Movable Dry Tubes Two movable dry tubes are normally stored in the reactor pool. These tubes can be used to place experiments at or near the core boundary.

Although a fixed mounting point is provided on the I-beam supporting the startup source drive mechanism above the pool surface, the tubes can be vertically mounted almost anywhere within the volume of the reactor pool. The tubes are locked when not in use. Both dry tubes are sealed at one end and partially filled with lead shot to maintain neutral buoyancy and a low center of gravity for vertical stability when the tubes are unsupported.

87

LEAD BARYTES CONCRETE M OaGRAPHITE Figure 3.26: Side Cutaway View of the BSF Thermal Column 88

i- 3 51..1 Q ~3I.

GRAPHITE ELEMENT

~~~~~---- -- -------- AeiD:;;I jSeet' 0 1'wi^sw."a;- l ISOTOPE PRODUCTION TUBE REFLECTOR PLUG Figure 3.27: Details of the Graphite Irradiation Elements, Tubes, L Plugs 89

The license for the OSURR allows construction of. special dry tubes and other experimental facilities and apparatus. These devices must be approved by the Reactor Operations Committee and deemed not to constitute an unreviewed safety question prior to use. If the committee deems that these devices do constitute an unreviewed safety question, review and approval by the appropriate office within the Nuclear Regulatory Commission is required prior to their use in the OSURR.

(_,

90

4.0 Normal Operating Characteristics 4.1 Core Loadina and Critical Mass Analysis of the OSURR core performance under normal conditions required use of various computer codes. Input data for the computer codes was based on 25 years of actual operating experience with the HEU-fueled OSURR core, and on LEU fuel characteristics as supplied by the RERTR program. The LEU fuel used in the OSURR core was described in Chapter 3.

Neutronic calculations considered, among others, the following design objectives:

1) maintain at least 1% dk/k shutdown margin with the highest-worth control rod and the regulating rod completely withdrawn to their highest position,

2) for a given power level, minimize the heat generated in the hot channel and

3) for a given power level, maximize the thermal neutron flux at the Central Irradiation Position (i.e., the central flux trap) and at the Beam Ports.

Utilizing the information on the standardized DOE fuel plate, and the specific design of the OSURR standard and control rod fuel assemblies, various core configurations were analyzed to meet the requirements noted above. These core configurations differ in the number and position of the standard fuel assemblies on the grid plate, as well as the number and position of the graphite irradiation elements (sometimes called graphite reflector blocks) on the grid plate. Figure 4.1 shows the numbering scheme for grid plate position, and will be utilized in specifying assembly positions on the grid plate in the remainder of this chapter.

The neutronics analysis used the LEOPARD code for the generation of four-group neutron diffusion parameters on a unit cell basis. The 2DB code was used for reactivity and neutron flux calculations. The computational predictions were tested (benchmarked) with data from the HEU-fueled OSURR core measurements and independent calculations utilizing the EPRI-CELL, DIF3D, and VIM codes.

ZMAlvtieAl results indicate that the OSURR core will contain between standard fuel elements and four control rod fuel ele

-is yfds a nominal 35U core loading of between respectively. Possible core configurations are sholn inrigure-74.2-through 4.4, which represent grid plate positions and the elements positioned in the available locations. The X's in each of these figures denote positions on the grid plate without either a fuel assembly or a graphite irradiation element (as discussed in Chapter 3, 91

West A a C D E 2

C C 3 CI F

South North A C C 5

6 G G G G 0 East Notes: (1) The control rod positions are shown by the letter "C",

and are fixed.

(2) The CIF position (water or graphite-filled central flux trap) is normally fixed.

(3) Graphite irradiation element positions (letter "G") can be removed. They are shown installed in the standard HEU core position.

(4) The core is reflected along the west and south edges by the extension pieces for the main and BSF thermal columns (graphite reflectors).

Figure 4.1: Numbering System for Denoting Positions on the Grid Plate of the OSURR 92

A B C D E l

2

!i.

3 E ,.

LEU Core F A

5 I ,

I 6

X's denote vacant grid positions

.A Bi C 0 E 2

3 LEU Core G A

5 6

t Figure 4.2: Possible OSURR Core Configurations, LEU Cores F and G 93

A B C D E I

2 C C

-, C A

1 5.53 keV)

Average Fast Neutron Flux 6.36 x 1012 nV 23

.Minimum Critical Mass 3.57 kg. 5U 235 Typical Core Loading (22 4.1 kg. U Elements)

Total Control Rod Worth 8.45 % Ak/k Minimum Shutdown Margin 1.00% Ak/k Temperature Coefficient -6.3 x 10-% Ak, voidC Core Average Void Coefficient -0.18% Ak/k/1% void (min.)

-0.45i Ak/k/1% void (max.)

123

Table 4.1: Summary of Reactor Data (continued)

Prompt Neutron Lifetime 6.6 x 10-- seconds Delayed Neutron Fraction 0.766i Reactivity Requirements Equilibrium Xenon 1.00% Ak:/k Temperature Feedback 0.20% Ak.k Burnup, Fission Product Buildup 0.50% Ak./k Experiments 0.70% Ak./k Reactor Control 0.20% Ak./k TOTAL EXCESS REACTIVITY 2.60% Ak /k Specific Design and Operating Character istics Void and Temperature Negative Coefficients Overpower Trip Limits (Lowest) 120%

Maximum Excess Reactivity 2.60% Ak/k Maximum Regulating Rod Worth 0.48% Ak/k Minimum Shim Safety Rod Worth 4.55% Ak/k Maximum Shim Safety Rod Worth 7.97% Ak/k Total-Control Rod Worth 8.45% Ak/k Maximum Movable Experiment Worth 0.40% Ak/k (Any Single Movable Experiment)

Total Movable Experiment Worth 0.60% Ak/k Total Experiment Worth In Core 0.70% Ak/k Minimum Startup Count Rate 2 counts/second 124

nmn..wmmmmmmd Table 4.1: Summary of Reactor Data (continued)

Control Systems Shim Safety Rods:

Number Three Composition Boron-Stainless Steel Boron Weight Percent 1.5 Cross-Sectional Shape Grooved Oblong Poison Section Length 26 Inches Regulating Rod:

Number One Composition Stainless Steel Cross-Sectional Shape Smooth Oblong

-Control Rod Stroke 24" Nominal Drive Speed 11 cm/min Scram Response:

Nominal Rod Drop Time 500 Milliseconds Nominal Magnet Release Time 30 Milliseconds Fuel Type Flat-Plate MTR Fuel Plate Characteristics:

Loading L 323U/plate Thickness Clad Material 6061 AiluminJ IuLU Clad Thickness 0.015" Fuel Meat Thickness 0.020"

.Fuel Alloy Uranium-Silicide (U3Si,)

Uranium Enrichment 19.5%

Active Fuel Length 24" (Nominal)

Fuel Elements:

Number 21 (Standard) 5 (Control Rod) 4 (Partial)

Plates/Standard Element 16 Fueled 2 Dummy 125

Table 4.1: Summary of Reactor Data (continued)

Plates/Control Rod Elemen t .10 Fueled 2 Guide Plates (Dummy)

Plates/Partial Element 4, 6, 8, 10 Fueled 14, 12, 10, 8 Dummy Side Plate Material 6061 Aluminum End Box Materials 6061 Aluminum Fastening Methods Swaging and Welding Experimental Facilities Beam Ports:

Number 2 Location North Core Edge Size 6" Inner, 7" Outer Length 6' 6" Orientation Perpendicular and Angle To Core Pneumatic Tube:

Number 1 Location North Core Edge Above Beam Ports Diameter 3" Main Thermal Column:

Extension Location West Core Edge I ., Dimensions 414" x 4'4" x 6' I I Stringer Materials Reactor-Grade Graphite Stringer Dimensions 4" x 4" x 57" (Removable)

Core End Shielding Lead (3")

Ii\

Outer Face Boral i Inch Boral Plate I Shielding Shield Door 2' Barytes Concrete BSF Pool Thermal Column:

Extension Location South Core Edge Shutter Boral Shield k Inch Boral Plate Shutter Cadmium Shield 0.030" Cadmium Sheet Column Shield Lead (3 Inches) 126

5.0 Auxiliary Systems 5.1 Introduction Several auxiliary systems aid in safe reactor operation. They provide the necessary support for a variety of operations including lighting, cooling, heating, electricity, etc. This chapter will discuss the features of these systems and their use.

5.2 Communication Communications are provided at various places around the reactor building. The communications systems include both audio and visual indications.

5.2.1 Control Room Intercom System A two-way intercom, located on the right-hand control console, provides the reactor operator with means to talk and listen to several stations located throughout and outside the Reactor building. These intercom stations also feature a local push-to-talk button that allows experimenters (or other personnel) to talk to the operator.

5.2.2 Building Phone System All phones located in the building are connected to the university for both internal and external calls. The control room has its own dedicated line.

5.3 Lighting Systems Three lighting systems are located inside and outside of the building to supply light for the personnel needs and safety requirements.

5.3.1 Normal Interior Lighting Several high-intensity overhead lights are located in the reactor bay and machine shop area. Additionally, a combination of fluorescent and incandescent lighting is provided in all office, classroom, and storage areas.

5.3.2 Emergency Interior Lighting In the event of power failure, four lamps, two at the front of the building and two at the rear, are activated. Each lamp has its own battery with a capacity sufficient for operation up to one hour.

Additionally, several flashlight are located around the building for use in an emergency, power failure, experiments, etc.

127

5.3.3 Exterior Lighting Four outdoor lamps are located on each corner of the building. These lights are automatically controlled by a light sensor affixed to the top of each lamp. There is also a light above the front entrance.

5.4 Building Services Several auxiliary systems are provided for use by reactor personnel and experimenters. These include electricity (120V and 230V outlets),

city water, 100# compressed air and natural gas.

5.4.1 Electrical, Water, Air and Gas Electricity is provided via several 120V outlets around the building.

Breaker panels located in the machine shop area and adjacent to the east service room provide remote isolation of these electric circuits.

Additionally, a few 240V service outlets are also provided.

Standard city water is provided for experiments and sanitary facilities. Although demineralized water is available (from the reactor demineralizer), it is usually not provided for most experimenters, but is mainly dedicated to the reactor system.

The university supplies compressed air (100#) and natural gas for the building. Several 'standard" laboratory outlets are provided throughout the building.

5.4.2 HVAC Heating and ventilation are provided throughout the building. The rooms, located in the southeast and southwest corners of the building, house the associated blowers, motors, heating units, plenums, and associated structures. Additionally, overhead gas heaters located above the front service door and rear entrance provide heat to the reactor bay.

Air conditioning is provided by an outside unit in conjunction with the ventilation system for the east offices in the building. Several auxiliary air conditioners are provided for individual rooms, such as the control room, that have extra cooling / humidity-control needs.

The control room (205) and room 104 have window-type air-conditioning units that exhaust to the outside environment. Room 109 has a window-type unit exhausting into the reactor bay area.

All building heating and air conditioning systems using air circulation fans are connected to the emergency switch located in the control for shutting off exhaust fans. This feature allows partial isolation of the building interior environment from the outside atmosphere, and may be used under certain emergency conditions.

128

5.5 Bridge Crane A An overhead crane, running on parallel tracks (north-south) provides capabilities for fuel handling, thermal column access, experiment movability, and other operations requiring movement of heavy objects.

The crane is rated at 3 tons and has a remote control box to allow operation of the crane from ground floor. Several hooks and attachments are provided for use with various experiments and experimental facilities as well as fuel handling.

5.6 Building Alarm System The reactor building is equipped with a security system in accordance with the NRC-approved security implementation procedures. The details of this system are discussed in the security procedures and are not subject to unrestricted disclosure. The general features of the security system include various devices and subsystems for detection of unauthorized entry, activity in certain restricted areas, and manually-actuated alarm switches.

5.7 Access Control II_

129

6.0 Radioactive Waste Management 6.1 Source Term Estimation 6.1.1 Liquid Effluents Under normal conditions, no liquid is released from the reactor pool, the primary coolant loop, or the secondary coolant loop. Events that result in significant fluid releases from these systems are considered accident conditions and are discussed in Chapter 8.

Liquid-borne radioactive materials in the reactor pool and primary coolant loop eventually pass through the water processing system. The primary radionuclide detected in the resin bed ion exchange demineralizer is 24 Na resulting from neutron-alpha reactions with 28A1 in the aluminum used in and around the reactor core. Current procedures require changing of the demineralizer cartridge at the point where it cannot maintain pool water conductivity requirements.

The replacement procedure allows for holding the spent demineralizer cartridge at the reactor building for several months prior to returning it to the manufacturer for regeneration. This holding period allows 24Na activity to decay to negligible levels.

Thus, under normal conditions, no liquid radioactive effluent will be produced by OSURR operations.

6.1.2 Gaseous Effluents The primary gaseous radionuclide produced by OSURR operations is 4Ar.

This isotope is produced whenever air is in contact with a neutron radiation field. Naturally-occurring 40Ar, which comprises over 99% of all argon, undergoes a neutron capture reaction to produce 41Ar, which decays by beta and daughter product (41K) gamma emission, with a half-life of 1.83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months . Argon is found in air at slightly less than 1; concentration under STP conditions.

Smaller concentrations of gaseous radioisotopes'will also'be produced from other activation products in air, experimental procedures, and a slight possibility of very small quantities of fission product gases released into the reactor room environment from dissolved fission product gases in the pool water. However, the quantities of these other sources are very small compared to "Ar production.

The water in the reactor pool also contains dissolved air. It is assumed that the dissolved air has an argon concentration equal to that found in atmospheric air. Some of this argon will activate and be released from the surface of the reactor pool into the building air.

Fast neutrons (with energies above about 10 MeV) can interact with the oxygen nuclei in the pool water via the, neutron-proton reaction and produce the 16 N isotope. This nuclide has a very short half-life (7 130

seconds), so very little of it will reach the surface of the pool because of decay during transit from the core to the surface. Further, if the circulating pumps in the primary loop of the cooling system are on, the water rising from the core will be dispersed into the lower regions of the pool, greatly increasing the effective transit time for a given volume of water from the core to the surface. However, since 16N can contribute to personnel doses as a source distributed on the surface of the pool as well as add to the radionuclide concentration in the air of the building, its production and distribution will be considered in a following section.

6.1.2.1 Argon Production in Experimental Facilities Production of 4'Ar can occur in the two beam ports, the rabbit facility, the central irradiation facility (CIF) tube, dry tubes mounted near the core, and any open stringers in the main or BSF pool thermal columns. When not in use, the beam ports are normally filled with shielding plugs, which effectively reduce to zero the volume of air in the portion of the beam ports normally exposed to neutron radiation. The main and BSF pool thermal columns are also filled with graphite stringers when not in use. Movable dry tubes are stored at a location away from the reactor core when they are not in use. Thus, in many cases, the only sources of 41 Ar during routine operations will be the CIF tube and the rabbit tube. A puff-type release can also occur from the rabbit carrier tube after it is withdrawn from the rabbit facility following irradiation and is opened.

Isotope production can be estimated from: (

At) = Nc*(1 -At, ) -At where decay constant t = time after removal of neutron flux t,= exposure time to neutrons a = microscopic cross-section for reaction of interest

= neutron flux in neutrons/cm2 /second N = total number of target atoms available for activation.
In using this equation, it is assumed that N remains constant; that is, there is no significant "burnup" of the atoms available for the reaction forming the activation product.
In the interests of conservatism, it is assumed that activity buildup of 41 Ar is sufficient to achieve saturation. That is, the irradiation time (ti) is equal to at least five half-lives of the activation product. For 41 Ar, with a half-life of 1.83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months , this corresponds to an activation time of 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . This is conservative in that few OSURR operations will involve irradiations of this length. Also, it is assumed that the release occurs immediately at the end of the irradiation, i.e., t = 0. The production equation simplifies to:
131
A = Naog where a and 4 are defined above. Goldman [1] reports a value of 0.61 barns (6.1x10-2 cm2) for the microscopic thermal neutron capture cross-section of 4OAr.
Values for neutron flux are based on thermal neutron flux measurements obtained in the various OSURR facilities at 10 kilowatt operation. It is expected that these thermal neutron fluxes will remain about the same or lower for an LEU-fueled OSURR core, and will be linear with reactor thermal power. Thus, the measured values for 10 kilowatt operation were multiplied by a factor of 50 to estimate the fluxes for 500 kilowatt operation. Table 6.1 shows data for 41 thermal neutron fluxes in the various experimental facilities. For "Ar production calculations for the main graphite thermal column and the two beam ports, the flux was taken to be half the peak flux, for reasons to be explained in a later paragraph.
The number of OAr atoms available for activation is a function of the volume of air in the experimental facility and the concentration of argon in air under STP conditions. Etherington [2] reports a concentration of 2.5x1017 atoms of argon per cubic centimeter of air under STP conditions. Using the isotopic abundance of 0.996 for 4'Ar, a concentration of 2.49x10 1 7 atoms of 40Ar per cubic centimeter of air at STP is estimated.
The volume of air in each experimental facility must be estimated.
j Since the experimental facilities of interest are either tubes or rectangular ducts, their volume is given by their cross-sectional area times their effective length. In this case, effective length is taken to be that length over which most of the 4"Ar production occurs. For facilities oriented either parallel or through the core, such as the CIF tube, the rabbit facility, and movable dry tubes, the effective length is taken to be the characteristic dimension of the core over which the experimental facility tube passes. For the CIF and dry tubes, the effective length is 24 inches, which is the vertical length of the active fuel'part of the core. For the rabbit facility, the effective length is the dimension of the side of the reactor core, which will be no more than 18 inches, since the grid plate is a 5 x 6 array, with each fuel element being 3 inches long. While some 41Ar production occurs in the regions beyond these boundaries, it is not as great as that produced within these regions. Conservatism is added by assuming that the flux through this region is uniform and equal to the peak neutron flux along the effective length, which is the value shown in Table 6.1. In actuality, the average neutron flux in the facility will be lower than the peak value, since the spatial distribution of flux in the vertical and horizontal directions shows a reduction in thermal neutron flux as the edge of the core is approached.
132
Table 6.1: Measured and Estimated Thermal Neutron Fluxes in Various Experimental Facilities of the OSURR Facility Measured Thermal Neutron Estimated Thermal Neutron Description Flux at 10 Kilowatt Power Flux at 500 Kilowatt Power CIF 2.02 x 101l 1.01 x 1013 Beam Port 1 1.27 x 10" 6.35 x 1012 Beam Port 2 9.74 x 10"' 4.87 x 101-Rabbit 4.58 x 1010 2.29 x 101-Thermal Column 1.33 x 101' 6.65 x 10l Dry Tubes 4.20 x 1010 2.10 x 101i Notes: (1) The thermal neutron 2 fluxes shown above are in units of thermal neutrons/cm /second, as reported by Horning
[:3].
(2) Neutron energy range for thermal flux includes all neutron with energies up to 0.6 eV.
'L)
(3) The flux shown for the rabbit facility is also that assumed for calculation involving the rabbit carrier tube.
(4) The flux shown for the thermal column was that measured in the main graphite thermal column at stringer position G-7 (central location). It is assumed to be the same in both the main and BSF graphite thermal columns.
(5) Estimated thermal neutron fluxes at 500 kilowatts were obtained by multiplying the measured flux at 10 kilowatts by 50.
(6) The thermal neutron fluxes shown for the movable dry tubes are assumed to be the same for both the 2" and 4" tubes. The tubes are assumed-to be mounted at the northeast corner of the core. The measured values at 10 kilowatts are those reported by Talnagi [4].
133 Q.,)
The effective length for the beam ports and graphite thermal column stringer positions, which converge at the core in a generally perpendicular direction, was estimated differently. Using a spatial neutron flux distribution measured in the open central stringer position in the main graphite thermal column, a characteristic length was determined for a neutron flux profile in air. This length was taken to be that required to achieve a 90% reduction in the initial (peak) neutron flux at the point closest to the core. For calculation of 4:Ar activity, the average neutron flux over this length was assumed to be one-half of the peak thermal flux occurring at the core end of the facility.
Table 6.2 shows the results for effective volumes of the various experimental facilities, and the resulting total number of 4Ar atoms available for activation. Using these results in the production equation leads to the source term estimates shown in Table 6.3.
Assuming an irradiation time sufficient to achieve saturation, the rate of radioisotope production is given by:
R AX where R = rate of isotope production in disintegrations/sec/sec X =decay constant, and A = saturation activity in disintegration/sec calculated from the production equation.
The rates of 41 Ar production at saturation in the various facilities are listed in Table 6.3.
The rabbit system can be activated and operate continuously for a set period of time. During operation, 4"Ar produced in the effective volume of the rabbit is continuously purged into the reactor room atmosphere, which is exhausted to the outside atmosphere by the building vent fan.
The 4Ar effluent concentration can be estimated from the production rate of "Ar and the volumetric flow rate of the rabbit blower system.
The name-plate capacity of the rabbit blower system is 150 cubic feet of air per minute, or 7.0792x104 cc/second. As shown in Table 6.2, the effective volume of the rabbit facility is 1737.36 cc. This implies a cycle time of 24.54 milliseconds for the effective volume. Assuming a constant neutron flux equal to that shown for the rabbit facility in Table 6.1, and the number of 4OAr atoms available for activation shown in Table 6.3, an irradiation 2 time of 24.54 milliseconds gives an isotope total of 4.22xO--'microcuries. This total must now be multiplied-by the total number of cycles of the rabbit effective volume per second (about 41). This results in a production rate of 1.72 microcuries of 41Ar per second in the exhaust of the rabbit blower.
