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Purdue University - Safety Analysis Report (Redacted)
	ML111890201
Person / Time
Site:	Purdue University
Issue date:	06/30/2008
From:	Joel Jenkins, Merritt E; Purdue University Research Reactor
To:	; Office of Nuclear Reactor Regulation
	Hardesty, D NRC/NRR/DPR/PRLB 415-3724
References
	Download: ML111890201 (268)
	v • d • e

SAFETY ANALYSIS REPORT for the PURDUE UNIVERSITY PUR-1 REACTOR LICENSE NUMBER R-87 DOCKET NUMBER 50-182 Prepared by:

J. H. Jenkins E. C. Merritt West Lafayette, IN 47907 30 June 2008

TABLE OF CONTENTS TABLE OF CONTENTS ....................................................................................................1 LIST OF FIGURES ......................................................................................................... V LIST OF TABLES ......................................................................................................... VII 1 THE FACILITY ..................................................................................................... 1-1 1.1 Introduction .................................................................................................... 1-1 1.2 Summary and Conclusions on Principal Safety Considerations .................... 1-1 1.3 General Description ....................................................................................... 1-1 1.4 Shared Facilities and Equipment ................................................................... 1-2 1.5 Comparison with Similar Facilities ................................................................. 1-2 1.6 Summary of Operations ................................................................................. 1-2 1.7 Facility Modifications and History .................................................................. 1-3 2 SITE CHARACTERISTICS ..................................... ~ .......................................... ;.. 2-1 .

2.1 Geography/and Demography ........................................................................ 2-1 2.1.1 Site Location and Description ............................ :......................................... 2-1 2.1.2 Population Distribution ......................... ~ ............. :......................................... 2-1 2.2 Nearby Industrial, Transportation and Military Facilities ................................ 2-7 2.3 Climatology and Meteorology ........................................................................ 2-8 2.3.1 General and Local Climate .......................................................................... 2-8 2.3.2 Weather ..................................................................................................... 2-11 2.3.3 Severe Weather ......................................................................................... 2-12 2.4 Hydrology .................................................................................................... 2-14 2.5 Geology and Seismology ............................................................................. 2-14 2.5.1 Regional Geology ...................................................................................... 2-14 2.5.2 Seismology ................................................................................................ 2-15 2.6 References .................................................................................................. 2-21 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS ......................... 3-1 3.1 Design Criteria ............................................................................................... 3-1 3.2 Meteorological Damage ................................................................................. 3-1 3.3 Water Damage .............................................................................................. 3-1 3.4 Seismic Damage ........................................................................................... 3-1 3.5 Systems and Components ............................................................................ 3-2 4 REACTOR DESCRiPTION ................................................................................... 4-1 4.1 Summary Description .................................................................................... 4-1 4.2 Reactor Core ................................................................................................. 4-4 4.2.1 Reactor Fuel ................................................................................................ 4-5 4.2.2 Control Rods .............................................................................................. 4-10 4.2.3 Neutron Moderator and Reflector ............................ ,................................. 4-10

4.2.4 Neutron Startup Source ............... :............................................................. 4-11 4.2.5 Core Support Structure .............................................................................. 4-11 PUR-1 SAR Rev 0, 6/08

4.3 4.4 4.5 Reactor Pool ................................................................................................ 4-11 Biological Shield .......................................................................................... 4-12 Nuclear Design ...................................................................................... ~ ..... 4-12 4.5.1 MCNP Model ............................................................................................. 4-12 4.5.2 Normal Operating Conditions ..................................................................... 4-15 4.5.3 Reactor Core Physics Parameters ............................................................. 4-2,4 4.6 Thermal-Hydraulic Design ........................................................................... 4-37 4.6.1 NATCON Code Description ....................................................................... 4-37 4.6.2 Fuel Element and Fuel Assembly Geometry .............................................. 4-37 4.6.3 Thermal Hydraulic Analysis Results .......................................................... 4-43 4.7 References .................................................................................................. 4-46 5 REACTOR COOLANT SYSTEMS ....................................................................... 5-1 5.1 Summary Description .................................................................................... 5-1 5.2 Primary Coolant System ................................................................................ 5-1 5.3 Secondary Coolant System ........................................................................... 5-2 5.4 Primary Coolant Cleanup System ................................................................... 5-2 5.5 Primary Coolant Makeup Water System ........................................................ 5-3 5.6 Nitrogen-16 Control System .......................................................................... 5-3 5.7 Auxiliary Systems Using Primary Coolant ..................................................... 5-3 6 ENGINEERED SAFETY FEATURES ................................................................... 6-1 6.1 Summary Description .................................................................................... 6-1 6.2 Detailed Descriptions ..................................................................................... 6-1 6.2.1 Confinement ................................................................................................ 6-1 6.2.2 Containment ................................................................................................ 6-2 6.2.3 Emergency Core Cooling System ................................................................ 6-2 7 INSTRUMENTATION AND CONTROL SySTEMS .............................................. 7-1 7.1 Summary Description .................................................................................... 7-1 7.2 Design of Instrumentation and Control Systems ............................................ 7-2 7.2-.1 Channel 1-Start-up Channel ...................................................................... 7-2 7.2.2 Channel 2-Log N and Period Channel ....................................................... 7-3 7.2.3 Channel 3-Linear Power ............................................................................ 7-3 7.2.4 Log-N Period Channel ................................................................................. 7-4 7.2.5 Channel 4-Safety Channel ...................................... ~ ................................. 7-5 7.3 Reactor Control System ........................................................... ~ .................... 7 7.4 Reactor Protection System ............................................................................ 7-8 7.5 Control Console and Display Instruments ................................................... 7-10 7.6 Radiation Monitoring Systems ..................................................................... 7-10 8 ELECTRICAL POWER SYSTEMS ...................................................................... 8-1 8.1 Normal Electrical Power Systems .................................................................. 8-1 8.2 Emergency Electrical Power System ............................................................. 8-1

9 9.1 AUXILLARY SYSTEMS ....................................................................................... 9-1 PUR-1 SAR Heating, Ventilation and Air Conditioning Systems ........................................ 9-1 ii Rev 0, 6/08

9.2 9.3 9.4 9.5 Fuel Storage and Handling ............................................................................ 9-1 Fire Protection Systems ................................................................................ 9-2 Communication Systems ............................................................................... 9-3 Possession and Use of Byproduct, Source and Special Nuclear Material. .... 9-3

, 10 . EXPERIMENT FACILITIES AND UTILIZATION ............................................ 10-1 10.1 Summary Description .................................................................................. 10-1 10.2 Experimental Facilities ................................................................................. 10-1 10.3 Experiment Review ...................................................................................... 10-1 10.4 References .................................................................................................. 10-1 11 RADIATION PROTECTION AND WASTE MANAGEMENT .......................... 11-2 11.1 Radiation Protection Program ...................................................................... 11-2 11.1.1 Radiation Sources .................................. ;................................................ 11-2 11.1.2 Radiation Protection Program .................................................................. 11-2 11.1.3 ALARA Commitment.. .............................................................................. 11-3 11.1.4 Radiation Monitoring and Surveying ........................................................ 11-3 11.1.5 Radiation Exposure Control and Dosimetry ............................................. 11-3 11.1.6 Contamination Control ............................................................................. 11-4 11.1.7 Environmental Monitoring ........................................................................ 11-4 11.2 Radioactive Waste Management.. ............................................................... 11-4 11.2.1 Radioactive Waste Management Program .............................................. 11-4 11.2.2 Radioactive Waste Controls .................................................. ~ .................. 11-4 11.2.3 Release of Radioactive Waste ................................................................. 11-5 11.3 Conclusions ................................................................................................. 11-5 12 CONDUCT OF OPERATIONS ........................................................................ 12-1 12.1 Organization ................................................................................................. 12.:.1 12.1.1 Structure .................................................................................................. 12-1 12.1.2 Responsibility ..................................................... :..................................... 12-2 12.1.3 Staffing, Selection a,nd Training of Personnel. .......................................... 12-2 12.1.4 Radiation Safety ...................................................................................... 12-3 12.2 Review and Audit Activities ......................................................................... 12-3 12.3 Procedures .................................................................................................. 12-4 12.4 Required Actions ......................................................................................... 12-4 12.5 Reports ........................................................................................................ 12-4 12.5.1 Annual Operating Reports ...................................................... ;................ 12-4 12.5.2 Non-Routine Reports ............................................................................... 12-5 12.5.3 Unusual Events ........................................................................................ 12-5 12.6 Records ................... ,................................................................................... 12-5 12.7 Emergency Planning ................................................................................... 12-5 12.8 Security Planning ......................................................................................... 12-6 12.9 Quality Assurance ....................................................................................... 12-6 12.10 Operator Training and Requalification ..................................................... 12-6

13 ACCIDENT ANALYSES ................................................................................. 13-1 13.1 Accident-Initiating Events and Scenarios .................................................... 13-1 PUR-1 SAR iii Rev 0,6/08

13.1.1 Maximum Hypothetical Accident .............................................................. 13-1 13.1.2 Insertion of Excess Reactivity .................................................................. 13-1 13.1.3 Loss of Coolant ........................................................................................ 13-3 13.1.4 Loss of Coolant Flow ............................................................................... 13-3 13.1.5 Mishandling or Malfunction of Fuel .............................. ;........................... 13-4 13.1.6 Experiment Malfunction ........................................................................... 13-4 13.1.7 Loss of Normal Electrical Power .............................................................. 13-5 13.1.8 External Events ........................................................................................ 13-5 13.2 Accident Analysis and Determination of Consequences ............................. 13-6 13.2.1 Maximum Hypothetical Accident (Failure of a Fueled Experiment) .......... 13-6 13.2.2 Insertion of Maximum Allowed Excess Reactivity .................................. 13-10 13.3 Summary and Conclusions ........................................................................ 13-14' 13.4 References ................................................................................................ 13-15 14 TECHNICAL SPECIFICATIONS .................................................................... 14-1 14.1 Summary Description of the Document.. ..................................................... 14-1

-14.2 Administrative Control of Technical Specifications ...................................... 14-1 15 FINANCIAL QUALIFICATIONS ..................................................................... 15-2 15.1 Financial Ability to Operate a Non-Power Reactor ...................................... 15-2 15.2 Financial Ability to Decommission the Facility ............................................. 15-2 APPENDIX 1: PUR-1 DRAWINGS .................................................................................1 APPENDIX 2: NATCON INFORMATION .......................................................................1 A 2.1 Hot Channel Factors in the NATCON Code Version 1.0 ................................ 1 Hot Channel Factors in the NATCON Code Version 2.0 .............................................. 1 Flow Rate in a Coolant Channel versus Power of a Fuel Plate .................................... 3 Bulk Coolant Temperature Rise versus Power ofa Fuel Plate ..................................... 7 Formulas for Hot Channel Factors ............................................................................... 9 Question 28 ................................................................................................................12 Question 29 ................................................................................................................13 Question 30 ...................................................................................................................14 Question 32 ................................................................................................................ 15 Question 33 ................................................................................................................16 Question 36 ................................................................................................................ 17 APPENDIX 3: FUEL SPECiFiCATIONS ........................................................................1

PUR-1 SAR iv Rev 0,6/08

LIST OF FIGURES Figure 2-1: State of Indiana, showing location of Tippecanoe County ........................ 2-2 Figure 2-2: Map of West Lafayette, Indiana, showing inset picture of the location of the*

Electrical Engineering Building on the Purdue Campus. 2 .................................. 2-3 Figure 2-3: Map of the Duncan Annex of the Electrical Engineering Building ............. 2-4 Figure 2:-4: Population of area surrounding PUR-1 reactor ......................................... 2-5 Figure 2-5: Population projection for 2030 in area surrounding PUR-1 reactor. 3........ 2-6 Figure 2-6: . Map showin~ the location of the Purdue University airport in relation to the location of PUR-1 ............................................................................................. 2-8 Figure 2-7: Map of the 1-Hz spectral acceleration for 2% probability of exceedance in 50 years for the Central and Eastern United States in standard gravity (g) .... 2-16 Figure 2-8:. Map of the 5-hertz spectral acceleration (SA) for 2% probability of exceedance in 50 years in the Central and Eastern United States in standard gravity (g).13 .................................................................................................... 2-17 Figure 2-9: Map of peak ground acceleration (PGA) for 2% probability of exceedance in 50 years in the Central, and Eastern united States in standard gravity (g).13 ... 2-18 Figure 2-10: Map of 1-hertz spectral acceleration (SA) for 10% exceedance in 50 years in the Central and Eastern United States in standard gravity (g).13 ................. 2-19 Figure 2-11: Map of peak grou'nd acceleration (PGA) for 10% probability in 50 years in

the Central and Eastern United States in standard gravity (g).13 ..................... 2-20 Figure 4-1: PUR-1 Pool Layout.. ................................................................................. 4-1 Figure 4-2: PUR-1 Grid Plate ...................................................................................... 4-5 Figure 4-3: Standard assembly can detail, showing wall spacers ............................... 4-7 Figure 4-4: Control assembly can detail, showing wall spacers .................................. 4-8 Figure 4-5: Standard fuel assembly ................................................... ~ ........................ 4-9 Figure 4-6: Control fuel assembly ............................................................................... 4-9 Figure 4-7: Plates are identified by unique serial numbers as shown here ............... 4-10 Figure 4-8: Dummy plates are differentiated from the fuel plates by a notch machined into the end of the plate ................................................................................... 4-10 Figure 4-9: PUR-1 Reactor pool temperature measurements from 1994-2007 .... :... 4-12 Figure 4-10: plate in X-Z plane .............................................................. :.................. 4-13 Figure 4-11: plate in Y-Z plane. (magnified) .............................................................. 4-13 Figure 4-12: plate in X-V plane, cutaway view .......................................................... 4-13 Figure 4-13: Model representation of standard assembly plate spacing detail showing wall spacers .................................................................................................... 4-14 Figure 4-14: Comparison of standard and control assemblies in the model. ............ 4-14 Figure 4-15: Model of regulating rod assembly ......................................................... 4-15 Figure 4-16: Representation of LEU core load .......................................................... 4-15 Figure 4-17: PUR-1 Core Layout and Bundle Drawings ........................................... 4-18 PUR-1 SAR v Rev 0,6/08

Figure 4-18: Axial Power Profiles in Plates 1348, 1228 and 1315 for Banked Rod Critical Configuration in PUR-1 ....................................................................... 4-21 Figure 4-19: Radial Power Profiles in Plates 1345, 1228 and 1315 for Banked Rod Critical Configuration in PUR-1 ....................................................................... 4-21 Figure 4-20: Calibration curve for SS-1 rod with calculated and measured values ... 4-26 Figure 4-21: Calibration curve for SS-2 rod with calculated and measured values ... 4-26 Figure 4-22: Calibration curve for RR with calculated and measured values ............ 4-27 Figure 4-23: Figure showing water regions for perturbation models ......................... 4-32 Figure 4-24: PUR-1 Core Layout with LEU Fuel. ...................................................... 4-33 Figure 4-25: Density of sub-cooled water at 1.5 atmospheres ................................... 4-35 Figure 4-26: Effect of Fuel Temperature, Water Temperature and Water Density Perturbations on Core Reactivity ......... : ........................................................... 4-36 Figure 4-27: Standard LEU fuel assembly................................................................ 4-37 Figure 4-28: Control LEU fuel assembly ................................................................... 4-38 Figure 9-1: MCNP model of in-pool fuel storage rack ................................................. 9-1 Figure 9-2: k-eff Values for Flooded Condition of Fuel Storage Facility ...................... 9-2 Figure 9-3: MCNP model of dry fuel storage facility .................................................... 9-2 Figure 12-1 ................................................................................................................ 12-1 Figure 13-1: Power and Clad Temperatures for 0.6%~k/k step insertion with scram. 13-

11 Figure 13-2:Power and Clad Temperatures for 0.6%~k/k step insertion with scram ... 13-12 Figure 13-3: Power and Clad Temperatures for 0.6%~k/k step insertion without scram .

.. ~ .......... ;................................ :........................................................................ 13-13 Figure 13-4: Power and Clad Temperatures for 0.6%~k/k slow insertion without scram .

...................................................................................................................... 13-14

PUR-1 SAR vi Rev 0,6/08

LIST OF TABLES Table 1-1: Summary of amendments and changes to the PUR-1 reactor facility ........ 1-3 Table 2-1: Population Data for Reactor Vicinity, centered on reactor location ............ 2-1 Table 2-2: Purdue University campus population detail for 1998-2008 ....................... 2-7 Table 2-3: Climatography Data for West Lafayette, Indiana; Mean and Extreme Temperatures .................................................................................................. 2-10 Table 2-4: Climatography Data for West Lafayette, Indiana; Mean Days Information 6 .. 2-10 Table 2-5: Precipitation Normals for West Lafayette, Indiana (Station: West Lafayette 6 NW) ........................................................................................., .................... 2-11 Table 2-6: Average and Maximum Wind Data Measured at Purdue University Airport for 1977-2006 ....... :............................................................................................... 2-11 Table 2-7: Tornados Reported in Tippecanoe County, Indiana between 01/01/1950 and 02/28/2008 ...................................................................................................... 2-13 Table 4-1: Summary of Design Parameters for PUR-1 ............................................... 4-3 Table 4-2: Summary of key reactor parameters for PUR-1 ......................................... 4-4 Table 4-3: Characteristics of the PUR-1 Fuel Plates .................................................. 4-6 Table 4-4: Channel Types and Thickness in PUR-1 Assemblies ................................ 4-9 Tab!e 4-5: Summary of control rod characteristics as listed in the PUR-1 Operations Manual. ........................................................................................................... 4-10 Table 4-6: Representation of fuel plates ................................................................... 4-13 Table 4-7: Bundle Powers Predicted by f7 and f6 Tallies in MCNP .......................... 4-19 Table 4-8: Plate Power (W) Computed from Heating Tallies in Bundles 4-4, 3-3 and 3-4 in PUR-1 Core with 190 Fuel Plates ................................................................ 4-20 Table 4-9: Axial and Radial Heating Profile for Plate 1348 of Bundle 4-4 for Banked Critical Configuration ....................................................................................... 4-22 Table 4-10: Axial and Radial Heating Profile for Plate 1228 of Bundle 3-4 for Banked Critical Configuration ....................................................................................... 4-23 Table 4-11: Axial and Radial Heating Profile for LEU Plate 1315 of Bundle 4-3 for Banked Critical Configuration .......................................................................... 4-24 Table 4-12: Comparison of HEU and LEU calculated control rod worths .................. 4-25 Table 4-13: Comparison of calculated and measured maximum reactivity insertion rates ................................................................................................................ 4-27 Table 4-14: Comparison of calculated and measured shutdown margins .................. 4-29 Table 4-15: Other core physics parameters .............................................................. 4-32 Table 4-16: Water and Fuel Coefficients for the PUR-1 Core ................................... 4-32 Table 4-17: Channel Types and Thickness in PUR-1 Assemblies ............................ 4-38 Table 4-18: Model Dimensions for the Thermal-Hydraulic Models ........................... 4-40 Table 4-19: Hot Chann~1 Factors for the LEU core ................................................... 4-40 Table 4-20: ONB Power With Hot Channel Factor FW to Account for Hydrodynamically PUR-1 SAR vii Rev 0, 6/08

Table Table Table Developing Flow, and the Effect of Viscosity Temperature Dependence on FBULK (Most Limiting Fuel Plate) ................................................................... 4-41 4-21: ONB Powers for HEU and LEU Cores ................................................... 4-44 4-22: Operating Conditions for PUR-1 as Determined by NATCON ............... 4-44 4-23: Key Power Levels for Reactor Operation and LSSS for PUR-1 ............. 4-45 Table 4-24: Power Levels at the Onset of Nucleate Boiling for HEU and LEU Cores ... 4-45 Table 5-1: PUR-1 water process system summary ..................................................... 5-1 Table 7-1: Summary of PUR-1 Instrumentation .......................................................... 7-1 Table 11-1: PUR-1 Personnel Exposuresfor2003-2007 .......................................... 11-3 Table 13-1: Peak power and clad temperature for trip and no-trip insertions of 0.6%

.ilk/k ...................................... ,.......................................................................... 13-3 Table 13-2: Dose Rates in the Reactor Room from a Failed Fuel Experiment (Power =

1 Watt) ............................................................................................................ 13-8 Table 13-3:Oose Rates at 100 meters ....................................................................... 13-9 Table 13-4: Peak power and clad temperature for insertions of 0.6% .ilk/k with scram .

...................................................................................................................... 13-11 Table 13-5: Peak power and clad temperature for trip and no-trip insertions of 0.6%

.ilk/k ............................................................................................................... 13-13

PUR-1 SAR viii Rev 0,6/08

... 1 1.1 THE FACILITY Introduction This report is submitted in support of the application for renewal of the operating license (R-87) for Purdue University Reactor (PUR-I) for a period of 20 years, and a power increase from 1 kW, to 10 kW continuous steady state operation, with a maximum allowed power of 12.5 kW.

The reactor is located in the Nuclear Laboratories in the Duncan Annex of the Electrical _

Engineering Building on the eastern edge of the campus in West Lafayette, Indiana.

The Duncan Annex is of brick and concrete block construction and was originally built as a high voltage laboratory. In 1962 the reactor was built in half of the existing high voltage laboratory, which was a high bay area. Offices, classrooms, and laboratories had been built in the remainder of the original building.

1.2 Summary and Conclusions on- Principal Safety Considerations The design power level for PUR-1 is 10 kW, and the reactor has operated safely since its construction in 1962. The analyses presented in this SAR support the continued operation of PUR-1, and also support the case for a power uprate to continuous operation at 10 kW, and maximum short term power of 12.5 kW .

The limiting safety system settings then of 110% and 120% power (11 kW and 12 kW, respectively) are then adequate to protect the reactor and fuel. Even in the unlikely case of a failure of the reactor protective system, the reactor is self-protecting, with a calculated maximum power level of 2.38 MW, and a maximum clad temperature of 133°C, which is still well below the safety limit of 530°C.

1.3 General Description The PUR-I is a 10 kW pool type reactor, previously licensed for operation at 1 kW, utilizing MTR type enriched fuel plates, which are graphite reflected, and light water moderated and cooled. It was designed and built by Loc;:kheed Nuclear Products of Lockheed Aircraft Corp.,. Marietta, Georgia.

The reactor is controlled by three blade-type control rods located in the core region of the reactor. There are two shim-safety rods made of solid borated stainless steel, utilizing a magnetic clutch between the blades and the lead screw operated drive mechanisms, and a regulating rod which is a screw operated direct drive and made of hollow stainless steel. Each control blade is protected by an aluminum guide plate on each side within the fuel assembly. .

PUR-1 SAR 1-1 Rev 0,6/08

... 1.4 Shared Facilities and Equipment The reactor facility is located within the former high voltage laboratory (Duncan Annex) in the Electrical Engineering Building. This space was converted prior to the construction of PUR-1 to also house classrooms and laboratories.

1.5 Comparison with Similar Facilities Similar research reactors are in use at the University of Missouri at Rolla, and the Ohio State University. The safe operating histories of these reactors, and PUR-1 demonstrate the reliability and safety of these systems. Both Missouri-Rolla and Ohio State are licensed for operation at much higher powers (200 kW for Rolla, and 600 kW for Ohio State), with similar reactor systems. The safe operation of these reactors at their respective higher powers also supports the case for an uprate for PUR-1.

1.6 Summary of Operations The PUR-1 reactor has been in operation since 1962. It is used for teaching and research to support the mission of Purdue University Nuclear Engineering, and the university as a whole~ The reactor operates about 90 times per year on average, and presently has three licensed senior operators. There have been periods, however,

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when there has only been one operator on staff.

PUR-1 SAR 1-2 Rev 0,6/08

1.7 Facility Modifications and History Table 1-1 summarizes the facility modifications and history.

Table 1-1: Summary of amendments and changes to the PUR-1 reactor facility.

May 1964 Amendment 1 Permit 10 kW Operation December 1965 Installation of pool traversing mechanism completed July 1966 Amendment 2 License Renewal October 1968 Change 1 Installation of stainless steel liner September 1969 Installation of air conditioner January 1972 Change 2 Change of pH of pool water February 1974 Change 3 Regeneration of demineralizer procedure change November 1978 Amendment 3 Technical Specifications August 1980 Amendment 4 Physical security plan February 1981 Installation of catwalk around air conditioner March 1981 Amendment 5 Physical security plan September 1982 Amendment 6 Technical specifications-revision 1:

Surveillance intervals, scram initiation October 1982 Amendment 7 Technical specifications-revision 2: Minor typographical modifications April 1983 Amendment 8 Technical specifications-revision 3:

Surveillance intervals, RCO qualifications August 1988 Amendment 9 License Renewal February 2000 Amendment 10 Technical specifications-revision 4: CORO members June 2007 Amendment 11 Possession limit increase August 2007 Amendment 12 HEU to LEU conversion order

PUR-1 SAR 1-3 Rev 0,6/08

2 SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Site Location and Description The PUR-I reactor is located in the Duncan Annex of the Electrical Engineering Building on the campus of Purdue University in West Lafayette, Tippecanoe County, in the State of Indiana, as shown in Figures 2.1, 2.2, and 2.3. The Lafayette-West Lafayette area is about 60 miles northwest of Indianapolis, the State Capitol, and about 140 miles south-southeast of Chicago, Illinois.

2.1.2 Population Distribution According to the 2000 census summary for Tippecanoe County, the total population was 148,955. The estimate of the county population for 2006 was 156,169 1 . Figure 2-4 shows the population within 1, 2, 4 and 8 km from the reactor location (summarized in Table 2-1) and Figure 2-5 shows the projected population in 2030. It should be noted that the population projections, given to Purdue by the Tippecanoe Area Planning Commission, use a *different system for calculation, and are not as accurate as the 2000 population report.

Table 2-1: Population Data for Reactor Vicinity, centered on reactor location .

Projected 2030 Circle radius Population (2000) Population 1 km 17,156 18,325 2 km 31,352 31,992 4km 60,828 68,628 6 km 93,761 117,271 8 km 117,285 165,374

PUR-1 SAR 2-1 Re~ 0, 6/08

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2-2 Rev 0, 6/08

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PUR-1 SAR 2-3 Rev 0, 6/08

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PUR*1 SAR 2-4 Rev 0 , 6/08

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2.2 Nearby Industrial, Transportation and Military Facilities The Purdue University enrollment for 2007-08 was 39,102 full and part-time students, and Purdue University employs approximately 15,3'04 faculty and staff members (2007-08). So, an approximate total campus population is approximately 54,400 at peak times. The data for the last ten years is detailed in Table 2-2: Purdue University campus population detail for 1998-2008. 4 98-99 99-00 00-01 01-02 02-03 03-04 04-05 05-06 06-07 07-08 Full-Time Students 32,788 33,725 33,907 34,442 34,563 34,867 34,745 34,968 35,497 35,549 Part-Time Students 4,090 4,037 3,964 3,766 4,001 3,980 3,908 3,744 3,731 3,553 Total Students 36,878 37,762 37,871 38,208 38,564 38,847 38,653 38,712 39,228 39,102 Total Faculty/Staff 12,888 13,144 13,411 13,831 14,052 14,329 14,636 14,966 15,217 15,304 Total Campus Population 49,766 50,906 51,282 52,039 52,616 53,176 53,289 53,678 54,445 54,406 Purdue University owns an airport (LAF) at the southwest edge of the West Lafayette Campus, as shown in Figure 2-6 below. It has two runways, the longest of which is Runway 10/28, which is 6600'x150', and the other is Runway 5/23, which is 4230'x1 00'.

The Purdue airport averages about 130,000 aircraft operations annually (115,000 for the calendar year 2007), and it is the second busiest airport in Indiana .

The university has a thriving Aviation Technology program, the flight instruction department of which constitutes the majority of the air operations at the Purdue Airport.

, During the fall and spring semester, the airport is the busiest, with an average of 750 takeoffs and landings per day. The maximum daily takeoffs and landings is 1000, and the minimum is zero (0), on Christmas day. During the summer, the airport sees and average of 30 flight operations per day.5 The students primarily use Runway 5/23 (outlined in red on the figure), the nearest end of which is 6344 feet (1 93 km) to the southwest (228°) from the Duncan Annex of EE.

Only light aircraft can use this runway, due to its length. The approach to this runway passes near the EE building. Larger aircraft must use the longer runway (10/28), which does not direct aircraft either on approach or takeoff near the EE building.

Due to the size restrictions of aircraft that use the 5/23 runway, and the heavy concrete block concrete and brick windowless construction of the Duncan Annex, the location of th , and the fact that no aircraft have crashed on this section of campus in istory the program, damage due to a crashed light aircraft does not pose' a significant threat to the safe operation of the reactor. Should an aircraft im

~ location of the reactor

_ in a tank of water will mitigate any potential imm iate mage, and the close proximity of the Purdue Fire Department will ensure that the public safety is maintained in the event of extensive fire damage or explosions .

PUR-1 SAR 2-7 Rev 0,6/08

r-?

Figure 2-6: Map showing the location of the Purdue University airport in relation to the location of PUR-1.2 2.3 Climatology and Meteorology 2.3.1 General and Local Climate The climate of the county is continental with hot summers and cold winters. The seasons are strongly marked, and the weather is frequently changeable. Climatological data available from NOAA for West Lafayette are summarized in

PUR-1 SAR 2-8 Rev 0,6/08

Table 2-3, Table 2-4, and Table 2-5. The tables show the conditions are measured at the West Lafayette station (West Lafayette 6NW), where the latitude is 40°28', the longitude is 8roo', and the ground elevation is 705 feet.

PUR-1 SAR 2-9 Rev 0,6/08

Table 2-3: Climatography Data for West Lafayette, Indiana; Mean and Extreme Mean* (OF)

Daily Daily Highest Temperatures6 .

Highest Month Extremes (oF' Dailyt Year Day Mean Year Dailyt lowest lowest Month Month Max Min Mean Year Day Mean* Year Jan' 31.5 15.0 23.3 68 1906 20 35.9 1990 -24+ 1985 20 7.7 1977 Feb 36.8 19.0 27.9 73 2000 26 38.9 1998 -23 1963 26 12.4 1978 Mar 48.4 29.1 38.8 87 1910 24 46.5 1973 -12 1960 1 29.6 1984 Apr 60.9 39.2 50.1 90 1930 11 55.7 1985 7 1982 7 44.6 1982 May 72.5 50.3 61.4 96+ 1911 27 68.9 1977 26+ 1966 10 ,56 1997 Jun ' 81.4 59.6 70.5 104 1934 1 75.0 1991 35 1992 22 65.8 1972 Jul 84.5 63.0 73.8 111 1936 14 77.9 1983 42+ 1972 5 70 1996 AUQ 82.5 60.6 71.6 103 1918 *5 77.7 1995 35 1965 29 66.7 1992 Sep 77.0 52.9 65.0 . 100 .1933 9 69.4 1978 25 1995 23 59.4 1993 Oct 64.8 41.6 53.2 90+ 1922 2 61.3 1971 18 1925 30 47.3 1988 Nov 50.0 32.2 41.1 78+ 1930 19 46.8 1999 -3 1930 28 32.9 1976 Dec 37.0 21.1 29.1 71+ 1982 3 38.6 1982 -22 1989 22 16.5 2000 Jul Jul Jan Jan Annual 60.6 40.3 50.5 111 1936 14 77.9 1983 -24+ 1985 20 7.7 1977

Derived from 1971-2000 serially complete data t Derived from station's available digital record: 1901-2001

+ Also occurred on an earlier date(s)

Table 2-4: Climatography Data for West Lafayette, Indiana; Mean Days Information 6 Month Days Days Mean Number of Days

Days Days Days Days Min ~ OaF Max ~ 100°F Max ~ gO°F Max ~ 50°F Max ~ 32°F Min ~ 32°F Jan 0.0 0.0 2.4 15.9 28.7 5.6 Feb 0.0 0.0 4.6 10.7 24.5 3.5 Mar 0.0 0.0 '13.1 3.3 20.5 0.1 Apr 0.0 0.0 24.3 0.2 7.8 0.0 May 0.0 0.4 30.7 0.0 0.6 0.0 Jun @ 4 30 0.0 0.0 0.0 Jul 0.1 5.7 31 0.0 0.0 0.0 Aug 0.0 3.2 31 0.0 0.0 0.0 Sep 0.0 1.7 30 0.0 0.3 0.0 Oct 0.0 @ 28.6 0.0 5.5 0.0 Nov 0.0 0.0 14.5 1.7 16.6 0.0 Dec 0.0 0.0 4.4 10 26.7 2.4 Annual 0.1 15 244.6 41.8 131.2 11.6

Derived from 1971-2000 serially com plete data

@ Denotes mean number of days greater than 0 but less than 0.05

The average annual temperature is about 50°F. The mean temperature in January, the coldest month, is 23°F, and in July, the warmest month, is 73.8°F. About nine days per PUR-1 SAR 2-10 Rev 0,6/08

year the temperature falls below zero, and about 137 days per year the temperature goes below freezing (32°F).

Table 2-5: Precipitation Normals for West Lafayette, Indiana (Station: West Lafayette 6 NW)

PRECIPITATION NORMALS(Total in Inches)

JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC ANNUAL 1.79 1.57 2.84 3.57 4.35 4.24 4.00 3.68 2.98 2.73 3.08 2.43 37.26 The average annual precipitation is 35.68 inches. July is the wettest month with 4.74 inches, and February is the driest month with 1.41 inches of precipitation. Prevailing winds are from the west or southwest during the winter and from the south during the summer. Wind velocity is highest in February and lowest in August. 1 2.3.2 Weather Wind conditions as measured at the Purdue University Airport (West Lafayette 6NW) are summarized over the period 1977 to 2006, and are detailed in Table 2-6. These data indicate an annual mean wind speed of 8.78 miles per hour and a maximum wind speed of 72 miles per hour. According to the Unified Building Code, 1985 edition, the

Purdue University lies in the maximum wind zone of 80 miles per hour, which translates to a wind load of 17 pounds per square foot. Buildings at Purdue University are designed to withstand this wind load.

Table 2-6: Average and Maximum Wind Data Measured at Purdue University Airport for

. 1977-20067 Avg. Win~

Direction Avg. Wind Max. Wind Month (degrees) Speed (mph) Speed t (mph)

January 216.91 10.27 64 February 205.39 9.84 62 March 198.20 10.51 62 April 195.53 10.36 63 May 193.11 8.73 72

June 191.52 7.55 62 July 200.37 6.90 57 Augl,lst 191.07 6.44 60 September 196.03 7.09 57 October 198.89 8.35 63 November 206.99 9.58 63

Denotes the compass direction the wind is blowing from (0° = North) .

t Max. Wind Speed is not the average maximum. It is the actual maximum that occurred during the 30 years of data.

PUR-1 SAR 2-11 Rev 0,6/08

December Annual 215.05 200.75 9.71 8.78 52 72 2.3.3 Severe Weather This region of the United States is subjected to tornado activity, primarily during the late spring and early summer months. Table 2.4 shows the tornados occurring in Tippecanoe County for the period from 1950 through February 2008. Thirty-eight tornados occurred over this period, which averages to less than one per year. The probability of damage due to a tornado is minimal.

PUR-1 SAR 2-12 Rev 0,6/08

Table 2-7: Tornados Reported in Tippecanoe County, Indiana between 01/01/1950 and

Location or COUNTY Date TIPPECANOE 02/28/2008 8 Time 6/13/1953 21 :00 Fujita Scale F1 Deaths 0

Injuries 0

Property Damage {$)

OK 2 TIPPECANOE 4/3/1956 17:00 F2 0 0 25K 3 TIPPECANOE 3/6/1961 6:05 F1 0 0 OK 4 TIPPECANOE 4/22/1963 20:15 F2 0 0 3K 5 TIPPECANOE , 6/10/1963 12:00 F1 0 0 3K 6 TIPPECANOE 4/11/1965 18:07 F4 0 10 OK 7 TIPPECANOE 9/14/1965 20:15 F2 0 0 250K 8 TIPPECANOE 6/24/1967 12:30 F2 0 0 3K 9 TIPPECANOE 5/15/1968 21:51 F2 0 0 3K 10 TIPPECANOE 3/19/1971 2:03 F2 0 0 25K 11 TIPPECANOE 5/29/1973 14:20 FO .0 0 OK 12 TIPPECANOE 6/12/1973 9:45 F1 0 0 25K.

13 TIPPECANOE 6/12/1973 10:29 F1 0 0 OK 14 TIPPECANOE 4/1/1974 16:32 F2 0 0 25K 15 TIPPECANOE 3/12/1976 14:25 F1 0 0 OK 16 TIPPECANOE 3/20/1976 15:20 F4 0 6 2.5M 17 TIPPECANOE 4/10/1978 12:15 F2 0 0 25K 18 TIPPECANOE 4/23/1978 18:00 F1 0 0 250K 19 TIPPECANOE 6/25/1978 15:55 FO 0 0 OK 20 TIPPECANOE 6/25/1978 17:00 F3 0 0 OK 21 TIPPECANOE 7/2/1978 12:15 F1 0 0 25K 22 TIPPECANOE 6/7/1980 16:15 F2 0 0 OK 23 TIPPECANOE 6/24/1981 19:47 F1 0 0 250K 24 TIPPECANOE 3/27/1991 18:10 FO 0 0 OK 25 Lafayette 4/26/1994 23:58 F4 3 70 5.0M 26 Lafayette 1/18/1996 14:30 FO 0 0 0 27 West Lafayette 7/4/1998 1 :30 F1 0 0 200K 28 Battle Ground 9/28/1999 18:54 F1 0 300K 29 Lafayette 6/11/2003 18:47 FO 0 0 0 30 West Lafayette 6/11/2003 19:00 FO 0 0 0 31 Lafayette 7/21/2003 4:00 FO 0 0 0 32 Romney 5/30/2004 19:26 FO 0 0 0 33 Dayton 5/30/2004 19:29 F2 0 0 1.0M 34 Dayton 7/26/2005 20:00 FO 0 0 10K 35 Romney 4/2/2006 20:25 F1 0 0 50K 36 Cairo 4/14/2006 18:58 FO 0 O* 0 37 Americus 4/14/2006 19:18 F1 0 0 30K 38 Odell 6/25/2006 14: 10 FO 0 0 3K TOTALS: 3 87 .10.004M

PUR-1 SAR 2-13 Rev 0,6/08

2.4 Hydrology Most of Tippecanoe County is covered by glacial drift. The drift ranges in thickness from a thin veneer to about 425 feet and was deposited upon a bedrock surface that was eroded by a preglacial drainage system. Much of the surface drift consists of glacial till. Water-laid cross bedded sand and gravel are associated with the till. The subsurface glacial deposits also include much till with interbedded sand and gravel.

Locally, clay deposits are as much as' 106 feet thick. Within the drift, five sheetlike water bearing units are differentiated in parts of the county. Ground water within these units occurs under artesian and water-table conditions. Locally these may occur within the same unit. 9 This area was repeatedly glaciated during the Pleistocene epoch.' Before glacial times, a giant drainage way, now know as the Teays River, flowed from the Appalachian Mountains across Ohio, and passed northwestward through the present site of Lafayette-West Lafayette. 1o Illinoian ice dammed the preglacial Teays River channel and ponded the relative small Glacial Lake Lafayette. An outlet channel, developed to drain this proglacial lake, was subsequently perpetuated as the present Wabash River drainage line southwestward from the Lafayette-West Lafayette area. 11 The elevation of the Purdue University campus is approximately 706 feet and the level of the Wabash River is approximately 510 feet. With this difference of over 100 feet the flow of both surface water and ground water is in a generally easterly and southerly direction toward the Wash River, which flows around two sides of the campus.

Any leakage of contaminated water from the PUR-I represents no potential hazard to either the West Lafayette or Purdue University water supply, since these flows are away from the well fields of both. The Wabash River represents a natural barrier between the reactor and the Lafayette well fields, so no potential hazard exists there.

2.5 Geology and Seismology 2.5.1 Regional Geology The county lies within the Tipton Till Plain of Indian and is a section of the Till Plains subprovince of the U. S. central Lowlands physiographic province. Most of the soils in this area are derived from the glacially deposited material. Extensive upland areas are covered with a thin mantle of loose deposits. A few areas are covered with soils of alluvial, colluvial or organic origin. 11 Glacial drift covers the bedrock to a depth ranging from a few feet to more than 300 feet. The underlying bedrock consisting of flint, shale, sandstone, and limestone of the Mississippian period, is exposed as rock terraces in the Wabsh Valley and on the upland in the western part of the county.10 Purdue University is located above an extensive glacial deposit of sand and gravel.

The land surfaces of Tippecanoe County are flat to rolling, except where the major

streams have cut deeply into the surface. The entire county lies within the drainage basin of the Wabash River and its tributaries. 9 PUR-1 SAR 2-14 The land slopes generally Rev 0,6/08

southwestward with the streams flowing westward. Two main tributaries, the Tippecanoe River and the Wild Car Creek enter the Wabash upstream from the campus. Minor tributaries include Little Pine Creek, Indian Creek, Burnetts Creek, Mott's Creek, Sugar Creek, Buck Creek, Wea Creek, and Flint Creek.

2.5.2 Seismology The three most significant seismic source zones which are closest to West Lafayette are:

1. The New Madrid area of southeastern Missouri;

2. The Wabash Valley Fault system of southwestern Indiana and southeastern Illinois;

3. The Anna, Ohio area.

Reasonable estimates of the maximum magnitude events which could occur in those areas give values of 7.4, 6.6 and 6.3 (body wave motion) for the seismic zones, respectively. Based on the distance from these zones (400, 200 and 200 km respectively) and attenuation curves, estimates for peak horizontal acceleration at West Lafayette for maximum magnitude events which could occur at these three seismic zones are approximately 5-15% G. 12 The figures that follow (Figure 2-7 through Figure 2-11) show the earthquake probabilities for the area.