134
Table 6.2: Calculated Air Volumes in Various Experimental Facilities.
of the OSURR Facility Cross-Sectional Effective Air Volume Description Area in cm2 Length in cm in cm' CIF 9.65 60.96 588.26 Beam Port 1 190.09 66.04 12553.84 Beam Port 2 190.09 66.04 12553.84
.Rabbit 45.60 38.10 1737.36 Thermal Column 103.23 66.04 6817.02 Rabbit Carrier 8.43 13.97 117.77 4" Dry Tube 61.58 60.96 3753.68 2" Dry Tube 21.29 60.96 1298.11 Notes: (1) When not in use, the two beam ports and thermal column have essentially no effective volume of air since they are filled with plugs or graphite stringers.
(2) When in use, inner volume of the main graphite thermal column is sealed with a boral-aluminum plate.
(3) The effective volume of the rabbit carrier tube may be lower than the value shown if it is stuffed with cotton, as is normally the case when it is used. The volume shown is the maximum available empty volume in the carrier tube.
(4) The effective volume of Beam Port 2 may actually be smaller than the value shown, since it intersects the face of the core at a non-perpendicular angle, which causes a sharper neutron flux gradient, reducing its effective length (see text for definition of effective length).
135
Table 6.3: Estimated 41 Ar Source Terms for the Various Experimental Facilities of the OSURR Facility Available 4 Ar Saturation 4Ar Saturation 4'Ar Description Atoms Activity In Activity Production Microcuries Rate (Lcuries/sec)
CIF 1.46 x 1020 2.43 x 104 2.56 Beam Port 1 3.13 x 1021 1.64 x 105 17.25 Beam Port 2 3.13 x 1021 1.26 x 105 13.25 Rabbit 4.33 x 1020 1.63 x 104 1.71 Thermal Column 1.70 x 102 9.32 x 103 0.98 Rabbit Carrier 2.93 x 105 1.11 x 103 0.12 4" Dry Tube 9.35 x 1020 3.24 x 10' 3.41 2" Dry Tube 3.23 x 102° 1.12 x 104 1.18 Notes: (1) The values assumed for thermal neutron flux in the two beam ports and the thermal column is one-half that shown in Table 6.1 for 500 kilowatt operation, as discussed in the accompanying text.
(2) For isotope production rate, the half-life was assumed to be 1.83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months , which gives a decay constant of 1.052 x 10-4 seconds.
(3) The estimate shown for the thermal column are assumed to be the same for the main and BSF thermal columns.
(4) All isotopic production calculation results shown above assume the fluxes given in Table 6.1 for 500 kilowatt operation, except as stated in Note 1 above.
136
6.1.2.2 Argon Production from Pool Water Estimation of the 4"Ar production from dissolved air in the water of the reactor pool begins with a calculation of the exposure time of water passing through the core. Section 4.8 noted that the average coolant velocity through the core is 6.5 cm/second, assuming a 500 kilowatt operating power and natural convection through the core. The length of the active fuel channel is 60.96 cm (24 inches), which gives a coolant transit time of 9.4 seconds, assuming a constant average velocity through the core. This is taken to be the exposure time of the water to the average flux throughout the core.
Based on measurements of the peak thermal neutron flux in the core region at a 10 kilowatt power level, and assuming linearity of thermal flux with reactor power, the peak thermal neutron flux in the core is assumed to be 1x10l 3 neutrons/cm2 /second. Measurements of the peak-to-average thermal neutron flux at 10 kilowatts indicate that the average thermal neutron flux throughout the core will be about 60% of the peak thermal flux, or 6x101 2 neutrons/cm'/second.
The volume flow rate of water through the core is the product of the coolant velocity and the total flow area. Assuming a core with 18 standard fuel elements and 4 control rod fuel elements, the total flow area is the product of the flow area of an individual coolant channel and the total number of channels in the core. Section 4.8 noted that the flow area of a single coolant channel is 1.964 cm2 . The total number of flow channels is assumed to be 364 (18 standard elements with 18 flow channel each, and 4 control rod fuel elements with 10 channels each). Thus, the total core flow area is 714.896 cm-, and the total volumetric flow rate is 4646.8 cm3 /second.
Now, the average out-of-core cycle time is given by:
T = Vp /V where Vp = total volume of the pool, and V = volumetric flow rate through the core.
If the dimensions given in Section 3.1.3.1 are used for the size of the reactor pool, a volume of 2.223x107 cm3 is obtained. Using the volume flow rate calculated earlier, an out-of-core cycle time of 4783.24 seconds is obtained. This can be thought of as a decay time for "Ar produced in the water of the pool.
The concentration of argon gas in the pool water can be predicted by Henry's Law. The dissolved concentration of a gas in contact with a liquid is proportional to the partial pressure of the gas and the temperature of the liquid. Dorsey [5] reports values for air at STP conditions in water that allow an estimation of 8.65x10'5 atoms of 40Ar 137
per milliliter of water, assuming a water temperature of 25 0 C (core inlet temperature).
The saturation activity of ClAr in the pool water may be predicted from:
A = Nca+(1 - e_"')/(l - e7xtItT) 0 where N = concentration of 4'Ar atoms in the pool water, a = neutron capture cross section for 40 Ar,
= physical decay constant of 4 1Ar,
= average thermal neutron flux in the core region, t = exposure time of water in the core, and T = average out-of-core cycle time.
Substituting appropriate constants in this equation yields an estimate of 79.02 disintegrations/second/cc. Dividing this estimate by the decay constant for 4Ar gives a calculated density of 7.512x10- atoms of "lAr/cc.
As water passes through the core it is heated, which reduces the solubility of air in the water. For this calculation, it is assumed that 25% of the dissolved argon is released from the water because of core heating. Some of this released argon will be redissolved as it mixes with cooler water in other regions of the pool. Measurements done at other reactors allow an estimate of 50% redissolving fraction.
Thus, the argon available for release to the building air is given by:
S= F1 (1- F2 )N41V where N4 1 = 4Ar concentration in the water at equilibrium, F, = release fraction from heating (assumed to be 25%)
F2 = redissolving fraction (assumed to be 50%), and V = volumetric flow rate through the core.
Substituting appropriate values in this equation leads to an available release term of 4.36x10 8 atoms of 4"Ar/second. This represents one component in the 4Ar release from the pool water.
Another release term arises from the tendency of dissolved gas at the surface of a liquid to escape to the air across the water-air boundary. Estimating the magnitude of this release term requires calculation of an effective exchange coefficient for argon (exchange coefficient being the amount of gas in a unit volume exchanged at the surface per unit time per unit area).
Other reactor facilities have analyzed this problem and provide possible exchange coefficients that appear to cover a wide range. For example, analyzing the gas exchange at the liquid-gas boundary in 138
terms of the diffusion coefficient of argon gas dissolved in water and the mean-square distance traversed by a molecule, an estimate of 2.35x10-3 cm/second is obtained. However, measurements made of the 4 Ar activity in the pool water of a TRIGA Mark III and subsequent analysis of these data indicate an exchange coefficient of about 2.9x10-4 cm/second. Further, Dorsey [5] reports approximately equal surface exchange coefficients for gases such as air, 02, and N2 . Assuming that the exchange properties of argon are similar to those Of these gases, an exchange coefficient of about 5.7x10-3 cm/second is possible. Note that these estimates vary by almost a factor of 10.
In the interest of conservatism, the largest exchange coefficient (5.7xlO-; cm/second) is assumed in this calculation. Using this, the release rate from gas exchange at the surface of the pool is given by:
S2 =O.93BN4 1 A5 where N41 = concentration of 41 Ar atoms in the pool water, B = exchange coefficient, and As = surface area of the pool (3.646x104 cm2 )
Using this equation a release rate of 1.45x108 atoms/second is obtained. Now, the total source term for 4"Ar released from the pool water is obtained by adding this to the previous estimate for dissolved argon:
S41 S 1 + S 2
= (4.36x10 8 + 1.45x10 8 ) atoms/second
= 6.81x108 atoms/second.
This is the source term for 4"Ar released from the pool water to be used later in estimating doses and isotopic concentrations. The source term assumes 500 kilowatt operation for a time sufficient to attain saturation activity.
6.1.2.3 Nitrogen-16 Production from Pool Water Section 6.1.2.2 above derived an exposure time of 9.4 seconds for water flowing through the core. The concentration of 16N atoms per cc of water leaving the reactor core can be estimated from the following modified form of the activation product production equation:
N = [Ccnpf (1-ext WX where N = concentration of 16N atoms leaving the core, C = concentration of oxygen atoms in the pool water, an,= n-p microscopic cross-section for 160,
= spectrum-averaged fast neutron flux (0.6-15 MeV),
t exposure time (9.4 seconds), and 139
X = decay constant for "N (9.761xl0O: second-).
it remains to find appropriate values to substitute into this equation.
First, the concentration of oxygen atoms in water can be taken to be approximately 3.3xl02 atoms of oxygen per milliliter of water. This value ignores dissolved air in the water, as the number of atoms of oxygen from the water molecules far outweighs the number from dissolved air in the water.
Next, the spectrum-averaged microscopic cross-section of 160 is taken to be about 0.021 millibarns, or 2.1x10-2 9 cm2 . This assumes an integration range of from 0.6 MeV to 15 MeV incident neutron energy.
Finally, a value must be assigned for the spectrum-averaged fast neutron flux. The cross-section threshold for the n-p reaction in 1CO is about 9.4 MeV, but this must be corrected for center-of-mass effects. When these are taken into account, the effective incident neutron energies are about 10.2 MeV. This relatively high threshold energy results in severe limitation of 16N production, since relatively few neutrons in the OSURR in-core neutron spectrum fell above this threshold. 2 Horning [3] reports a value of 8.4x10lC neutrons/cm2 /second for neutrons above 0.5 MeV (sometimes called the "fission" component) at the central irradiation facility (assumed to be the peak flux) for the HEU-fueled OSURR. Assuming that the LEU-fueled OSURR in-core neutron spectrum is about 15% "harder" (based on experience of other core conversions), and that the power is increased by a factor of 50, the effective neutron flux above 0.5 MeV is assumed to be 4.83x101 2 neutrons/cm2 /second.
Substituting values into the production equation yields an estimate of 2.01x10 6 atoms of 16N per milliliter of water leaving the reactor core.
This is assumed to be an equilibrium concentration, given the very short half-life of the isotope. If the volume flow rate through the core calculated-in Section 6.1.2.2 (4646.8 cc/sec) is multiplied by this concentration, a rate of 9.34x109 atoms of ioN per second are released from the core. Multiplying this by the decay constant for °N and converting to activity units gives a release term of 24.64 millicuries of 16N per second released from the top of the core.
6.2 Liquid Effluent Waste Management 6.2.1 Pool Water Monitoring Normally, no water is released from the reactor pool. The water level of the pool is. visually checked prior to each startup of the reactor.
The reactor safety system has a reactor trip function should the pool water fall below a setpoint. The water process system has a water 140
inlet valve controlled by a second water level sensor switch to add makeup water to the reactor pool.
Concentration of gamma-emitting radionuclides in the reactor pool water is checked as part of routine maintenance and surveillance activities. Gamma dose rates above the pool are monitored continuously by an area radiation monitor (ARM). An additional ARM monitors dose rates in the vicinity of the primary heat exchanger. A third ARM monitors dose rates in the area where the reactor pool demineralizer is located. This unit traps most of the 24Na activity contained in the reactor pool water. After each reactor shutdown, the on-contact dose rate of this demineralizer is measured and recorded. If necessary, the area .is posted and access to it is controlled.
6.2.2 Secondary Loop Coolant Monitoring Under normal conditions, no radionuclide concentrations should be present in the secondary coolant. However, to assure this, the dose rate at the corrosion product trapping filter near the secondary coolant pump is surveyed routinely as part of surveillance and maintenance activities.
6.2.3 Liquid Effluent Releases If significant radionuclide concentration in either the primary or secondary coolant is suspected (above that which is routinely encountered), appropriate procedures are invoked to determine the radionuclide identity, concentration, and release pathway. Based on these tests, necessary corrective action can be taken.
Should release of all or part of the coolant inventory be deemed necessary for repair and/or maintenance activities, appropriate procedures based on the results of the radionuclide assay will be followed to assure compliance with regulations specified in 10CFR, part 20. In most cases, immediate release of the pool water to the city sewage system will be allowed. If not, the fluid will be held until sufficient decay time has elapsed to reduce radionuclide activities to permissible release levels. Otherwise, alternate storage/disposal methods and procedures will be followed.
6.2.4 Cooling System Maintenance Operations 6.2.4.1 Draining, Blowdown, and Purging At the lowest point in the secondary loop of the cooling system, a trap and drain valve is available for drawing a small sample of secondary coolant. Additional coolant draining can be done at this point to remove larger volumes of fluid.
Maintenance of the secondary coolant chemical and fluid properties can be achieved by intermittent blowdown procedures. A small amount of fluid can be withdrawn from the drain valve and replaced with fresh 141
fluid at the surge tank charging port. The entire volume of secondary coolant can be purged, if necessary, and refilled from the, charging port.
An isotopic assay will be performed on all fluid withdrawn from the secondary coolant. If significant quantities of radionuclides are detected, they will be identified and quantified. Based on these data, appropriate procedures will be followed prior to release of any secondary coolant, and tests will be conducted to determine the primary-secondary leakage path. Appropriate repair and maintenance actions can then be taken.
6.2.4.2 Tertiary Loop Effluent Holdup There is a very small probability that the city water supply used in the tertiary coolant loop could be contaminated by primary coolant.
The probability of significant levels of contamination being present is low, since it would require a primary-secondary-tertiary leakage path. However, to assure that tertiary loop radionuclide concentration is known, part of the routine surveillance and maintenance activities will include a radionuclide isotopic assay on tertiary loop effluent during or immediately following a reactor operation involving use of the tertiary loop, 6.3 Gaseous Effluent Waste Management 6.3.1 Effluent Monitoring System Gaseous radionuclides are detected by an effluent monitoring system.
This system extracts a sidestream of the air ejected from the reactor building by the building ventilation fan. The sampled air is introduced to a shielded volume containing a double-sided pancake-type Geiger-Mueller detector. Detector output is counted on a rate meter in the control room, and recorded on a panel-mounted stripchart recorder.
System response is calibrated for 41 Ar activity. Detector count rate is noted in the control room. The count rate for the derived air concentration (DAC) of 4"Ar is posted at the recorder. Total 4"Ar production is tabulated on a yearly basis and compared with permissible limits.
6.3.2 Blower Effluent Monitor An on-line monitoring system continuously samples the radionuclide content in the rabbit blower exhaust stream. This system takes the exhaust stream through a shielded volume containing a beta scintillator detector. System response for 4'Ar is determined and the detector output recorded on a panel-mounted stripchart recorder.
Instantaneous 41Ar concentration can be obtained, as well as integrated totals over various recording times.
142
6.3.3 Release Points The primary gaseous effluent release point is from the building ventilation fan located at the top of the north wall of the building.
This vent is about 30 feet above floor (ground) level. The ventilation fan creates a building exhaust stream that has been measured at a volume flow rate of approximately 1000 CFM, or about 4.72xl0-cc/second. Assuming a building volume of 70,000 cubic feet, an exchange time of 70 minutes is obtained.
Other release pathways are available, such as through open building doors, windows, the vent fan in the control room, and the fume hood in room 104. However, the total capacity of these release pathways is small compared with the building vent. In addition, the pathways are normally unavailable for release during reactor operation, since building doors are closed, the fume hood is not operated continuously and is used sparingly, and windows, being located in offices and classrooms, are usually closed since these areas are serviced by the building HVAC systems. Only the vent fan in the control room is used a significant amount of the time, and the control room is normally isolated (door is closed) from the main reactor room.
Within the confines of the building, release points for gaseous effluents can be identified. For non-vented experimental facilities, gaseous effluent release is limited since the facilities are either plugged or closed when not in use. Any venting of gaseous radioisotopes will occur at the point where the facility exits from the reactor pool or shielding wall. These points are identified in Table 6.4. The exhaust point for the rabbit blower is located at the end of the exhaust pipe at the top of the building, near the north U
wall.
6.3.4 Estimated Releases in the Restricted Area 6.3.4.1 Types of Releases Release of 41 Ar from experimental facilities can occur as either a puff or, for a vented facility such as the rabbit or the surface of the pool, a continuous stream. Section 6.1.2 discussed the estimated source terms for 4 1 Ar from experimental facilities and 41 Ar and 16N from the surface of the pool, assuming 500 kilowatt operation. The following sections will analyze individual release scenarios and their consequences. These analyses concern releases made within the confines of the reactor building, which is defined as a restricted area.
6.3.4.2 Puff Release from the Rabbit Saturation levels of 41Ar can build up in the effective volume of the rabbit facility during a long reactor operation with the rabbit system blower turned off. Table 6.3 shows a saturation activity of 16.3 millicuries of 4"Ar being in the rabbit volume under these conditions.
The release scenario assumes that the rabbit system blower is then 143
Table 6.4: Release Points of the Various Experimental Facilities of the OSURR Facility Release Point Location Release Aperture Description Description CIF Top of Reactor Pool, Open Aluminum Tube, About 20' Elevation 1.38" Diameter Beam Port 1 North Reactor Bay, Open Port, Flush With About 5' Elevation Shield, 7.125" Diameter Beam Port 2 North Reactor Bay, Open Port, Flush With About 4' Elevation Shield, 7.125" Diameter Rabbit Building North Wall, Open Aluminum Tube, 3" About 28' Elevation Diameter Thermal Column West Side of Building, Open Stringer, 4" x 4" 1'-6' Elevation Square Opening 4" Dry Tube Top of Reactor Pool, Open Aluminum Tube, About 20' Elevation 3.486" Diameter 2" Dry Tube Top of Reactor Pool, Open Aluminum Tube, About 20' Elevation 2.05" Diameter Notes: (1). Elevations shown above are referenced to the floor of the reactor building.
(2) Aperture release points for the two beam ports assume that one or the other is open for an experiment, with no shielding plugs or other apparatus restricting access to the interior of the tube.
(3) The aperture for the thermal column assumes a single stringer position completely opened.
144
activated and the entire activity is instantaneously and perfectly mixed with the 70,000 cubic feet of air in the reactor building. &
Diluting the 16.3 millicuries of 4"Ar in the building atmosphere leads to a concentration of 8.22x10-6 microcuries of "lAr per cc of air in the building. Table I of Appendix B, 10CFR20, shows a DAC of 3 xlO-E -
microcuries/cc for "Ar, which assumes a 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months working year as an available averaging time. The estimated release for this scenario exceeds the DAC by a factor of about 2.74 times. Thus, an effective exposure time of about 730 hours0.00845 days 0.203 hours 0.00121 weeks 2.77765e-4 months at the predicted concentration from the puff release would be allowed under DAC yearly restrictions, assuming that the released concentration remains constant over this time. This assumption, however, is very conservative since the concentration will diminish with time as a result of building purging and radioactive decay.
An accurate analysis of building concentration requires consideration of the reduction in concentration as a function of radioactive decay and building purging. Treatment of this problem is similar to analysis of a radioactive material passing through a biological system, where radionuclide concentration decreases after initial introduction because of physical decay and elimination from the system by purging processes. This leads to the concept of the effective half-life, defined as follows; Te = (Td x Tp)/(Td + Tp) where Te = effective half-life Td Tp
= half-life from radioactive decay, and
= half-life from building purging.
U From the building exhaust rate of 1000 CFM, and a building volume 70,000 cubic feet, a purging time of 70 minutes is obtained. A relatively simple analysis of the inflow and outflow of the building, assuming an equilibrium condition, shows that the value for Tp in the above equation should be 70 minutes. Using this, and assuming a value of 1.83 hours9.606481e-4 days 0.0231 hours 1.372354e-4 weeks 3.15815e-5 months for the radiometric half-life of 4"Ar, an effective half-life of 42.75 minutes is obtained for 4'Ar in the reactor building.
Now, if one assumes that most 4"Ar activity is lost after five effective half-lives have elapsed (213.75 minutes), the average concentration of 4"Ar in the reactor building during this time-can be obtained by integrating the time-dependent concentration over this time:
t:
Cave = C,' e- tdt / (t - t,)
where Cave = average concentration of 4"Ar in the Reactor building during the time interval from t1 to t2 ,
145
cc= initial 41 Ar concentration, t = time at the beginning of the release (0 minutes),
t = time at the end of the release (213.75 minutes),
A = decay constant (0.693/T, = 1.62xlO min-)'.
Performing this integration leads to the following expression for average concentration:
C.,.= (Co/Xt) [1-e t)
Substituting t = 213.75 minutes, and appropriate values for Co and X, an average concentration of 2.3x10-6 microcuries of 41Ar per milliliter of building air is obtained. This average 41Ar concentration is below the DAC limit of 3x10-6 p.Ci/ml. This calculation is conservative in that it assumes a saturation activity in the rabbit effective volume being available for a puff release.
6.3.4.3 Continuous Release from the Rabbit Section 6.1.2.1 estimated a source term of 1.72 microcuries of 4"Ar per second being produced in the exhaust of the rabbit blower at 500 kilowatts. Expressing the concentration buildup of the isotope in the air of the reactor building, accounting for losses from radiological decay and building purging, leads to an equation similar in form to the production of a radioactive material by neutron irradiation, assuming a constant term for isotope production:
C(t) = P(1 - eXt)/(XV) where C(t) = time-dependent concentration of 4"Ar in the building air at time t after starting the rabbit blower, P = production rate of 4"Ar in the rabbit blower exhaust stream, t = time the blower has been running, X = effective half-life defined earlier, and v = building volume.
Figure 6.1 shows the time-dependent behavior of the 4"Ar concentration in the building air. The concentration approaches an equilibrium value when t becomes large. For conservatism, assume that this equilibrium value has been reached. Substituting appropriate constants in the above equation leads to an estimate of 3.21x10-6 microcuries of 4"Ar per milliliter of air in the building at equilibrium. This about 7% above the DAC limit for a restricted area, which limits the exposure time at this concentration to about 1869 hours0.0216 days 0.519 hours 0.00309 weeks 7.111545e-4 months per 2000-hour working year.
Essentially, the rabbit blower may run with the reactor at full power for most of the working year. It is difficult to postulate a reactor operation involving continuous rabbit blower operation for anything on the order of 1869 hours0.0216 days 0.519 hours 0.00309 weeks 7.111545e-4 months at a time.