The way, in which the reactor facility was constructed by modifying an existing building with no reinforcing bars tied into the original structure, the reactor pool can be y considered a free standing unit in the event of any seismic activity. The reactor pool consists of steel cylinders containing compacted magnetite sand between the cylinders and the 1/3 inch carbon steel tank. The inside of this tank was later lined with 1/16 inch stainless steel. With these barriers to contain the reactor pool water and considering the reactor pool as a free standing unit it is highly unlikely that any reactor water would be lost during any severe seismic activity. '

PUR-1 SAR 2-15 Rev 0,6/08

SA 15 2%50yr PE , 2008 4S'N

.1

\

1.20 0.93 0.72 0.S6 lS'N 0.43 0.33 0.26 S

'. 0.20 A 0.1S 0.12 9 lO'N ,.DO 0.09 0.07 km ~ .(

0.06 0.04 2S'N

~

l00'W

=~~'-_~~ T-~'~

9S'W

____ 9O'W

__- r__

85Yi

~~~~~~~::~0.

8O'W

' I, i5'W 0.03

~iO'02~~---

W Figure 2-7: Map of the 1-Hz spectral acceleration for 2% probability of exceedance in 50 years for the Central and Eastern United States in standard gravity (g). '3

PUR* ' SAR 2* 16 Rev 0, 6/0 8

SA 0.2-s 2%/50year PE, 2008 45 'N ~ \ ..

. -.....', 1 .

3.00 o 2.40 1.90 1.52 1.21 3S'N 0.99 o.n

0.50 0.61 0.39 0.31 0.25 0.20 0.16 0.12 S

A 9

0.10 0.08 km 'ill ~ ,. 0.06 25'N ~;1_.,-.~0~_-r-:500~O~-,--_~~c~J-;;.;i-:,:0~.070'~5;.;--~ W l00'W 95'W 8O'W 7S'W 90'W 8S"W Figure 2-8: Map of the 5-hertz spectral acceleration (SA) for 2% probability of exceedance in 50 yea rs in the Central and Eastern United States in standard gravity (g) -"

PUR-1 SAR 2* 17 Rev 0, 6/08

PGA with 2%/50 yr PE , 2008 40'N 1.20 0.93 0.72 0.56 R 35'N 0.43 0 0.33 C

086 k 0.20 0.15 P 0.12 G 30 'N .. 0.09 A 0.07 0.06 0.04 9

" '-l- 0.03 10'~

15' ~

8(r~

1!J!iI "'P\\ '~ .'J7 IIOAn;.eo,. .... .e ... _ .... _ ....

Figure 2-9: Map of peak ground acceleration (PGA) for 2% probability of exceedance in 50 yea rs in the Central and Eastern united States in standard gravity (g)."

PUR*1 SAR 2*18 Rev 0, 6/08

CEUS 1s SA 10%/50yr 2008 4S 'N ,

"\ , ."" 71

')

4O'N 0.35 0.30 0.25 3S'N 0.20

.... ...'> ( 0.16

~

, - 0.14 C 0.12 k

.i....) ( 0.10 0.07 S

lO'N

J.. ~~ '

(;--~, l-

"A

'r, 0.06 0.04 A

km 1'. *

, -I*

0.03 9 0.02 0 500 2S 'N

" 80'1'1

'.of 75'1'1 70'1'1 l00"W 9S'W 9O'W 85'1'1

[!!ZjJJ . 11 17 12 _ " ' . _ ...... _ _ _ _

Figure 2-10: Map of 1-hertz spectrat acceteration (SA) for 10% exceedance in 50 years in the Central and Eastern United States in standard gravity (g)"

PUR-1 SAR 2-19 Rev 0 , 6/08

CEUS PGA 10%/50 years , 2008 0.35 0.30 0.25 R 3S'N ,'- 0.20 0.16 0

C

k 0.14

'"" 0.12 0.10 P 0.07 G 30'N ~ 0.06 A 0.04 0.03 9 0.02 2S'N 70' Vl 7S' W

'DOW 9S'W 8S'W 8O'W 90' W Figure 2-11: Map of peak ground acceleration (PGA) for 10% probability in 50 yea rs in the Central and Eastern United States in standard gravity (g)-"

PUR* ' SAR 2-20 Rev 0, 6/08

2.6 1

References U.S. Census Bureau, http://www.census.gov 2 United States Geological Survey, http://www.usgs.gov (1973).

3 Tippecanoe County Area Planning Com~ission, (2008) 4 Purdue Data Digest, http://www.purdue.edu/datadigesU 5 Private Communications with Betty Stansbury, Airport Administrator, Purdue University, (2008).

6 National Climatic Data Center, NOAA-NESDIS, http://www.ncdc.noaa.gov/oa/climate/normals/usnormals.html. (2004) 7 Indiana State Climate Office, Purdue University, http://www.iclimate.org. (2008).

8 National Climatic Data Center, NOAA-NESDIS, http://www.ncdc.noaa.gov/oa/ncdc.html. (2008) 9 Resenshein, Joseph S., "Ground-Water Resources of Tippecanoe County, Indiana',

State of Indiana, Indiana Department of Conservation, Division of Water Resources Bulletin No.8, 1958.

10 11 Ulrich, H.P., Barnes, T.E., and Krantz, B.A., "Soil Survery, Tippecanoe County, Indiana Series 1940, No.22', 1959.

Maarouf, Abdelraham M., and Melhorn, Wilton N., "Technical Report No. 61', Purdue University Water Resources Research Center, June, 1975.

12 Braile, L.W. Professor, Purdue University, private communications.

13 Petersen, Mark D., et aI., "Documentation for the 2008 Update of the United States National Seismic Hazard Maps", Open File Report 2008-1128, U.S. Geological Survey, (2008) .

PUR-1 SAR 2-21 Rev 0, 6/08

3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.1 Design Criteria The Duncan Annex of the Electrical Engineering Building is of brick, concrete block and reinforced concrete construction which was originally designed as a large high voltage laboratory. It was subsequently subdivided into offices, classrooms and laboratories.

The reactor i ted in the southwest corner on the round floor in a h ba area of the building.

The outside air supply and exhaust are both passed through HEPA filters. The reactor room is maintained at negative air pressure (minimum 0.05 inches of water). All doors to the reactor room have foam rubber seals.

The only floor drain to the sewers is sealed except for a vent opening. This vent is raised about two feel above the floor and has a filtered inverted opening. Condensate from the air conditioner is released to this drain through an opening 12.0 feet above the floor.

3.2 Meteorological Damage According to the Unified Building Code, 1985 edition, the Purdue University lies in the maximum wind zone of 80 miles per hour, which translates to a wind load of 17 pounds per square foot. Buildings at Purdue University are designed to withstand this wind load.

3.3 Water Damage The Electrical Engineering Building, and all of the Purdue Campus, lie well above any flood plain. In ~ of operation, there has been no standing water in the reactor room. The top _ of the reactor pool stand above floor level, and the sides of the pool are made of 15 inches of reinforced concrete In the unlikely event that this wall were to break, it would only result in 3 inches of standing water on the reactor room floor. None of the reactor instrumentation would be harmed by such an unlikely occurrence.

3.4 Seismic Damage The way, in which the reactor facility was constructed by modifying an existing building

with no reinforcing bars tied into the original structure, the reactor pool can be considered a free standing unit in the event of any seismic activity. The reactor pool PUR-1 SAR 3-1 Rev 0,6/08

consists of steel cylinders containing compacted magnetite sand between the cylinders and the 1/3 inch carbon steel tank. The inside of this tank was later lined with 1/16 inch stainless steel. With these barriers to contain the reactor pool water and considering the reactor pool as a free standing unit it is highly unlikely that any reactor water would be lost during any severe seismic activity.

3.5 Systems and Components, Should the need arise for an emergency shutdown of the reactor, the 2 shim-safety control rods can be scrammed automatically by instrumentation, or manually via two scram buttons: one located within easy reach of the operator on the control panel, the*

other outside the main personnel access door to the reactor room.

During emergency conditions, the ventilation systems can be shut off by a console switch, and the sealed room will prevent the rapid spread of contamination. During an emergency the air conditioner and the valve on the drain from the condensate holdup tank. are shut off with the same switch that shuts off the ventilation system. The condensate would then be held until it is tested by Radiological Control before it is released to the sewer. If contamination is found, it would be disposed of as radioactive liquid waste.

PUR-1 SAR 3-2 Rev 0,6/08

4 REACTOR DESCRIPTION 4.1 Summary Description The 10 kW PUR*1 reactor described herein was designed and constructed by Lockheed Nuclear Products of Lockheed Aircraft Corp. of Marietta . Georgia. It was designed for continuous steady-state operation at 10 kW , but previously licensed for operation at 1 kW .

PUR-1 is a heterogeneous, pool-type non-power rea ctor. T he core is cooled by natural convection of light water, moderated by light water, and reflected by water and graphite.

The reactor is located near the bottom of a water*filled tank surrounded and supported by a concrete shielding structure as shown in Figure 4- 1. An aluminum grid plate structure supports the reactor and control mechanisms at the bottom of the pool, with additional support of the control mechanisms provided by a fixture at the top of the pool.

Three detectors used for monitoring reactor cond itions are located in fixed positions next to the reactor core . And the startup detector istlocated in a tube affixed to a fuel element in the core , which allows the detector to be removed from the neutron flux when the reactor is at power .

The reactor core is composed of sixteen fuel elements positioned in holes in the I. aluminum grid plate. The grid plate contains a rectangular matrix of holes to allow the cha nging of fuel eleme nt locations and the insertion of graphite reflector elements to PUR-1 SAR 4*1 Rev 0 , 6/08

Reactivity of the reactor core is changed by the operator moving the control rods that are suspended from fail-safe electromagnets. The ionization chambers used for sensing neutron and gamma-ray fluxes are located near the core. The control console, from which the' operator can observe the reactor pool and top structures, is located adjacent to the reactor, and consists of typical read-out and control instrumentation.

Heat removal is achieved by natural convection, with a general flow up through the nozzle at the bottom of the fuel assemblies. The reactor is located within a 6400 gallon cylindrical water tank, 17 feet deep and 8 feet in diameter.

Safety and other operational characteristics of this reactor system are similar to other reactors using the MTR type fuel assembly. The power and flux level of the PUR-I are of adequate range and the experimental facilities are sufficiently flexible to encompass a wide variety of training and research experiments. The reactor is designed so that a minimum of restrictions are imposed on the experimenter, and the console can be readily operated by one person .

Safety is an overriding requirement in a training reactor. Self-limiting features of the PUR-1 core, coupled with carefully designed control instrumentation, assure the highest degree of safety. The safety record of this facility, demonstrated over the past 46 years give proof that the design, construction, and installation of the reactor system, coupled with the administrative control over operation, maintenance, afld utilization, are more than adequate to provide protection for the public he,alth and safely.

Table 4-1 and Table 4-2 summarize the key design parameters for the PUR-1 .

PUR-1 SAR 4-2 Rev 0,6/08

Table 4-1: Summary of Design Parameters for PUR-1 DESIGN DATA Value Fuel Meat Dimensions Width (mm)

Thickness (mm)

Height (mm)

Fuel Plate Dimensions Width (mm)

Thickness (mm)

Height (mm)

Cladding Composition 6061 AI Cladding Thickness (mm) 0.381 Dummy Plate Composition 6061 AI Dummy Plate Dimensions Same as Fuel Standard Fuel Assemblies Number of standard assemblies 13 Number of plates per standard assembly 14 Control Fuel Assemblies Number of control assemblies 3 Number of plates per control assembly 8 Total plates in core (fuel and dummy).

Fuel plates in core (current)

Dummy plates in core (current, expected) 16 Plate spacing in standard assemblies (mm) 3.66 Plate spacing in control assemblies (mm) 4.60

PUR-1 SAR 4-3 Rev 0,6/08

Table 4-2: Summary of key reactor parameters for PUR-1 .

REACTOR PARAMETERS Fresh core excess reactivity (%Llklk)

Shutdown margin (%Llk/k) 1 "Calculated OAi

-1.80 1

Control rod worth (%Llk/k)

Shim-safety 1 3.93 Shim-safety 2 2.22 Regulating Rod 0.27 MaXlmum

. . .Insertion reactivity

. . rate (%MJ k*s Shim-safety 1 1.7SE-02 Shim-safety 2 8.75E-03 Regulating Rod 4.66E-03 Avg. coolant void coefficient ( %~k )2 -1.93E-1+/-7%

k *%void Coolant temperature coefficient (%~k

__ ) 3 -9.05E-3+/-9%

k*oC Fuel temperature coefficient (%~k)

__ 4 -8.05E-4+/-10%

k*oC Effective delayed neutron fraction (%) 0.784 Neutron lifetime (lJs) 81.3 4.2 Reactor Core The PUR-1 core layout is a sixteen assembly (4x4 array), heterogeneous, light-water moderated, graphite reflected, water cooled reactor fueled with LEU plate-type fuel.

The core layout is shown in Figure 4-2. Each of the thirteen standard fuel assemblies in the core can hold up to 14 fuel plates, or a mixture of fuel and dummy plates. The three control elements each hold up to eight fueled plates.

Twenty graphite reflector assemblies surround the core, 6 of which contain a cylindrical aluminum tube normally filled with graphite. These 6 elements comprise the irradiation facility. The graphite can be removed from these tubes and replaced with experiment capsules which can then be irradiated with normal reactor operation.

1 From the 2008 Annual Report.

2 Calculated for the representative range of 0-0.6% void .

3 Calculated for the representative range of 20-30°C.

4 Calculated for the representative range of 20-12rC.

PUR-1 SAR 4-4 Rev 0,6/08

Figure 4-2: PUR-1 Grid Plate The reactor is controlled by three control rods located in the core region of the reactor.

There are two shim-safety rods made of solid borated 304 stainless steel , utilizing a magnet clutch between the blades and the lead screw operated drive mechanisms, and

a regulating rod , wh ich is a screw operated direct drive and made of hollow stainless steel. Each control blade is protected by an aluminum guide plate on each side with in the control fuel assemblies.

Each of the standa rd assemblies and the control assemblies are co ntai ned in a 6061 aluminum container. The standard graphite assembli es and the irradiation facility graphite assembli es are contai ned in simi lar 6061 aluminum co ntainers. Th e startup neutron source is located outside the core in a simil ar 6061 aluminum container.

4.2.1 Reactor Fuel Th e reactor is fueled by standard MTR LEU plates installed in 2007 during the conversion of PUR-1 from HEU to LEU fuel. The LEU fuel is silicide dispersion standard MTR plates manufactu red by BWX-Technologies (BWXT) of Virgi nia. Dumm y aluminum plates identical in size to the fuel plates were also manufactured by BWXT, and these are used in place of fuel in the reactor in some locations. The assembly ca ns that contai n th e plates were manufactu red by General Atomics, of Ca lifornia.

The fuel and dummy plates are inserted into both the sta ndard and control assembly cans individually, each one contained with in its own slot. These slots control the plate spacing to close tolerances over the length of the plates , which is a significant improvement over the old design where the plates were pi nned together at the four corners. The nominal fuel plate information is detailed in Table 4-3 .

PUR-1 SAR 4-5 Rev 0, 6/08

Table 4-3: Characteristics of the PUR-1 Fuel Plates Fuel Type Fuel "Meat" Design Detail U3Si2- AI U3Si2- AI 1

Fuel Plate Dimensions Width (mm)

Thickness (mm)

Height (mm)

Fuel Meat Dimensions Width (mm)

Thickness (mm)

Height (mm)

Cladding Type 6061 AI Cladding:

Along width (mm) 3.63 (min)

Along thickness (mm) 0.381 Figure 4-3 and Figure 4-4 show the design of the wall spacers that control the plate locations and channel thicknesses for the standard and control elements .

PUR-1 SAR 4-6 Rev 0; 6/08

Figure 4-3: Standard assembly can detail; showing wall spacers .

PUR-1 SAR 4-7 Rev 0,6/08

Figure 4-4: Control assembly can detail, showing wall spacers.

Figure 4-5 and Figure 4-6 show the cross-sectional view of the standard and control assemblies for the PUR-1 reactor. Figure 4-5 shows the handle assembly at the top of the assembly can, which can be removed for insertion and removal of the fuel plates.

The Control assemblies have no such handle .

PUR-1 SAR 4-8 Rev 0, 6/08

Figure 4-5: Standard fuel assembly.

The nominal plate-la-plate spacing, and the nominal plate-ta-wall spacing is shown in Table 4-4.

Table 4-4: Channel Types and Thickness in PUR-1 Assemblies Plate-to-plate (mils) Plate-lo-wall (mils)

Standard Control Standard Control Dimension 144+/-1S 181+/-15 127+/-8 127+/-8 Fuel and dummy plates are .!!.Q;!S!!.~ nUI11bers, and the dummy plates are differentiated by , as shown in Figure 4-7 and Figure 4-8.

PUR-1 SAR 4-9 Rev 0, 6/08

~I I

- - -\---'d-lf-++t-I I

! : I I

___________ ! ILI

_ _ _ _ /1

~

Figure 4-7: Plates are identified by unique serial numbers as shown here.

4.2.2 Control Rods Table 4-5: Summary of control rod characteristics as listed in the PUR-1 Operations Manual.

CONTROL RODS Number of Regulating Rods 1 - 304 stainless steel, hollow Number of Shim Safety Rods 2 - Boron-stainless steel, solid Operating Rates Regulating Rod 17.7 in/min Shim Safety Rod 4.4 in/min Scram Less than 1 second (from signal to complete insertion).

Size Regulating rod (inches) %x2~

Shim safety rods (inches) % x 2 % x 25 %

Maximum rate of reactivity change 11k Regulating rod 1.1x10 k*s Shim safety rods 11k 9.9x10 k*s Average rate of reactivity charge 11k Regulating rod 5.8x10 k*s Shim safety rods 11k 5.4x10 k*s 4.2.3 Neutron Moderator and Reflector

PUR-1 is moderated by the light water of the reactor pool. A graphite reflector surrounds the four sides of the reactor, and light water reflects the top and bottom. The PUR-1 SAR 4-10 Rev 0,6/08

graphite in the reflector assemblies is nuclear grade, manufactured by Union Carbide, and contains less than 1 part per million boron.

4.2.4 Neutron Startup Source A 5 Curie Plutonium-Beryllium neutron startup source is located next to the core, and is removable when the reactor reaches criticality. A drive mechanism controlled from the operator's console raises and lowers the source as needed.

4.2.5 Core Support Structure The reactor is supported on an aluminum grid plate typical of all Lockheed MTR reactors. This grid plate controls the placement of fuel within the core, and can be used to locate experiments that would be placed outside the core. A drawing of the grid plate is located in the Appendix 1.

The grid plate is manufactured with 6061 aluminum, and has not shown any degradation over its greater than 46 years of service. It is expected that the grid plate will continue to function as designed.

4.3 Reactor Pool

cylindrical tank 17 4 inches deep and 8 The pool is contained in a in diameter. The core is located to one side to give additional experimental space. The pool has a welded stainless steel liner.

The supports for the drive mechanisms for the control rods, the fission chamber and the source, and the neutron detectors are fastened to the support plate at the top of the tank. A traversing mechanism was mounted on the top of the reactor pool wall after the reactor was built. A light weight, portable aluminum bridge can be placed across the pool for maintenance and fuel handling operations.

The average pool temperature in recent PUR-1 operating history is 26°C. The figure below shows measured pool temperatures from 1993 to 2006 .

PUR-1 SAR 4-11 Rev 0,6/08

35.0 PUR-1 Reactor Pool Temperature Measurements 30.0 u

Cl 25.0 CI)

~

Q. , ,

E

~

20.0 15.0 +----~---~---...:..,-.---~---~---~---~

Dec-93 Dec-95 Dec-9? Dec-99 Dec-D1 Dec-D3 Dec-05 Dec-O?

Date

,/

Figure 4-9: PUR-1 Reactor pool temperature me.asurements from 1994-2007.

4.4 The Biological Shield The water level is normally 4 inches below the top of the tank.

Shielding over the core is provided by 13 feet of light water. This reduces the radiation level to a calculated less than 1 mrem/hr when the core is 'operating at 1 kW. The concrete biological shield is designed for a maximum radiation level at any point along its outer lateral surface of 0.1 mr/hr at 1 kW.

4.5 Nuclear Design Neutronics analysis was performed with the Monte Carlo N-Particl~ Code (MCNP5),

published by Los Alamos National Laboratory. Thermal-Hydraulic analysis was performed with NATCON, a natural convection analysis code written by Argonne National Laboratory (ANL). Transient analysis was performed using PARET, also written by ANL.

4.5.1 MCNP Model The MCNP model of the fuel plates along various dimensions is shown in Table 4-6.

PUR-1 SAR 4-12 Rev 0,6/08

The materials used in the model are standard for the LEU MTR plates, with the addition of 20 parts per million of a boron-equivalent to the 6061 cladding material to account for impurities in the alloy.

Figure 4-11: plate in Y-Z plane.

(magnified)

Figure plate in X-Z plane Figure 4-12: plate in X-Y plane, cutaway view.

PUR-1 SAR 4-13 Rev 0, 6108

Figure 4-13: Model representation of standard assembly plate spacing detail showing wall spacers.

The representations of the complete standard and shim safety control fuel assemblies

for the core are shown in Figure 4-14 and Figure 4-15 respectively . There are one or two dummy plates in the standard assemblies, and no dummy plates in the control assemblies, which are the plates without the center fuel material in the figures . The assembly cans are identical in size and composition .

Figure model.

As stated before , the control assemblies contain no dummy plates, but do contain a guard plate made of 6061 aluminum between the fuel elements and the control rods to prevent any mechanical damage to the fuel from insertion of the control rods. The shim safety control rods themselves, made of borated stainless steel, are oblong shaped plates inserted down the center of the assemblies. The regulating rod is similar to the shim-safety control assemble shown in the figure above , but reflects the fact that the rod is hollow, and filled with water.

PUR- 1 SAR 4- 14 Rev 0 , 6/08

Figure The entire core model is shown in Figure This im age shows the present configuration of th e core , the ori"ntotio of the irradiation facilities , and th e startup neutron source Figure 4-16: Representation of LEU core load 4.5.2 Normal Operating Conditions 4.5.2.1. PUR-1 Core Power Distribution The PUR-1 LEU fu el plates are fabricated with U, Si, diiwersion has a nominal loading of _ U-235. A fresh core with "dummy" aluminum plates was evaluated in the models p".. e-nt"d h,,,ein PUR- 1 SAR 4-15 Rev O. 6/08

shows the core layout modeled in MCNP, and drawings of control and 14-plate fuel assemblies. The numbers in parenthesis in the core layout drawing indicate either the number of fuel plates in the 14-plate fuel assemblies (the remainder are dummy plates of aluminum) or the label of the control assembly (shim or regulating rods).

Table 4-7 compares the heating by assembly and in non-fueled components for the LEU core with the rods banked at 53.5 cm. The banked rod critical configuration was found to have the highest peak power density in the analysis of the HEU core.

Elements 3-4, 4-3, and 4-4 are noted as having the highest average plate powers, and are therefore of interest for thermal-hydraulics analysis. Figure Q27 -1 provides a schematic of the core layout and the plate orientations.

PUR-1 SAR 4-16 Rev 0,6/08

Table 4-8 compares the power in individual plates in elements 4-4, 3-4, and 4-3 for the reactor with the banked rods critical configuration. The tallies were summed over the fuel meat in each fuel plate, all clad, coolant, and the bundle can. The plates are numbered from left-to-right in these elements (see the element drawings in Figure 4-17). It can be seen that plate 1348 in assembly or bundle 4-4 has the highest power (8.07 W at 1 kW). This plate is adjacent to the large water hole that the SS1 rod *falls into, and nearer the center of the reactor than plate 1355 on the other side of the water hole. Plates 1228 (bundle 3-4) and 1315 (bundle 4-3) face the center-line of the core, and have roughly equal power (6.51 and 6.41 W at 1 kW).

Figure 4-18 and Figure 4-19 compare the local-to-average axial power density profiles for fuel plates 1348, 1228, and 1315 for the banked rods critical configuration. Plate 1348 has the highest peak power density of all the plates in the LEU core and was evaluated in the thermal-hydraulics analyses. The slight "pinching" of the axial power profile due to the insertion of the SS1 shim rod is evident.

The local-to-average axial power density profiles in plates 1228 and 1315 are nearly identical due to their symmetric positioning across the core center-line. For these two plates, the highest peak/average density (1.712) occurs in plate 1228. The peak power density in plate 1228 of standard element 3-4 is about 22% lower than that of plate 1348. Furthermore, the coolant channel thickness in the standard fuel assembly is narrower than that in the control assembly (144 vs. 181 mils). This reduction in channel thickness brings the channel thickness closer to the optimum value (~ 100 mil), which will result in a higher ONB power with natural convection cooling. Therefore, plate 1348 is the most limiting of all fuel plates.

A higher water-to-fuel ratio at the edges of the fuel plates induces a power density profile along the width of the fuel plates. This is shown in Figure 4-19 for plates 1348, 1228, and 1315 with the rods in the banked critical position. The radial segments are numbered from bottom-to-top in the MCNP model (see the plate orientations indicated in Figure 4-17). It is expected that the power density will be higher at the edge of the plate closest to the core center-line.

Lastly, Table 4-9, Table 4-10, and Table 4-11 provide the axial and radial power profiles for plates 1348, 1228, and 1315, respectively. These data were used for thermal-hydraulic analysis of the LEU-fueled PUR-1.

PUR-1 SAR 4-17 Rev 0,6/08

Purdue LEU Core Layout . plates)

Element 4-4

155 1

~

I~

1£ I

Figure 4-17: PUR-1 Core Layout and Bundle Drawin gs.

PUR* 1 SAR 4-18 Rev 0 , 6/08

Table 4-7: Bundle Powers Predicted by f7 and f6 Tallies in MCNP.

2-2 (RR)

Power (W)

. 1000W 38.19 Average Plate Power (W) 4.77 Power (W) 10 kW 381.9 Average Plate Power (W) 47.7 2-3 61.85 4.76 618.5 47.6 2-4 (SS2) . 45.65 5.71 456.5 57.1 2-5 45.93 3.83 459.3 38.3 3-2 62.56 4.81 625.6 48.1 3-3 78.09 6.01 780.9 60.1 3-4 78.42 6.03 784.2 60.3 3-5 61.20 4.71 612 47.1 4-2 63.52 4.89 635.2 48.9 4-3 80.78 6.21 807.8 62.1 4-4 (SS1) 60.09 7.51 600,9 75.1 4-5 62.29 4.79 622.9 47.9 5-2 52.32 4.03 523.2 40.3 5-3 63.97 4.92 639.7 49.2 5-4 63.49 4.88 634.9 48.8 5-5 47.64 3.97 476.4 39.7

Inter-assembly water Graphite reflector Grid plate Water reflector (pool)

SS1 3.78 9.52 2.27 18.00 0.23 37.8 95.2 22.7 180 2.3 SS2 0.17 1.7 RR 0.03 0.3 Total 1000.0 10000

PUR-1 SAR 4-19 Rev 0,6/08

Table 4-8: Plate Power (W) Computed from Heating Tallies in Bundles 4-4,3-3 and 3-4 in PUR-1 Core with. Fuel Plates.

Bundle 4-4 Plate 1345 Meat Plate 1346 Meat Power (W at 1 kW) 7.00 7.09 Power (W at 10 kW) 70 70.9 Plate 1347 Meat 7.40 74 Plate 1348 Meat 8.07 80.7 Plate 1355 Meat 7.74 77.4 Plate 1356 Meat 6.98 69.8 Plate 1357 Meat 6.59 65.9 Plate 1358 Meat 6.43 64.3 Clad 0.36 3.6 Water 2.19 21.9 Can 0.24 2.4 Totar " .. . *60.09' . . '" ,"r " 600.9 .:,,:.;~ "l

" I' ",

Bundle 3-4 Plate 1215 Meat 5.80 58 Plate 1216 Meat 5.56 55.6 Plate 1217 Meat 5.49 54.9 Plate 1218 Meat 5.50 55 Plate 1219 Meat 5.61 56.1 Plate 1220 Meat 5.84 58.4 Plate 1222 Meat 5.91 59.1 Plate 1223 Meat 5.76 57.6 Plate 1224 Meat 5.73 57.3 Plate 1225 Meat 5.76 57.6 Plate 1226 Meat 5.92 59.2 Plate 1227 Meat 6.12 ' 61.2 Plate 1228 Meat 6.51 65.1 Clad 0.48 4.8 Water 2.19 21.9 Can 0.24 2.4

- I Total 78.42 784.2 Bundle 4-3 Plate 1315 Meat 6.41 64.1 Plate 1316 Meat 6.16 61.6 Plate 1317 Meat 6.07 60.7 Plate 1318 Meat 6.04 60.4 Plate 1319 Meat 6.09 60.9 Plate 1320 Meat 6.28 62.8 Plate 1322 Meat 6.21 62.1 Plate 1323 Meat 5.92 59.2 Plate 1324 Meat 5.76 57.6 Plate 1325 Meat 5.68 56.8 Plate 1326 Meat 5.65 56.5 Plate 1327 Meat 5.68 56.8 Plate 1328 Meat 5.84 58.4 Clad 0.49 4.9 Water 2.19 21.9 Can 0.24 2.4 8Q7.1 .. i',,/~'

,i, '

Total '." ,

~ 80.71 . ,

,. I PUR-1 SAR 4-20 Rev 0,6/08

"r-------------------------------------------------------,

f

I "

!* h.L~""O:"'----------_;__ P,...

~

I0 l JAa IEI..... n' 4-4

_ _ PI... 1221 (EI . ....

1 n' ).<II l---------------,;:..------.-j

.9 I __ PI... 131$(EI. ....n" *ll

- L-____________________________________________________--"

Figure 4-18: Axial Power Profiles in Plates 1348, 1228 and 1315 for Banked Rod Critical Configuration in PUR-1

_ 1',.,. IJAa rE_ , 4-4) 1

'~ __ 1',.,. 1m (E........').<I)

......n' ..JI f"

_ PI... ,ll~rE

..I ". ~

~

I L ~

~

Figure 4-19 : Radial Power Profiles in Plates 1345, 1228 and 1315 for Banked Rod Critical Configuration in PUR-1

4-2 1 Rev 0, 6/08

Table 4-9: Axial and Radial Heating Profile for Plate 1348 of Bundle 4-4 for' Banked Axial Segment 1

Critical Configuration.

z-Iow (cm) 1.88595 z-high (cm) 5.88645 Power (W at 1 kW) 3.04 a

0.99%

2 5.88645 9.88695 4 0.86%

3 9.88695 13.88745 5.17 0.76%

4 13.88745 17.88795 5.95 0.70%

5 . 17.88795 21.88845 6.76 0.66%

6 21.88845 25.88895 7.2 0.64%

7 25.88895 29.88945 7.47 0.63%

8 29.88945 33.88995 7.71 0.62%

'9 33.88995 37.89045 7.62 0.63%

10 37.89045 41.89095 6.97 0.65%

11 41.89095 45.89145 6.41 0.68%

12 45.89145 49.89195 *5.63 0.74%

13 49.89195 53.89245 3.85 0.86%

14 53.89245 57.89295 1.72 1.25%

15 57.89295 61.89345 1.2 1.48%

Total 80.7 0.21%

y-Iow y-high Local!Average Radial Segment (cm) (cm) Power Density a 1 24.0838 24.6253 1.672 0.48%

2 24.6253 25.1668 1.604 0.48%

3 25.1668 25.7082 1.572 0.48%

4 25.7082 26.2497 1.554 0.49%

5 26.2497 26.7912 1.523 0.49%

6 26.7912 27.3327 1.519 0.49%

7 27.3327 27.8742 1.506 0.49%

8 27.8742 28.4157 1.496 0.50%

9 28.4157 28.9571 1.493 0.50%

10 28.9571 29.4986 1.496 0.50%

11 29.4986 30.0401 1.518 0.50%

1Positions correspond to MCNP model of PUR-1 .

PUR~1 SAR 4-22 Rev 0,6/08

Table 4-10: Axial and Radial Heating Profile for Plate 1228 of Bundle 3-4 for Banked Axial Segment 1

z-Iow (cm) 1.88595 Critical Configuration.

z-high (cm) 5.88645 Power (W at 1 kW) 2.36 a

1.07%

2 5.88645 9.88695 3.11 0.94%

3 9.88695 13.88745 3.95 0.83%

4 13.88745 17.88795 4.65 0.77%

5 17.88795 21.88845 5.23 0.73%

6 21.88845 25.88895 5.57 0.70%

7 25.88895 29.88945 5.84 0.70%

8 29.88945 33.88995 5.98 0.69%

9 33.88995 37.89045 5.76 0.69%

10 37.89045 41.89095 5.4 0.71%

11 41.89095 45.89145 4.98 0.74%

12 45.89145 49.89195 4.36 0.80%

13 49.89195 53.89245 3.42 0.89%

14 53.89245 57.89295 2.55 1.05%

15 57.89295 61.89345 1.95 1.17%

Total 65.11 0.22%

y-Iow y-high Local/Average Power Radial Segment (cm) (cm) Density a 1 24.0838 24.6253 1.424 0.52%

2 24.6253 25.1668 1.320 0.52%

3 25.1668 25.7082 1.262 0.53%

4 25.7082 26.2497 1.229 0.54%

5 26.2497 26.7912 1.194 0.54%

6 26.7912 27.3327 1.193 0.55%

7 27.3327 27.8742 1.178 0.54%

8 27.8742 28.4157 1.178 0.55%

9 28.4157 28.9571 1.189 0.54%

10 28.9571 29.4986 1.215 0.54%

11 29.4986 30.0401 1.297 0.54%

1Positions correspond to MCNP model of PUR-1 .

PUR-1 SAR 4-23 Rev 0, 6/08

Table 4-11: Axial and Radial Heating Profile for LEU Plate 1315 of Bundle 4-3 for Axial Segment 1

Banked Critical Configuration.

z-Iow (cm) 1.88595 z-high (cm) 5.88645 Power (W at 1 kW) 2.33 a

1.07%

2 5.88645 9.88695 3.03 0.96%

3 9.88695 13.88745 3.89 0.85%

4 13.88745 17.88795 4.63 0.79%

5 17.88795 21.88845 5.11 0.75%

6 21.88845 25.88895 5.43 0.72%

7 25.88895 29.88945 5.67 0.70%

8 29.88945 33.88995 5.74 0.70%

9 33.88995 37.89045 5.58 0.70%

10 37.89045 41.89095 5.23 0.73%

11 41.89095 45.89145 4.9 0.76%

12 45.89145 49.89195 4.36 0.81%

13 49.89195 53.89245 3.42 0.90%

14 53.89245 57.89295 2.65 1.02%

... 15 57.89295 61.89345 2.1 1.13%

Total 64.07 0.23%

y_low 1 y-high ' Local!Average Radial Segment (cm) (cm) Power Density a 1 16.3894 16.9309 1.279 0.54%

2 16.9309 17.4724 1.201 0.55%

3 17.4724 18.0138 1.172 0.56%

4" 18.0138 18.5553 1.147 0.56%

5 18.5553 19.0968 1.144 0.57%

6 19.0968 19.6383 1.151 0.56%

7 19.6383 20.1798 1.174 0.55%

8 20.1798 20.7213 1.199 0.55%

9 20.7213 21.2627 1.246 0.54%

10 21.2627 21.8042 1.313 0.52%

11 21.8042 22.3457 1.435 0.51%

1Positions correspond to MCNP model of PUR-1.

4.5.3 Reactor Core Physics Parameters For the model, the material composition for the plates was U3Si2-AI, and 6061 aluminum cladding. An addition of 20 parts per million of a boron-equivalent was added to the 6061 cladding material to account for impurities in the alloy. The MCNP calculated maximum thermal neutron fluxes in the fuel region are 2.01 E1 0 n/(cm2*s) peak in the fuel region, and 1.38E1 0 n/(cm2*s) average in the fuel region.

Eigenvalue calculations were performed with* MCNP5, typically using from 25 to 50 million neutron histories. These calculations yielded a reactor keff with a 1-0' uncertainty of +/-17 pcm (0. 017% ~k/k) for 25 million histories; the uncertainty was reduced to +/-12 pcm (0. 012% ~k/k) when 50 million histories were employed. The 1-0' uncertainty of PUR-1 SAR 4-24 Rev 0,6/08

the eigenvalue calculations reduces by the square root of the number of histories. The uncertainty of the reactivity feedback coefficients can be calculated as the square root of the sum. of the squares of the eigenvalue calculations. Based on the eigenvalue calculations with 25 to 50 million histories, the uncertainties of the water temperature and water void coefficients were found to be on the order of 15% to 30% when calculated over the expected 10 to 20°C perturbations of water temperature.

It was decided that the calculational uncertainty should be reduced by extending the number of histories in the MCNP eigenvalue calculations. Eigenvalue calculations for the nominal core state and certain other cases which exhibit only small perturbations to the core keff were performed with 200 to 300 million neutron histories. This reduced the 1-0' uncertainty of the eigenvalue calculations to around 5 pcm (0.005% ~k/k).

4.5.3.1. Control Rod Worths and Excess Reactivity Excess reactivity of the LEU core in PUR-1 was determined to be 0.00468 (0.47%) bok/k in a 190 fuel plate core (16 dummies), including a reactivity bias of 0.32% bok/k. This value is within the Technical Specification limit of 0.6% for excess reactivity.

Control rod worths were calculated for the core and compared with measured data. The calculated and measured control rod worth values are shown in Table 4-12. The calculated calibration curves were done utilizing MCNP5 for the core, and Figure 4-20, Figure 4-21, and Figure 4-22 show the calculated control rod worth curves for each of the control rods, along with the measured values for each of the rods.

Table 4-12: Comparison of HEU and LEU calculated control rod worths.

LEU Calculated (l1k/k) LEU Measured (l1k1k)

Shim Safety 1 0.0377+/-0.0003 0.0393 Shim Safety 2 0.0189+/-0.0003 0.0222 Regulating Rod 0.0023+/-0.0003 0.0027

PUR-1 SAR 4-25 Rev 0,6/08

, 00%

0.50%

55-1 Rod Calibration Curve

.. 0 000%

<50%

o

1.()O%

o

~- 1.5O%

2.00%

o

2.50%

o

300% o o

0 0

~.50%

4 .00%

0 20

" Rod POSition (em) 50 60 o 55-1 Calculated o 55-1 Measured Poly. (SS*' Calculated)

Figure 4-20: Calibration curve for 85-1 rod with calcu lated and measured values.

>.00%

55*2 Rod Calibration Curve 0.50",1, o c

c ~ c 0.00% C o o

o

,* -0.50%

o

1.00%

-1 .50% 0

-2.00%

0 20 50 60 70 Rod Position (em) 55-2 Calculated o 5$-2 Measured Poly. (5$*2 Calculated)

(I Figure 4-21: Calibration curve for S8-2 rod with calcu lated and measured values.

PUR-1 SAR 4*26 Rev 0, 6/08

0.35%

0.30%

Reg Rod Calibration Curve Q

0.25% Q a

020%

0 15% ,

a O.l~;'

, a 0.05% a 0.00%

0

" 30 Rod POSition (em) 50

'" 70

<> RR Calc ulated C RR Measured Po ly. (RR Calculated )

Figure 4-22: Calibration curve for RR with calcu lated and measured values.

Using the rod worth cu rves in preceding figures, the maximum reactivity insertion rates were determined by finding the maximum slope, or rate of cha nge, of the curve. The comparison of calculated and measured maximum reactivity insertion rates for PUR-1 are shown in Table 4-13.

Table 4-13: Comparison of calcu lated and measured maximum reactivity insertion rates.

Maximum Reactivity Insertion Rates for Control Rods

( dk k .s 1

Calculated Measured Shim-safety 1 1.7SE-04 2.31E-4 Shim -safety 2 8.75E-05 1.42E-4 Regulating Rod 4.66E-05 5.39E-5 4.5.3.2. Shutdown Margin The shutdown margin was calculated using MCNP5 and has been measured. These data are shown in

PUR-1 SAR 4-27 Rev 0, 6/08

Table 4-14. These values meet the technical specification (TS 3.1.a) .

PUR-1 SAR 4-28 Rev 0, 6/08

Table 4-14: Comparison of calculated and measured shutdown margins .

55-2 Worth (Ak/k) kexcess (Ak/k)

Shutdown Margin (Aklk)

Calculated

-1.89%

0.351%

-1.58%

Measured

-2.22%

0.42%

-1.80%

4.5.3.3. Other Core Physics Parameters Reactivity coefficients and reactor kinetics parameters were. calculated for the PUR-1 model core. These values are used to estimate the core reactivity response to changes in properties* of the fuel temperature, coolant/moderator temperature, or coolant/moderator density. They are therefore essential for analyses of reactivity-induced transients. These calculations were performed with the MCNP5 code using the same core model as for the core design and power distribution analyses.

The reactor kinetics parameters evaluated for PUR-1 were the effective delayed neutron fraction, ~eff, and the prompt neutron lifetime, e. The effective delayed neutron fraction is calculated using two eigenvalue calculations from MCNP5. Normal calculations of keff include both prompt and delayed neutrons. A additional calculation of keff is performed with delayed neutrons turned off in MCNP, yielding a keff that depends only on prompt neutrons, which is denoted by k:;/mPI. The effective delayed neutron fraction is then defined as kProlllPI p~tf .=1-~ k.*

~tf The delayed neutron fraction was calculated using the formulation:

Prompl k (t k - kPromPI f3en* - _e_.. - -~.( t ' f~I.. *

.. kett ken.

The eigenvalue calculations were performed using MCNP5. The value for 13eff is calculated as shown below:

keff = 1.00379 +/- 0.00010 k;omPI = 0.99589 +/- 0.00015 13eff = 1 - 1.00379/0.99589 = 7.870x10- 3 +/- 1. 79x1 0-4 (2.3%)

The bias of the PUR-1 MCNP model is ,1Pbias= 0.32% ,1k/k. This was determined by comparing the results of eigenvalue calculations for several cases with the control rods at measured critical positions. Accounting for the bias in the core model, the 13eff is calculated as:

PUR-1 SAR 4-29 Rev 0,6/08

j3biased

~fl

=

k ..

eft

-~ . _(kP':Olllpl Pbws eft keft - ~Pbias The resulting biased value is obtained then as:

-~ .)

Pbws =

k. _kprOlllpl

~ft efl keft - ~Pbias p:;nsed = (1.00379 - 0.99589)1(1.00379 - 0.0032) = 7.895x1.0-3 +/- 1.79x10-4 (2.3%)

The reactivity insertion *accident analyses were performed using the unbiased Peff values, which are slightly smaller (0.3%) than the effective delayed neutron fraction determined by accounting for the bias in the MCNP model. Consequently, the reactivity insertion accident analyses were performed with a value of Peff that gives more conservative results.

The prompt neutron lifetime is calculated using the "1/v insertion method," where a uniform concentration of a 1/v absorber such as 10S is included at a very dilute concentration everywhere in the core and the reflector. The prompt neutron lifetime, e, is calculated by I = N->O lim

where k1 is the keff of the system with a uniform concentration, N1 , of a 1/v absorber, and Oa is the infinitely-dilute absorption cross section of the absorber for neutrons at speed v. For this effort, the 10S absorption cross section is assumed to be oa=3837 barns for a neutron speed of v=2200 m/s.

Reactivity coefficients provide an estimate of the reactivity response to changes in state properties, given as:

~p=a\.*Ax where ax is the reactivity coefficient due to a unit change in property x, and l1x is the value change for property x. The reactivity coefficients are calculated assuming that simultaneous changes in multiple state properties are separable. These are calculated from core eigenvalue calculations with independent perturbations to state properties, as shown here:

PUR-1 has operated at a maximum power of 1 kW and is cooled by natural convection.

The nominal conditions for the core used in the reactor design model assumed fresh fuel (i.e., no fission products given the low burnup of the PUR-1), isothermal conditions PUR-1 SAR 4-30 Rev 0,6/08

of 20°C, and impurities in the fuel, clad, and graphite based on the best-available data .

The reactivity coefficients are calculated assuming separability of the reactivity feedback effects due to changes in fuel temperature, water temperature, and water density. After establishing a nominal state based on the conditions ,given above and a critical rod configuration determined in the ~eactor design analysis, cases with perturbations to the temperatures or water density were evaluated.

The fuel temperature changes are assumed to occur uniformly throughout the reactor (i.e. the same temperature perturbation in all fuel plates), while the water temperatures or densities were perturbed in different zones of the reactor depending on the proximity to the fuel plates. In the time immediately after the initiation of a reactivity induced transient, the water inside the fuel assembly can (shown in light blue in Figure 4-23) would heat up and have a feedback effect on the transient. However, the water between the cans and also in the space between the control rod and guard plates (shown in orange in Figure 4-23) would take some time to be heated as a result of a power increase from the transient due to the slow water circulation time with natural convection cooling. It would take even longer for the water in the reactor tank to be heated. Consequently, cases were evaluated with the water temperature or density perturbed:

within the fuel assembly (light blue regions in Figure 4-23)

within the fuel assembly and the water between the assemblies (orange regions in Figure 4-23), and

within the fuel assembly, the water between the assemblies, and the pool or reflector water (all of the water in the MCNP5 model).

For the evaluation of the reactivity induced transients, only the reactivity coefficients calculated by perturbing the fuel assembly water will be used. A summary of the values determined in the analyses of the HEU and LEU reactor physics parameters and reactivity coefficients is presented in Table 4-15 .

PUR-1 SAR 4-31 Rev 0,6/08

Figure 4-23: F~~~~~~n~g~w~a~ter~r~e~g~iO~n~s:-t~~~~b"tia'n models Table 4-15: Other core physics parameters.

LEU (calculated)

Ofuel

(%M)

k. ~C Omoderalor a void

(%lIk

( %Ilk )

k . °C

)

-8.05E-04

-9.05E-03

-1.93E-01 k '%l'Oid

,ff 0.784%

I ~s) 81.3 Table 4-16: Water and Fuel Coefficients for the PUR-1 Core.

PUR-1 LEU Core Design Water Temperature Coefficient (aw tor) 201030 C -9.05t E-05 ilore +/- 19%

PUR-1 LEU Core Design Water Void Coefficient Ca v I o to 0.60% void -t .933E-03 601% void +/- 1 7%

PUR-1 LEU Core Desian Fuel Temperature Coefficient 20 to 127 °c -8.053E-06 "pre

'" . ,)

+/- 10%

PUR*1 lEU Core Desian Effective Delaved Neutron Fraction elll 0.00784 +/- 0.00008

For the LEU core, two different critical rod configurations were determined from the reactor design analysis. A rod configuration with the SS2 rod inserted at 43 em, and the PUR-1 SAR 4-32 Rev 0, 6/08

551 and regulating rods fully withdrawn, was found to result in the largest peak-to-average power density. This critical rod configuration was modeled in the LEU core reactivity coefficients and kinetics parameters calculations.

Figure 4-24: PUR-1 Core Layout with LEU Fuel.

Figure 4-26 shows the effects of the fuel and water temperature , and water density perturbations on the core reactivity. The core has a negative fuel temperature coefficient because of the Doppler effect on the U-238 capture resonances in the fuel.

The fuel temperature coefficient is smaller than that from water temperature or density effects, but it is a non-negligible parameter for the LEU accident analyses.

The behavior of the LEU core reactivity due to water temperature and density perturbations was quite similar to that for the HEU core. The LEU core does have a slightly harder neutron spectrum under nominal conditions, so the spectrum hardening due to the water temperature increase should have a smaller feedback effect on the core reactivity . On the other hand, coolant voiding in the LEU core results in greater neutron leakage because of the harder spectrum , so the negative reactivity feedback effect due to the reduced water density is greater for the LEU core.

Table 4-16 gives more detail for the reactivity feedback coefficients for water temperature and density, and fuel temperature perturbations for the LEU core. As can be seen, the water temperature coefficient becomes more negative as the water temperature is increased. The temperature coefficient calculated over the range from 20°C to 30°C for the "fuel assembly" water should be used for the accident analyses. It should be noted that the water temperature coefficient over this range was actually larger than the HEU core. For the water density feedbacks , the coefficient calculated over the range from 40°C to 60°C is the most conservative value for the accident analyses.

PUR-1 SAR 4-33 Rev 0 , 6108

Temperature coefficients of reactivity were calculated assuming separability of the reactivity feedback effects due to water temperature, water density (void), and fuel temperature. This is a common practice and makes it easy to understand the inherent shutdown mechanisms that are responsible for affecting reactivity-induced transients.

Not only are the feedback coefficient treated as separable, the reactivity coefficients were also calculated under non-isothermal conditions because different regions of the reactor will heat up at different rates during a power increase.

Three distinct regions containing water in the PUR-1 were considered.

1. The water between the fuel plates; this is called the "fuel assembly" water.

2. The water between the fuel element cans and also the water in the control elements between the control rod guard plates; this is called the "inter-assembly" water.

3. The water in the reactor tank; this is called the "reflector" water.

Water temperature coefficients were calculated by adjusting the temperature of the water in each of these regions and calculating the impact on the core reactivity.

Reactivity feedback coefficients due to the perturbation of the fuel assembly water temperature, the fuel assembly plus inter-assembly water temperature, and fuel assembly, inter-assembly, and reflector water temperature were calculated. It is important to note that only the feedback effect due to heating of the fuel assembly water (between the fuel plates) was considered in the accident analyses, because only this water would experience an immediate heating due to power increases during the transient. The other regions containing water would take much longer to heat up.

Heating the water increases the thermal motion of the hydrogen atoms in the water.

The result is to increase the energy of neutrons which are in "thermal equilibrium" with the hydrogen moderator, thus hardening the neutron spectrum in the reactor. At higher neutron energies, the U-235 fission cross section is reduced. Thus, there is a negative reactivity feedback effect-due to heating of the water between the fuel plates.

Increasing the temperature of the reflector (tank) water was found to have a positive reactivity feedback effect over a small temperature range from nominal conditions. This is due to a decrease in the neutron absorption in the reflector as the temperature is increased and the neutron spectrum hardens. For the HEU core, the feedback coefficient due to heating all water in the reactor tank from 20 to 30°C was calculated to be 2.38x10-3 +/- 2.11 x1 0- 3 % L1k/k/oC. However, it would take a long time for any transient to heat the reflector water, so it is judged that there are no significant safety issues related to the positive reactivity feedback coefficient when all the water is heated.

It should be noted that for larger increases in the temperature of the reflector water, the reactivity feedback effect is negative. -

The void coefficient was also converted to the unit of J"...p/oC. The reason for expressing the void coefficient in the alternative units was to facilitate a comparison of the water

void and temperature reactivity feedback effects, which were treated as separable. This comparison is illustrated in Table 4-15 and Table 4-16. When the reactor coolant within PUR-1 SAR 4-34 Rev 0,6/08