146
C H -- ---
'-4 ba /
C(X) 3.21 pCi/ml (Rabbit Blower Only)
., /C(") 3.62 pCi/ml (Pool Surface Only)
C(X) = 6.83 pCi/ml (Rabbit + Pool Surface) 0j UK 4J I 0 024 487 29 6 2 120 144 168 192 216 240 264 288 312 336 Exposure Time In Minutes Figure 6.1: 41Ar Production Curve 147 C C. C
Another way of analyzing continuous 4'Ar release from the rabbit is to set C(t) in the above equation equal to the DAC limit (3x10-6 microcuries per milliliter) and solve the equation for the rabbit blower operation time (t). Doing this, a blower operation time of about 10033 seconds is obtained. Thus, the rabbit blower may operate about 2.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months before a DAC level of "Ar is reached in the building air. Most rabbit operations involve blower run times on the order of 20 minutes or less.
6.3.4.4 Puff Release from the Rabbit Carrier Tube Reactor operations involving rabbit irradiations require use of a carrier tube to insert and remove samples from the rabbit facility.
Each irradiation therefore involves a puff release from the carrier tube when it is opened. Since this is a commonly-performed operation, it is analyzed separately here.
Table 6.3 lists a source term for saturation 4"Ar in the rabbit carrier tube of 1.11x103 microcuries. Assuming that this source is instantaneously and perfectly mixed with the air in the reactor building, a concentration of 5.6x10-7 microcuries per milliliter results. This is less than the DAC of 3x10- microcuries/ml allowed for 4'Ar in a restricted area. Therefore, puff releases from the rabbit carrier tube are allowable even if the activity in the tube has reached saturation levels (which is unlikely in almost all conceivable rabbit operations).
6.3.4.5 Puff Releases from Other Experimental Facilities Using the source term estimates from Table 6.3, a similar analysis for puff-type releases from the other experimental facilities was done.
Table 6.5 shows the estimates for initial and average concentration of 4"Ar in the reactor building atmosphere. The average concentration assumes an averaging time equal to five times the effective half-life of 4"Ar in the building. The last column shows the relationship between the average concentration in the building air and the allowable DAC limits for a restricted area (factor = average concentration/DAC).
Note that puff releases from the rabbit, the rabbit carrier tube, the thermal column, and the 2 inch dry tube result in average "Ar concentration less than restricted area DAC for releases averaged over five effective half-lives. Thus, under these conditions, no operational limits need to be established.
Table 6.6 shows operational limits on facility use for a 2000-hour work year. The last column in Table 6.6 shows the length of operation time necessary to attain a DAC level of "Ar in the building, while the other column assumes a saturation level released in the initial puff and predicts the allowed exposure duration at the resulting concentration in the building.
Assuming that a single saturation-level puff release occurs over a time equal to about five effective half-lives (213.75 minutes),
148
. Table 6.5: Estimated 4'Ar Concentrations for Puff Releases of Saturation Activities of 4 1Ar from the Various Experimental Facilities of the OSURR Facility Initial 41Ar Average "Ar DAC Factor For Description Concentration Concentration Average (Lcuries/ml)~(curies/ml) Concentration CIF 1.23 x 10-5 3.44 x 106 1.15 Beam Port 1 8.27 x 10-5 2.31 x 10- 7.70 Beam Port 2 6.36 x 10- 1.78 x 10-- 5.93 Rabbit 8.22 x 10-' 2.30 x 10-E 0.77 Thermal Column 4.71 x 10-5 1.32 x 10- 0.44 Rabbit Carrier 5.60 x 10-7 1.57 x 10-' 0.05 4" Dry Tube 1.63 x 10 5 4.56 x 1O-6 1.52 2" Dry Tube 5.65 x 10-6 1.58 x 10-6 0.53 Q.)
Notes: (1) The initial concentrations shown above assume instantaneous and perfect mixing with the building air.
(2) The average concentrations shown above assume losses of initial activity from radioactive decay and building purging. The effective half-life is taken to be that derived in the accompanying text. The release is averaged over five effective half-lives.
(3) The DAC factor shown above is calculated by dividing the average concentration by the DAC for "Ar in a restricted area (3xlO-1 Rcuries/ml).
149 U
Table 6.6: Operational Limits and Activation Time Estimates for the Various Experimental Facilities of the OSURR Facility Limit of Full-Power Activation Time Required To Description Hours of Operation Attain Average 4"Ar Per Year Concentration Equal to DAC CIF 1739 329 minutes Beam Port 1 260 10 minutes Beam Port 2 337 14 minutes Rabbit- 2608 No Limit Thermal Column 4545 No Limit Rabbit Carrier 45000 No Limit 4" Dry Tube 1316 170 minutes 2" Dry Tube 3774 No Limit Notes: (1) Where the limits on full-power operation exceed 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months per calendar year, there are no limits on facility use under the assumptions of this analysis.
(2) The limits expressed above assume a puff release at saturation activity levels in a completely voided effective volume of the facility.
(3) The limits on full-power operation are in fact effective exposures times allowed at the estimated concentration. DAC limits are not exceeded if exposure times are less than or equal to this value.
150
dividing this release time into the calendar.year hours of operation limit indicates how many releases of this type may be made during the calendar year. For example, the releases per calendar year for Beam Port 1, which has the lowest hourly limit, yields about 73 puff-type U
releases per year to stay within DAC limits. Similarly, for puff releases from the rabbit, about 732 releases are allowed. In all cases, based on previous operating history, it is unlikely that the total puff releases from these facilities during a calendar year will exceed these estimates.
41 6.3.4.6 Continuous Release of Ar from the Pool Water Section 6.1.2.2 estimated a release rate from the pool water for 4"Ar of:
S41 = 6.81x108 atoms/second.
Multiplying this release rate by the radiological decay constant for 4"Ar (1.0519x10-4 second-) and converting activity to microcuries leads to an estimate of 1.936 microcuries/second released from the pool.
Using the concept and equation developed in Section 6.3.4.3 for continuous discharge of 41 Ar from the rabbit blower, an equilibrium concentration of 3.62x10-6 microcuries of 41 Ar per milliliter of building air. This is about 21% above the DAC for "Ar in a restricted area, and thus would limit reactor operation (effective exposure time) to 1657 full-power equivalent hours each calendar year. Since this would average out to about 6.6 full-power equivalent hours of operation each day, it is unlikely that OSURR operation would result
& X in exceeding restricted area DAC for 41Ar.
6.3.4.7 Combined Continuous Release from the Pool & Blower The release rate of 1.936 microcuries of UAr per second calculated above for the pool water can be added to the rabbit blower source term estimated in Section 6.1.2.1 of 1.72 microcuries/second to yield a combined continuous source term of 3.656 microcuries of 41 Ar added to the building air per second. Using the equation developed in Section 6.3.4.3 and substituting appropriate constants gives an equilibrium concentration of 6.83x10-¢ microcuries of 41 Ar per milliliter of building air. This is a factor of about 2.28 above the DAC allowed in a restricted area. Using an averaging time of 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months , an effective exposure time of 878 hours0.0102 days 0.244 hours 0.00145 weeks 3.34079e-4 months is allowed. Thus, operation with the reactor at full power and the rabbit blower running is restricted to this number of hours each working calendar year. It allows an average of about 3.5 full-power equivalent hours each day of combined full power reactor operation with the rabbit blower running continuously.
Solving the characteristic equation for run time, assuming the production rate above, leads to a run time of about 2141 seconds, or about 35 minutes, to attain a DAC level in the building air. Again, most rabbit operations will involve run times of 20 minutes or less.
151
6.3.4.8 Continuous Release of i5N from the Pool Water Section 6.1.2.3 provided calculations to show that the release rate of 16N from the core is about 2.01x106 atoms of l1N per milliliter of water per second, or a total of 9.34x107 atoms of 14N per second (24.64 millicuries/second). This source term must be diluted to account for delay in traversing the distance from the top of the core to the surface of the pool.
Using the average coolant velocity through the core of 6.5 cm/second noted in Section 4.8 and assuming a constant average coolant velocity from a point immediately above the core to the surface of the pool, the total transit time is obtained by dividing the distance from the top of the core to the surface by the velocity. Given that the minimum depth of water in the reactor pool is 15 feet (457.2 cm), a total transit time of 70.34 seconds is obtained. This estimate is conservative in that it ignores delay times resulting from pool water mixing and dispersion by the cooling system pump. Essentially, it assumes the cooling system dispersion pump is off, but the reactor is at full power. This condition is prohibited by the reactor safety system, which initiates a reactor trip if the power rises above 100 kilowatts with the primary coolant pump off.
The 16N concentration at the pool top can thus be estimated from:
C CoeX where C = concentration of '6N atoms at the pool surface, X = decay constant for 16N, Con concentration of 16N atoms immediately above the core, and t transit time from the core to the pool surface.
Substituting appropriate constant in this equation yields a concentration of 2096.9 atoms of 16N per milliliter of pool water at the surface of the pool.
As the '6 N-bearing water reaches the surface of the pool, it spreads across the surface in the shape of a disk, forming an area source of radiation and a release interface to the building air. For the purpose of this calculation, assume that the disk has a radius of 85 cm, which is about the width of the reactor pool. The time the water takes to spread across this area, assuming a 6.5 cm/sec constant velocity is t = r/v = (85 cm)/(6.5 cm/sec) = 13.07 seconds.
During this distribution time, the concentration of 16 N decays from that initially available in the rising plume from the core. The average concentration of N across the surface of the disk source given by:
152
N = - JNoe AIdt = "I1 -
C)t where t spreading time across disk surface, A =decay constant for 16N, and N initial "6N concentration at the pool surface.
From earlier calculations, we find that N = 2096.9 atoms/ml, X 9.761x10-2 second-', and t = 13.07 seconds. Performing the calculation gives an average disk source concentration of 1184.8 atoms of 16N per milliliter in the disk source at the surface of the pool.
For estimation of the release rate of gaseous radionuclides to the air of the building, the number of 16N atoms diffusing from the surface of the disk source to the air must be estimated. Dorsey [5] reports an escape velocity of 9x10-3 cm/second for nitrogen atoms from water.
Multiplying this by the average concentration of nitrogen in the disk source gives:
S = Cv = (1184.8 atoms/cc) (9x10-3 cm/sec)
= 10.66 atoms/cm2 /second.
The total release rate of atoms of '6N to the building air would be the area of the disk (85 cm equivalent radius) times the emission rate per unit area noted above. The release rate to the building air is thus 2.4196xlO 5 atoms of i6N per second.
As the ioN atoms enter the reactor room air, their concentration is reduced by dilution into the volume of the building, exhausting through the ventilation fan, and radioactive decay. Since the half-life of 16 N is very short (7.1 seconds) compared to the building purge time (70 minutes, or an effective half-life for air in the building of 70 minutes), the radioactive half-life will dominate the effects of concentration reduction resulting from decay. The rate of buildup of 16N in the building air is given by:
d(VN) = S -i+ 4/
dt where S = release rate of '6N atoms to the air, X= decay constant of 16N, q = building ventilation rate, N = concentration of 16 N in the building air, and V = building volume.
Under equilibrium conditions, the time rate of change of the concentration is zero, and the above equation can be solved for N:
N = S/(XV + q) 153
Substituting known values in this equation gives an equilibrium concentration of 1.25xlO-1 nuclei of 16N per milliliter of building air.
Converting this to activity units gives a concentration of 3.3xlO' microcuries of 16N per cc of building air. Table I of Appendix B, 10CFR20 does not have a listing for 1'N. However, it states that, "Any single nuclide not listed above with decay mode other that alpha emission or spontaneous fission and with radioactive half-life of less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ", has a DAC of lx10- gCi/ml. The calculated value for 16N concentration is well within this bound.
For this calculation, assume that the volume of the building is represented by a hemisphere with an effective radius of 980 cm (about 32 feet). The dose rate resulting from the 16N concentration calculated above, dispersed in this volume, can be estimated by:
R = KjNR/2K 2 where K1 = 3.7x104 photons/second/microcurie, N = 1°N concentration in the building air, R = effective building radius, and K2 = 1.6x105 photons/sec-cm2 /rad-hr.
Performing this calculation yields an estimate of 3.75x10-7 rads/hr in the building from dispersed 16N, or about 0.4 microrads/hr.
6.3.4.9 Actual 4"Ar Releases i The actual releases of 41Ar into the restricted area can be calculated from an effluent monitor near the intake of the building exhaust fan.
It will be assumed that the concentration of "Ar measured by this monitor is representative of the concentration in the reactor building. This is a conservative assumption in that the outlet for the rabbit is very near the intake for the effluent monitor, which will result in a higher reading than that of the rest of the building. In the half-year period from 1/1/99 to 6/30/99, the effluent monitor at the building exhaust fan measured a net count of 1,923,402 counts. The calibration for this monitor is 19.3 counts/second corresponds to 3x10-6 pCi/ml. Assuming that 1/2 of the work year is 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months , this gives a concentration of 1,923,402 counts _
3x10 6 pOCi/ ml =
8 pi / ml 8.30Ox10-~L 1000 hr
3600sec//hr 19.3 counts/sec This is well below the DAC limit for 4"Ar in restricted areas of 3xlO-1 gCi/ml.
kJ 154
6.3.5 Releases from the Restricted Area Once gaseous radionuclides are released to the building atmosphere, they begin to be discharged to the outside environment by the building ventilation fan. As noted in the earlier analyses, this fan has a measured capacity of 1000 CFM, which results in a building purge time of 70 minutes. The exhaust point is approximately 32 feet above ground level, at the roofline of the building along the north wall. The exhaust stream exits the building parallel to the ground.
The following sections will consider the dilution factors available for the building releases and analyze several release cases.
6.3.5.1 Dilution Factor Radionuclides contained in the building exhaust stream will mix with the outside air in the lee of the building. The dilution resulting from this mixing effect can be described as:
AD = AqV(x) where AD = effective exposure concentration in curies/m 3 ,
q = building exhaust rate in m3 /second,
, 3
=(x)dilution factor at distance x, in sec/M , and A = activity concentration in the exhaust stream.
The dilution factor is computed for the lee of the building (x=0)j and assumes that the release is made from the roofline of the building.
Further, assume that the wind velocity is steady at the time of the release and is equal to 1 m/sec. The dilution factor can be written U
as:
% (0) l=/[ (0.5) (s) (u)]
where u = wind velocity in m/second, and s = building cross-sectional area normal to the wind direction in m.
Assuming that the prevailing westerly winds are blowing at the time of the release, a normal cross-sectional area of 201.6 m2 is presented to the wind. Substituting values into the above equation indicates a dilution factor of %V(O) = 9.921x10-3 second/m 3 .
The building exhaust rate is assumed to be 1000 CFM, or 0.47195 m 3 /second. Using this for q in the above equation and the value of
%V(0 ) calculated above, values for A can be substituted for various release cases. In the puff release cases analyzed in the following sections, the exhaust stream activity concentration is taken to be the average concentration over the release period shown in Table 6.5, column 3.
155 (
6.3.5.2 Puff Release from Various Facilities wo Using appropriate data in the equations derived above, various concentrations in the air on the lee side of the reactor building were calculated. The results are shown in Table 6.7. The Effluent Concentration Limit for 4"Ar in unrestricted areas is specified in Appendix B, Table II, Column 1 of 10CFR20 as lxlO-6 microcuries/ml.
To maintain releases to the outside air within Effluent Concentration Limits, either the source term must be reduced or operational limits imposed so that when the releases are averaged over the permitted averaging time, the average concentration does not exceed the limit.
For releases to unrestricted areas, 10CFR20 specifies an averaging time of one year. Table 6.8 shows the results of calculations based on these limits.
If one assumes that the releases occur over a time equal to 213.75 minutes (five effective half-lives of 4"Ar in the building), dividing the operational limits shown in Table 6.8 by this gives an approximate number of puff-type releases of this type allowed in a year. About 228 releases are allowed for Beam Port 1, and 295 releases are permitted for Beam Port 2. Considering Beam Port 1 as the most restrictive case, 228 releases each year would require an average of 4.38 releases of this type each week. It is very unlikely that OSURR operations would result in this frequent a release rate, so it is unlikely that unrestricted area limits will be exceeded.
It should be noted that the calculations shown above are conservative t) in that they assume a saturation activity source term in the puff releases, with the entire beam tube volume being air void. Generally, lower initial source activities will be available for release, since reactor operation times less than that necessary to achieve saturation activities of "Ar (about 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months ) are more common than those exceeding this duration, and many, if not most, reactor operations will have much less than the full volume of the beam tube voided. Thus, lower overall concentrations in the building and in the outside air will result.
6.3.5.3 Continuous Release from the Rabbit Blower The calculations shown in Section 6.3.4.3 indicate that an equilibrium concentration of 3.21x10-6 microcuries/ml of 4"Ar in the air of the reactor building will result from continuous operation of the rabbit blower. Equilibrium concentrations will be present after about five effective half-lives (213.75 minutes). At equilibrium, the equation used in Section 6.3.5.1 will predict equilibrium concentrations in the air on the lee side of the building. Substituting appropriate values leads to an estimate of 1.5xlO 6 microcuries/ml of 4"Ar. This is above the limit for 4"Ar in an unrestricted area, which means that full-power reactor operation is limited to 5844 hours0.0676 days 1.623 hours 0.00966 weeks 0.00222 months per year. This is well above the amount of time actually spent running at full-power in a year.
156
Table 6.7: Estimated 41Ar Concentrations in the Lee of the Reactor Building for Puff Releases of 4"Ar from the Various Experimental I Facilities of the OSURR Facility Average Exhaust Outside 41Ar DAC Factor For Description Concentration Concentration Average
(>curies/ml) (gcuries/ml) Concentration 8
CIF 3 . 4 3 xlOf. 1.61x10- 1.16 Beam Port 1 2.31xlO-- 1.08x10' 10.80 Beam Port 2 1.78xl -Q 8.33x.O1- 8.33 Rabbit 2.30x10-6 1.08x10 8 1.08 Thermal Column 1.32x10-6 6.18xl10 9 0.62 Rabbit Carrier 1.56x10-7 7.30x10-10 0.07 4" Dry Tube 4.57x10-6 2.14x10-8 2.14 2" Dry Tube 1.58x10-6 7.40x10 9 0.74 U
Notes: (1) The average exhaust stream concentration was taken from column 3 of Table 6.5. It assumes a puff release from the building and a purge time equal to five effective half-lives of 4"Ar in the building.
(2) Of the above cases, releases from the CIF, beam ports, rabbit, and 4" dry tube result in outside concentration greater than the effluent concentration limit over the averaging time.
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Table 6.8: Operational Limits for the Various Experimental Facilities of the OSURR to Maintain Unrestricted Area Effluent Concentration Limits on the Lee Side of the Reactor Building Facility Limit of Full-Power Description Hours of Operation I i; Per Year i . I CIF 7556.9 Beam Port 1 811.7 Beam Port 2 1052.3 Rabbit 8116.7 F I
-Thermal Column No Limit Rabbit Carrier No Limit 4" Dry Tube 4096.3 2" Dry Tube No Limit Notes: (1) The limits expressed above assume a puff release at saturation activity levels in a completely voided effective volume of the facility.
(2) The limits on full-power operation are in fact effective exposures times allowed at the estimated concentration.
Effluent Concentration Limits are not exceeded if exposure times are less than or equal to this value.
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6.3.5.4 Continuous Release from the Pool Water Section 6.3.4.6 noted that a building equilibrium concentration for 4Ar of 3.62xl0-6 microcuries per milliliter would result from releases from the pool water. The equation shown in Section 6.3.5.1 predicts an outside air concentration of 1.62x1O-8 microcuries of "Ar per milliliter of air, which is also above the limit for 4"Ar in an unrestricted area. This restricts operation to 5411 hours0.0626 days 1.503 hours 0.00895 weeks 0.00206 months per year.
Section 6.3.4.8 gave an estimate of 3.3x10-5 pCi/ml for 16N restricted area concentration. Using the calculation from Section 6.3.5.1, this results in a unrestricted area concentration of l.55xlO-10 RCi/ml. This well below the Effluent Concentration Limit for unrestricted areas of lxlO09 RCi/ml. Most likely, the actual concentration will be far below that calculated since the seven second half-life of 16 N will result in most of it decaying away before it reaches ground level.
6.3.5.5 Combined Pool Water and Rabbit Blower Releases If the source terms for continuous 41Ar production from the rabbit blower exhaust and pool water are added, the outside air concentration at equilibrium is estimated to be about 3.20xl0- microcuries of 4"Ar per milliliter. The combined 41Ar release restricts the reactor to 2739 hours0.0317 days 0.761 hours 0.00453 weeks 0.00104 months of full-power operation per year, which is still well above the number of hours actually run.
41 6.3.5.6 Actual Ar Released Using the effluent monitor data for 1/l/99 to 6/30/99 along with the calculation method shown in Section 6.3.5.1 gives an 41Ar outside air concentration of 8.87xlO-1 p.Ci/ml (assuming that half of a year is 4383 hours0.0507 days 1.218 hours 0.00725 weeks 0.00167 months ). This is well below the limit of lx10-8 .Ci/ml given for "Ar for unrestricted areas.
6.3.6 Steps to Limit Release Levels Although the calculations noted in the preceding sections are conservative, it is possible to take steps to limit releases even further. This section will discuss some of these actions and their effects.
6.3.6.1 Reducing Effective Irradiated Volumes In most cases, the volumes considered in the preceding calculations will be larger than those normally irradiated during routine operations. For example, the two beam ports are normally filled with shielding plugs, and the thermal column is filled with graphite stringers when it is not being used in an experiment, which essentially eliminate their effective volumes. Devices or samples being irradiated in the dry tubes, thermal columns, or CIF will also 159
reduce the effective irradiated volumes. Also, the dry tubes are J~ usually stored away from the core when not in use, so their effective volumes are not irradiated during routine operations. Air voids in experiments placed in these facilities can be limited by packing them with inert, non-activating materials (e.g., the rabbit carrier tube is packed with cotton). Only the rabbit and CIF effective volumes are irradiated when they are not in use during routine operations.
6.3.6.2 Facility Purging The rabbit facility can be purged at any time during normal operations. The other experimental facilities can also be purged by insertion of apparatus to circulate gas through the effective volume of the facility. Purging the experimental facility limits the buildup of 41Ar in the effective volume, thereby reducing the initial concentration that might be released in a puff-type expulsion of air from the facility. Limiting the purging rate can keep equilibrium concentrations of 4"Ar in the restricted area within DAC limits.