the fuel element (i.e., between the fuel plates, which is the region of interest for accident analyses in the PUR-1) is heated, the reactivity feedback from the water temperature increase is slightly larger.

The unit conversion to !1p/oC was accomplished by equating the void (water density)

. perturbation to a corresponding water temperature perturbation at 1.5 atmospheres.

This is the water pressure in the PUR-1 core, which is at the bottom of a 15 foot tank of water (the actual tank depth is 17 feet, but the distance to the bottom of the core to the waterline is 15 feet. The density of water at 1.5 atmospheres as a function of temperature is shown in Figure 4-25.

1.005 1.000 0.995 ~

",- 0.990 ~

E

..!:! 0.985

~

CI

~ 0.980

~ 0.975 ~

c III 0.970 ~

0.965 ~

0.960 ~ .,.

0.955 o 20 40 60 Temperature (DC) 80 100 120 Figure 4-25: Density of sub-cooled water at 1.5 atmospheres.

MCNP allows for two adjustments on neutron scattering reactions based on the temperature of the medium. For neutron energies above 4 eV, the code adjusts the elastic scattering cross sections of nuclei in the medium using a free gas thermal treatment. The code user can specify the temperature of each cell within the model and the cross sections are adjusted if the specified temperature differs from the temperature of the processed nuclear data in the cross section library.

For modeling elastic and inelastic scattering events for neutrons below 4 eV, an S(a,~)

treatment is employed. These data are available in the MCNP libraries for certain materials; light water and graphite are of interest for the PUR-1 analyses. S(a,~) data for graphite at 20°C were employed. Furthermore, S(a,~) which had previously been evaluated for light water at 20, 30, 60, 100, and 150°C were also used .

PUR-1 SAR 4-35 Rev 0,6/08

0.2%

0.0%

-0.2%

-0.4%

Fuel Temperature Perturbati on PUR 1 190-Plale lEU Core

=.

~ -0.6%

-0 .8%

.1 .0%

-1.2%

-1.4% -

-1 .6%

250 300 350 "0 500 550 600 65' Fuel Temperature (K) 0.2%

PUR 1 190-PI ...

Water Temperature Perturbation

- LEU C0 ..

0.0%

-0.2% ::--"

-0.4%

.(l.S% ----

, ~ -0.8% -------

___ F... W ....

r--- __ -------

-1 .0% F... . ' ... W . ...

-L?% f--- __ F.... ... . RHW .... ---------

-1.4% -------

-'.6% -------

60 '00 02' '"

Water Temperature ("C)

Water Density Perturbation 0.2% - -

PUR 1 190 PI ~t e LEU Core 0.0%

-0 .2%

.0.4%

.(l.S%

-0.8% ------- -........

_ _ F ... W . ...

-1 .0 %

-1.2%

__ f.*. . W....

_ _ f ... * ...

REFW_

-1 .4% -

-1 .6 %

20) 60 80 '00 Effe(:llve Water Temperature (oC) 02' Figure 4-26: Effect of Fuel Temperature , Water Temperalure and Water Density Perturbations on Core Reactivity.

PUR* ' SAR 4-36 Rev 0, 6/08

4.6 Thermal-Hydraulic Design In this section, the results of the thermal-hydraulic analyses are discussed in order to demonstrate that the PUR-1 LEU core design provides the cooling capacity necessary to ensure fuel integrity under all anticipated reactor operating conditions. Analyses for behavior under hypothetical accident scenarios are presented in Section 13.

4.6.1 NATCON Code Description Thermal-hydraulic analyses were performed using the computer code NATCON 1 ,2, which can be used to analyze the steady-state thermal-hydraulics of plate type fuel in a research reactor cooled by natural convection. The reactor core is immersed in a pool of water that is assumed to be at a constant average temperature.

NATCON computes coolant flowrate, axial temperatures in the coolant and fuel plate surface and centerline, and the approach to onset of nucleate boiling (ONB). Other safety related parameters such as the Onset of Nucleate Boiling Ratio (ONBR) and Departure from Nucleate Boiling Ratio (DNBR) are calculated as well. And an automatic search for the power at ONB can be performed.

Flow is driven by density differences in the coolant that are the result of coolant heating by the fuel. Resulting buoyant forces are counter-balanced by viscous forces that result from the flow. Hot channel factors may also be introduced for determining safety margins. NATCON v2.0 documentation is included as Appendix 1 of this document. It includes information on the calculation of hot channel factors, inputs, and use of the code.

4.6.2 Fuel Element and Fuel Assembly Geometry In PUR-1, each fuel plate element is loaded into its own assembly container, or can, as described in Section. Cross section views of the two different types of assemblies, standard and control, are shown in Figure 4-27 and Figure 4-28 .

,- I- -(

... P""

r- r- r- r- r- r- ~ ~

~ i;'

~ I-- I'-

V \

II \

1\ I 1\ I i' .... V i-' i"'"

I--- ~

....r- r- r- r- r- r-II II' II Figure 4-27: Standard LEU fuel assembly.

PUR-1 SAR 4-37 Rev 0,6/08

Figure 4-28: Control LEU fuel assembly.

Two types of channels are encountered in the PUR-1 fuel assemblies. One is the channel between plates, and the other is the channel between the last plate of an element and the assembly can wall. The plate-to-plate and plate-to-assembly wall channel thicknesses are fixed by the spacers on the wall of the assembly. It should be noted that the plate-to-wall channel is heated on only one side, so it can be conservatively assumed that half of the heat from the fuel plate associated with the fuel channel heats the coolant. Table 4-17 summarizes the channel types and thicknesses

inPUR-1 .

Table 4-17: Channel Types and Thickness in PUR-1 Assemblies Plate-to-plate (mils)

Standard Control Plate-to-wall (mils)

Standard Control LEU Asssemblies 144+/-15 181+/-15 127+/-B 127+/-B This is the smaller of the two boll heads on the HEU elements.

In the thermal-hydraulic analyses, the peak power plates identified in Section 4.5.2 were analyzed using NATCON . The relative power densities in each fuel plate were obtained from detailed MCNP5 criticality calculations, also described in Section 4.5.2 . In the NATCON analysis, the relative axial power profiles of the individual plates were utilized in each respective case.

Hot Channel Factors are used by NATCON to account for dimensional variations inherent in the manufacturing process , as well as variations in other parameters that affect thermal-hydraulic performance. The geometry dimensions used in the NATCON models for the HEU and LEU models are shown in

PUR* , SAR 4*38 Rev 0 , 6lO8

Table 4-18. And the hot channel factors are listed in Table 4-19 .

PUR-1 SAR 4-39 Rev 0,6/08

Table 4-18: Model Dimensions for the Thermal-Hydraulic Models Number of Axial Nodes Number of Plates Thermal Conductivity (W/m*K)

FuelMe~

14 80 Clad 180 Pool Temperature Cc) 30 Inches mm Fuel Meat Height Width Thickness Channel Height 25.110 637.79 Width 2.832 71.933 Channel Thickness Clad Thickness 0.015 Distance assembly can extends above fuel plate 0.450 Table 4-19: Hot Channel Factors for the LEU core Hot Channel Factors Type of Tolerance FBULK FFILM Uncertainty FFLUX Tolerance Fraction Coolant Temp. Film Temp Heat Flux Rise Rise Fuel meat thickness Random 0.000 1.000 1.000 1.000 U-235 HomoQeneity Random 0.200 1.000 1.200 1.200 U-235 Mass per plate Random 0.030 1.014 1.030 1.030 Power Density Random 0.100 1.046 1.100 1.100 Channel Thickness Random 0.110 1.180 1.111 1.000 Flow Distribution Random 0.200 1.200 1.000 1.000 Random Uncertainties Combined 1.322 1.251 1.226 Power Measurement FQ Systemic 0.500 1.500 Flow friction factor Fw Systemic 0.048 1.048 Heat Transfer Coeff. FH Systemic 0.200 1.200

1. Plate-to-plate channel, 20 mil uncertainty on 181 mil channel thickness.

The following calculations were done to determine the core power distribution and the ONB power for the new design of the PUR-1 LEU fuel assemblies. The power distribution analyses discussed below show that the LEU plate 1348 is the most limiting.

"The ONB power calculation given below includes the impact of the power density variation along the width of the plate itself

For the most limiting LEU plate, 1348 (located plate-to-plate in control assembly 4-4 as PUR-1 SAR 4-40 Rev 0, 6/08

shown in Figure 4-23), the ONB power is found to be 94.2 kW, using a 20 mil uncertainty on the 181 mil thick channel. The 20 mil uncertainty used is conservative compared to the 15 mil uncertainty given in Table 4-17 for the evaluated core design.

This ONB power is shown in Table Q27-. Reducing the channel thickness increases the ONB power because the new LEU design is approaching the optimum channel thickness (== 100 mil) which gives the highest ONB power.

Table 4-20: ONB Power With Hot Channel Factor FW to Account for Hydrodynamically Developing Flow, and the Effect of Viscosity Temperature Dependence on FBULK (Most Limiting Fuel Plate)

LEU Plate 1348 LLJ Channel Type Plate-to-Plate, 20 mil uncertainty, on 181 mil Channel Hot Channel Factor FW 1:048 Hot Channel Factor FBULK 1.322[8]

Hot Channel Factor FFILM 1.357[8]

ONB Power (kW) 94.2

~P Due to Wall Shear, Pa 14.88 Inlet + Outlet Loss ~P, Pa 2.01 Total Friction ~P, Pa 16.89

~P Due to Buoyancy, Pa 16.89 Flow rate in hot channel, 9/S 16.28 Reynolds Number at Channel Outlet 698

1. Other hot channel factors used for LEU plate 1348 are taken from Table 4-26 of the conversion proposal: FQ = 1.500, FH = 1.200, FFLUX = 1.226

2. The ratio of apparent friction parameter Capp for developing laminar flow to the friction parameter C for fully developed laminar flow, Capp/C is 1.0985 for the 197 mil thick channel in the LEU core. This ratio was calculated using Eqs. (341) and (576) of Shah and London [Ref. 2]. FW in the NATCON code input was set to 1.09850.5 because the frictional pressure drop is multiplied by a factor FW2 by the code.

3. For the LEU core (20 mil uncertainty on 197 mil thick channel), the hot channel factor for bulk coolant temperature rise, FBULK, calculated ignoring the temperature-dependence of water viscosity is 1.308 (calculated using the formula in the revised Appendix 1).

4. The value of FFILM that accounts for the power density

, variation over the plate width.

5. FBULK has changed to 1.322 due to 20 mil uncertainty on 181 mil channel thickness in the LEU core design as shown in Table 4-19. The FFILM in Table 4-19 (1.251) was increased by a factor of 1.085 to account for the power density variation along the width of plate 1348 (shown in Table 4-19).

This results in FFILM = 1.357 (=1.0850x1.251) that is used here.

Table 4-19 shows the hot channel factors for the new LEU core design, which were calculated using the equations in the revised Appendix 1 that includes the effect of temperature-dependent water viscosity. A conservative uncertainty of 20 mil (rather than 15 mil shown in Table 4-4) on the 181 mil channel thickness (the most limiting fuel plate 1348 in control assem bly 4-4) was used in find ing the hot channel factors. To calculate the ONB power, the NATCON code was run using (i) the hot channel factors shown in Table 4-19, PUR-1 SAR 4-41 Rev 0,6/08

(ii)

(iii)

(iv)

(v) a total o f . fuel plates, a channel thickness of 181 mil, a radial power factor of 1.5414 (=the ratio of 8.07 kW power in plate 1348 to 5.2356 kW at 1 kW per average plate), and the axial power shape for plate 1348 shown in Table 4-9.

To account for the power density variation along the width of plate 1348, the FFILM in Table 4-19 (1.251) was increased by a factor of 1.085 (= 1.672/1.5412 = the maximum-fo-average power density ratio variation over the width of plate 1348, shown in Table 4-9). This results in FFILM =1.085x1.251 = 1.357 that was used in NATCON to calculate the ONB power.

The NATCON code calculates the Oarcy-Weisbach friction factor f = C/Re for fully developed laminar flow, using a* built-in table of the parameter C for different aspect ratios of the rectangular channel cross section. To account for the increased pressure drop due to hydrodynamically developing laminar flow in the channel, an apparent value of the parameter C averaged over the channel length, called Capp , was calculated using Eq. (576) of Shah and London [Ref. 2]. The ratio Capp/C was found to be 1.0897 at a Reynolds number of 800 at the exit of the 181 mil channel in the new LEU design. Since the NATCON code multiplies the fully developed friction factor by FW 2 , the hot channel factor FW equals 1.044 (= 1.0897°*5). A higher value of 1.048 for FW was used in the NATCON calculation to be conservative.

The ONB power is found to be 94.2 kW which is shown in Table 4-20 for comparison.

The margin to incipient boiling shown in Table 4-28 was calculated at the present operating power of PUR-1 (i.e., 1 kW), and it is the smallest value of the temperature difference (T ONB - T w) over the coolant channel length in the hottest channel where T w is cladding surface temperature with all hot channel factors applied,and T ONB is the local onset-of-nucleate-boiling temperature. This basically gives an idea of how far below the onset of nucleate boiling condition the reactor is operating. This definition can be written as an equation as follows:

where T(z) Bulk coolant temperature at axial position z in the channel heated by the .plate power of PopFr FaiN and applying the global hot channel factors for flow and Nusselt number of Fw and Fh T wall(Z) = Cladding surface temperature at axial position z in the channel heated by a plate power of PopFr FaiN and applying the global hot channel factors for flow and Nusselt number of Fw and Fh PUR-1 SAR 4-42 Rev 0,6/08

q"(z) p(z)

=

Tincp(p(z), q"(z)F flux) -

Heat flux at position z for the plate power of PopFr Fa/N and applying the global hot channel factors for flow and Nusselt number pf Fw and Fh Absolute pressure in the channel at axial position z Onset of nucleate boiling temperature at absolute pressure p(z) and heat flux q"(z)Fflux Pop = Operating power of the reactor (e.g., 1 kW for PUR-1)

N = Number of fuel plates in the core (e.g., 190 for PUR-1 LEU core)

To = Coolant temperature at the channel inlet Fw = Hot channel factor for flow in the channel Fa = Hot channel factor for reactor power Fh = Hot channel factor for Nusselt number Fr = RPEAK = Radial power factor of the plate cooled by the channel Ffi1m = FFILM = Hot channel factor for temperature drop across the coolant film on cladding surface Fflux = FFLUX = Hot channel factor for heat flux Fbu1k = FBULK = Hot channel factor for bulk coolant temperature rise in the channel I

4.6.3 Thermal Hydraulic Analysis Results The NATCON/ANL V2.0 code was used to determine the thermal-hydraulics performance of the PUR-1. First, the code was used to compute the power at which the ONB is reached for the plates being examined in each of the HEU and LEU cores. This was done to identify the limiting channel. Then the limiting channel was evaluated under nominal operating conditions for both the HEU and LEU cores. The ONB results provide verification that the Safety Limit (SL) and Limiting Safety System* Settings (LSSS, trip points) of the Technical Specifications will indeed assure safe operation of PUR-1 for the LEU core. .

4.6.3.1. NATCON Analyses The reactor pool temperature varies throughout the year from about 22°C to 30°C depending on the ambient temperature and humidity conditions in the reactor room. In all of the following calculations, the higher value 30°C was used.

The power search function of NATCON was used to determine the power level at the Onset of Nucleate Boiling for both the HEU and LEU cores. Table 4-21 provides a summary of the ONB powers for each of the cases analyzed. For the HEU core, the limiting channel/plate was the plate-to-plate (P-T-P) case, with an ONB power of 76.3 kW. For the LEU core, the limiting channel/plate was plate 1348 with the plate-to-plate channel, which had an ONB power of 96.1 kW .

PUR-1 SAR 4-43 Rev 0,6/08

Table 4-21: ONB Powers for HEU and LEU Cores Channel Type ONS Power (kW)

LEU Plate 1348 P-T-P 96.1 P-T-W 187.8 LEU Plate 1215 P-T-P 165.6

1. Plate-to-Plate spacing

2. Plate-to-Wall spacing Using NATCON, the thermal-hydraulics parameters of the HEU and LEU cores at the nominal operating conditions were also calculated. All hot channel factors are included in these calculations. These results are shown in Table 4-22.

Table 4-22: Operating Conditions for PUR-1 as Determined by NATCON LEU (Plate 1348) LEU (Plate 1348)

P-T-P (1 kW) P-T-P (10 kW)

Max. Fuel Temp. rC) 32.3 39.1 Max. Clad Temp. rC) 32.3 39.1 Coolant Inlet Temp. (DC) 30.0 30.0 Coolant Outlet Temp. (DC) 32.0 36.0

Margin to incipient boiling (DC)

Coolant Velocity (mm/s)

Coolant Mass Flow Rate (kg/s) 78.49 5.41 0.0019 70.0 17.4 0.0062 4.6.3.2. Safety Limits for the LEU Core In PUR-1, the first and principal physical barrier protecting against the release of radioactivity is the cladding of the fuel plates. The 6061 aluminum alloy cladding has an incipient melting temperature of 582°C. However, measurements (NUREG 1313) on irradiated fuel plates have shown that fission products are first released near the blister temperature (-550°C) of the cladding. To ensure that the blister temperature is never reached, NUREG-1537 concludes that 530 cC is an acceptable fuel and cladding temperature limit not to be exceeded under any conditions of operation., As a result, PUR-1 has proposed a safety\ limit in its Technical Specifications requiring that the fuel and cladding temperatures should not exceed 530 cC.

4.6.3.3. Limiting Safety System Settings for the LEU Core Limiting safety system settings (LSSS) for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. When a limiting safety system setting is specified for a variable on which a safety limit have been placed, the setting must be chosen such that the automatic protective actions will correct the abnormal situation befor'e a safety limit is reached. Table 4-23 shows the maximum power, the LSSS and operating power for PUR-1.

PUR-1 SAR 4-44 Rev 0,6/08

Table 4-23: Key Power Levels for Reactor Operation and LSSS for PUR-1 Maximum Power Level Including 50% Uncerta,inty Limiting Safety System Settings Power Level 1kW 1.8 kW 1.2 kW 10 kW Proposed 18.75 12.0 kW Operating Power Level 1.0 kW 10.0 kW During steady-state operation, peak clad temperatures are maintained far below 530°C, as well as below the temperatures required for ONB (see Table 4-22). NATCON was used to determine the minimum power for ONB for the HEU and LEU cores in the limiting channels, as well as the thermal-hydraulic parameters at these calculated powers. The results of these calculations are shown in Table 4-24.

Table 4-24: Power Levels at the Onset of Nucleate Boiling for,HEU and LEU Cores.

LEU Power level i

at ONB (kW) 96.1 Max. Fuel Temp. (0C) 112.6 Max. Clad Temp. (0C) 112.5 Coolant Inlet Temp. (0C) 27.0 Coolant Outlet Temp. (0C) 42.6

Onset of Nucleate Boiling Ratio Coolant Velocity (mm/s)

Coolant Mass Flow Rate (kg/s) 1.0 53.3 0.019 The licensed operating power level of PUR-1 is 1 kW. The LSSS scram setting of 120%

power (1.2 kW for 1 kW, 12.0 kW for uprate) is well below the power level of 96.1 kW in the LEU core, at which ONB would occur in the respective limiting channels. Thus, the present LSSS on power at 12 kW (120% normal operating power) will easily protect the reactor fuel and cladding from reaching the Safety Limit under steady state operations.

Chapter 13 (Accident Analyses) analyzes two hypothetical transients based on values of the' Technical Specifications for the LEU core. These transients are: (1) Rapid insertion of the maximum reactivity worth of 0.3% ~k/k of all moveable and non-secured experiments, and (2) Slow insertion oft reactivity at the ma~imum allowed rate of 0.04%

~k/(k*s) due to control blade withdrawal.

For the case of the rapid insertion, of 0.3% ~k/k, the reactor scram was initiated based on the power level trip, assuming failure of the period trip. For the case of the slow insertion of 0.04% ~k/(k*s), scram was initiated on the second power level trip, assuming the first power level trip failed. The reason for this is that the period trip is never reached for the case of this slow reactivity insertion .

Thus the selected LSSS is a conservative setting which ensures that the maximum fuel PUR-1 SAR 4-45 Rev 0,6/08

f

and cladding temperatures do not reach the safety limit of 530°C for the range of accident scenarios that were analyzed. In summary; the selected LSSS will protect the reactor fuel and cladding from reaching the safety limit of 530°C under any condition of operation.

However, a NATCON thermal-hydraulic calculation for the LEU plate 1348 was performed assuming a hypothetical pool temperature of 35°C, and a hypothetical inlet loss coefficient of 10.0 (increased from 0.5), while applying all six hot channel factors of the case H. The ONB power was found to be 79.3 kW, indicating a .large margin compared to the PUR-1 operating power of 1 kW.

4.7 References R. S. Smith and W. L. Woodruff, "A Computer Code, NATCON, for the Analyses of Steady-State Thermal-Hydraulics and Safety Margins in Plate-Type Research Reactors Cooled by Natural Convection," Argonne National Laboratory (ANL),

ANLIRERTRITM-12, Dec 1988.

2 M. Kalimullah, "NATCON v2.0 Instructions", Argonne National Laboratory (ANL),

ANLIRERTR, July 2006 .

PUR-1 SAR 4-46 Rev 0,6/08

5 5.1 REACTOR COOLANT SYSTEMS Summary Description The water process system includes a 30 GPM water pump, a water filter,. a demineralizer, flow meter, a chiller, conductivity cells that measure the pool water before and after passing through the demineralizer and appropriate valves. The details of the reactor coolant system are summarized in Table 5-1: PUR-1 water process system summary.

WATER PROCESS SYSTEM Water Capacity, Reactor Pool 6,400 gal.

Pump Type Centrifugal Capacity At 100°F/11 0 ft Head 30 gpm Ion Exchanger Mixed bed (replaceable cartridge type)

Rated 'Capacity 15, 000 grains as CaC0 3 Flow Rate 0-20 gpm Design Pressure 100 psi

Heat Exchanger Cooling Capacity Design Pressure Tube Shell and tube (water chiller) 36,000 Btu/hr (10,550 W) 150 psi 225 psi Shell Filter Cartridge Type Maximum Flow 80 gpm Pressure Drop 2 psi 5.2 Primary Coolant System The process system is provided with the reactor to control the pool .water quality and temperature. The purification system is designed to limit corrosion and coolant activation by the use of microfilters and ion exchange resins.

The process system is assembled in one unit and is mounted on a base structure near one wall of the reactor room. The components include a Circulating pump, a cartridge-type demineralizer, a Cuno filter, a refrigeration compressor-condenser unit, a heat exchanger, flowmeter, conductivity indicator, thermostat, wiring, piping and valves to

complete the system. City water is available through a water supply tank to provide a source of makeup pool water. City water is also provided to the compressor-condenser PUR-1 SAR 5-1 Rev 0,6/08

unit as a secondary cooling medium .

The reactor pool system is arranged in the following order: Water from the pool is drawn out from the scupper drain or suction line via PVC pipe leading to the circulating pump; a second source of water for the pump is a water supply tank supplied with city service water and controlled by a float valve. Ball valves for water shutoff and a vacuum cleaning connection are provided in, the pump supply line. From the pump a pipe with ball valve installed leads first to the filter and then to a demineralizer. An adjustable by-pass or throttling valve is inserted in the system to regulate water flow through the demineralizer. A flow indicator and a conductivity indicator are installed as a check on water purity and flow rate from the demineralizer. The water next flows through a -stainless steel heat exchanger. The water from the heat exchanger is then returned to the reactor pool. A magnetrol water-level control is located in the reactor pool; this unit controls a solenoid valve in the line from the water -supply tank to ensure that the prescribed pool water level is maintained.

5.3 Secondary Coolant System The secondary coolant The water next flows through a Ross Model 302 stainless steel heat exchanger of 36,000 Btu/hr capacity which serves as a water chiller; included with this system is a thermostatically controlled Copelametic Model W300H compressor-condenser unit with a water cooled condenser. The water from the heat exchanger is returned to the reactor pool.

Experimentally, no temperature increase has been observed with the pool thermometer following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of operation at 1 kW. However, based on the mass of water as 1.85

-104 Kg, calculations indicate that the temperature increases after operating the PUR-I at power lever of 10 kW would be 0.465°C/hour (this takes no credit for heat loss to the surrounding sand and gravel or loss by evaporation). The capacity of the chiller system is 10,550 Watts, which will be able to maintain pool temperature as required.

The chiller is designed with three loops to prevent the ~pread of contamination in an emergency. The pool water passes through the primary loop while a freon refrigerant is in the secondary loop. The third loop uses campus water to remove the heat and is discharged into the ca-mpus sewer system. The chance of contamination passing through the three loop system is small.

5.4 Primary Coolant Cleanup System From the pump a pipe with ball valve installed leads first to the cartridge-type Cuno filter and then to a Barnstead BO-10 demineralizer. An adjustable by-pass or throttling valve is inserted in the system to regulate water flow through the demineralizer. A flow indicator and a conductivity indicator are installed as a check on water purity and flow rate from the demineralizer.

PUR-1 SAR 5-2 Rev 0,6/08

5.5 Primary Coolant Makeup Water System The water level in the pool must be maintained at least 13 feet above the reactor core during operation. During periods when there are no planned operations, the water level is still maintained, at approximately 13 feet. On average, the addition of about 40 gallons of water per week is required to maintain that water level to make up for evaporation. The primary water makeup system consists of a 20 gallon, gravity driven reservoir that is filled with water from the university water system, that is processed through deionizers consisting of each of an anion, cation and mixed-bed tank.

This tank is filled as required, water is added as needed through a solenoid controlled float switch at the top of the reactor pool. Records are kept for the amount of water used with each fill, and maintained at the reactor facility.

5.6 Nitrogen-16 Control System The main possible source of nitrogen-16 is from the fast neutron interaction with oxygen in the pool water. The nitrogen must then diffuse to the surface of the pool before it is released to the atmosphere. In normal operation, no strong currents are established in the reactor pool and with the short half-life (7.14 seconds), the nitrogen decays before reaching the surface. No nitrogen-16 has been observed in the reactor room.

5.7 Auxiliary Systems Using Primary Coolant PUR-1 has no auxiliary systems that use primary coolant.

PUR-1 SAR 5-3 Rev 0, 6/08

6 6.1 ENGINEERED SAFETY FEATURES I

Summary Description PUR-1 is located In a high-bay room that serves the function of confinement in case of emergency. The room is kept at negative air pressure (minimum of -0.05 inches of water) at all times when the reactor is operating, and air intake and. outlet are through HEPA filters. The ventilation system can be shut down during emergency situations to restrict release of contamination to the surrounding environ'ment.

The LEU MTR fuel type serves as containment 'for fission products produced in the fuel during reactor operation. This fuel type is widely used, and has shown excellent fission product retention under operating conditions much harsher than what is experienced at PUR-1, both in burnup and water quality. In the unlikely event that radioactive fission products were to leak from the fuel plates, they would be contained in the reactor pool water, which is monitored for contamination regularly, and also filtered through the water process system."

6.2 . Detailed Descriptions 6.2.1 Confinement The Duncan Annex of the Electrical Engineering Building is of brick, concrete block and reinforced concreted construction which was originally designed as a large high voltage uentl subdivided into offices classrooms and laboratories.

Figure 3.1 shows the floor plan of the Nuclear Engineering laboratories, including the reactor room.

The outside air supply and exhaust are both passed through HEPA filters. The reactor room is maintained at negative air pressure (minimum -0.05 inches of water). All doors to the reactor room have foam rubber seals. Steam heat is used to heat the room and a room air conditioner circulates and cools the reactor room air.

The only floor drain to the sewers is sealed except for a vent opening. This vent is raised about two feel above the floor and has a filtered inverted opening. Condensate from the air conditioner is released to this drain through an opening 12.0 feet above the floor.

During emergency conditions the exhaust systems is shut off and the sealed room will prevent the rapid spread of contamination. During an emergency the air conditioner and the valve on the drain from the condensate holdup tank are shut off with the same switch that shuts off the exhaust system. The condensate is held until it is tested by Radiological Control before it is released to the sewer. If contamination is found, it is disposed of as radioactive liquid waste .

PUR-1 SAR 6-1 Rev 0,6/08

6.2.2 Containment The PUR-1 reactor does not require containment.

6.2.3 Emergency Core Cooling System PUR-1 is a low power pool-type reactor, and in no credible accident scenarios would an Emergency Core Cooling System be necessary. .

PUR-1 SAR 6-2 Rev 0,6/08

7 7.1 INSTRUMENTATION AND CONTROL SYSTEMS Summary Description The PUR-1 instrumentation is summarized in Table 7-1. Is consists of 3 operating channels, and 1 safety channel. The three operating channels are the startup range channel, a log power channel, and a linear power channel. The safety channel has its own detector, as well as monitoring the log power channel for period.

Table 7-1: Summary of PUR-1 Instrumentation.

Instrumentation Detail Startup Channel Detector Fission chamber Range Source level to 10-6 Nt Indicators Log countrate meter Range Decade Scaler Log-N Period Channel CIC Detector 104 to 10 10 n/cm 2 -sec Range Log-N meter Indicators 10-4 to 300 Nt Range Period Meter Indicators -30 to +3 sec Range Power Channel Detector BF3 Ion Chamber Range 10 4 to 10 10 n/cm 2-sec Indicator Linear Level Meter Range o to 100 Nt Safety Channels Detector (period) CIC Range 10 4 to 10 10 n/cm 2 sec Detector (level) BF3 Ion Chamber Range 10-3 to 150 Nt

PUR-1 SAR 7-1 Rev 0,6/08

Rod Indicators Detectors Position Regulating Range Shim Safety 10-turn precision potentiometer 0-2500 ohms 10-turn precision potentiometer 0-2500 ohms Voltmeters Range 0-70 CM Indicators 0-70 CM Digital Voltmeter Coarse Range Fine Range Area Monitors (3)

Detectors Scintillation Detectors Range 0.05 to 50 mr/hr Indicators Local & Remote Meters 7.2 Design of Instrumentation and Control Systems The design of the instrumentation and control system for the Purdue University reactor is based on systems in use at such reactors as the Bulk Shielding Reactor, the Tower Shielding Reactor, and the Pennsylvania State University Reactor .

The instrumentation system consists of three operational channels and one safety channel. The operational channels include a counting rate channel, a log-N period channel, and a linear level channel. The *Iog-N period channel sends a signal to a composite safety amplifier and the sigma bus to provide a fast period scram. Scram systems are interconnected into these channels to effect reactor shutdown in the event of an emergency or abnormal condition. An annunciator and alarm system are included to indicate specific trouble.

7.2.1 Channel 1-"Start-up Channel The startup channel is used to monitor the neutron flux. The channel consists of a fission chamber, a preamplifier; a pulse amplifier, a scaler for accurate counting, a log count rate and period amplifier, a log count rate recorder, and shares a period recor:der with Channel #2. The range of this equipment if from 10 to 104 counts/second with periods from -30 to +3 seconds. In addition to the outputs shown on the recorders, readout is also provided by a log count rate meter and a period meter shared with Channel #2 on the console and instrument panel. The complete reactor power range may be monitored by this instrument by appropriate repositioning of the detector by means of the fission chamber drive mechanism. The fission chamber may be raised into a cadmium shield by means of a drive mechanism similar to the control rod drive units. The controls and position indication for this drive are located on the console.

Two set points, specified in the Technical Specification and based on the reactor period,

provide for a reactor setback and trip in the event of a short reactor period .

PUR-1 SAR 7-2 Rev 0,6/08

The first activity in the reactor is indicated by means of the counting rate channel. At start-up, the fission chambe~ is placed near the reactor core (lower limit). Neutron-produced pulses from the fission chamber are amplified and counted on the scaler, an electronic high-speed counting device with a mechanical register. (Smaller pulses produced by any agency other than neutrons, such as gamma radiation, are rejected by a pulse height discriminator circuit.) These amplified pulses are also fed into the log count rate meter where they are integrated and the average counting rate displayed on a logarithmic indication meter calibrated. for four decades (1 to 10,000 counts per second).

The average counting rate is also displayed on a remote indicating meter and recorded on a 10 millivolt Speedomax recorder provided with four-cycle logarithmic paper. When the control rod has been pulled out far enough to produce an effective reproduction constant greater than one (1.0), the reactor power level increases exponentially (at low power levels) and a straight line is drawn on the logarithmic recorder chart. The slope of this line is the pile period, or the time requi~ed (in seconds) for the power level to change by a factor of e (approximately 2.718).

The reactor period is displayed on a pile period meter located on the log count rate meter and is recorded on the period recorder until the lower limit of the log-N period channel range is reached.

When the counting rate channel is near the limit of its counting range, the log-N period and linear servo channels become the principal means of controlling the reactor, and the fission chamber is withdrawn to a region of lower neutron flux (to keep the recorder on scale).

7.2.2 Channel2-Log N and Period Channel The log N channel indicates the reactor power level over the range from 0~0001 to 300 percent power level. The detector for this channel is a compensated ionization chamber followed by a log N amplifier plus period instrumentation with outputs to the log N .

recorder and to the period instrumentation with outputs to the log N recorder and to the period recorder shared with Channel #1. Indication, in addition to the recorders, -is provided by log N and period meters on both console and instrument rack, with the console period meter shared with Channel #1. This channel is not 'on scale' at startup, but will be indicating before the range of the fission is exceeded.

A reactor trip will be initiated if this channel indicates power levels in excess of 120% of the licensed power. Two set points, specified in the Technical Specifications and based on th,e period, provide for a reactor setback or trip in the event of a short reactor period.

In the event of the loss of high voltage to the compensated ion chamber, a trip will be initiated.

7.2.3 Channel 3-Linear Power

The linear level channel is capable of measuring neutron flux in a reactor operating range from shutdown to > 100 kilowatt. The sensing element is BF3 ionization chamber PUR-1 SAR 7-3 Rev 0,6/08

coupled to a micro-microammeter. The range of the instrument is adjustable by means of a range switch located on the instrument (instrument panel) from 0-10.0x10-12 to 0-10.Ox1 0-4 amperes. Detector characteristics, however, limit its maximum output to 1,0-4 amperes. This channel will thus read from startup to full power by adjustment of the range switch. The output is recorded on the linear power recorder on the instrument panel and indicated on meters on the console and on the instrument panel.

This channel has two set points that will initiate a reactor set back at either zero or 100% range. These set points insure that the instrument is kept on range at all times during reactor operation. This is also a 120% range set point that will initiate a reactor trip.

The linear servo channel consists of a BF3 ionization chamber (whose output current is proportional to neutron flux) and its power supply, a micro-microampere amplifier, a Speedomax recording controller with a position-adjusting type control unit, and the regulating rod drive unit. Because this channel is the most precise channel in measuring neutron flux level, it is used as the means to obtai,n uniform neutron flux density.

The BF3 ionization chamber provides a DC current which is proportional to the neutron flux at the chamber. This current is amplified by the micro-microampere amplifier which provides a 0-10 millivolt input signal to the recording controller. The controller provides a signal to the servo control unit which, through a drive unit, drives the control rod the necessary amount in the correct direction to maintain the power level at the control point set on the controller. Whenever the power level exceeds the control point setting the control rod is inserted into the reactor the amount necessary to decrease the reactivity and restore the power level to the control point setting. Conversely, whenever the power level drops below the control point setting, the control rod is withdrawn the amount necessary to increase the reactivity and restore the power level to the control point setting.

In practice, the operator sets the red control pointer to the desired power level and selects the range on the micro-microampere amplifier which will provide a reading on the linear recorder equivalent to the power level reading desired on the log-N recorder.

The rods are then withdrawn until the black indicating pointer coincides with the red control pointer at which time the power is leveled off manually and the control unit is placed on automatic operation.

7.2.4 Log-N Period Channel The log-N period channel includes a compensated ionization chamber (whose output current is proportional to neutron flux only) and its power supply; a log-N amplifier and a Speedomax recorder which records log-N; and a composite safety amplifier whose components are a period safety preamplifier, sigma amplifier and magnet amplifier.

The compensated ionization chamber supplies a d-c current (proportional to the neutron flux at the chamber) which is amplified by the log-N amplifier. One output of the log-N amplifier is the logarithm of the power level which is indicated on the power level meter PUR-1 SAR 7-4 Rev 0, 6/08

and recorded on the log-N recorder. This output is also differentiated and indicated on the pile period meter. (The pile period is the number of seconds necessary for the power level to increase or decrease by a factor of e.) Positive periods from infinity to three seconds, and negative periods from infinity to thirty seconds are indicated.

7.2.5 Channel 4-Safety Channel This channel utilized a BF3 ion chamber and feeds directly into the safety amplifiers.

The sensitive range of this instrument is from a few percent to at least 150 percent of power, linearly. Its output is indicated on the instrument chassis (instrument panel).

The purpose of this channel is solely to provide a trip at the measure value as specified in the Technical Specifications.

The level safety channel, in conjunction with the safety circuit of the log-N period channel, functions to shut down the reactor immediately by dropping both safety rods whenever the power level increases above 120% Nf and/or an abnormally short period occurs. (Set for a period of 7 seconds.)

In the level safety channel a BF3 ionization chamber supplies a d-c current proportional to the reactor power level to the composite safety amplifier. Since the current produced by neutron flux at high levels of operation is much greater than that produced by gamma radiation or other sources, this chamber is not compensated. Each magnet amplifier supplies current to an electromagnet which holds a safety rod. The output of each

sigma amplifier is connected to the sigma bus which is connected to the input of each magnet amplifier. When there is no signal from the ionization 'chamber, the sigma amplifiers maintain the potential of the sigma bus at 37 volts. However, if the positive period should become abnormally short (7 seconds) or if the neutron flux (power) level should become dangerously high the sigma bus potential is increased which causes the magnet current in both magnet amplifiers to be quickly reduced to zero, thus dropping both safety rods into the reactor core. This is what is meant by a "fast scram."

The same result is also achieved if the sigma bus potential should increase for any other reasons, since the sigma bus is connected to the input of each magnet amplifier.

(Note that a fast scram may be initiated automatically by a short period or a high power level.) Indicator lamps in the annunciator system located on the control console indicate whether a fast scram is due to high level or to period. '

7.3 Reactor Control System Four control channels are incorporated into the reactor control system to raise or lower the two shim-safety rods, the regulating rod and the fission chamber. A fifth control channel is provided to raise and lower the source. The control rods may always be lowered with a jam circuit being the only restriction; however, permissive circuits are included in the raise circuits of the control rods and fission chamber to prevent raising the rods under the following circumstances:

Indicator PUR-1 SAR Condition 7-5 Rev 0,6/08

Source Missing The log count rate resprder must indicate the presence of a source of neutrons by indicating a count rate of at least two counts per second. A source missing indicator shows when this condition is not fulfilled.

Log Cou nt Rate A relay in the log count rate recorder prevents raising of the Recorder On control rods or the fission chamber when the log count rate recorder power is turned off, thus prohibiting the disabling of the source missing circuit. The log count rate recorder may be turned off after the log-N recorder comes on scale. This is accomplished by a relay in the log-N recorder circuit.

Source Drive in Interlock circuits on the source drive switches prevent raising Operation the control rods or fission chamber when the source is being raised or lowered.

A jam circuit is incorporated in the drive circuits to operate a jam indicator light on the console in the event of a mechanical jam in the drive. This indication alerts the operator to the possibility of cable kinking in the source and fission chamber drive units, or mechanical friction in the rod drives.

7.3.1.1. Shim Safety Rod Drive System

The shim-safety rod to be driven is selected by pushing the desired drive indicator lamp .

This connects the drive system for that rod and clears any other drive circuit that may be energized. Electrical interlocks prevent the raising of more than one control rod or the fission chamber simultaneously.

The reactor operator controls the rod with the raise-lower switch on the control console.

The raise-lower switch operates mercury relays which control the rod drive motors.

Indicator Condition Upper Limit The drive unit is at the upper limit of its travel.

Shim The drive unit is two-thirds out. This point is where the shim-safety rods are set during critical experiment fuel loading.

Engage The shim-safety rod is attached to the drive electromagnet.

Lower Limit The drive unit is at the lower limit of its travel.

Drive The drive unit is connected to the raise-lower switch.

7.3.1.2. Regulating Rod Drive System The regulating rod drive system is activated by pushing the reg rod drive indicator lamp.

This connects the regulating rod system to the raise-lower switch and clears any other drive circuit that may be energized. Electrical interlocks prevent raising more than one

control rod or the fission chamber simultaneously .

PUR-1 SAR 7-6 Rev 0,6/08

The reactor operator controls the regulating rod with the raise-lower switch on the control console. The raise-lower switch operates mercury relays which control the rod drive motors.

The regulating rod may be controlled by the servo amplifier if the reactor operator pushes the servo-permit indicator. The servo system is turned off by pushing the servo indicator.

The following indicator lights are provided for the regulating rod:

1. Upper Limit

2. Servo

3. Servo Permit

4. Lower Limit

5. Drive 7.3.1.3. Fission Chamber Drive System The fission chamber drive system is activated by pushing the fission chamber drive indicator lamp. This connects the fission chamber drive system to the raise-lower switch and clears any other drive circuit that may be energized. Electrical interlocks prohibit raising the fission chamber'while the control rods are being driven .

The reactor operator controls the fission chamber with the raise-lower switch on the control console. .

The following indicator lights are provided for the fission chamber:

1. Upper Limit

2. Lower Limit

3. Drive 7.3.1.4. Source Drive System The source is raised or lowered by pushing either the source raise or source lower indicator lamp. The source drive switches are of the momentary type and must be pushed and held on until it is desired to stop the source drive or one of the limits is reached.

The following indicator lights are provided for the source drive:

1. Upper Limit

2. Source-Raise 3 . Source-Lower

4. Lower Limit PUR-1 SAR 7-7. Rev 0, 6/08

7.3.1.5. Gang Lower System When the reactor is to be shut down, it is permissible to lower all control rods simultaneously by pushing the gang lower indicator light and thus activating the gang lower switch and relay. The gang lower relay directly energizes the lower drive relays for the control rods. The gang lower relay is de-energized by momentarily placing the raise-lower switch into either its raise or lower position. In addition, the gang lower relay may be energized by the setback system described in Section 1.5.5.2.

7.3.1.6. Rod Position Indicating System Four coarse rod position indicators are provided which continuously monitor the positions of the two shim safety rods, the regulating rod and the fission chamber with a resolution of 2 centimeters. A ratiometer is used as a fine position indicator which may be switched to anyone of the control rods or fission chamber to provide a position indication with a resolution of 0.1 millimeter. The rod position indicator measures a voltage across a potentiometer connected to the drive unit. The reference voltage for all drive unit potentiometers is supplied by a 68 volt power supply.

7.4 Reactor Protection ,System Two types of scrams are included in the control system to effect shutdown of the reactor in the event of emergency conditions. A fast scram can be accomplished when a short, period signal or a high power-level signal removes magnet current. The fast scram operation is described in Section 7.2.5.

A slow scram is initiated by various safety circuits located in the reactor system. The slow scram circuit opens the input power line to the magnet amplifier power supplies located in the composite safety amplifiers.

I A setback system is included in the reactor control system to insert the control rods into the reactor without producing a scram. This minimizes the necessity of repeating the startup procedure.