The experimental facilities can also be purged with nitrogen gas. When irradiated by neutrons, nitrogen undergoes very little neutron capture reactions leading to radioactive products. Replacement of air in the experimental facilities with nitrogen thereby limits the concentration of 4"Ar in the building air. Nitrogen gas can be introduced to the facility prior to irradiation, or in a continuous stream by gas lines inserted during the time the facility is in use.
6.3.6.3 Limiting Facility Releases Release of "Ar from irradiated air volumes can be reduced by sealing the facilities against leakage to the atmosphere for a time sufficient to allow decay of the isotope. For example, the boral plate on the outer surface of the main graphite thermal column has a rubber gasket along its inner edge, allowing it to be hermetically sealed against the outer surface of the facility. If the experiment allows, the facility may be kept sealed until "4Ar concentration has been reduced by radioactive decay. Similarly, the CIF can be left plugged for a time to reduce 4'Ar activity, if allowable under the conditions of the experiment. For neutron activation experiments resulting in relatively long-lived radioisotopes, overnight decay periods are generally acceptable. In this case, for example, 4lAr activity in the CIF effective volume would be reduced by about 99% from that initially present at the time the irradiation ended.
6.3.6.4 Ventilation System Control The building ventilation systems can be turned off by a switch in the control room. Such an action can thereby limit releases of gaseous radionuclides to the unrestricted area. If more rapid purging of the building is desired, ventilation rates can be increased by activating portable fans and allowing them to exhaust out of opened doors or windows. Any such releases will be recorded, and the calculated 160
release concentrations of gaseous radioisotopes averaged into the-total yearly release to the unrestricted area.
6.3.7 Estimated Doses Release of 4"Ar from the experimental facilities of the OSURR will result in accumulated doses to persons exposed to the resulting isotope concentration in the air. The whole-body gamma ray dose to a person surrounded by and immersed in a semi-infinite cloud of radioactive gases can be approximated by:
D = 900EAD where D = dose rate in rads/hr, E = photon energy in MeV, and AD = effective exposure concentration in curies/m3 Earlier calculations and tables listed values for AD given various types of releases. Assuming a gamma photon energy of 1.3 MeV, exposure rates can be estimated. Table 6.9 summarizes the results of these calculations.
6.4 Solid Radioactive Waste Management Operation of the OSURR will generate very little solid low-level radioactive waste. The primary source of low-level solid waste will be the demineralizer cartridge in the reactor pool water processing system. Current procedures call for cartridges to be kept on-site for a decay period sufficient to reduce the activity of short-lived radionuclides (e.g., 24 Na) to negligible 1-oe levels. If higher-power U-operation of the OSURR results in additional radioisotopes being present in the resins of the demineralizer cartridge with longer half-lives, a bulk radioassay will have to be performed to determine specific and total activity in the cartridges prior to disposal.
Disposal of this low-level waste is handled by the OSU Radiation Safety Section and comprises a few cubic feet per year.
Spent fuel assemblies might also be classified as solid radioactive waste. These are stored in the fuel storage pit at the east end of the reactor pool, unless otherwise approved by the Reactor Operations Committee and the Nuclear Regulatory Commission. After suitable decay times have elapsed, spent fuel assemblies are returned to the Department of Energy for ultimate disposal. Because of their isotopic content and activity inventory, used fuel assemblies are not considered low-level radioactive waste, and are therefore handled separately from other waste forms generated by the laboratory. Spent fuel shipment is performed in accordance with approved procedures that meet appropriate federal, state, and local requirements.
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Table 6.9: Estimated Dose Rates Resulting From 4"Ar Concentrations As A Result Of Releases From Various Experimental Facilities of the OSURR Facility Type of Restricted Area Unrestricted Area Description Release Dose Rate (rads/hr) Dose Rate (rads/hr)
CIF Puff 4.02 x 10-3 1.36 x 10--
Beam Port 1 Puff 2.70 x 1O- 1.26 X lo0-Beam Port 2 Puff 2.08 x 10-' 9.75 x 10--
Rabbit Puff 2.69 x 10- 1.26 x 10--
Rabbit Continuous 3.76 x 10-3 1.76 x 10--
Pool Water Continuous 4.24 x 10-3 1.90 X iO-'
Pool & Rabbit Continuous 7.99 x 10-- 3.74 x 10--
Thermal Column Puff 1.54 x 10-3 7.23 x 10-'
Rabbit Carrier Puff 1.83 x 10-4 8.54 x 10-'
4" Dry Tube Puff 5.34 x 10-3 2.50 x 10-5 2" Dry Tube Puff 1.85 x l0-, 8.66 x 10-i Notes: (1) The puff releases calculated above assume saturation activities have been attained. The continuous discharge from the rabbit facility assumes a blower operation time sufficient to attain equilibrium concentrations in the building.
(2) The averaging time for the concentration used in the calculations is assumed to be five effective half-lives of 41 Ar in the building (213.75 minutes).
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6.5 Liquid Radioactive Waste Management Section 6.1.1 noted that no liquid-borne radioactive materials are discharged from the OSURR during normal operation. However, certain maintenance and repair activities may result in liquid discharges.
An alternative to demineralizer cartridge regeneration by the manufacturer or replacement of the resins by OSURR personnel would involve on-site regeneration of the cartridges by OSURR staff. In. this event, some liquid-borne radionuclides would result. These eventually would have to be released from the reactor building as liquid radioactive waste.
Prior to release of liquid radioactive waste, the isotopic content of the material and specific activities of radioisotopes present in the liquid must be determined. The liquids may then be kept in a holding tank to allow decay to reduce the total activity inventory, or make it available for dilution or treatment prior to release.
6.6 Byproduct Materials Management Operation of the OSURR will result in the production of radioactive materials as part of experimental procedures. Production of radioactive materials by neutron activation, also called byproduct materials, may be a deliberate result of the experiment (as would be the case in isotope production or neutron activation experiments), or incidental to the experiment (as would occur in materials damage studies or medical experiments). In either case, the radionuclides so produced must be handled safely.
Laboratory procedures are available for survey and assay of all materials irradiated in the OSURR. Handling and storage of activated materials is also governed by laboratory procedures. These procedures fall within the overall university guidelines for working with radioactive materials, which themselves are designed to meet or exceed the requirements specified in 10CFR20.
Typical byproduct materials would include a variety of beta and gamma-emitting radioisotopes of various half-lives, generally formed by thermal neutron-induced activation of parent (target) nuclei. A few radionuclides are formed by fast-neutron capture reactions, such as 58 Co (from the neutron-proton reaction with 58 Ni) and 24Na (from the neutron-alpha reaction with 27 Al). A few byproduct materials emit alpha particles. Half-lives can range from seconds up to years. Most byproduct materials are in solid form, but a few are liquids, and very few are gases. Experimental procedures call for activation targets to be encapsulated, where possible. Induced activity can result in dose rates ranging from a few tenths of a millirem/hour up to a few rem or tens of rem per hour. Radiation safety procedures are followed for dealing with sources producing intense radiation fields and significant dose rates.
163
Radioactive materials are typically handled in the northeast corner of w J the building, which is designated as the radioactive materials handling and storage area. Other areas of the building may also be used for handling and storage if they are so designated and posted.
Handling and storage procedures are available to assure safety in these activities. Storage of irradiated materials, if necessary, can be facilitated by using appropriate shielded containers. A variety of lead storage containers, of various shapes and sizes, are available in the laboratory.
Solid byproduct materials, if they contain essentially no transuranic radionuclides, are disposed of in designated containers. Transuranic materials are handled separately, but the quantities are generally very small. A record is kept of materials placed in the disposal container. The container is collected when required by Radiation Safety Section personnel, and added to the university total waste inventory. Liquids are released, if permitted within the framework of 10CFR20 limits, into a designated disposal sink ("hot" sink). Records are kept of liquid disposals. Liquid sources may be evaporated and disposed of as dry materials, if the radionuclides they contain are non-volatile. Gaseous radioactive materials may be vented to the air, if such procedures do not exceed averaged annual effluent release concentration limitations, or the containers holding the gases, if tight, can be disposed of in the waste container.
Miscellaneous radioactive waste, such as contaminated gloves, tools, apparatus, or absorbent pads are considered as solid byproduct waste materials and are disposed of in the waste container. The items are surveyed for dose rate prior to disposal.
The area of the building where the radioactive waste disposal container is located is surveyed for gamma dose rates are part of the area radiation surveys performed routinely at the laboratory. If necessary, the area is posted as a Radiation Area or High Radiation Area. The container is posted as a Radioactive Materials storage area.
6.7 Chapter 6 References (1] David T. Goldman, "Chart of the Nuclides", The General Electric Company, Schenectady, N.Y., 1965.
(2] Harold Etherington, Editor, "Nuclear Engineering Handbook", First Edition, McGraw-Hill Book Company, Inc.,
New York, 1958.
(3] Nicholas C. Horning, "Measurement of the Neutron Spectra in the Beam Ports and Thermal Column of The Ohio State University Research Reactor", M.Sc. Thesis (unpublished),
The Ohio State University, Columbus, OH, 1976.
(4] Joseph W. Talnagi, "Investigation of Perturbation Induced 164
by Neutron Detectors in the Spatial and Energy Dependent Neutron Flux of The Ohio State University Research Reactor", M.Sc. Thesis (unpublished), The Ohio State University, Columbus, OH, 1979.
[5] Noah E. Dorsey, "Properties of Ordinary Water-Substance In All Its Phases: Water-Vapor, Water, And All The lees",
Hafner Publishing Co., New York, NY, 1940.
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7.0 Radiation Protection Q 7.1 Radiation Sources During Operation Operation of the OSURR will create a source of radiation in the form of particles emitted from the core. In all conceivable cases, gamma ray emission and neutron radiation will be the primary components of radiation fields observed external to the core. Beta and alpha particles will be entirely absorbed by the materials of the pool walls, water covering the core, and materials of the core itself.
The source of gamma rays includes both prompt gammas from the fission of 235U, and decay gammas from the fission product inventory in the core. Neutrons will also result from 235 U fission and delayed neutron emitters in the fission product inventory. The intensity of the prompt neutron and gamma is proportional to reactor power, while the intensity of the delayed neutrons and decay gammas is a function of the power history of the core and, if the reactor is shutdown, the elapsed time between the time of observation of the shutdown field and the time of the last reactor shutdown.
7.1.1 Direct Exposures Exposure to radiation directly from the core can occur in several ways. First, gammas and neutrons from the core can penetrate the walls of the pool or water covering the core and cause a dose to be accumulated by an exposed individual. A person can also be exposed to Q core radiation through an opened experimental facility. These cases will be considered in this section.
7.1.1.1 Gamma Dose From The Core Through Shield Walls For this calculation, the core was modeled as a sphere with the sample volume as the rectangular solid that is the actual core geometry. This approach is considered valid since it is used in the Engineering Compendium on Radiation Shielding 11) for calculating the gamma doses from the core of the Bulk Shielding Reactor (BSR), which is similar in design to the OSURR. In the OSURR model, the effective spherical volume of the core is covered with 15 feet of light water, surrounded on all other sides by 1 foot of light water, then surrounded by 6 feet of barytes concrete. Figure 7.1 shows the representation of the core and shielding materials used in this calculation.
The energy spectrum of fission-induced gamma rays was considered to have a representative set of 8 lines, centered at energies of 1, 2, 4, 6, and 8 MeV. This spectrum was also used to represent the gamma ray energy spectrum of fission product gamma emission.
The intensity of the gamma source distributed through the core volume is estimated by considering the energy released per fission in the form of gamma radiation. For 500 kilowatt operation, the fission rate can be approximated by the following formula:
166
R = PK:K K,,
where P = reactor power in megawatts, K, = 106 joules/MW-second (conversion factor),
K2 = 1 fission/200 MeV (conversion factor), and K3 = lMeV/1.6xlO 3 joules (conversion factor).
Using P = 0.5 megawatts, this calculation gives a fission rate of 16 approximately 1.56x0O fissions per second. Using a volume of 88490.2 3
cm for the effective volume of the core, the fission rate density is estimated to be 1.77x10'1 fissions/second/cm3.
The intensity of the volume-distributed gamma source is calculated from:
S, = DK4 where D = fission rate density calculated above, K4 = 8.6 MeV/fission (conversion factor for gamma energy released per fission event), and S.: = gamma source intensity.
This calculation gives a gamma source intensity of 1.52x10 12 MeV/second/cml.
As the gamma rays are released from the core, they interact with the various materials making up and surrounding the core. The absorption of the gamma rays in each of these materials is a function of the W energy of the gamma ray, the thickness of the material, and the absorption properties of the material. The materials to be considered include the water, aluminum, and uranium making up the core, the surrounding water, and the barytes concrete making up the walls of the pool.
The absorption properties of the core can be estimated by averaging the linear attenuation coefficients of the various materials, weighted by the percentage volume of each material. For this calculation assume that 65% of the core volume is light water, 25.5% is aluminum, and 9.5% is uranium.
At 1 MeV, K4 = 8.6 MeV/fission, which gives a gamma source intensity of 1.52x10' 2 MeV/second/cm3 . At 2, 4, and 6 MeV, K4 is 5.41, 1.65, and 0.392 MeV/fission, respectively. These give gamma source intensities of 9.57xlO", 2.91xlO", and 4.35x101 0 for the 2, 4, and 6 MeV components, respectively.
Now, combining the gamma ray point kernel, the exponential buildup function, and the basic point kernel integral, and also assuming a constant source, spherical geometry, and buildup occurring in the concrete shield (pool walls) only (this is reasonable since water is between the core and pool walls, and buildup in the water is small 167
compared with that in the pool walls), the following equation can be
used is estimate the gamma flux

+(z) = S /2) (R,/ (R. + z]) CAE.(4z(l + a) + (1 - A)E.(.Lz(1 + B))]
where S = gamma source intensity, p = attenuation factor for core or pool wall, R, =core (source) radius, z distance from the core, and El = integral function calculated from:
E.(x) = x f (e /t )dt 0
In the above equation, A, B, and a depend on the material properties.
This is the form of the shielding equation used in the Engineering Compendium on Radiation Shielding [1] to analyze the shielding requirements for the BSR at Oak Ridge.
Once the gamma ray flux is found with the above equation, the dose rate at a particular location can be estimated from:
D = (E)*(3600 sec/hr)/5.4xl0' MeV/rem/g 4(Z)air Using this, the following estimates are obtained for dose rate at the outside edge of the reactor pool wall at 500 KW operating power:
1 MeV : 1.2x10-5 mrem/hr, 2 MeV : 6.6x10 3 mrem/hr, 4 MeV : 1.3x10 2 mrem/hr, 6 MeV : 1.9x10-2 mrem/hr, and 8 MeV : 6. X10 2 mrem/hr.
These results show that very little dose rate from the gamma rays emitted from the core will be present at the outside of the pool walls. The dose that does appear is a result of the higher-energy gamma rays. Almost all of the lower-energy gammas are absorbed by the materials surrounding the core.
7.1.1.2 Gamma Dose Through an Experimental Facility Several experimental facilities are available which lead up to or into the core. These include the CIF, two beam ports, and movable dry tubes. To estimate gamma doses from opened experimental facilities, the core is modeled as a sphere with a volume equal to that of the actual core volume in the form of the rectangular solid that it 168
actually is. This source is assumed to emit Nrf(e) photons per unit time per unit solid angle. The polar angle, 0, is the angle between the direction of emission and the axis of the tube which forms the experimental facility. The function f(0) is normalized such that Nc is the number of photons emitted into the half-space directed toward the shield per unit area per unit time. Assuming a cosine emission from the source yields the following equation to predict the gamma flux at a distance z from the core along the tube axis:
O(z) = zNo[l-(l+R2 /z 2
) ] -1/2 where Nc, is the number of gamma rays emitted from the surface of the core per unit area per unit time (after some have been absorbed by materials in the core), R is the radius of the irradiation tube, and z is the distance along the axis of the tube. This equation was used to estimate the dose rates from streaming through the CIF and beam ports.
Note that this estimate does not do an exact calculation for buildup of gamma dose from secondary emissions or scattering of gamma rays from the walls of the irradiation facility. Such a calculation would involve a more complex gamma photon transport analysis using Monte Carlo techniques.
Calculation of No requires consideration of the absorption properties of the materials making up the core. The total gamma absorbing effect of the core can be estimated by a weighted average of the linear attenuation coefficients of the materials composing the core. The average is weighted by the percent volume of each material in the core. For this calculation, the core volume was considered to be formed of 65i light water, 25.5% aluminum, and 9.5% uranium. The fission rate source intensity was estimated in Section 7.1.1.1, and assuming 500 KW operation and using the earlier equations, we can calculate:
No= FK/2v4L, where F = fission rate = 1.77xlO1l fissions/cm3 /second, K = MeV/fission released as gamma rays, vc= effective volume of the core, and
= core attenuation factor.
The CIF is a hollow aluminum tube 1.5 inch outer diameter, extending from the top of the pool into the geometric center of the core, about 16 feet (488 cm) long. For this calculation, it is considered to be empty of shielding plugs or experimental apparatus, and is filled with air at STP conditions.
Using equations derived earlier, the gamma flux for 1 MeV, 2 MeV, 4 MeV, 6 MeV, and 8 MeV gamma rays was estimated. The following dose rate estimates for streaming out of the open top of the CIF tube at 500 KW operation were obtained:
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1 MeV : 2.1 rem/hr 2 MeV : 78.3 rem/hr 4 MeV : 23.7 rem/hr 6 MeV : 5.2 rem/hr 8 MeV : 2.2 rem/hr TOTAL : 111.5 rem/hr This is a high dose rate at the exit of the CIF tube. A later section will discuss protection measures to keep doses as low as possible.
Dose rates from the CIF and beam ports one hour after shutdown were also estimated. Dose rates after shutdown result from gamma rays emitted by the decay of fission products in the core. The intensity of the gamma source in the core after shutdown (sometimes called the shutdown field) is a function of the power history of the reactor and the subsequent decay time following shutdown. For this calculation, it was assumed that the recent power history would include an irradiation at 500 kilowatts for a duration of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , followed by a shutdown decay period of one hour. The fission product activity can be approximated by:
A = 1.4P[(t 0 2 - (t+T) 0 - 2]
where A = fission product activity in curies, P = operating power of the reactor in megawatts, t = shutdown time in days, and T = operating time in days.
Also, assume an average decay gamma ray having an energy of 0.7 MeV per disintegration.
Performing the calculation for fission product activity, a total of 3.4x1024 MeV/second is being generated in-the source. Using the equations developed earlier, a dose rate of 13.52 rem/hr at the open top of the CIF is obtained. For the two beam ports, which are considered to be hollow aluminum tubes with a 6 inch inner diameter and 7 feet long, with no shielding plugs or experimental apparatus installed, the dose rate at the open end of the beam ports will be about 1192 rem/hour. Again, protection methodologies are discussed-in a later section.
7.1.1.3 Neutron Dose Through an Experimental Facility An opened experimental facility will allow neutrons to leak from the core to the exit of the experimental facility. A precise estimate of the neutron flux at the exit of an opened experimental facility is difficult, since it would require modeling of neutron transport 170
phenomena along the length of the experimental facility. These transport effects would depend on the nature of the neutron source, geometry of the experimental facility relative to the core, and the (X) materials making up and surrounding the tube through which the neutrons are passing. Effects such as neutron absorption, elastic and inelastic scattering events, and in-scattering of neutrons from the surrounding materials would have to be taken into account. However, a rough estimate of the neutron flux and dose at the exits of the experimental facilities can be obtained from considering geometry effects alone.
First, consider an experimental facility for which the neutron flux and energy distribution has been measured at the core end of the facility. The energy-dependent neutron flux is usually represented as a(E) neutrons/cm2 /second for a given reactor power. If the cross-sectional area of the experimental facility is known, the area-distributed neutron source can be collapsed to an equivalent point source given by A~a(E), where A, is the cross-sectional area of the experimental facility.
Now, the neutron flux passing through a unit area of a sphere-surrounding this equivalent point source is simply Asa(E)/4nr2 , with r the length of the experimental facility from the core end to the exit.
Table 7.1 shows the geometric parameters used in the neutron dose calculations.
Horning [2] has measured the energy-dependent neutron flux in various experimental facilities at 10 KW for the HEU-fueled OSURR. Table 7.2 summarizes the data used in the flux estimations. For these calculations, the flux measurements were normalized to 1 watt reactor power. Etherington [3] gives neutron fluence to dose conversion factors as a function of neutron energy. Table 7.3 lists the conversion factors used in these calculations.
Using the geometry equation developed earlier, and the data shown in Tables 7.1 and 7.2, estimated neutron fluxes at the exits for various experimental facilities for a 1 watt operating power were calculated.
The results are shown in Table 7.4. Using the dose conversion factors shown in Table 7.3, equivalent doses were estimated. These are shown in Table 7.5.
7.1.1.4 Gamma Dose At The Pool Top Through Pool Water Gamma radiation from the core interacts with the pool water above the core. Some of the gamma photons will travel through the 15 feet of water above the core and give rise to a dose rate at the surface of the pool. Using an approach like that taken in Section 7.1:1.1, the gamma dose was estimated for gamma rays of energy 2, 4, 6, and 8 MeV.
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Table 7.1: Geometric Parameters of Various Irradiation Facilities Facility Name Cross-Sectional Area Effective Length CIF 9.65 cmX 487.7 cm Beam Port 1 190.09 cm2 198.1 cm Beam Port 2 190.09 cm' 198.1 cm Thermal Column G-7 103.23 cm2 144.8 cm 172
Table 7.2: Measured Core-End Neutron Flux in Various Irradiation Facilities Facility Name Maxwellian 1/E Neutron Fission Total Neutron Flux Flux Neutron Flux Neutron Flux (n/cm /sec) (n/cm2 /sec) (n/cmj/sec) (n/cm2/sec) 1 1 CIF 1.94x10 1.20x10 8.40x101' ; 1 3.97x10 Beam Port 1 1.22x1011 5.46x10' 0 4.08x10°' 2.17x10 1 Beam Port 2 9.47x10 1 0 3.26x101 0 2.67x10°' 1.54x10 1 Thermal Column 1.29x101 0 5.42x109 4.96x109 1.88x10 0 Position G-7 Notes: (1) The measurements shown above are at the core end of the experimental facility for the HEU-fueled OSURR operating at 10 KW.