'When a scram or setback condition occurs, or when other trouble arises, a buzzer alarm will sound, and a lamp or lamps will be lighted on the control console, indicating the source of the trouble. An annunciator acknowledge button is used to turn off the buzzer. If the trou'ble results in a scram, a scram reset switch must be actuated before magnet power can be reapplied. In the event of trouble other than a scram, the corresponding switch-lamp is pushed to extinguish the lamp and reset the annunciator system after the trouble has been corrected. If the trouble is not corrected, the indicator light will remain lighted. An annunciator test switch is provided for checking the lamps and the buzzer. An evacuation alarm horn is installed in the reactor instrument racks.

This horn is activated by pushing the alarm button on the reactor console.

7.4.1.1. Slow Scram System

Opening of a fault switch actuates a slow scram channel which causes primary power to be removed from the two magnet amplifiers located in the composite safety amplifiers, PUR-1 SAR 7-8' Rev 0,6/08

thus resulting in the shim-safety rods dropping into the core. The following conditions result in a slow scram.

Slow Scram Type Condition Scram - Console Manual pushbutton on console depressed.

High Level - Console Monitor High radiation level at the console radiation monitor.

High Level - Process System Monitor High radiation level at the process system radiation monitor .

High Level - Rod Drive Area Monitor High radiation level at the rod drive area radiation monitor.

CIC Power Supply Trouble Failure of the high voltage power supply for the compensated ion chamber.

\

Composite Safety Amplifier Trouble Failure of oneor more of the circuits in either composite safety amplifier.

The following scram conditions are initiated by relay type meters located on the auxiliary scram panel. The scram set points are determined and set by the reactor supervisor prior to operation .

Scrams Conditions Low-Level Period Period circuit in the log count rate meter indicates a short period.

High-Level Period Period circuit in the log-N period amplifier indicates a short period.

High-Level Log-N The log-N channel indicates a high power level.

High-Level - Linear Level The linear level channel indicates a high power level.

The following permissive circuits are located in the slow scram circuit and must be in operation before the shim-safety rods can be raised:

1. The 10g~N selector switch must be in the operate position.

2. The log-count-rate meter switch must be in the use position.

3. The key switch must be on.

4. The scram reset button must be pushed and the magnet power lamp energized .

PUR-1 SAR 7-9 . Rev 0,6/08

, 7.4.1.2. Setback System Four setback circuits are included in the reactor control system. Actuation of anyone of these circuits will result in the shim-safety rods and the regulating rod being driven into the core to their lower limits unless the trouble is cleared and the circuit reset before the lower limits are reached. The .setback set points are determined and set'by the reactor supervisor prior to operation. The following setback conditions are initiated by relay type meters located on the auxiliary scram panel:

Low Level Period Period circuit in the log count rate meter indicates a short period.

High-Level Period Period circuit in the log-N period amplifier indicates a short period.

High-Level - Linear Level The linear level channel indicates a high power level.

High-Level - Safety Amplifier The level safety amplifier indicates a high power level.

7.5 Control Console and Display Instruments The reactor console is designed to provide maximum visibility of the instruments and accessibility to the controls and indicators. All indicators and controls necessary for startup and shutdown operations are located in one group in front of the operator.

Colors for the indicator lights on the console show. the operator the status of the reactor at a glance. All trip and warning indicators are red or yellow. Operating procedure, as well as interlock, keep the operator from withdrawing the control rods when a warning indicator is showing.

7.6 Radiation Monitoring Systems Three scintillation type area. monitors are installed in the reactor area to monitor the radiation level at the top of the pool, in *the pool water flowing through the process system, .and at the reactor console. Each area monitor is connected into the slow scram circuit and is equipped with a local lamp and alarm that is activated if the radiation level exceeds the set point. Three remote area monitor meters are mounted on the reactor instrument racks to provide the operator with an indication of the radiation level at the area monitor sites .

PUR-1 SAR 7-10 Rev 0,6/08

8 8.1 ELECTRICAL POWER SYSTEMS Normal Electrical Power Systems The instrumentation and controls of PUR-1 use the standard building electrical power system at 120 Volts on a dedicated circuit. The water process system uses three-phase power at 240 Volts. Room lighting and HVAG are also powered by building power on their own circuits. .

8.2 Emergency Electrical Power System There are no emergency electrical power systems for any of the systems associated with PUR-1. Loss of electrical power during operation will de-energize the magnet amplifiers on the shim-safety rods, which will drop them into the core and shutdown the reactor similar to a scram, leaving the reactor in a safe condition under any credible operating or accident scenario.

Emergency lighting for the room will be activated upon loss of building power, enabling the operators to ensure the safe shutdown condition of the reactor (i.e. shim-safety rods inserted) and direct personnel in the reactor room as necessary .

(

PUR-1 SAR 8-1 Rev 0, 6/08

9 9.1 AUXILLARY SYSTEMS Heating, Ventilation and Air Conditioning Systems The outside air supply and exhaust are both passed through HEPA filters. The reactor room is maintained at negative air pressure (minimum 0.05 inches of water). All doors to the reactor room have foam rubber seals. Steam heat is used to heat the room and a room air conditioner ci rculates and cools the reactor room air.

9.2 Fuel Storage and Handling a one quarter to reduce fuel stored there. There is also a dry storage facility , located within the facility. but outside of the reactor room.

An MCNP model of the in~pool storage racks was constructed. Two cases for the LEU fuel were examined. One case was run with standard LEU assemblies in all of the 18 position s (which is not possible with the anticipated LEU inventory, but was run as a limiting case). and no cred it was taken for the y." BORAL plate between the two rows.

Thi s first case had a calcu lated eigenval ue of 0.7660+/-0.0046. The second case was modeled with the y." BORAL plate with a boron density of 23.8E21'tom'/~. and 16 standard assemblies with 14 fuel plates in each. The eigenvalue for this calculation was determined to be 0.3319 +/-0.00178. Both of these bracketing cases are below the TS 5.3 limit of O.B. thus TS 5.3 will be met. T he geometry of the model is shown in Figure 037-1 below.

Figure 9-1: MCNP model of in-pool fuel storage rack .

The storage of the extra LEU plates is in the dry storage facility. Parametric eigenvalue calculations have been performed with MCNP5, va rying the storage plate spacing in a hypothetical flooded condition to ensure that a critical configu ration could not be achieved. These results are shown in Figure 9-2. The maximum keff obtained in these

calculations was 0.41, which is less than the Techn ical Specification (TS 5.3.1) requirement of less than O.B. The model used for the dry storage is shown in Figure PUR-1 SAR 9-1 Rev 0, 6/08

9-3.

0.45 k*values for Extra LEU Plates in a Flooded Storage Facility 0.4 f-0.35 * * -

0.3 0.25

~

0.2

0. 15 0.1 0.05 0

0 02 0.4 06 08 1.2 1.4 Plate Spacing (em)

Figure 9-2: k-eff Values for Flooded Condition of Fuel Storage Facility.

Figure 9-3: MCNP model of dry fuel storage facility.

Wilh all of Ihe LEU fuel plates in Ihe inventory located in the dry storage facility , the calculated eigenvalue of the slorage rack is 0.015+/-0.0003. Should the storage fiood with water, Ihe eigenvalue becomes 0.26+/-0.004. In both of Ihese conditions, the TS limit of keH<O.B is met. Parametric analyses were performed with spacing of the pi ales within the tubes , and the above listed cases are the worst case. All movement of fuel is performed under the direction of licensed senior reactor operators.

9.3 Fire Protection Systems A fire extinguisher, maintained by the Purdue Fire Department, and appropriate for the types of fires that could be encountered in the reactor room , is located within the reactor room.

PUR-, SAR 9-2 Rev 0, 6/08

9.4 Communication Systems A telephone is located at the console from which the operator can contact the Senior Operator on call, or any other assistance that may be necessary to support operations.

Dialing 911 from the console phone will direct a call to Emergency Services.

9.5 Possession and Use of Byproduct, Source and Special Nuclear Material Purdue University has a 8roadscope license, and all byproduct material generated with the reactor falls under this license, as described in Chapter 11 .

PUR-1 SAR 9-3 Rev 0,6/08

10 EXPERIMENT FACILITIES AND UTILIZATION 10.1 Summary Description PUR-1 experimental facilities consist of experiment locations within the graphite reflector 'on one side of the reactor, and drop tubes located as close as next to the reflector boundary, and as far away as 30 inches. All of the drop tubes are dry air tubes, and the in-reflector facilities are aluminum tubes normally filled with graphite, which can be replaced with experiment capsules.

10.2 Experimental Facilities The aluminum sample tubes in the irradiation facility are filled with graphite unless they are being used. Replacing the graphite in the six sample tubes with air, as would be expected if samples were inserted, reduces the calculated eigenvalue of the LEU core from 1.00731 +/-0.0002 to 0.996+/-0.0002 without any reactivity bias, but with the estimated bias of 0.32%, these values would be 1.0041 and 0.993, respectively. .

The effect of replacing graphite in the irradiation facility with an aluminum sample holder on the reactor power distribution used in the thermal-hydraulic analyses is well covered by the' global hot channel factor of 1.5 for the reactor power measurement uncertainty used in all the reported analyses. The reactor power hot channel factor was used in

addition to (1) the radial power factor (used to account for plate-to-plate power variation), (2) the factor used to account for power density variation along the width of the hot plate, and (3) the axial power profile. While replacing the graphite in the irradiation facility with an aluminum sample holder will cause a change in the reactor power distribution, this change will be small compared with the hot channel factors already applied in the thermal-hydraulic analysis. Therefore, no additional thermal-hydraulic analysis is needed.

Other experimental facilities include dry air drop tubes. Two tubes are located next to the core, one 5/8" in diameter, and the other 1.75" in diameter. A 3" PVC drop tube is located in the pool, and a 5" stainless steel drop tube is available that can be located in any grid location. An analysis of the flooding of the 5" drop tube is included in the safety analysis in Chapter 13.

10.3 Experiment Review Review of new experiments is performed by the reactor staff, and then by the Committee on Reactor Operations, as required by the Technical Specifications.

10.4 . References

PUR-1 SAR 10-1 Rev 0,6/08

11 RADIATION PROTECTION AND WASTE MANAGEMENT 11.1 Radiation Protection Program Purdue University has a structured radiation safety program. Policies for the program are determined by the University Radiation Safety Committee, which has the mission to ensure the safety of the University and community in the utilization of all radioactive materials and radiation producing devices at the University by faculty, staff or students.

This includes all teaching, research, and outreach programs. The program is administered by the Radiation Safety Officer (RSO) and his staff, as part of Radiological and Environmental Management (REM). The staff is equipped with radiation detection instrumentation to determine, control and document occupational radiation exposures at the reactor facility, and all 'laboratories using radioisotopes at the university under the By-product License 13-02812-04 (Broadscope).

Natural background radiation levels in the West Lafayette area result in an average exposure. of about 100 mrem/yr. On the basis of normal reactor use, the maximum potential non-reactor room dose would be less than 1 mrem/yr, so there should be no significant contribution to the background radiation in unrestricted areas.

11.1.1 Radiation Sources 11.1.1.1. Reactor Radiation from the reactor core if the primary source of radiation directly related to reactor operations. Radiation exposure rates from the reactor core are reduced to acceptable levels by the water in the pool and concrete shielding.

11.1.1.2. External Sources Sources of radiation associated with reactor use include radioactive isotopes produced for research, activated components of experiments and activated samples.

11.1.2 Radiation Protection Program Purdue University "Executive Memorandum No. B-14" establishes the University administrative structure for radiation protection, and a "Radiation Safety Manual" is published and maintained by REM, which contains the rules and radiation safety procedures for all laboratories using radioisotopes and/or ionizing radiation, including the reactor. Routine surveys are performed of the reactor room and include analysis of the reactor pool and reactor room air by personnel from REM ..

The University ha,s a variety of detecting and measuring instruments for monitoring potentially hazardous ionizing radiation. The instrument calibration procedures and techniques ensures that any credible type of radiation and any significant intensities will be detected promptly and measured correctly .

All reactor-related personnel are required to attend a radiation safety training session PUR-1 SAR 11-2 Rev 0, 6/08

before they begin work at the reactor. Written procedures have' been prepared that address routine health physics monitoring at the University's research reactor facility, and all reactor personnel are trained in these as well.

11.1.3 ALARA Commitment The University is committed to the principle of ALARA (As Low as Reasonably Achievable), and REM makes every effort to keep doses to* a minimum. All unanticipated or unusual exposures are investigated.

11.1.4 Radiation Monitoring and Surveying 11.1.4.1. Fixed-Position Monitors The PUR-I has 3 fixed-position radiation area monitors (RAM) with adjustable alarm set points and 1 continuous air monitor (CAM) ill the reactor room. The CAM air filters are changed and analyzed semi-monthly.

11.1.4.2. Experimental Monitoring Wipe tests of exposed surfaces of the reactor room are made monthly. Water samples are taken and counted monthly. All samples and material removed from the reactor are checked for levels of activity and wipe tests made for loose contamination .

11.1.4.3. Personnel Monitoring TLD badges and TLD finger rings are assigned to all approved reactor personnel. In addition, self reading pocket dosimeters and dose rate instruments are used to administratively keep occupational exposures below regulator limits in 10 CFR 20.

Students and visitors are provided self reading pocket dosimeters.

11.1.5 Radiation Exposure Control a~d Dosimetry The reactor staff always operates under the prinCiples of ALARA, and personnel exposures are maintained at levels that are well below those required by 10 CFR 20.

Personnel exposures for the recent operating history are detailed in Table 11-1 below.

Table 11-1: PUR-1 Personnel Exposures for 20~3-2007 Number of Individuals Exposure Range (rem) 2003 2004 2005 2006 2007 0 0 0' 2 3 1

< 0.1 1 1 1 0 2 0.10-0.25 0 0 0 0 0 0.25-0.75 0 0 0 0 0 0.75-1.0 0 0 0 0 0

~ 1.0 0 0 0 0 0, PUR-1 SAR 11-3 Rev 0,6/08

11.1.6 Contamination Control

-Wipe tests of exposed surfaces of the reactor room are made monthly. Water samples are taken and counted monthly. All samples and material removed from the reactor are checked for levels of activity and wipe tests made for loose contamination.

11.1.7 Environmental Monitoring 11.1.7.1. Airborne Effluent Argon Argon-31 is produced by thermal neutron activation of Argon-40 in the air.

No detectable traces of Ar-41 from air dissolved in the water or in the isotope irradiation tubes has been observed. No detectable tritium has been observed in the pool water.

Even at the design power levels of 10 kW the neutron flux is too low to produce detectable quantities.

The main possible source of nitrogen-16 is from the fast neutron interaction with oxygen in the pool water. The nitrogen must then diffuse to the surface of the pool before it is released to the atmosphere. In normal operation, no strong currents are established in the reactor pool and with the short half-life (7.14 seconds), the nitrogen decays before reaching the surface. No nitrogen-16 has been observed in the reactor room.

A continuous air monitor (CAM), which utilizes a GM tube as a detector, is in operation in the reactor room to indicate long term levels of radiation and to monitor any radioactive particulates released to the air in the room. No airborne radioactivity has been measured at PUR-1. No airborne effluents have been released from the PUR-1

~acility.

11.1.7.2. Perimeter monitoring TLD badges are located at the boundaries of the facility, and are checked for exposure every other month. The doses to these badges are at background levels. Therefore, facility operation has not resulted in public exposures greater than those specified in 10 CFR 20.

11.2 Radioactive Waste Management 11.2.1 Radioactive Waste Management Program Only very limited contaminated materials are generated by PUR-1. Any contaminated material is disposed of under the Purdue University 8roadscope license, as described previously.

11.2.2 Radioactive Waste Controls Every effort is made to limit the generation of radioactive waste. Very limited amounts of radioactive materials have been produced at PUR-1 in the 46 years of operation.

PUR-1 SAR 11-4 Rev 0,6/08

11.2.3 Release of Radioactive Waste Disposal of radioactive material is under the 8roadscope license, as described previously. No wastes have been released to the environment in an uncontrolled manner.

11.3 Conclusions

PUR-1 SAR 11-5 Rev 0,6/08

12 CONDUCT OF OPERATIONS 12.1 Organization 12.1.1 Structure The reactor facility is an integral part of the Schools of engineering at Purdue University as shown in Figure 6.1 The reactor supervisor has direct responsibility for the operation of the PUR-I. He is responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, Technical specifications, and other applicable regulations.

Pres'ident r----- purduebriiv~rsity ProvQst Leve! '1 ------1 P~rd~ Un!versiiy

,Radiati.or,i, Safety Committ~e D~an*

College of Engine~rin$1, Head T

I I

I NU<<:I'ear i:nlJineerinlJ I

Radiation Laboratory' Oi recto r' .C9mmlttee,'On

$.~fety o.ffi Cer Level 2 - R.ea~tor

Operations' R¢actor, Sup,ervi~oJ Level'3

'ReactOr 'Ope rat Io'ns

- - - - " Primarily,Administration

- - - Prlmarily'Safety Figure 12-1 PUR-1 SAR 12-1 Rev 0,6/08

12.1.2 ~esponsibility The Provost (Level 1) is the individual responsible for the facility's licenses or charter.

The Laboratory Director (Level 2) or the designated alternate is responsible for overall reactor facility operation. The Reactor Supervisor (Level 3) shall be responsible for the day-to-day safe operation of the PUR-1. The Reactor Supervisor is be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, including the technical specifications and other applicable regulations.

In all matters pertaining to the operation of the reactor and the administrative aspects of these technical specifications, the Laboratory Director' (Level 2) [or the Reactor Supervisor (Level 3) in the absence of the Laboratory Director] shall report to and be directly responsible to the Head of the School of Nuclear Engineering, and the Provost (Level 1). In all matters pertaining to radiation safety they will work with the Radiation

'\, Safety Officer.

12.1.3 Staffing, Selection and Training of Personnel At the time of appointment to the position, the Level 1 Licensee (Insert Level One Title) shall receive briefings sufficient to provide an understanding of the general operational and emergency aspects of the reactor facility.

At the time of appointment to the position, the Laboratory Director (Level 2) shall have a minimum of five years of nuclear experience. The individual shall have a recognized baccalaureate or higher degree in an engineering or scientific field. Education or experience that is job-related may be substituted for a degree ona case-by-case basis.

The degree may fulfill four years of the six years of nuclear experience required on a one-for-one time basis. The individual shall receive appropriate facility-specific training based upon a comparison of the individual's background and capabilities with the responsibilities and duties of the position. Because of the educational and experience requirements of the position, continued formal training may~ not be required.' The Laboratory Director shall possess a valid Senior Operator License, and meet the certifications requirements of the licensing agency.

Reactor Supervisor (Level 3) ;. At the time of appointment to the active position, the reactor supervisor shall have a minimum of five years of nuclear experience. He shall \

have a baccalaureate degree or equivalent experience in an engineering or other scientific field. The degree may fulfill four years of experience on a one-for-one basis.

The reactor supervisor shall possess a valid Senior Operator License. During periods when the Reactor Supervisor is absent, these responsibilities may be delegated to a Senior Reactor Operator (Level 4)

Licensed Senior Operator (Level 4) - At the time of appointment to the active position, a senior operator shall have a minimum of a high school diploma or equivalent and should PUR-1 SAR 12-2 Rev 0,6/08

have four years of nuclear experience. A maximum of two years of experience may be fulfilled by related academic or technical training on a one-for-one time basis. He shall hold a valid NRC Senior Reactor Operator's license.

Licensed Operator - At the time of appointment to the active position, an operator shall have a high school diploma or equivalent. He shall hold a valid NRC Reactor Operator's license.

Operator Trainee - An operator trainee shall have all the qualifications to become a licensed operator except for possessing an operator's license.

12.1.4 Radiation Safety The reactor The reactor facility is an integral part of the Schools of engineering at Purdue University as shown in Figure 6.1 The reactor supervisor has direct responsibility for the operation of the PUR-I. He is responsible for assuring that all operations are conducted* in a safe manner and within the limits prescribed by the facility license, Technical specifications, and other applicable regulations.

In all matters pertaining to the administrative aspects of the operation of the reactor, the reactor supervisor reports directly to the Head of the School of Nuclear Engineering or his designated alternate. Financial budgets for the operation of the reactor are handled through the School of Nuclear Engineering .

In all matters pertaining to radiation safety the reactor supervisor is responsible to the Radiological Control committee, usually through the Radiological Control Officer. The Radiological Control Committee was established by the President of the University under an executive memorandum A-50. The duties of the committee, revised under executive memorandum B-14, include responsibility, from the standpoint of safety, for all University programs involving radioactivity or producing radiation. The committee determines the policies and reviews all applications for use of radioactive materials and radiation on the Purdue campus. The Radiological Control Program is administered by the Radiation Control Officer and his staff. The Radiological Control Committee also have several subcommittees reporting directly to it, including the Committee On Reactor Operations (CORO).