(2) Neutron energy ranges are as follows:
Maxwellian: 0 - 1.85x10-7 MeV 1/E: 1.85x10-7 - 0.5 MeV Fission: 0.5-15 MeV (3) Measured neutron fluxes are as reported by Horning [2].
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Table 7.3: Neutron Fluence to Dose Conversion Factors Neutron Energy Range Dose Conversion Factor in in MeV rad/neutron/cm 0 - 1.85xlO' 3.5 x 10
1.85xlO-' - 0.5 1.0 x lO-',
0.5 - 15 3.5 x 10-'
Notes: (1) Dose equivalents are approximate values derived from visual examination of the curves shown in Etherington
[3].
(2) Doses are for soft-tissue equivalent materials, based on a first-collision dose delivery model.
174
Table 7.4: Estimated Neutron Flux at the Exit of Various Irradiation Facilities Facility Maxwellian 1/E Neutron Fission Total Name Neutron Flux Flux Neutron Flux Neutron Flux (n/cm 2 /sec/w) (n/cm2 /sec/w) (n/cm'/sec/w) (n/cm:/sec/w)
CIF 62.6 38.7 27.1 128.2 Beam Port 1 4702.6 2104.6 1562.7 8364.5 Beam Port 2 3650.3 1256.6 1029.2 5936.1 Thermal 505.4 212.4 19.4 736.6 Column Position G-7 Notes: (1) The estimates shown above are at the exit of the experimental facility for the HEU-fueled OSURR normalized to 1 watt operating power. C)
(2) Neutron energy ranges are as follows:
Maxwellian: 0 - 1.85x10-' MeV 1/E: 1.85x10 0.5 MeV Fission: 0.5-15 MeV 175 Q."
Table 7.5: Estimated Neutron Dose at the Exit of Various Irradiation Facilities Facility Maxwellian 1/E Neutron Fission Total Name Neutron Dose Dose Neutron Dose Neutron Dose (rad/hr/W) (rad/hr/W) (rad/hr/W) (rad/hr/W) 5 CIF 7.89x10 1.39X1 05 3.41x10 4 4.34x10 4 Beam Port 1 5.93x10 3 7.58x10-4 1.97x10-2 2.64xl1--
Io Beam Port 2 4.60x10-3 4.52x10-4 1.30x10-2 1.81X1O--
Thermal 6.39x10 4 7.65x1O 5 2.44x10-4 9.60x10 4 It Column.
Position G-7
.t ,
II Notes: (1) The estimates shown above are at the exit of the experimental facility for the HEU-fueled OSURR normalized to 1 watt operating power.
(2) Neutron energy ranges are as follows:
Maxwellian: 0 - 1.85xlO- MeV 1/E: 1.85xl0' - 0.5 MeV Fission: 0.5-15 MeV 176.
The following dose rate estimates were obtained:
2 MeV : 0.08 mrem/hr 4 MeV : 10.60 mrem/hr 6 MeV : 32.24 mrem/hr 8 MeV 97.40 mrem/hr TOTAL : 140.32 mrem/hr So, at the surface of the pool, with the reactor at 500 KW, the direct gamma dose exceeds the limits for a high radiation area. This will require protection methods to be discussed in a later section.
It should be noted that this estimated dose rate is only at the surface of the pool directly above the core. During normal operation, this point is difficult to access. The most likely position for a person at the pool top closest to the core would be standing at the side of the pool, with the lower half of their body shielded by the 1-foot thick pool wall. Thus, the most likely whole-body dose rate to which an individual would be exposed is less than that estimated above.
7.1.2 Indirect Exposures Radiation exposure can occur from sources produced by the reactor.
Chapter 6 considered the doses from production of Ar and N that escapes from the reactor pool surface and/or experimental facilities U
and mixes with the air in the building. This section will consider other indirect sources.
7.1.2.1 Nitrogen-16 At The Surface Of The Pool Chapter 6 discussed the production of 16N isotope in the core. Section 6.1.2.3 estimated that about 24.64 millicuries of 16 N per second are released from the top of the core for 500 KW operation. Further, Section 6.3.4.8 considered the effects of a rising plume of 16 N-bearing water to the surface of the reactor pool and spreading across the surface into the shape of a disk. The average concentration of leN atoms in the disk source thus produced was estimated to be 1184.8 atoms of 1'N per milliliter. This estimate accounts for decay of the isotope during its rise to the surface of the pool and spreading into a disk source.
The dose rate from a disk source of 16N can be estimated from the following:
D = [A.N/2RK][1-E 2 (ph)I, where D = dose rate near the pool surface, 177
h = thickness of disk source (see below),
= linear attenuation coefficient of 6 MeV photons in water, K = conversion factor (1.6x105 photons/cm /sec/rad/hr),
X = decay constant-for 'ON, N = average -6N concentration in the disk source, and E- = second exponential integral (see Section 7.1.1.1).
The thickness of the disk source can be estimated from:
h = vct/A, where h = thickness of the disk source, v, = volume flow rate of the I'N-bearing water, tr = time to spread into a disk source, and A, = area of the disk source.
Substituting appropriate values into the source thickness equation gives a value of 2.48 cm for the disk source thickness. Using this to evaluate the integral term, and substituting other constants in the above equation for D yields an estimate of 2.02 mrad/hr from the disk source of 16N at the surface of the pool. This source intensity is considerably less than that from direct penetration of the water layer above the core by gamma rays from the core.
7.1.2.2 Gaseous Effluents
: Operation of the reactor produces gaseous radioisotopes which diffuse into the air of the reactor building, and ultimately to the outside atmosphere. Chapter 6 considered the production of these gaseous sources, and the dose consequences both within and outside the building. Table 6.9 summarizes the dose calculation for various types gof 4 Ar releases. Section 6.3.4.8 estimated the dose consequences of iON release from the surface of the pool.
7.1.2.3 Cooling and Process System Activation Products circulation of the reactor pool water through the core will induce activation of any trace elements contained in the water. Also, neutron-alpha reactions occurring in the aluminum materials in the core will produce some 24 Na, part of which will diffuse into the pool water. As the reactor pool water is circulated through the primary cooling loop and the water processing system (when it is activated),
some of these activated elements will become trapped in the internals of these systems, giving rise to a dose rate around the components of these systems. Prediction of the amount of materials deposited in these systems and the resulting dQse rates is difficult, as it depends on the power history of the reactor and the relative probability of elements being retained in the cooling or process system components.
Radiation Protection methods for these systems will be discussed in a later section.
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7.1.2.4 Beam Port Plug Activation During operations not using the beam ports as experimental facilities, shielding plugs are inserted to reduce neutron and gamma exposures through the ports. These plugs have aluminum end caps and sides. The end cap of the shielding plug in Beam Port 1 is exposed to the higher neutron flux, since it rests directly adjacent to the core. Activation of the aluminum in this plug will produce radioisotopes which can produce a dose rate to personnel exposed to the shielding plug upon its removal from the beam port.
It is possible to estimate the dose rate produced from the radioisotopes in the end cap of the beam port plug. First, the basic neutron activation equation can be written as:
A = NiC (l -e-Xt')e-xt where Ai= activity of isotope i at time t after irradiation in
- disintegrations per second, Ni= number of "target" atoms of the parent element for isotope i exposure to the neutron flux, 01 = cross-section of the neutron reaction producing isotope i,
= neutron flux of energy necessary to produce isotope i in the target, to =
t X=
=
decay constant of the activation product, irradiation time, and decay time.
C)
To estimate N', assume that the aluminum end cap is made of pure aluminum in the shape of a disk 6 inches in diameter and 0.25 inches thick. The volume of such a cap is approximately 15.2 cm3 . Aluminum has a density of 2.7 grams/cm3 , which gives an end cap mass of 41.04 grams.
Now, the number of atoms of aluminum available for irradiation is:
N1 = mFlF..NA/M where m = total target mass, NA = Avogadro's Number, M = molecular weight of the target element, F= elemental purity of the sample for the target element, and F = isotopic fraction of the target element that the parent isotope of the activation product represents.
Since we assume that the end cap is made of pure aluminum, F. is taken to be 1. Similarly, F~'is also 1, since all of the naturally-occurring aluminum is 27Al. Taking NA to be 6.123x102 3 atoms/mole, and M to be 26.98 grams/mole, the number of aluminum atoms exposed to the neutron 179
flux is thus 9.162x10 2 - atoms.
Now, when aluminum is exposed to a polyenergetic neutron flux, several neutron reactions can occur. These are listed below:
27 Al(n,y)228A:
a = 0.235 barns A = 0.301 minutes' En = thermal 27 Al (np) 2 'Mg:
a = 0.05 barns
= 7.3x10-2 minutes' E, = 3 MeV 27 24 Al(n,a) Na:
a = 0.03 barns X = 7.7x10-4 minutes-En = 7 MeV In the above list, En represents the threshold neutron energy required to induce the reaction listed.l i h For this calculation, it is assumed that the OSURR is operated at 500 kilowatts for 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , then is shut down. Neutron flux in Beam Port 1 at 10 kilowatts has been measured by Horning [2]. Assuming linearity with reactor power, the following values are assumed for neutron flux:
Thermal (0-0.185 eV) : 6.35x1012 nv Fast (0.5-15 MeV) : 2.04x1012 nv The value for fast flux shown above was used to estimate both :Mg and 4
2 Na activity. The calculation for isotope activities, for various decay times, is shown in Table 7.6.
Using the estimated activities, dose rates can be predicted for the various isotopes at different decay times. Dose rate can be predicted from the following approximation:
R = 6CE/D2 where R = dose rate in R/hr, C = isotope gamma activity in curies, E = gamma energy in MeV, and D = distance from the source in feet.
180 I
Table 7.6: Activity Estimates for Shielding Plug End Cap 0 Irradiation Time: 30 Hours Operating Power : 500 kilowatts Material: Pure Aluminum Size : 0.25" thickness, 6" Diameter Disk Mass : 41.04 grams 2 21 24 Decay Time 8Al Activity Mg Activity Na Activity (min.) (microcuries) (microcuries) (microcuries) 0 3.70x10' 2. 53x100 1.1 4 xlOE 10 1.82x106 1.22x106 1.13xlO 60 5.21x100 3.18xl0 1 . 09x10 360 2.48x10-4 ' 9.96x10-6 8.64x10 5 720 1440 0
0 3, 92x10-"
6. 07x10 40 6.55x10 5 3.76x10 5 U) 181 U
This approximation is valid only for gamma emissions and if the distance from the source is such that the source subtends a small enough solid angle to be considered a point. Generally, a distance of about 5 times the maximum dimension of the source is sufficient to achieve a quasi-point source geometry. For the beam port plug end cap, this will be about three feet. Thus, at three feet, in air, the dose rates from the various activation products can be estimated. The following gamma decay properties were assumed:
28 Al:
E= 1.7789 MeV (100% abundance) 27 Mg:
E= 0.8438 MeV (72% abundance)
E72 = 1.0144 MeV (28% abundance) 24 Na:
E7l = 1.3685 MeV (100% abundance) 2 = 2.7539 MeV (100% abundance)
E-r The results of the calculations are shown in Table 7.7. These results indicate that the major contributor to long-term radiation exposure from shield plug activation is 24Na. The shorter half-life isotopes will only contribute to doses received a short time after the end of i j the irradiation, and can thus be reduced or eliminated by allowing sufficient decay time before removing the shielding plug from the beam port. The remaining dose rate from 24Na can be handled by shielding the plug in a cask, or increasing distance from the end cap of the plug.
An overhead crane handling system is available for this type of operation.
7.1.2.5 Experimental Sample Activation Using the equations developed in the preceding section, it is possible to estimate the doses from activation of elements in experimental samples and apparatus, provided that information of the elemental makeup of these samples is available. However, this information is not always known precisely, so it is difficult to accurately estimate the induced activity and resulting doses rates from irradiation samples in all cases. Experimental procedures require that an activity and dose rate estimate be provided based on the best available information about the material to be irradiated. Upon removal from an experimental facility, dose rate surveys are performed to assess the gamma + beta and gamma-only dose rates. If possible, an isotopic assay is also done to determine the type and quantity of induced activation products.
Records of all in-pile irradiations are required to be maintained.
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Table 7.7: Dose Rate Estimates for Shielding Plug End Cap Irradiation Time: 30 Hours Operating Power : 500 kilowatts Material: Pure Aluminum Size: 0.25" thickness, 6" Diameter Disk Mass: 41.04 grams Source Distance: 3 feet Absorber: air 2 24 Decay Time SAl Dose Mg Dose Na Dose Total Dose (min) Rate Rate Rate Rate (mr/hr) (mr/hr) (mr/hr) (mr/hr) 0 43900 1504 3133 48537 10 2157 725 3109 5991 60 0 .19 2992 3011- U 360 0 0 2375 2375 720 0 0 1800 1800 1440 0 0 1034 1034 183 0..
7.2 Protection Strategy The approach taken to radiation protection at the NRL is to keep radiation exposures to personnel as low as reasonably achievable. This approach, sometimes designated as the ALARA concept, incorporates methods and procedures to reduce radiation exposure to individuals working within the confines of the reactor building. The ALARA approach to radiation protection includes use of methods and procedures involving shielding of radiation sources and/or personnel, increasing distance between an exposure point and a radiation source, reducing time a person might be exposed to a given dose rate, containment of sources, and careful, thoughtful, advance planning when working in an area in which might exist a radiation field.
Administrative controls maintained at the NRL are designed to compliment the ALARA approach. All experiments involving reactor use are reviewed by a qualified individual (i.e., one who holds a current, valid Senior Operator License for the OSURR). Requests for reactor use involving radioisotope production or any other potential radiation exposure of significance are examined for indications or estimates of what the level of radiation hazard might be for the experiment, and methods and/or procedures to reduce the potential hazards. If a radiation accident scenario is possible, an experimenter must provide a methodology for dealing with such a scenario and steps to be taken to mitigate its consequences. Conduct of all experiments is subject to the approval and concurrence of the Senior Operator On Duty during the reactor operation. If this individual determines that a particular reactor operation poses a hazardous or potentially unsafe condition, the run may be terminated at his/her discretion.
Production and use of radioactive materials within the reactor building fall under the guidelines issued by the university's Radiation Safety Section. These guidelines in turn fall within the regulatory framework of 10CFR20, and in most cases provide for mare stringent controls than those specified in these regulations. In addition, the NRL provides in-house procedures that fall within the guidelines of both the university and federal regulations.
7.3 Protection Methodology 7.3.1 Instruction and Training Persons working in the reactor building on a daily basis are trained in appropriate radiation protection concepts. In addition, these persons are trained to assist in responding to abnormal radiological conditions in the reactor building. Certification as to having received such instruction is provided as required by 10CFR19. Persons in this category include regular NRL operations, research, and maintenance staff, selected personnel from the University Radiation Safety Section, and student assistants working at the NRL facility.
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Individuals not working in the reactor building on a daily basis, but involved in ongoing research and instructional activities at the laboratory on a routine basis, are provided instruction on self-protection against radiation exposure as per 10CFR19 requirements.
Extended training in radiological emergency procedures is not given to these persons. Individuals in this protection category include faculty cand students involved in research projects and instructional activities at the reactor building, NRL administrative and clerical staff, and custodial workers.
Occasional visitors to the NRL, such as commercial vendors, onetime visitors for tours and demonstrations, and non-routine experiment personnel, are given instruction in basic procedures such as wearing of dose monitors, signing in and out of the building, radiation and radioactive material storage areas. These persons are escorted by an individual more fully trained in radiation protection.
7.3.2 Access Control The reactor building is designated as a Restricte Certain areas of the reactor building. Qre desiina te d as iermanent radioact stora e O .}--this area, no eating-,--
Xdf Wrm^n§g, smoking, or app y3.ng of c metics is allowed. These restrictions also apply to the catwalk areas along the top of the reactor pool. In addition, other areas of the reactor building may be posted as radioactive materials storage areas, in which the same precautions must be observed.
-Areas may-also-be-po-sted -as-Radiation--Areas.-Appropriate-restricti-ons-------------
and precautions must be observed in these areas. Occasionally, an area may be posted as a High Radiation Area. Again, appropriate precautions must be observed, and access to this area limited.
The sign-in procedure for access to the reactor building requir es certain information~ f an admitted person. This includes their 3Visitor's logbooks are maintained as part of the permanent records o the NRL.
7.3.3 Personnel Monitoring Persons working full time in the reactor building are required to wear a film badge. Badge numbers may be noted in the visitor's logbook for 185
persons not registered in the university film badge record system.
Film badges used at the NRL contain gamma, beta, and neutron-sensitive materials for monitoring doses at various tissue depths.
For monitoring extremity doses, ring badges based on TLD dosimeters are available from the Radiation Safety Section. Both regular film badges and ring badges are read on a monthly basis.
Immediate dose indications are provided by wearing a self-reading pocket dosimeter. These devices, based on an air ionization and capacitor discharge effect, provide the most convenient indication of exposure in near-real time. Several pocket dosimeters are available at the NRL. These are used to monitor the exposures of individuals in classes at the reactor and are recorded in the Visitor's Logbook.
7.3.4 Area Monitoring A variety of area radiation monitoring systems are available at the NRL. Section 3.7 discussed the Area Radiation Monitor system installed at the NRL, which provides continuous indications (local and in the control room) of the dose rate at various locations in the reactor building. Other devices are also available to provide information on area dose rates.
A beta-gamma counter is mounted at the sign-in desk at the front of the building. This unit is sometimes called a "frisker" and provides indications of gamma and beta activity. This device uses a detachable Geiger-Mueller probe which can be used to survey areas of a person's
@ body for evidence of contamination, and it provides an indication of overall radioactivity in the area of the front door. It can be used to survey incoming and outgoing shipments of radioactive materials and devices, and can also serve as a monitor for airborne and gaseous radioactive material in the entire reactor building.
A series of four TLDs are mounted at various points in the main reactor bay. These badges, processed every three months, provide a long-term average of area dose in the vicinity of the badges.
Similarly, film badges are mounted at eight different locations outside the building.
7.3.5 Survey Instruments Several portable survey instruments are available at the NRL for routine monitoring of radiation sources. These include Geiger-Mueller and air-ionization chambers for gamma and beta + gamma detection and detectors) for neutron detection. These survey instruments provide enough range and flexibility to monitor most expected radiation sources produced by OSURR operation. For those sources not.detectable with these instruments, the Radiation Safety Section can provide additional support.
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7.3.6 Shielding Earlier sections discussed the radiation shielding features of the OSURR. Additional shielding for sources is obtainable. Normally, the beam ports are plugged with shielding materials which, when coupled with closed beam port shutters, reduce exposures at the beam port.
exits to negligible levels. For storage of irradiated beam port plugs, a movable shielding cask is available. This cask has a maximum lead thickness of about eight inches.
A large lead storage chest is located in the northeast corner of the building. This chest is used to store calibration sources and highly radioactive irradiated samples and devices. The lead thickness of this storage chest'is about four inches. It has a motor-driven, crane-liftable lid to prevent easy access to the sources inside.
Smaller lead containers are available for storage of contained radioactive sources. These containers, commonly called "pigs", vary in wall thickness from a fraction of an inch up to several inches.
7.3.7 Administrative Controls Safe operation of the OSURR and performance of associated experiments depends on reliable and conscientious observance of established procedures and protocols by the staff of the NRL. Experiments involving production of radioisotope sources of significant quantities and intensities, use of experimental facilities such as beam ports and dry tubes, and the use of radiation sources apart from the reactor, are subject to the approval of appropriate NRL or university administrators. Such experiments and uses must be in accordance with regulations specified in 10CRF20 or other applicable regulations.
Administrative controls are established to allow NRL personnel discretion in approving and carrying out experiments and operations a safe manner. NRL personnel may themselves specify appropriate procedures to be incorporated in an experimental program to assure radiological safety. If such procedures are unavailable, the experiment may be denied on that basis. If such procedures are not observed during the experiment, the Senior Reactor Operator on duty during the operation may terminate the experiment at his or her discretion.
7.4 Chapter 7 References
[1] R. G. Jaeger et al., "Engineering Compendium on Radiation Shielding", The Internation Atomic Energy Agency, Vienna, Springer-Verlag, 1968-1975.
[2] Nicholas C. Horning, "Measurement of the Neutron Spectra in the Beam Ports and Thermal Column of The Ohio State University Research Reactor", M.Sc. Thesis (unpublished),
Ohio State University, Columbus, OH, 1976.
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9.0 Administrative Controls 9.1 Organization 9.1.1 Structure The Ohio State University Research Reactor is a part of the College of Engineering administered by the Engineering Experiment Station. The organizational structure is shown in Figure 9.1.
9.1.2 Responsibility The Director of the Engineering Experiment Station (Level 1) is the contact person for communications between the U.S. Nuclear Regulatory Commission and The Ohio State University.
The Director of the Nuclear Reactor Laboratory (Level 2) will have overall responsibility for the management of the facility.
The Associate Director (or Manager of Reactor Operations, Level 3)
E shall be responsible for the day-to-day operation and for ensuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and technical specifications. During periods when the Associate Director is absent, his/her responsibilities are delegated to a Senior Reactor Operator (Level 4).
9.1.3 Support Groups The Radiation Safety Section (RSS) of The Ohio State University provides both support and audit functions for the OSURR. They conduct and review results of area and smear wipes of the Reactor Laboratory, assure proper posting of Radiation and High Radiation Areas, assist in calibration of radiation monitoring instrumentation, inventory and wipe sealed sources, and file required SNM Reports. They are available as needed for response to unplanned emergencies at-the reactor or for planned experiments that require additional monitoring. They are on the call list to respond to nuclear emergencies.
The Ohio State University Police Department (OSUPD) is the local law enforcement agency responsible for facility security. The OSUPD is equipped to provide all actions normally associated with a full-service police department, including arrest and prosecution when warranted, and special tactics and investigative units.
The Ohio State University Office of Environmental Health and Safety is available to assist in the monitoring and disposal of non-radioactive hazardous materials from the OSURR.
Fire protection and ambulance services are provided by the Clinton Township Division of Fire or the Cities of Columbus and Upper Arlington.
238
Provo:St Vice President for Business and Finance Dean, College of Engineering Reactor Operations Committee Director, Engineering Experiment Station (Level 1)
Director, Nuclear Director, Radiation Reactor Laboratory Safety Section (Level 2)
Associate Director, Nuclear Reactor Laboratory (Level 3)
Senior Reactor Operator (Level 4)
Reactor Operations Staff Solid Lines Paths of Direct Responsibility Dashed Lines ----- Paths of Information Figure 9.1: Administrative Organization 239
9.2 Training Program The OSU-NRL shall maintain a RO/SRO requalification program reviewed and approved by the NRC. It will be designed to demonstrate Operator and Senior Operator competence, and to satisfy the requirements of 10CFR55.33(c), 10CFR55 Appendix A, and ANSI/ANS-15.4-1988 Selection and Training of Personnel for Research Reactors. The Manager of Reactor Operations (or Associate Director) shall serve as Training Coordinator and shall be responsible for the implementation, coordination and operation of the Requalification Program including the training of new operators.
9.3 Recordkeeping and Reporting Requirements Requirements for record keeping and reporting are contained -in Section 6 of the Technical Specifications for the OSURR.
9.4 Emergency Planning and Preparedness The OSURR shall maintain an Emergency Plan reviewed and approved by the NRC. It shall be based on the requirements of Appendix E to 10CFR50 and the criteria set forth in Revision 1 to Regulatory Guide 2.6 and ANSI/ANS 15.16-1982 "Emergency Planning for Research Reactors." The plan shall include a site and facility description; normal and emergency organizational structures; an emergency classification system; an emergency response plan; and provisions to maintain emergency equipment. There shall be Emergency Plan training, drills, and periodic review.
9.5 Internal Reviews and Audits 9.5.1 OSURR Reactor Operations Committee 9.5.1.1 Responsibilities and Authority There shall be a Reactor Operations committee (ROC) which shall review and audit reactor operations to assure the facility is operating in a manner consistent with public safety and within the terms of the facility license. The Committee advises the Director of the NRL, and is responsible to the Provost of The Ohio State University (Figure 9.1).
9.5.1.2 Committee Membership Committee members shall be appointed annually by the Provost of The Ohio State University. The Committee shall be composed of at least nine members including ex-officio members. The Director and Associate Director of the Nuclear Reactor Laboratory, and the Director of the Radiation Safety Section shall be ex-officio voting members of the Committee. The remaining Committee members shall be faculty, staff, 240
and student representatives of The Ohio State University, having professional backgrounds in engineering, physical, biological, or medical sciences, as well as knowledge of and interest in applications of nuclear technology and ionizing radiation.
9.5.1.3 Committee Meetings The Committee shall meet at least twice a year. They should meet every six months. A quorum shall consist of at least 50 percent of the members.
9.5.1.4 Subcommittees The chairperson may appoint a subcommittee from within the committee membership to act on those matters which cannot await the regular semi-annual meeting. The full committee shall review the actions taken by the subcommittee at the next regular meeting.
A three member subcommittee shall meet annually to perform an audit of NRL operations and records or to review the results of an independent audit. At least two individuals on the Audit Subcommittee shall be ROC members. The third may be a staff member from the Reactor Laboratory or another individual appointed by the ROC chairperson. Each person should serve for three consecutive audits, at which time he or she should be replaced by a new member. In this way, each subcommittee should consist of two holdovers and one new member. The member serving for his or her second audit should be the Audit Subcommittee Chairperson.
9.5.2 Experiment Approval All proposed experiments utilizing the reactor shall be evaluated by the experimenter and a licensed Senior Reactor Operator to assure compliance with the provisions of the utilization license, the Technical Specifications and 10CFR20. If, in the judgement of the Senior Reactor Operator, the experiment meets with the above provisions, is an approved experiment, and does not constitute a threat to the integrity of the reactor, it may be approved for performance. When pertinent, the evaluation shall include considerations of:
(1) the reactivity worth of the experiment, (2) the integrity of the experiment, including the effects of changes in temperature, pressure, or chemical composition, (3) any physical or chemical interaction that could occur with the reactor components, and (4) any radiation hazard that may result from the activation of materials or from external beams.
241 ,
Prior to performing an experiment not previously approved for the reactor, the experiment shall be reviewed and sanctioned by the Reactor operations Committee. Their review shall consider at least the following information:
(1) the purpose of the experiment, (2) the procedure for the performance of the experiment, and (3) the safety evaluation previously approved by a licensed Senior Reactor Operator.
9.5.3 Additional Oversight 9.5.3.1 The Radiation Safety Section (RSS)
The RSS has direct lines of communication to the Director of the NRL, the Provost of The Ohio State University (through the VP for Business and Finance) and the NRC. (See Figure 9.1) This combined with their review and audit functions at the NRL provide additional control over actions that may affect the health or safety of the public.
9.5.3.2 Operations Manager The Manager of Reactor Operations (or Associate Director) as the designee of the Director of the NRL has direct oversight responsibility for Reactor operations and experiments. He or she may terminate any operation at any time if it is deemed to be or likely to adversely affect the health or safety of the public.
9.5.3.3 SRO On-Duty In the absence of the Operations Manager the SRO on duty has oversight responsibility and shall make decisions regarding safe operation of the OSURR.
9.6 Security 9.6.1 Security Plan The Ohio State University Research Reactor shall implement security procedures based on Regulatory.Guide 5.59 designed to meet the requirements of 10CFR50, 70, and 73. These procedures shall describe the mechanisms and organization to protect special nuclear material against sabotage and to detect theft and attempted theft.
9.6.2 Security Organization The Director of the NRL and the Chief of Police of The Ohio State University Police Department (OSUPD) share the responsibility for 242
implementation and operation of the security procedures at the NRL.
Day to day security is the responsibility of the NRL staff during normal working hours and whenever an individual authorized access to the NRL is present in the building. At other times, the OSUPD is the local law enforcement agency (LLEA) responsible for facility security.
Several other LLEA's also have jurisdiction at the reactor site and would respond if requested by the OSUPD. These include the Clinton Township Police, the Franklin County Sheriff, and in case of civil emergencies, the Ohio State Highway Patrol on order of the Governor.
9.7 Quality Assurance 9.7.1 Quality Assurance Program The OSURR shall establish and maintain a quality assurance program based on ANS-15.8 ANSI N402-1976 to provide adequate confidence that safety-related items will perform satisfactorily in service. As a minimum, safety-related items included in the plan shall be those identified in the Limiting Conditions for Operation section of the Technical Specifications (Section 3).
9.7.2 Program Requirements 9.7.2.1 Program Requirements The responsibility for the QA program is delegated to the Director of the NRL as shown in the "Administrative Organization" (Figure 9.1).
His/her designee, normally the Associate Director, shall be 4 responsible for the daily implementation of the program. The Reactor Operation Committee has responsibility for independent review and audit functions associated with the program. Reviews may include experimental equipment and the design of safety-related items.
9.7.2.2 Records and Documents All activities affecting safety-related items identified in Section 3 of the Technical Specifications shall be identified and documented.
Procedures shall be established to control the development, revision, and use of documents and drawings which are safety-related. Records of quality assurance activities shall include inspection and test results; QA reviews by the ROC, and analyses of modifications and design changes. Retention requirements shall be established for these records and will include duration, location, and responsibility.
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10.0 Financial Qualification
~ 10.1 Financial Ability to Operate a Non-Power Reactor Since the issuance of the original license in 1960, The Ohio State University has provided funding to assure safe operation of the OSURR.
This provides reasonable assurance that operating costs for the next five years will be made available. The current budget allocation provides funding for about 1.9 FTEs plus adequate funds for supplies and equipment. Funding for personnel includes benefits. No direct funds are required for the typical overhead costs of heat, lighting, water, electric, or space utilization.
10.2 Financial Ability to Decommission In a letter dated July 30, 1990 the OSURR submitted its Decommissioning Funding Plan as required by 10CFR50.33(k). It provides the assurance that The Ohio State University will make funds available for eventual decommissioning.
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APPENDIX A TO FACILITY OPERATING LICENSE NO. R-75 Technical Specifications And Bases For The Ohio State University Pool-Type Nuclear Reactor Columbus, Ohio Docket No. 50-150
TABLE OF CONTENTS I
.0 I N T R O.....................................................
D U C T ION ;
1.1 Scope .................................
1.2 Application ....................................................
1.2.1 Purpose ..................... I 1.2.2 Format ..................... 1 1.3 Definitions ......... .. 1 2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS (LSSS' ... 7 2.1 Safety Limit .
2.2 Limiting Safety System Settings.
3.0 LIMITING CONDITIONS FOR OPERATION . . . 9 3.1 Reactor Core Parameters.. 9 3.1.1 Reactivity.......................... ....................... 9 3.2 Reactor Control and Safety System ...... ....................... 10 3.2.1 Control Rod Drop Times.............. ....................... 10 3.2.2 Maximum Reactivity Insertion Rate... ....................... 11 3.2.3 Minimum Number of Scram Channels.... ........................ 11 3.3 Coolant System ........................ ....................... 14 3.3.1 Pump Requirements................... ....................... 14 3.3.2 Coolant Level....................... ....................... 14 3.3.3 Water Chemistry Requirements........ ....................... 15 3.3.4 Leak, or Loss of Coolant Detection.. ....................... 15 3.3.5 Primary and Secondary Coolant Activity Limits ............. 15 3.4 Confinement Isolation .... ............... ..................... 16 3.5 Ventilation Systems ........................................... 16 3.6 Radiation Monitoring Systems and Radioactive Effluents .. 17 3.6.1 Radiation Monitoring ...................................... 17 3.6.2 Radioactive Effluents ..................................... 18 3.7 Experiments ........................ .......................... 19 3.7.1 Reactivity Limits ......................................... 19 3.7.2 Desian and Materials ...................................... 20 4.0 SURVEILLANCE REQUIREMENTS . . .................................... 21 4.1 Reactor Core Parameters ................... ................... 21 4.1.1 Excess Reaztivity and Shutdown Margin ..................... 21
4.1.2 Fuel Elements ................. ...................... 2; 4.2 Reactor Control and Safety Systems ... i.................... . . .
4.2.1 Control Rods ............................................. Z2 4.2.2 Reactor Safety System ..................................... 22 4.3 Coolant System ............................................. 3 4.3.1 Primary Coolant Water Purity ................................. 23 4.3.2 Coolant System Radioactivity.. 24 4.4 Confinement .............................. 24 4.5 Ventilation System ........................................... 24 4.6 Radiation Monitoring Systems and Radioactive Effluents . .......
..5 4.6.1 Effluent Monitor .. 25 4.6.2 Rabbit Vent Monitor ....................... .
4.6.3 Area Radiation Monitors (ARMs) .. 25 4.6.4 Portable Survey Instrumentation ........................... 26 5.0 DESIGN FEATURES. ......................... 27 5.1 Site and Facility Description . .......................... 27 5.1.1 Facility Location. .......................... 27 5..2 Exclusion and Restricted Area .......................... 27 5.2 Reactor Coolant System . .......................... 27 5.2.1 Primary Coolant Loop. .......................... 27 5;2.2 Secondary and Tertiary Coolant Loc)PS ...................... 27 5.3. Reactor Core and Fuel . .......................... 27 5.4 Fuel Storage . .......................... 28 6.0 ADMINISTRATIVE CONTROLS. ,......................... 29 6.1 Organization . ..................... .. 29 6.1.1 Structure . ........ I................. 29 6.1.2 Responsibility. ,......................... 29 6.1.3 Staffing .......................- 29 6.1.4 Selection and Training of Personnel ....................... 31 6.2 Review and Audit .. 31 6.2.1 Composition and Qualifications of the ROC .31 E.2.2 ROC Meetinas .............................................. 31 6.2.3 Sub-Committees ......................... .................. 32 E.2.4 ROC Review and Approval Function .32 6.2.5 ROC Audit Function .... 33 is
6.3 Procedures ....................................... .
6.3.1 Reactor Operating Procedures .............................. 3 6.3.2 Administrat-ve Procedures .................................-.
6.4 Experiment Review and Approval ............................... 3 6.4.1 Definitions of Experiments ................................ 8 6.4.2 Approved Experiments ....................................... 35 6.4.3 New Experiments ............................. ......... 36 6.5 Required Actions.. .................................... .36 6.5.1 Action To Be Taken In the Event A Safety Limit Is Exceeded 36 6.5.2 Action To Be Taken In The Event Of A.Reportable Occurrence 36 6.6 Reports................ .. 38 6.6.1 Operating Reports .............................. 38 6.6.2 Special Reports ............................................ 38 6.7 Records .............................. 40 6.7.1 Records to be Retained for a Period of at Least Five Years 40 6.7.2 Records to be Retained for at Least One Requalification Cycle 40 6.7.3 Records to be Retained for the Life of the Facility .......41
1.0 INTRODUCTION
1.1 Scope This document constitutes the Technical Specifica -ions for Facility License No. R-75 and supersedes all prior Technical Specifications.
ncluded are the "Specifications" and the "Basest for the Technical Specifications. These bases, which provide the technical support fcr the individual technical specifications, are included for information purposes only. They are not part of the Technical Specifications, and they do not constitute limitations or requirements to which the licensee must adhere.
This document was, written to be in conformance with ANSIIANS-15.1-1990. The content of the Technical Specifications includes:
Definitions, Safety Limits, Limiting Safety.System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls.
1.2 Application 1.2.1 Purpose These Technical Specifications have been written specifically for The Ohio State University Research Reactor (OSURPR).
The Technical Soecifications represent the agreement between the licensee and the U.S. Nuclear Regulatory Commission on administrative controls, equipment availability, and operational parameters.
Specifications are limits and equipment requirements for safe reactor operation and for dealing with abnormal situations. They are typically derived from the Safety Analysis Report (SAR). These specifications represent a comprehensive envelope for safe operation. Only those operational parameters and equipment requirements directly related to preserving that safe envelope are listed.
1.2.2 Format The.format cf this document is in general accordance with ANSI/ANS-
'5. -1990.
1.3 Definitions Admninistrative Controls - those oraaniza-tsnac2 an6 Procedural requirements established by the Comrazssian and/cr the facility management.
ALARA - as low as is reasonably achievable.
Channel - the combination of sensor, line, amplifier, and output devices which are connected for the purpose c_ measuring the value of a parameter.
Channel Calibration - an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the measured parameter. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip settings, and shall be deemed to include a channel test.
Channel Check - a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the same variable.
Channel Test - the introduction of a signal into the channel for verification that it is operable.
Cold Clean Core - when the core is at ambient temperature and the reactivity worth of xenon is negligible.
Commission - the U.S. Nuclear Regulatory Commission (or NRC).
Confinement - a closure on the overall facility which controls the movement of air into it and out of it through a controlled path.
Containment - a testable enclosure which can support a defined C pressure differential and which is. normally closed.
Control Rod - a device fabricated from neutron absorbing material which is used to establish neutron flux changes.
Control Rod Fuel Element - a fuel element capable of holding a control rod.
Controls - mechanisms used to regulate the operation of the reactor.
Core - the general arrangement of fuel elements and control rods.
Critical - when the effective multiplication factor (keyf) of the reactor is equal to unity.
Direct Supervision - in visual and audible contact.
Excess Reactivity - that amount of reactivity that would exist if all control rods were removed from the core.
2
Exclusion Area - that area around the reactor building in which the licensee has the authority to determine all activities as per 10CFRI00.3.
Experiment - any operatiorr, or any apparatus'-.0device, or material installed in or near the core or which could conceivably have a reactivity effect on the core and which itself is not a core component or experimental facility, intended to investigate non-routine reactor parameters or radiation interaction parameters of materials..
Experimental Facility - any structure or device associated with the reactor that is intended to guide, orient, position, manipulate, or otherwise facilitate completion of experiments.
Explosive Material - any material that is given an Identification of Reactivity (Stability) index of 2, 3, or 4 by the National Fire Protection Association in its publication 704-M, Identification System for Fire Hazards of Materials, or is enumerated in the Handbook for Laboratory Safety published by the Chemical Rubber Company (1967).
Facility - the Reactor Building including offices and laboratories.
Fueled Experiment - any experiment that contains U-235 or U-233 or Pu-239, not including the normal reactor fuel elements.
ticensee - The Ohio State University.
Limiting Conditions for Operation (LCO) - the lowest functional
-capability or performance levels of equipment required fur safe operation.of the facility. LCO are administratively established constraints on equipment and operational characteristics..
Limiting Safety System Settings (LSSS) - settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for.a variable on which a safety limit has been placed, the setting shall be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.
Measured Value - the value of a parameter as it appears on the output of.a channel.
Movable Experiment - one for which it is intended that all or part of the experiment may be moved in relation to the core while the reactor is operating.
Nuclear Regulatory Commission - (NRC).
3
Onset of Nucleate Boiling - (ONB). -
Operable - a component or system which is capable of performing its intended functions in a normal manner.
Operating - a component or system which is performing its intended function.
Protective Action - the initiation of a signal or the operation of equipment within the reactor safety system in response to a variable or condition of the reactor facility having reached a specified limit.
Reactivity Limits - those limits imposed on reactor core excess reactivity based upon a reference core condition.
Reactivity Worth of an Experiment - the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter an experiment's position or configuration.
Reactor - the combination of core, permanently installed experimental facilities, control rods, and connected control instrumentation.
Reactor Operating - whenever the reactor is not secured or shutdown.
Reactor Operations Committee - (ROC). C)
Reactor Operator (RO) - an individual who is licensed to manipulate the controls of the reactor in accordance with 10CFR55.
Reactor Safety Systems - those systems, including their associated input channels, which are designed to initiate automaticreactor protection or to provide information for initiation of manual protective action.
Reactor Secured - whenever (1) all shim/safety rods are fully inserted, (2) the console key is in the OFF position and is removed from the lock, and (3) no in-core work is in progress involving fuel or experiments or maintenance of the core structure, control rods, or control rod drive mechanisms.
Reactor Shutdown - when the reactor is subcritical by at least 1_
delta k/k in the cold clean core condition.
Regulating Rod - a low reactivity-worth control rod. used primarily to maintain an intended power level. Its position may be varied, either by manual control or by the automatic servo-controller.
4
Reportable occurrence - any of the conditions described in Section 6.5.2 of these specifications.
Restricted Area - the Reactor Building to which access is controlled for purposes of protection of individuals fi5m exposure to radiation and radioactive materials.
Safety Analysis Report - (SAR).
Safety Channel - a measuring or protective channel in the reactor safety system.
Safety Limits (SL) - limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity.
Scram - the rapid insertion of the shim/safety rods into the reactor for the purpose of quickly shutting down the reactor.
Scram Time - the elapsed time between reaching a limiting safety system setting and the time when a control rod is fully inserted.
Secured Experiment - any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces JIiiust be substantially greater than those to which the experiment might be subjected from the normal environment of the experiment or
'by forces which can result from credible malfunctions.
'Senior Reactor Operator (SRO) - an individual who is licensed to birect the activities of reactor operators. Such an individual may also operate the controls of the reactor pursuant to 10CFR55.
Shall, Should, and May - the word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word "may" to denote permission, which is neither a requirement nor a recommendation.
Shim/Safety Rods - high-reactivity ,worth control rods used primarily to provide coarse reactor contrQl. They are connected electro-magnetically to their drive mechanisms and have scram capabilities.
Shutdown Margin - the shutdown reactivity necessary to provide confidence that the reactor canrbe made subcritical by means of the control and safety systems withlthe, most reactive shim/safety rod and the regulating rod in the most reactive position (fully withdrawn) and that the reactor will remain subcritical without further operator action.
5,
Standard Fuel Element - an element to be used or stored in the core, fuel storage pit or other approved area, but not a control rod element.
Startup Source - a spontaneous source of neutrons which is used to provide a channel check of the startup (fission chamber) channel and provide neutrons for subcritical multiplication during reactor startup.
Surveillance Time Intervals - The average over any extended period for each surveillance time interval shall be closer to the normal surveillance time, e.g. for the two year interval the average shall be closer to two years rather than 30 months.
two-year (interval not to exceed 30 months).
annually (interval not to exceed 15 months).
semiannually (interval not to exceed 7-1/2 months).
quarterly (interval not to exceed 4 months).
monthly (interval not to exceed 6 weeks).
weekly (interval not to exceed 10 days).
daily (shall be done during the same working day).
Any extension of these intervals shall be occasional and for a valid reason and shall not affect the average as defined.
True Value - the actual value of a parameter.
Unscheduled Shutdowns - any unplanned shutdown of the reactor caused C) by actuation of the reactor safety systems, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation. They do not include those shutdowns resulting from expected testing operations, or planned shutdowns, whether initiated by controlled insertion of control rods or planned manual scrams.
6
2.0 SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS (LSSS) 2.1 Safety Limit Applicability: This specification applies to the melting temperature of the aluminum fuel cladding.
Objective: The objective is to assure that the integrity of the fuel cladding is maintained.
Specification: The reactor fuel temperature shall be less than 550 0C.
Bases: The melting temperature of aluminum is 660 0 C (1220 OF). The blister threshold temperature for UtJSi: dispersion fuel has been measured as approximately 550 -C. (ANL/RERTR/TM-10, October 1987, NRC NUREG 1313). Because the objective of this specification is to prevent release of fission products, any fuel whose maximum temperature reaches 550 'C. is to be treated as though the safety limit has been reached until shown otherwise.
2.2 Limiting Safety System Settings Applicability: This specification applies to the following items associated with core thermodynamics:
(1) Reactor Thermal Power Level and (2) Reactor Coolant Inlet Temperature.
Objective: To assure that the fuel cladding integrity is maintained.
Specification:
(1) Steady state power level shall not exceed 500 kW thermal.
(2) Reactor safety systems settings shall initiate automatic protective action so that reactor thermal power level shall not exceed 600 kW (120; of full power) during a transient.
(3) Reactor safety systems settings shall initiate automatic protective action so that core inlet water temperature shall not exceed 35 0 C.
Bases: The criterion for these safety system settings is established as the fuel integrity. If the temperature of the clad is maintained below that for blister threshold then cladding integrity is
maintained. This is the case for a power level of 600 kW and a core inlet temperature of 35 OC (normal inlet temperature is 20-25 °C).
The maximum credible accident analysis is provided in Section 8.4.3.2 of the Safety Analysis Report. The maximum credible accident assumes steady state operation at 600 kW and a transient to 750 kW.
The maximum temperature of the cladding reaches 91 OC (SAR 8.4.3.3>
One may also reference SAR Sections 4.8.1, 4.e.2 for an estimate of cladding temperature during steady state operation at 500 kW ;5E.5 0
C).
8.
3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 3.1.1 Reactivity Applicability: These specifications apply to the reactivity condition of the reactor and the reactivity worths of the shim/safety rods and regulating rod under any operating conditions.
Objective: To ensure safe shutdown of the reactor and that the safety limits are not exceeded.
Specification: The reactor shall be operated only if the following conditions exist:
(1) The reactor core shall be loaded so that the excess reactivity, including the effects of installed experiments does not exceed 2.6% delta k/k under any operating condition.
(2) The minimum shutdown margin under any operating condition with the maximum worth shim/safety rod and the regulating rod full out shall be no less than 1.0% delta k/k.
(3) The total reactivity worth of the regulating rod shall be less than 0.7% delta k/k.
(4) All core grid positions internal to the active fuel boundary shall be occupied by a standard, control, regulating rod, instrumented, or blank fuel element; or by an experimental facility.
(5) The moderator temperature coefficient shall be negative and shall have a minimum absolute reactivity value of at least 2x10-5/oC across the active core at all normal operating temperatures.
(6) The moderator void coefficient of reactivity shall be negative and shall have a minimum value of at least 2.8x10-3
/1% void across the active-core.
Bases:
-(1) The maximum allowed excess reactivity of 2.6% delta k/k provides sufficient reactivity to accommodate fuel burnup, xenon buildup, experiments, control requirements, and fuel and moderator temperature feedback (Section 4.2 of the SAR).
Also, calculations show that this excess reactivity assures that the maximum temperature of the surface of the cladding will be well below the blister threshold of the U3 Si2 fuel 9
during a design basis accident (SAR 8.4.3.2).
(2) The minimum shutdown margin ensures that the reactor can be shutdown from any operating condition and remain shutdown after cooling and xenon decay even with the highest worth rod and the regulating rod fully withdrawn.
(3) Limiting the reactivity worth of the regulating rod to a value less than the effective delayed neutron fraction assures that a failure of the automatic servo control system cannot result in a prompt critical condition.
(4) The requirement that all grid positions be filled during reactor operation assures that the volume flow rate of primary coolant which bypasses the heat producing elements will be within the range specified in Section 4.8 of the SAR.
Furthermore, the possibility of accidentally dropping an object into a grid position and causing increase of reactivity is precluded.
(5) A negative moderator temperature coefficient of reactivity assures that any moderator temperature rise will cause a decrease in reactivity. The U3Si2 fuel also has a significant negative temperature coefficient of reactivity due to the Doppler broadening of neutron capture resonances in 238U, but no credit is taken for this effect in our safety analyses.
(6) A negative void coefficient of reactivity helps provide reactor stability in the event of moderator displacement by (a-)
experimental devices or other means.
3.2 Reactor Control and Safety System 3.2.1 Control-Rod Drop Times Applicability: This specification applies to the time from the receipt of a safety signal to the time it takes for a shim/safety rod to drop from fully withdrawn to fully inserted.
Objective: To ensure that the reactor can be shutdown within a specified period of time.
Specification: The reactor will not be operated unless the drop time of each of the three shim/safety rods is less than 600 msec.
Bases: Control rod drop times as specified ensure that the safety limit will not be exceeded in a short period transient.
The analysis for this is given in Section 4.3.3 of the SAR.
10
3.2.2 Maximum Reactivity Insertion Rate Applicability: T'his applies to the maximum positive reazctl_:tv insertion rate by the most reactive shim/safety rod and the regulating rod simultaneously.
Objective: To ensure the reactor is operated safely and the safety limit is not exceeded due to a short period.
Specification: The reactor will not be operated unless the maximum.
reactivity insertion rate is less than 0.02i delta k/k per second.
Basis: This maximum reactivity insertion rate assures that the Safety Limit will not be exceeded during a startup accident due to a short period generated by a continuous linear reactivity insertion.
3.2.3 Minimum Number of Scram Channels Applicability: This specification applies to the reactor safety system channels.
Objective: To stipulate the minimum number of reactor safety system channels that shall be operable to ensure the Safety Limits are not exceeded by ensuring the reactor can be shutdown at all times.
Specification: The reactor shall not be operated unless the safety system channels described in the following table are operable.
Reactor Safety System Minimum Function Component Required
1. Core Hi0 Inlet Temp. 1 Slow scram if temp. > 350 C
2. Reactor Thermal power 2 Fast scram if thermal power >
level (Safety Channels) 600 kW, as indicated on calibrated ionization chamber channels.
3. Reactor Period 1 Fast scram if period < 1 sec
4. Reactor Thermal power 1 Slow scram if coolant system level/coolant system pumps pumps not on by > 120 kW thermal power
5. Coolant Flow Rate .1 Slow scram if coolant system has no flow kprimarv) by >
120 kW thermal power 11
Reactor Safety System Minimum Function t '
Component Required
6. Pool Water Level 1 Slow scram if pool level < 20 feet (15 feet above core)
7. Switches 6 Slow scram if any one switch is not properly set at the
a. Magnet Power Key "On" position indicated in quotes
b. Effluent Monitor (Also prohibits startup) counter in "Count"
c. Period Generator Switch "Off"
d. LOG-N Amp Calibrate Switch "Norm"
e. LOG-Period Amp Calibrate Switch "Norm"
f. Effluent Monitor Compressor "On"
8. Recorders 5 Slow scram if power is lost to any one of the listed
a. LOG-N recorders
b. Linear Level
c. Start-Up Channel
d. Period
e. Effluent Monitor
9. Manual Scrams 5 Slow scram upon activation of any one manual scram switch
a. Control Room Console
b. Pool Top Catwalk
c. BSF Catwalk
d. Rabbit/BP Area
e. Thermal Column/BP Area
10. Compensated Ion Chambers 2 Slow scram if voltage drops below operational specifications 12
Reactor Safety System Minimum Function Component Required
11. Safety Set Points On 4 Slow scram if associated Recorders recorder values are exceeded
a. Period < 5 sec
b. Linear Level > 120i of licensed power
c. Start-Up Channel < 2 cts/sec (may be bypassed if Kff < 0.9)
12. Safety System 2 Slow scram in case of a safety amp fault or if system is discontinuous
13. Backup Shutdown Mechanisms 3 Rod drop will occur for any control rod which has excess magnet current > 100 ma Bases:
1. Assures safety limit is not exceeded; core inlet temperature is same as cooling system outlet
2. Assures safety limit is not exceeded
3. Assures safety limit is not exceeded
4. Assures coolant system pumps are functional before raising power > 120 kW
5. Assures there is always primary coolant flow when greater than 120 kW
6. Assures there is enough primary coolant for natural convection cooling
7. Assures nuclear instrumentation is in proper mode for operation
8. Assures information is available for observation by the reactor operator during operation, and is recorded if required as a record of reactor operations
9. Assures that the reactor can be shut down bv the reactor operator in the control room or at other locations near experimental facilities if deemed necessary by other reactor 13
staff
10. Assures shutdown if nuclear instrumentation fails
11. Assures backup shutdown capability from short period or high power level. Assures shutdown if count rate is too low to provide meaningful startup information. The startup interlock may be bypassed if K, is < 0.9
12. Assures all components of the safety system are installed and operational
13. Assures that any control rod exhibiting excess magnet current will be released and fall to the bottom due to gravity 3.3 Coolant System 3.3.1 Pump Requirements Applicability: This specification applies to the operation of pumps for both the primary and secondary coolant loops.
Objective: To ensure that both pumps are functioning whenever the
.eactoris operated above 120 kW.
Specification: The- reactor will not be operated above 120 kW unless both the primary and secondary coolant pumps are activated and there is flow in the primary coolant loop.
Bases: Having both pumps operating and flow in the primary loop will ensure there is adequate cooling of the primary coolant so the Safety Limit is not exceeded.
3.3.2 Coolant Level Applicability: This specification applies to the height of the water in the Reactor Pool above the core.
Objective: To ensure there is adequate primary coolant in the Reactor Pool and sufficient biological shielding above the core.
Specification: The reactor shall not be operated unless there is 20, feet cf water in the reactor pool and 15 feet of water above the core.
Bases: With the pool full of water to a level of 20 feet there is adequate primary coolant for natural convection cooling. With 15 feet of water above the core there is sufficient shielding at the 14
licensed power level. Section 7.l.1.4 of the SAR discusses this shielding.
_.3.3 Water Chemistry Requirements Applicability: This specification applies to the purity of the primary coolant water.
Objective: To-minimize corrosion of the cladding on the fuel elements, and to reduce the probability of neutron activation of ions in the water.
Specification:
(1) The conductivity of the pool water shall not exceed the limit of 2.0 . mho/cm.
(2) The pH of the pool water shall not exceed 8.0.
Bases: Operation in accordance with these specification ensures aluminum corrosion is within acceptable limits, and that the concentration of dissolved impurities that could be activated by neutron irradiation remains within acceptable limits.
3.3.4 Leak, or Loss of Coolant Detection Applicability: This specification applies to the capability of detecting and preventing the loss of primary coolant.
Objective: To ensure there is adequate primary coolant in the Reactor Pool and sufficient biological shielding above the core when the reactor is operating.
Specification: The pool water level shall be at least 15 feet above the top of the fuel in the core.
Bases: The same system that functions to scram the reactor on low pool level will also be used as-the detection system for this specification. Design criteria of the cooling system to prevent large losses of pool water due to siphoning are discussed in Section 3.2.2.1 of the SAR.
3.3.5 Primary and Secondary Coolant Activity Limits Applicabilitv: This specification applies to the buildup of radioactive materials in the secondary coolant system.
Objective: To ensure there is a level low enough so as not to exceed 10CFR20 limits if coolant is released to the sanitary sewer system.
Specification: The primary and secondarY coolant system shall be monitored for the buildup of radicactivity and analyzed at least semiannually for increase in the concentration cf radionuclides.
Basis: The basis for this specification is to ensure releases are legal and consistent with the ALARA Princioal.
3.4 Confinement Isolation Applicability: This specification applies to the capability of isolating the reactor building from the unrestricted area outside.
Objective: To prevent the exposure of the public to airborne radioactivity exceeding the limits of 10CFR20, and the ALAfA principle.
Specification: The reactor shall not be operated unless the following conditions are met:
(1) Ventilation fan operating (2) Reactor Building bay door closed (3) Reactor Building front and rear personnel doors closed (4) Office windows closed Bases: By having the capability to isolate the Reactor Building, the release of airborne radioactive material may be confined and limited to the extent analyzed in the revised SAR of September 1987.
3.5 Ventilation Systems Applicability: This specification applies to all heating, ventilating, and air conditioning systems that exhaust building air to the outside environment.
Objective: To provide for normal ventilation and the reduction of airborne radioactivity within the reactor building during normal reactor operation and to provide a way to turn off all vent systems quickly in order to isolate the building for emergencies.
Specification:
(1) An exhaust fan with a capacity of at least 1000 cfm shall be operable whenever the reactor is operating.
'21 This fan, as well as al! other heating, ventilating, and air conditioning systems shall have the cauabi.ity to be shut _of from a single switch in the control room.
Bases: In the unlikely event of a release of Fission products or Other airborne radioactivitv, the vent-iation sYstem will reduce radioactivitv inside the reactor buildina cr be able to be isolated.
An analysis of fission oroduct release is f66 in section 8.4,4 of the SAR.
3.6 Radiation Monitoring Systems and Radioactive Effluents 3.6.1 Radiation Monitoring Applicability: This specification applies to the availability of radiation monitoring equipment which shall.be operable during reactor operation.
Objective: To assure that monitoring equipment is available to evaluate radiation levels in restricted and unrestricted areas and to be consistent with ALARA.
Specification:
(1) When the reactor is operating, the building gaseous effluent monitor shall be operatina and have a readout and alarm in the control room. It may be used in either the "normal" mode or "sniffer" mode.
(2) When the reactor is operating and the rabbit experimental facility is used, the rabbit monitoring system shall be operating and have a readout and alarm in the control room.
(3) When the reactor is operating, the following Area Radiation Monitors (ARMs) shall be operating and have both local and control room readouts and alarms.
a. Pool Top
b. Primary Cooling System C. Beam Port/Rabbit Area d.- Thermal Column Area; (4) Portable survey instrumentation shall be available whenever the reactor is operating to measure beta-gamma exposure rates and neutron dose rates.
(5) Portable instruments, surveys, or analyses may be substituted for the instruments in the above sections (3.6.1.1, 3.6.1.2, or 3.6.1.3) for periods up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Read-out and alarms from these temporary instruments shall be reported to the reactor operator on duty at least once per hour.
Bases:
17
(11 The gaseous effluent monitor will detect Ar-41 levels in the reactzr building. During "normal" mode operation it will sample and monitor a-r Just-befcre it is released from the reaztcr building. (SAR 6.3.1) Durina "'sniffer" mode of operation it may be used for short periods to monitor in and around experimental facilities to determine local Ar-41 levels.
(2) The rabbit stack monitor is used with the rabbit since the rabbit system uses air as its transport mechanism and Ar-41 production takes place. This monitor will provide warning if Ar-41 levels being released in the building are too high (SAR 6.3.2 and 6.3.4.3).
(3) The ARMs provide a continuing evaluation of the radiation levels within the Reactor Building (SAR 3.7) and provide a warning if levels are higher than anticipated.
(4) The availability of survey meters enables the Reactor Staff to independently confirm radiation levels throughout the building.
(5) In the event of instrument failure short term substitutions will enable the safe continued operation of the Reactor.
3.6.2 Radioactive Effluents Applicability: This specification applies to the monitoring of radioactive effluents from the facility.
Objectives:
(1) To ensure that liquid radioactive releases are safe and legal.
,2) To ensure that the release of Ar-41 beyond the site boundary does not result in concentrations above the Annual Average Concentration (10CFR20.1302 b(i)) for unrestricted area.
(3) To assure that the release of Ar-41 in the restricted area does not result in concentrations above the DAC.
Specifications:
(1) The release rate for radioactive liquids beyond the site boundary shall not exceed the limits as specified in 10CFR20 at the point of release.
(2) The concentration of Ar-41 at the point of release into the 1i
unrestricted area shall not exceed the unrestricted area Annual Average Concentration (AAC) (IOCFR2C.1302 b(i)) when averaged over one year or 10 x AAC when.averaged over one day.
(3) The concentration of Ar-41 in the restricted area shall nst exceed the DAC when averaged over a 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months work year.
Bases:
(1) The basis for this specification is found in Section 6.2 of the Safety Analysis Report.
(2) The basis for this specification is found in Section 6.3 of the Safety Analysis Report.
(3) The basis for this specification is found in Section 6.3 of the Safety Analysis Report and IOCFR20.1003.
3.7 Experiments 3.7.1 Reactivity Limits Applicability: This specification applies to experiments to be
installed in or near the reactor and associated experimental facilities.
Objectives: To prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.
Specification:
(1) The absolute value of the reactivity worth of any single secured experiment shall not exceed..0.7; delta k/k.
(2) The absolute value of the reactivity worth of any single movable experiment shall not exceed 0.4+ delta k/k.
1%3) The absolute value of the reactivity worth of all movable experiments shall not exceed 0.6i delta k/k.
(4) "-The absolute value of the reactivity worth of experiments having moving parts shall be designed to have an insertion
rate less than 0.05i delta k/k per second.
(5) ,,The absolute value of the'zeactivity worth of any movable experiment that may be oscillated shall have a reactivity
,change of less than 0.05; delta k/k.
(6) The total reactivity worth of all experiments shall not be 19
greater than 0.7* delta k/k.
Bases:
(;) The bases for specifications 1, 2, 3, and 6 are found in Section 8.4.3.2 of the SAR which evaluates a step insertion of reactivity from an experiment.
(2) The bases for specifications 4 and 5 allows for certain reactor kinetics experiments to be performed but still limits the rate of change of reactivity insertions to levels that have been analyzed. Section 8.4.3.2 of the SAR evaluates a step insertion of reactivity from an experiment.
3.7.2 Design and Materials Specification:
(1) No experiment shall be installed that could shadow the nuclear instrumentation, interfere with the insertion of a control rod, or credibly result in fuel element damage.
(2) All materials to be irradiated in the reactor shall be either corrosion resistant or doubly encapsulated within corrosion resistant containers.
(3) Explosive materials shall not be allowed in experiments, except for neutron radiographic exposures of items performed outside of the core and experimental facilities. The amount of explosive material contained in capsules used for radiographic exposures shall not exceed 5 grains of, gunpowder.
Bases:
(1) Specification 1 assures no physical interference with the operation of the reactor detectors, control rods, or physical damage to fuel element will take place.
(2) Limiting corrosive materials in Specification 2, and explosives in Specification 3 reduces the likelihood of damage to reactor components and/or releases of radioactivity resulting from experiment failure.
(3) Limiting explosive materials to neutron radiographic exposures done outside of the core and experimental facilities reduces the likelihood of damage resulting for this experimental failure.
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4.0 SURVEILLANCE REQUIREMENTS 4.1 Reactor Core Parameters 4.1.1 Excess Reactiv'tv and Shutdown Margin Applicability: This specification applies to surveillance requirements for determining the excess reactivity of the reactor core and its shutdown margin.
Objective: To assure that the excess reactivity and shutdown margin limits of the reactor are not exceeded.
Specifications:
(1) Whenever a net change in core configuration, for which the predicted change in reactivity is > 0.2; delta k/k, involving grid position is made, both excess reactivity and shutdown margin shall be determined.
(2) Both shutdown margin and excess reactivity shall be determined annually.
Bases: A determination of excess reactivity is needed to preclude operating without adequate shutdown margin. Moving a component out of the core and returning it to its same location is not a change in the core configuration and does not require a determination of excess reactivity.
4.1.2 Fuel Elements Applicability: This specification applies to surveillance requirements for determining the physical condition of the reactor fuel.
Objective: To ensure that visible deterioration, corrosion, or other physical changes to the fuel elements are detected in a timely.
manner.
Specification: All fuel elements, both in-core and out, shall be visually inspected at least once every five years, by inspecting at least one fifth'of the elements annually.
Basis: If the -water purity is continuously maintained within specified limits,, it is projected that chemical corrosion of the fuel glad wjj4j proceed slowly. However, faults in the basic materials or fabz'itation could lead to loss of cladding integrity.
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4.2 Reactor Control and Safety Systems 4.2.1 Control Rods Applicability: This specification applies to the surveillance requirements for the shim safety rods and the regulating rod.
Objective: To assure that all rods are operable.
Specifications:
(1) The reactivity worth of the shimt safety rods and regulating rod shall be determined annually and prior to the routine operation of any new core configuration.
(2) Shim safety rod drop and drive times and reaulating rod drive time shall be determined annually or after maintenance or modification is completed on a mechanism.
(3) The shim safety rods and regulating rod shall be visually inspected annually for indication of corrosion and indication of excessive friction with guides.
Bases: The reactivity worth of the rods is measured to assure the required shutdown margin and reactivity insertion rates are maintained. It also provides a means for determining the reactivity of experiments. Measuring annually will provide corrections for burnup and after core changes assures that altered rod worths will be known prior to continued operations.
The visual inspection of the rods and measurements of drive and drop times are made to assure the rods are capable of performing properly. Verification of operability after maintenance or modification of the control system will ensure proper reinstallation.
4.2.2 Reactor Safety System Applicability: This specification applies to the surveillance requirements for the Reactor Safety System.
Objective: To assure the reactor safety system channels will remain operable and prevent safety limits from being exceeded.
Specification:
(1) A channel check of each measuring channel shawl be performed daily when the reactor is operating.
(2) A channel test of eazh measuring channel shall be perfcrmed 22
prior to each day's operation or prior to each operation extending more than one day.
(31 A channel calibration of the reactor power level measuring channels shall belhade annually. (Linear, Level and LOG-N.!
(4) A channel calibration of the Level and Period Safety Channels shall be made annually. Channel tests are done on these before each day's operation.
(5) A channel calibration of the following shall be made annually
a. Core inlet temperature measuring system
b. Pool water level measuring system
c. Coolant system pumps measuring system
d. Primary coolant flow measuring system (6) The control room manual scram shall be verified to be operable prior to each day's operation. All other manual scram switches shall be tested annually.
(7) Other scram channels shall be tested/calibratedannually.
(e) Any instrument channel replacement shall be calibrated after installation and before utilization.
-C, (9) Any instrument repair or replacement shall have a channel test prior to reactor operation.
Bases: The daily channel tests and checks will assure that the scram channels are operable. Appropriate annual tests or calibrations will assure that long term functions not tested before daily operation are operable.
4.3 Coolant System 4.3.1 Primary Coolant Water Purity Applicability: This specification applies to the conductivity of the primary coolant water.
Objective: To assure high quality pool water.
Specification: The conductivity and pH of the pool water shall be measured weekly.
Bases: This assures that changes that might increase the corrosion rate are detected in a timely manner and that the concentrations of impurities that might be made radioactive do not increase sianificantlv.
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4.3.2 Coolant System Radioactivity Applicability: This specification applies tc the radioactive material in the primary coolant or secondary coolant.
Objective: To identify radionuclides as potential sources of release to the sanitary sewer system.
Specification: Primary and secondary coolant shall be analyzed for radioactivity quarterly or before release.
Bases: Radionuclide analysis of the pool water ox secondary coolant allows for determination of any significant buildup of fission or activation products and helps assure that radioactivity is not permitted to escape to the tertiary system in an uncontrolled manner.
4.4 Confinement Applicability: This specification applies to the surveillance requirements for building confinement.
Objective: To assure that the building closure capability exists.
Specification: A monthly test shall be made to assure that the building exhaust fan, bay door, front and rear personnel doors, and office doors and windows are operable.
Bases: Monthly surveillance of this equipment will verify that the confinement of the reactor bay can be maintained if needed.
4.5 Ventilation System Applicability: This specification applies to the surveillance requirements for the building ventilation system.
Objective: To assure that the ventilation system functions satisfactorily.
Specification:
(1) Ventilation fans and closures shall be checked for proper operation on a quarterly basis.
(2) The shutoff switch for all fans and air conditioning systems shall be tested an a quarterly basis.
Bases: This surveillance will assure that during normal operations the airborne radioactivity will be minimized inside the building and 24
that the building can be isolated quickly if necessary to prevent uncontrolled escape of air-borne radioactivity to the unrestrizted environment.
4.6 Radiation Monitoring Systems and Radioactive Effluents 4.6.1 Effluent Monitor Applicability: This specification applies to the sUrveillance requirement of the effluent monitor.
Objective: To assure the effluent monitor is operational and providing accurate effluent readings.
Specification: The effluent monitor shall have a channel calibration annually and a channel test before each days operation.
Bases: The calibration will assure effluent release estimates are accurate and the test will assure the monitor is operable whenever the reactor is operating.
4.6.2 Rabbit Vent Monitor Applicability: This specification applies to the surveillance requirements of the rabbit vent monitor.
Objective: To assure the monitor is operational and providing meaningful information about effluent releases from the rabbit into the reactor building.
Specification: The monitor shall have a channel calibration annually and a channel test before each day's reactor operation.
Bases: The calibration will assure effluent releases inside the building are accurately estimated and the test will assure the monitor is operable before the rabbit is used.
4.6.3 Area Radiation Monitors (ARMs)
Applicability: This specification applies to the area radiation monitoring equipment.
Objective: To assure that radiation monitoring equipment is operable whenever the reactor is operating.
Specification: A channel test of the ARMs shall be completed before each day's operation and a channel calibration shall be completed annually.'
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Bases: Calibration annually will insure the required reliability and a check on days when the reactor is operated will detect obvious malfunctions in the system.
4.6.4 Portable Survey Instrumentaticn Applicability: This specification applies to the portable survey instrumentation available to measure beta-gamma exposure rates and neutron dose rates.
Objective: To assure that radiation survey instrumentation is operable whenever the reactor is operating.
Specification: Beta-gamma and neutron survey meters shall be tested for operability each day the reactor is to be operated and shall be calibrated annually.
Bases: Tests on days when the reactor is operated will detect obvious detector deficiencies and an annual calibration will assure reliability.
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5.0 DESIGN FEATURES 5.1 Site and Facility Description 5.1.1 Facility Location The reactor and associated equipment is housed in a building at 129e Kinnear Road, Columbus, Ohio. The minimum free air volume of the building housing the reactor will be a 70,000 ft. There is an exhaust fan with dampers providing for contral of release of airborne radioactivity. It is in the area of The Ohio State University Research Center.
5.1.2 Exclusion and Restricted Area The fence surrounding the Research Center shall describe the exclusion area. The restricted area as defined in 10CFR20 shall -
consist of the Reactor Building.
5.2 Reactor Coolant System 5.2.1 Primary Coolant Loop Natural convective cooling is the primary means of heat removal from, the core. Water enters the core at the bottom and flows upward through the flow channels in the fuel elements.
5.2.2 Secondary and Tertiary Coolant Loops The secondary coolant loop removes heat from the primary coolant.
The secondary coolant (ethylene glycol and water) passes through two separate heat exchangers to remove heat if necessary. Heat is removed from the first by an outside fan-forced dry cooler.. City water flow through the secondary side of an additional heat
--exchanger-makes--up. -the tertiary-loop it-provides- additional cooling for the secondary coolant.
5.3 Reactor Core and Fuel Up to 30 positions on the core grid plate are available for use as fuel element positions. Control rod fuel elements occupy four of these positions and one is reserved for the Central Irradiation Facility flux trap. Several arrangements for the c ea critical core have been investigated. Approximate 1y 3 standard fuel elements in addition to the control rod ue e ements are required. Partial elements, core filler elements, and graphite elements may be utilized in various combinations to achieve the proper K excess.
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The reactor fuel is The DOE Standard uraniurri-silicide tU.Si ; with a U-235 enrichment of less than 210. Tt is flat plate fuel with a "meat" thickness of 0.020" and aluminum cladding of 0.015". Standard fuel elements have a total of 16 fueled plates and 2 outer pure aluminum plates. The control rod fuel elements have eight of the inner fuel plates removed to allow the control rods to enter. Pure aluminum guide plates are on the inside of this gap. The outer two plates for each control rod assembly are fueled. Partial elements are also available with 25, 40, 50, and 60 percent of the nominal loading of a standard element. These partial fuel elements are prefabricated by the vendor with fixed numbers of plates.
(1)
References:
NRC NUREG 1313 ANL/RERTR/TM-i0 ANL/RERTR/TM-11 5.4 Fuel Storage The fuel storage pit, located below the floor of the reactor pool and at the end opposite from the core, shall be flooded with water whenever fuel is present and shall be capable of storing a complete core loading. When fully loaded with fuel and filled with water Ks.-
shall not exceed 0.90, and natural convective cooling shall ensure that no fuel temperatures reach a-po-int at which ONB is possible.
5.5 Fuel Handling Tools All tools designed for or capable of removing fuel from core positions or storage rack positions shall be secured when not in use by a system controlled by the supervisor of reactor operations, or the-senior reactor operator on duty.
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6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure The Ohio State University Research Reaztor is a part of the College of Engineering administered by the Engineering Experiment Station.
The organizational structure is shown in Figure 6.1.
6.1.2 Responsibility The Director of the Engineering Experiment Station (Level 1) is the contact person for communications between the U.S. Nuclear Regulatory Commission and The Ohio State University.
The Director of the Nuclear Reactor Laboratory (Level 2) will have overall responsibility for the management of the facility.
The Associate Director (or Manager of Reactor Operations) (Level 3) shall be responsible for the day-to-day operation and for ensuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license and Technical Specifications. During periods when the Associate Director is absent, his responsibilities are delegated to a Senior Reactor Operator (Level 4).
6.1.3 Staffing During Reactor Operations:
(1) Two or more personnel, at least one of whom is a licensed reactor operator, shall be in the building during all reactor operations. The second shall be capable of following simple written instructions for shutting down the reactor.
(2) At least two licensed operators should be in the building during any extended operations (longer than 60 minutes).
(3) Two persons, one of whom shall be a licensed senior reactor operator, shall be in the building for the first start-up of the day.
(4) TWC persons, one of whom shall be a licensed senior reactor operator, shall be in the building during start-up after an unplanned shutdown.
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.Pov~o: St Vice President for Business and Finance Dean, College of Engineering Reactor Operations Committee Director, Engineering Experiment Station I .v-(Level 1) 1 4.
Director, Nuclear Director, Radiation Reactor Laboratory Safety Section (Level 2)
Associate Director, Nuclear Reactor Laboratory (Level 3)
Senior Reactor Operator (Level 4)
Reactor Operations Staff Solid Lines Paths of Direct Responsibility Dashed Lines Paths of Information Figure 6.1: Administrative Organization 30
5) During all operations, a licensed operator shall be in the control room either as console cnerator or directina the activities cf a student operator or trainee.
(6) A minimum of three people shall be present during fuel handling. one shall be a licensed senior reactor operator, and one shall be at least a licensed reactor operator.
6.1.4 Selection and Training of Personnel The selection, training, and requalification of operations personnel shall meet or exceed the requirements of American National Standard for Selection and Training of Personnel for Research Reactors, ANSI/ANS-15.4-1988.
6.2 Review and Audit There shall be a Reactor operations Committee (ROC) which shall review and audit reactor operations to assure the facility is operating in a manner consistent with public safety and within the terms of the facility license. The Committee advises the Director of the NRL, and is responsible to the Provost of The Ohio State University.
6.2.1 Composition and Qualifications of the ROC Committee members shall be appointed annually by the Provost of The Ohio State University. The Committee shall be composed of at least nine members including ex-officio members. The Director and Associate Director of the Nuclear Reactor Laboratory, and the Director of the Office of Radiation Safety shall be ex-officio voting members of the Committee. The remaining Committee.-members shall be faculty, staff, and student representatives of The Ohio State University (but not part of the staff of the Reactor Lab),
having professional backgrounds in engineering, physical, biological, or medical sciences, as well as knowledge of and interest in applications-.of nuclear technology and ionizing radiation.
6.2.2 ROC Meetings The Committee shall meet at least twice each year. It should meet on or about six month intervals. A quorum shall consist of at least 50 percent of the members. Ex-officio members shall be counted in the quorum as follows:
(1) The Provost is an ex-officio member. Since the Provost is not appointed as a member of the ROO, the Provost is not required to act as a member, is not counted as a merber when counting 31
a quorum, but does have the right to vote.
!Q) Ex-officio members who are under the authority of the Provost serve in the same capacity as those who are appointed by the Provost, i.e., they nave the right to vote and are counted as members when counting a quorum.
(3J Ex-officio members, if any, who are not under the authority of the Provost, have the right to vote, but have no obligation to participate. Accordingly, they are not counted as members when counting a quorum.
(4) All ex-officio members hold membership by virtue of their office. They cease to be members when they cease to hold office.
6.2.3 Sub-Committees The chairperson may appoint a Subcommittee from within the Committee membership to act on behalf of the full committee on those matters which cannot await the regular semi-annual meetings. The full Committee shall review the actions taken by the Subcommittee at the next regular meeting.
6.2.4 ROC Review and Approval Function The responsibilities of the ROC include, but are not limited to the following:
(1) Review and approval of experiments utilizing the reactor facilities (2) Review of procedures (3) Review and approval of all proposed changes to the license and technical specifications (4) Determination of whether a proposed change, new test, or experiment would constitute an unreviewed safety question or require a change in the technical specifications per 10CFR50.59 (5) Review of audit reports (6) Review of abnormal performance of plant equipment and operating abnormalities having safety significance as) Review of unusual occurrences and incidents which are remOrtable under 10CFR19, 20, 2', and 50, or Section 6.6.4 of this document, and 32
(8) Review of violations of technical specifications, license, or procedures having safety significance.
Relative to item (1), responsibilitv for re-.7ie6 of experiments a.n a day-to-day basis shall lie with the Director of the Nuclear Reactor Laboratory or his designee. This day-to-day review shall determine whether a specific experiment has previously been approved in the generic sense by the ROC. A semi-annual report of performed experiments shall be provided for ROC review.
Relative to item (2), the NRL Director or his designee shall be responsible for approval of procedures or changes to procedures on a day-to-day basis. He shall provide a summary of all procedure changes to the ROC for their review.
A complete set of minutes of all Committee and Subcommittee meetings, including copies of all documentary material reviewed, and all approvals, disapprovals, and recommendations shall be kept.
Minutes or reports of all Committee meetings or Subcommittee meeting should be disseminated to the Committee members prior to the next regularly scheduled meeting and should be read for approval as the first item on each agenda. A copy of the minutes, or any reports reviewed, should also be forwarded to the Director of the Engineering Experiment Station in a timely manner.
6.2.5 ROC Audit Function A three member Subcommittee shall meet annually to perform an audit of NRL operations and records or review the results of an independent audit completed by another qualified agency. At least two individuals on the Audit Subcommittee shall be ROC members. The third may be a staff member from the Reactor Laboratory or another individual appointed by the ROC chairperson. No member shall audit a function that he is responsible for performing. Each person should serve for three consecutive audits, at which time he or she should be replaced by a new member. In this way, each Subcommittee should consist of two holdovers and one new member. The member servina for his or her second audit should be the Audit Subcommittee Chairperson. The following items shall be audited:
(1) Reactor operations for adherence to facility procedures, Technical Specifications, and license requirements (2) The requalification program for the operating staff, (3) The facility Emergency Plan and implementing procedures, 4
j ) The facility Security Plan and implementing Procedures, and 33
(51 The results of actions taken to correct any deficiencies that affect reactor safety, and
,6J Conformance with the AIARA Policy and the effectiveness of radiologic control.
Deficiencies found by the Audit Subcommittee that affect Reactor Safety shall be reported immediatelxy to the Director of the Engineering Experiment Station. A written report of audit findings should be submitted to the Director of the Engineering Experiment Station and the full Reactor Operations Committee within three months of the audit's completion.
6.3 Procedures 6.3.1 Reactor Operating Procedures Written procedures, reviewed and approved by the Director, or his/her designee, and reviewed by the ROC, shall be in effect and followed. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use of independent judgement and action should the situation require such. All new procedures and changes to existing procedures shall be documented by the NRL staff and subsequently reviewed by the ROC. At least the followina items shall be covered:
(;' Startup, operation, and shutdown of the reactor, (2) Installation, removal, or movement of fuel elements, control rods, experiments, and experimental facilities, (3) Actions to be taken to correct specific and foreseen potential malfunctions of systems or components including responses to alarms, suspected cooling system leaks, and abnormal reactivity changes, (4) Emergency conditions involving potential or actual release of radioactivity including provisions for evacuation, re-entry, recovery, and medical support, (5'1 Preventive and corrective maintenance procedures for systems which could have an effect on reactor safety, (6). Periodic surveillance of reactor instrumentation and safety systems, area monitors, and radiation safety equipment, (7) Implementation Of Security, Emergency and Operator training and requalification plans, and (8) Personnel radiation protection.
34
6.3.2 Administrative Procedures Procedures shall also be written and maintained to assure complianze with Federal regulatiopsf the facility licerlsei, and commitments made to the ROC or other advisory or governing bodies. As a minimum, these procedures shall include:
(1) Audits, (2) Special Nuclear Material accounting, (3) Operator requalification, (4) Record keeping, and (5) Procedure writing and approval.
6.4 Experiment Review and Approval 6.4.1 Definitions of Experiments Approved experiments are those which have previously been reviewed and approved by the ROC. They shall be documented and may be included as part of the Procedures Manual. New experiments are those which have not previously been reviewed, approved, and performed.
Routine tests and maintenance activities are not experiments.
6.4.2 Approved Experiments All proposed experiments utilizing the reactor shall be evaluated by the experimenter and a licensed Senior Reactor Operator to assure compliance with the provisions of the utilization license, the Technical Specifications, and 10CFR Parts 20 and 50. If, in the judgement of the Senior Reactor Operator, the experiment meets with the above provisions, is an approved experiment, and does not constitute a threat to the integrity of the reactor, it may be approved for performance.. When pertinent, the evaluation shall include considerations of:
(1) The reactivity worth of the experiment (2) The integrity of the experiment, including the effects of changes in temperature, pressure, or chemical composition (3) Any physical or chemical interaction that could occur with the reactor components, and
,4) Any radiation hazard that may result from the activation of materials or from external beams 35
6.4.3 New Experiments
~u /
Prior to performing an experiment not previously approved for the reactor, the experiment shall be reviewed and approved by the Reactor Operations Committee. Committee review shall consider the following information:
(1) The purpose of the experiment, (2) The procedure for the performance of the experiment, and (3) The safetv evaluation previously reviewed by a licensed Senior Reactor Operator.
6.5 Required Actions 6.5.1 Action To Be Taken In the Event A Safety Limit Is Exceeded (1) The reactor shall be shut down, and reactor operations shall not be resumed until authorized by the NRC.
(2) The safety limit violation shall be promptly reported to the Director of the Reactor Laboratory.
(3) The safety limit violation shall be reported by telephone to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
(4) A safety limit violation report shall be prepared. The report )
shall describe the following:
a. Applicable circumstances leading to the violation including, when known, the cause and contributing factors,
b. Effect of the violation upon reactor facility components, systems, or structures and on the health and safety of personnel and the public, and
c. Corrective action to be taken to prevent recurrence.
(5) The report shall be reviewed by the Reactor Operations Committee and shall be submitted to the NRC within 14 working days when authorization is sought to resume operation of the reactor.
6.5.2 Action To Be Taken In The Event Of A Reportable Occurrence A reportable occurrence is-any of the following conditions:
(1) operating with any safety system setting less conservative than stated in these specifications, 36 -
(2) Operating in violation of a Limiting Conditior. for Operation established in Section 3 of these specifications.
(3) Safety system component malfunctions bror:oher component or system malfunctions during reactor operation that could, or threaten to, render the safety system incapable of performing its intended function.
(4) An uncontrolled or unanticipated increase in reactivity in.
excess of 0.4% delta k/k, (5) An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of an unsafe condition in connection with the operation of the reactor, and (6) Abnormal and significant degradation in reactor fuel and/or cladding, coolant boundary, or confinement boundary (excluding minor leaks) where applicable that could result in exceeding prescribed radiation exposure limits of personnel and/or the environment.
(7) Any uncontrolled or unauthorized release of radioactivity to the unrestricted environment.
In the event of a reportable occurrence, the following action shall be taken:
(1) The reactor conditions shall be returned to normal, or the reactor shall be shutdown, to correct the occurrence.
(2) The Director of the Reactor Laboratory shall be notified as soon as possible and corrective action shall be taken before resuming the operation involved.
(3) A written report of the occurrence shall be madelwhich shall include an analysis of the cause of the occurrence, the corrective action taken, and the recommendations for measures to preclude or reduce the probability of recurrence.,This report shall be submitted to the Director and the Reactor Operations Committee for review and approval.
(4) A report shall be submitted to the Nuclear Regulatory Commission-in accordance with Section 6.6.2'pf these specifications.
37 I
6.6 Reports Reports shall be made to the-Nuclear Regulatory Commission as fCllows:
6.6.1 Operating Reports An annual report shall be made by September 30 of each year to the Director, Office of Nuclear Reactor Regulation, NRC, Washington, DC 20555, with a copy to the NRC, Region III, in accordance with 10CFR 50.4, providing the following information:
(1) A narrative summary of operating experience (including experiments performed) and of changes in facility design, performance characteristics, and operating procedures related to reactor safety occurring during the reporting period.
(2) A tabulation showing the energy generated by the reactor (in kilowatt hours) and the number of hours the reactor was in use.
(3) The results of safety-related maintenance and inspections.
The reasons for corrective maintenance of safety-related items shall be included.
(4) A table of unscheduled shutdowns and inadvertent scrams, including their reasons and the corrective actions taken.
(5) A summary of the Safety Analyses performed in connection with changes to the facility or procedures, which affect reactor safety, and performance of tests or experiments carried out under the conditions of Section 50.59 of 10CRF50.
(6) A summary of the nature and amount of radioactive gaseous, liquid, and solid effluents released or discharged to the environs beyond the effective control of the licensee as measured or calculated at or prior to the point of such release or discharge.
(7) A summary of radiation exposures received by facility personnel and visitors, including the dates and times of significant exposures.
6.6.2 Special Reports (1) A telephone or telegraph report of the following shall be submitted as soon as possible, but no later than the next working day, to the NRC Region !II Office:
38
(a) Any accidental offsite release of radioactivity above authorized limits, whether or not the release resulted in property damage, personal injury, -r known exposure.
(b) Any exceeding of the safety limit as defined in Secticn 2.1 of these-specifications.
(c) Any reportable occurrences as defined in Section 6.5.2 of these specifications.
(2) A written report shall be submitted within 14 days to the Director, Office of Nuclear Reactor Regulation, US NRC, Washington, DC 20555 with a copy to the NRC Region III, in accordance with 10CFR 50.4, of the following:
(a) Any accidental offsite release of radioactivity above permissible limits, whether or not the release resulted in property damage, personal injury, or known exposure.
(b) Any exceeding of the safety limit as defined in Section 2.1.
(c) Any reportable occurrence as defined in Section 6.5.2 of these specifications.
(3) A written report shall be submitted within 30 days to the Director, Office of Nuclear Reactor Regulation, US NRC, Washingtonj DC 20555, with a copy to the NRC, Region III Office in accordance with 10CFR 50.4, of the following:
(a) Any substantial variance from performance specifications contained in these specifications or in the SAR, (b) Any significant change in the transient or accident analyses as described in the SAR, and (c) Changes in personnel serving as Director, Engineering Experiment Station, Reactor Director,' or Reactor Associate Director.
(4) A report shall be submitted within nine months after initial criticality of the reactor or within 90 days of completion of the startup test program, whichever is earlier, to the Director, Office of Nuclear Reactor Regulation, U.S. NRC, Washington, DC 20555, with a copy to the NRC, Region III upon receipt of a new facility license, an amendment to license authorizing an increase in power level or the installation of a new core of a different fuel element type or design than previously used.
39
The report shall include the measured values of the operating conditions or characteristics of the reactor under the new conditions, and comparisons with predicted values, including the following:
(a) Total control rod reactivity worth, (b) Reactivity worth of the single control rod of highest reactivity worth, and (c) Minimum shutdown margin both at ambient and operating temperatures.
(d) Excess reactivity (e) Calibration of operating power levels (f) Radiation leakage outside the biological shielding (gi Release of radioactive effluents to the unrestricted environment..
6. 7 Records Records or logs of the items listed below shall be kept in a manner convenient for review, and shall be retained for as long as indicated.
6.7.1 Records to be Retained for a Period of at Least Five Years (1) normal plant operation, (2) principal maintenance activities, (3) experiments performed with the reactor, (4) reportable occurrences, (5) equipment and component surveillance activity, (6) facility radiation and contamination surveys,
'7) transfer of radioactive material, (8) changes to operating procedures,. and (9) minutes of Reactor Operations Committee meetings.
8.7.2 Records to be Retained for at Least One Requalification Cycle 40
Regarding retraining and requalification of licensed operations personnel, the records of the most recent complete requalification cycle shall be maintained at all times the :ndividual is emplyved.
6.1.3 Records to be Retained for the Life of the Facility (1) gaseous and liquid radioactive effluents released to the environment, (2) fuel inventories and transfers (3) radiation exposures for all personnel, (4) changes to reactor systems, components, or equipment that may affect reactor safety, (5) updated, corrected, and as-built drawings of the facility.
(6) records of significant spills of radioactivity, and status, (7) annual operating reports provided to the NRC, (e) copies of NRC inspection reports, and related correspondence 41

ML053400287

Text

SUMMARY

SUMMARY

1.0 INTRODUCTION

References:

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