Qualifications for the reactor supervisor, reactor staff, and Radiological Control Officers are established in Section 6.1 of the Technical Specifications. The duties of the reactor staff are also specified in Section 6.1 of the Technical Specifications.

12.2 Review and Audit Activities Review of the facility activities shall be performed by the CORO. Audits of the facility activities shall be performed under the cognizance of the CORO but in no case by the personnel responsible for the item audited. Individual audits may be performed by an

individual who needs not be an identified CORO member. These audits shall examine the operating records, procedures, retraining of the facility staff, results of actions taken PUR-1 SAR 12-3 Rev 0,6/08

to correct deficiencies, facility emergency plan, and the facility security plan at intervals designated in the Section 6.2 of the Technical Specifications.

12.3 Procedures Written procedures for all reactor operations are prepared by the reactor staff and reviewed by the Committee On Reactor Operations (CORO). Those proposed changes to procedures, equipment or systems that change the original intent or use and/or are non-conservative or those that involve an unreviewed safety questions as defined in Section 50.59, 10 CFR are reviewed and approved by CORO before being implemented.

12.4 Required Actions In the event of a Safety Limit Violation, the following actions shall be taken:

a. The reactor will be shut down immediately and reactor operation will not be resumed without authorization by the Commission.

b. The Safety Limit Violation shall be reported to the Director of the appropriate NRC Office of Inspection and Enforcement (or designee), the Laboratory Director and to the CORO not later than the next work day .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the CORO. Tbis report shall describe (1) applicable

, circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and '(3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the CORO and the Reactor Supervisor within 14 days of the violation, in support of a request to the Commission for authorization to resume operations.

12.5 Reports 12.5.1 Annual Operating Reports A report covering the operations for the previous year is submitted to the director of the appropriate NRC office by March 31 st of each year. It includes:

1. Changes in Plan design and operation

2. Power generation

3. Unscheduled shutdowns

4. Maintenance

5.

PUR-1 SAR Changes, tests and experiments 12-4 Rev 0,6/08

6. Radioactive effluent releases .

12.5.2 Non-Routine Reports In the event of a reportable occurrence, notification shall be made within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and/or telegraph to the Director of Regional Regulatory Operations Office, followed by a written report within ten days. The written report on these Abnormal occurrences, and to the extent possible the preliminary telephone and/or telegraph notification shall:

1. Describe, analyze and evaluate safety implications,

2. Outline the measures taken to assure that the cause of the condition is determined,

3. Indicate that corrective action taken to prevent repetition of the occurrence and/or similar occurrences involving similar components or systems,

4. Evaluate the safety implications of the incident in light of cumulative experience obtained from the record of previous failures and malfunctions of similar systems and components.

12.5.3 Unusual Events A written report shall be forwarded within 30 days to the director of the appropriate Regulatory Operations Office in the event of: '-

1. Discovery of any substantial errors in the transient or accident analysis or in the methods used for such an analysis, as described in the Hazards Summary Report on the bases for the Technical Specifications.

2. Discovery of any substantial variance from performance specifications contained in the Technical Specifications.

3. Discovery of any condition involving a possible single failure which, for a system designed against assumed single failures, could result in a loss of the capability of the system to perform its safety function.

12.6 Records Records to be maintained for the life of the facility are stored in filing cabinets in the Reactor Supervisor's office, or in storage cabinets in facility spaces. Records to be kept for five years are also kept in the Reactor Supervisor's office. Recent documentation also has electronic versions kept on private network servers. Security associated documentation is not stored on any servers, and is kept in a secure repository.

12.7 Emergency Planning

The PUR-I reactor Emergency Plan (EP) includes the guidelines, policy, and organization required to mitigate the consequences of an emergency.

PUR-1 SAR 12-5 Specific Rev 0,6/08

implementation procedures are provided for each type of emergency in the standard operating procedures for the PUR-I reactor. A revised EP has been approved by the NRC, and is maintained as a separate document from this Safety Analysis Report.

12.8 Security Planning There is a Physical Security Plan (PSP) for the PUR-I reactor facility which describes the physical protection system and the security organization which provides protection against radiological sabotage and detection of theft of special nuclear material from the facility and related laboratories. The physical security plan was submitted as Amendment 4 and Amendment 5 of the reactor license, and is on file with the NRC. It is withheld from public disclosure pursuant to 10 CFR 2. 790 (d).

An audit of the PSP is performed biennially, which is overseen by the Committee on Reactor Operations, and the PSP is revised as necessary when changes are recommended in the audit process, or as other external conditions change. A revised PSP has been approved by the NRC, and is maintained as a separate document from this Safety Analysis Report.

12.9 Quality Assurance The Laboratory Director and the Reactor Supervisor are responsible for maintaining quality assurance for continued safe operation of the PUR-1 reactor. This includes operator performance monitoring through the requalification and training program, continuous evaluation of existing operating procedures, development of new operating procedures, testing and calibration of reactor equipment, repair of existing equipment, and installation of new equipment. The Committee on Reactor Operations (CORO) provides review, audit and oversight functions as defined in the Technical Specifications for these activities. Checklists and other appropriate documentation, including operations logbooks, are kept to maintain a record of these activities.

12.10 Operator Training and Requalification The PUR - I reactor facility has an NRC-approved operator requalification program that all licensed reactor operators and senor reactor operators must complete as a condition for renewal of their licenses. Persons who are preparing to take the NRC operator's licensing examination participate in essentially the same training program as well as receive extensive 'hands on' reactor operations training at the console. All licensed operators at the PUR-I participate in the program and must satisfactorily complete this program during each license renewal period. Each licensed operator or senior operator includes in his/her license renewal application a statement that he/she has satisfactorily completed the requirements of the requalification program. The requalification program is divided into three major areas which are designed to provide assurance that all operators maintain competence in all aspects of the licensed activities. The three areas are as follows:

1. Lectures followed by examination of various parts of the reactor operations PUR-1 SAR 12-6 Rev 0,6/08

2.

Technical Specifications, emergency plans, and security plans. Special lectures or makeup studies are used to retrain those operators who demonstrate deficiencies in any part of the examinations.

An annual written examination is used to verify the operator's overall knowledge level of reactor operations.

3. An annual evaluation of the operator's performance on the reactor console to .

actual and/or simulated plant conditions.

At the present time there are only two licensed operators for the facility, both are senior operators and both annually teach courses in reactor operation and the facility.

Therefore, in a letter dated March 23, 1982, the facility has received permission to suspend the retaining program for the staff. It will again be reinstated at such a time as the reactor staff is increased .

PUR-1 SAR 12-7 Rev 0,6/08

13 ACCIDENT ANALYSES In this chapter, details of the analysis of various accident scenarios are presented. The results of some of these analyses validate the safety system settings established in the Technical Specifications for the PUR-I. The potential effects of the accidents on the health and safety of the staffand public are analyzed.

13.1 Accident-Initiating Events and Scenarios 13.1.1 Maximum Hypothetical Accident In the scenario of this accident it is assumed that a capsule containing irradiated fissile material breaks and a portion of the fission product inventory becomes airborne. The consequences of the release are analyzed for both the reactor staff and general public.

Since the potential impact of this postulated accident is greater than in any other accident analyzed, the failure of a fueled experiment is designated as the maximum hypothetical accident of the PUR-I.

13.1.2 Insertion of Excess Reactivity This accident scenario characterizes the reactor response to an insertion of the maximum allowable excess reactivity for the PUR-1 reactor, 0.6% ~k/k. This transient is examined in four different scenarios, a fast (step) and. slow insertion, and each of these with and without scram. In the with scram case, it is assumed that the first scram signal (likely from short period) fails, and the reactor trips on power. The power scram is set for 120% power (12 kW), and it is also assumed that thereis a 50% uncertainty in power measurement (actual scram happens at 18 kW), and that the scram signal is delayed by 0.1 s before power is cut to the electromagnets that hold the shim-safety rods at their respective heights. It is also assumed that only the least reactive shim-safety rod, SS-2, is able to be inserted into the core (stuck rod scenario).

For the no scram case both rods are stuck at their fully raised position out of the core, which would be similar to the reactor being critical on one shim-safety rod, and that rod being ejected from the core (not a credible accident).

Tbe results for these accidents are summarized in

PUR-1 SAR 13-1 Rev 0,6/08

Table 13-1 .

PUR-1 SAR 13-2 Rev 0,6/08

Table 13-1: Peak power and clad temperature for trip and no-trip insertions of 0.6%

Llk/k.

Time of T clad max (DC)

Po Pmax Peak at at Peak SCRAM Reactivity Inserted (kW) (kW) Power (s) t=O Power Maximum YES 0.6% ~k/k step 10 40.3 0.187 49.94 49.73 49.94 YES 0.6% ~k/k over 10 s 10 18.4 7.83 49.94 42.30 49.98 NO 0.6% ~k/k step 10 2388 672 49.94 133.08 133.08 NO 0.6% ~k/k over 10 s 10 2388 680 49.94 133.08 133.08 13.1.3 Loss of Coolant The reactor pool is designed to prevent the possibility of an unintentional drainage. It is const nk with n filled with Therefore, a sudden loss of coolant is considered to be extremely remote. If the pool drained instantaneously, while the reactor was operating, the loss of water (moderator) would shut the reactor down.

The most severe problem identified in this accident scenario if the removal of decay heat during and after loss of coolant. There is no. danger of significant fuel overheating as long as the core stays immersed and heat can be removed by the water. If the core were to become uncovered, heat transfer would occur by natural convection of ambient air. For this case, the amount of heat removed is proportional to the cladding.

temperature. Decay heat generation after reactor shutdown is shown in Figure 7.1.

According to this Figure, the decay power of the PUR-I immediately after the shutdown

. from full power (1 kW) is about 63 watts. The decay power rapidly decreases as indicated in Figure 7.1, being about 35 watts after 1 minute of cooling. At these power levels, no heating problem exists.

If any accident which is reasonably conc~ivable, the leakage of water from the reactor pool is expected to be rather slow. In such a case the radiation area monitor .mounted directly above the core would detect any additional radiation coming from the core due to the decreasing pool water level. The pool water level is checked during daily routine operations. It is concluded, that a slow leak of pool water would be discovered early and specific actions could be taken to mitigate it consequences.

It is concluded that no adverse consequences are to be expected to the health and safety of the public or the staff from this type of accident.

13.1.4 Loss of Coolant Flow

The Purdue University Reactor is cooled by natural convection, with peak flow rates at the onset of nucleate boiling (ONS, 98.11 kW) of 5.41 cm/s, and nominal flow rates at 10 PUR-1 SAR 13-3 Rev 0,6/08

kW of 1.65 cm/s. The only consideration for a loss of coolant flow (LOCF) scenario would be a blockage in a channel. The power density of PUR-1 at 10 kW is such that the individual plate power is very low, and conduction of the heat to adjacent channels would occur, and plate temperatures would remain well below ONS temperatures of 112°C, which is still more than 400°C below the safety limit for the fuel. Therefore this accident is not analyzed.

13.1.5 Mishandling or Malfunction of Fuel Fuel element maneuvers are always conducted in the reactor pool. They are removed from the core and moved into the storage space, one at a time, using a hand-held fuel handling tool. Annually a fuel element is removed from the pool for inspection. A fuel element weighs about 4.5 kg (9.9 Ib) in air and only about 2.8 kg (6.2 Ib) in water.

Therefore, even if a fuel element should fall from the handling tool during its transfer it is not heavy enough to cause any considerable damage. The most severe consequence likely to occur would be some denting of the end fittings since the fuel element, being an elongated object, would tend to fall in water in a rather upright position.

The PUR-I Standard Operating Procedures define administrative steps which are intended to prevent a fuel handling mishap. They are:

1. All fuel handling is done in accordance with written procedures.

2. Loading operations are done by qualified personnel under direct supervision 3.

of a Senior Operator.

It is concluded that no adverse consequences are to be expected to the health and safety of the public or the

~

staff from this type of accident.

13.1.6 Experiment Malfunction In this section an analysis is performed to assess the hazard associated with the failure of an experiment in which fissile material has been irradiated in the reactor. In the scenario of this accident it is assumed that a capsule containing irradiated fissile material breaks and a portion of the fission product inventory becomes airborne. The consequences of the release are analyzed for both the reactor staff and general- public.

Since the potential impact of this postulated accident is greater than in any other accident analyzed, the failure of a fueled experiment is designated as the maximum hypothetical accident of the PUR-I. This accident is analyzed in Section 13.2.1.

The flooding of the 12.7 cm drop tube when placed in position G6 as described in the SER 'NUREG-1283 resulted in a reactivity insertion of 0.246% 6k/k, which is below the insertion resulting from the failure of a moveable experiment examined in the conversion proposal. This accident scenario is within the envelope of the examined

accident cases, and therefore was not analyzed .

PUR-1 SAR 13-4 Rev 0,6/08

13.1.7 Loss of Normal Electrical Power The loss of normal electric power at PUR-1 will shutdown the reactor by simulating a scram with a loss of power to the electromagnets that hoist the shim-safety rods. This action will shutdown* the reactor from any conceivable operating condition. Since the reactor is cooled by natural convection, there are no shutdown and decay heat issues.

There is adequate heat capacity in the reactor pool to address shutdown and decay heat loads. Therefore, this accident scenario is not addressed.

13.1.8 External Events 13.1.8.1. Fire or Explosion .

The reactor building* is a steel frame structure with concrete block and brick construction. Additionally, and the reactor itself below the floor r room. e materia s surrounding the reactor core are concrete, steel and earth. There is limited combustible material in the reactor room, but it is virtually impossible to exclude all burnable materials. Therefore, some small fire possibility exists. There is portable fire-fighting equipment available, and PUR-1 staff are trained in use of that equipment. Purdue Fire Department is located on campus, and are available for assistance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day.

In general, personnel judgement is used in deciding the response to a a fire. If the fire is small and can be easily controlled, little action is necessary to* safely shutdown the reactor and control the fire. If the fire is of larger magnitude, the reactor will be secured and outside assistance obtained.

Because of the large volume of water surrounding the core, damage to the fuel is not likely even in a severe fire scenario. Even if the fire involved the control systems, power to those circuits could be cut at the wall of the reactor room, and the control rods would drop and safely shut down the reactor.

Explosions external to the building could affect the reactor and its systems, however explosions external to the reactor pool will have little effect on the reactor due to its location and the construction of the pool. The amount of water covering the core,will aid in the reduction of risk to pyrotechnic devices thrown into the pool, either serving to quench the devices, or absorb most of the energy from the detonation.

With these facts, and the location of the reactor in the pool, damage from fire or explosion is not likely.

13.1.8.2. Acts of Sabotage There is a security plan in place, and security systems designed to protect all of the controlled areas in the reactor facility. Purdue Police, a fully functioning police department with arrest authority, is available for assistance, and is staffed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day. The PUR-1 Security Plan is reviewed and approved by the U.S. NRC, and PUR-1 SAR 13-5 Rev 0,6/08

provides for in-house procedures to ensure facility protection .

Riots or acts of civil disobedience directed against the reactor facility would be recognizable is sufficient time to ensure safe shutdown of the reactor. Personnel at the*

reactor could quickly alert local law enforcement, and quickly come under their protection. Then access to the area could be restricted until it is safe to resume normal operations.

13.2 Accident Analysis and Determination of Consequences 13.2.1 Maximum Hypothetical Accident (Failure of a Fueled Experiment)

In this analysis the consequences of a failed experiment generating 1 W were studied.

The capsule containing the experiment is assumed to break as it is removed from the reactor. The fission products expected to become airborne are the noble gases and elemental iodine. Other fission products and actinides are not volatile at the temperature 9which is essentially at room temperature) at which the experiment would be performed. The amount of noble gases and radioiodine is assumed to be that specified in Ref. 1, i.e. 100 % of the noble gases and 25 % of the iodine inventory. If the experiment were to break in the reactor poor a credit for the absorption of iodine in water can be taken. However, this is not considered in this analysis.

A conservative assumption that the irradiation time was infinite was made in the analysis. There fore, the fission inventories used in the analysis for some long-lived radionuclides, e.g.Kr-85 or even 1-131, are overly conservative. Furthermore, it was assumed' that the fission products are instantaneously released and uniformly distributed in the Reactor Room air. The free volume of the Reactor Room is approximately 424 m3 .

The external dose rate (in mrem/hr) due to y and l3-radiation was calculated using the relationships given in Lewis 1 where X = radionl!clide concentration (Cilcm 3 ). Er = average v-energy per disintegration (MeV) and

. 11-DfJ = 9.43 x 10 x X x E!l (0.1 )

Where E fJ = the average l3-energy per disintegration (MeV). The dose rate, iJfJ represents the skin dose. The dose rate to the thyroid (in rem/hr) due to inhalati.on of the radioiodines is give by DT = DCFxBx X

where DCF =dose-conversion factor for the thyroid (rem/Ci), B =breathing rate (cm 3/hr)

PUR-1 SAR 13-6 Rev 0,6/08

and X = radioiodine concentration (Ci/cm 3 ). The standard breathing rate recommended is 1.25x1 06 cm 3/hr.

The calculated saturation activity for each respective radioisotope and its concentration is the Reactor Room after experiment failure is shown in Table 7.5 for an experiment of 1 W. Also shown in this table are the calculated dose rates for the whole-body, skin and thyroid. .

PUR-1 SAR 13-7 Rev 0,6/08

Isotope 1-131 Table 13-2: Dose Rates in the Reactor Room from a Failed Fuel Experiment As (Ci)

X (Ci/cm3) 2.66E-02 1.57E-11 E-gamma (MeV) 3.71 E-01 (Power = 1 Watt)

E-beta (MeV) 1.97E-01 DCF 1.00E+06 DR-gamma DR-beta DR-thyroid (rem/Ci) (mrem/hr) (mrem/hr) (rem/hr) 5.49E+OO 2.55E+OO 1.96E+01 1-132 3.75E-02 2.21 E-11 2.40E+OO 4.48E-01 6.60E+03 5.00E+01 8.16E+OO 1.82E-01 1-133 5.31 E-02 3.13E-11 4.77E-01 4.23E-01 1.80E+05 1.41 E+01 1.09E+01 7.04E+OO 1-134 5.94E-02 3.50E-11 1.94E+OO 4.55E-01 1.10E+03 6.41 E+01 1.31 E+01 4.82E-02 1-135 4.96E-02 2.77E-11 1.7BE+OO 3.0BE-01 4.40E+04 4.64E+01 7.02E+OO 1.52E+OO Kr-83m 5.90E-03 1.39E-11 2.60E-03 1.03E-02 3.41 E-02 1.18E-01 Kr-85m 1.27E-02 3.00E-11 1.51 E-01 2.23E-01 4.27E+OO 5.50E+OO Kr-85 2.53E-03 5.97E-12 2.11 E-03 2.13E-01 1.19E-02 1.10E+OO Kr-87 2.00E-02 4.27E-11 1.37E+OO 1.05E+OO 6.09E+01 4.08E+OO Kr-88 3.12E-02 7.36E-11 1.74E+OO 3.41 E-01 1.21 E+02 2.07E+01 Kr-89 3.96E-02 9.34E-11 1.60E+OO 1.33E+OO 1.41 E.+02 1.02E+02 Xe-131m 2.53E-04 5.97E-13 2.00E-02 1.40E-01 1.13E-02 6.88E-02 Xe-133m 1.35E-03 3.18E-12 3.26E-01 1.55E-01 9.79E-01 4.07E-01 Xe-133 5.4BE-02 1.29E-10 3.00E-02 1.46E-01 3.66E+OO 1.55E+01 Xe-135m 1.77E-02 4.17E-11 4.22E-01 9.74E-02 1.66E+01 3.35E+OO Xe-135 5.23E-02 1.23E-10 2.46E-01 3.22E-01 2.86E+01 3.27E+01 Xe-137 5.31 E-02 1.25E-10 1.50E-01 1.37E+OO 1.77E+01 .1.41 E+02 Xe-138 5.70E-02 1.31E-10 1.10E+OO 8.00E-01 1.36E+02 8.66E+01 7.11E+02 4.55E+02 2.84E+01 With a v-dose rate in the Reactor Room as high as 50 mrem/hr anyone of the radiation ara monitors would cause an automatic reactor shutdown and audible and visual alarms in the control room. From the past experience, it is know that the reactor building can be evacuated within 1.5 minutes. Therefore, it is assumed in the following analysis that the exposure time to the members of the reactor staff is 1.5 minutes. The resulting radiation doses are: whole-body 22.3 mrem, skin does 13A mrem, and the thyroid dose 1A3 mrem.

The radioactive material content, including fission products, of any singly encapsulated experiment should be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation will not result in doses in excess of 10% of the equivalent annual doses stated in 10 CFR 20. This dose limit applies to persons occupying (1) unrestricted areas continuously for two hours starting at time of release or (2) restricted areas during the length of time required to evacuate the restricted area.

The radioactive material content, including fission products, of any doubly encapsulated experiment or vented experiment should be limited so that the complete release of all gaseous, particulate, or volatile components from the encapsulation or confining boundary of the experiment could not result in (1) a dose to any person occupying an

unrestricted area continuously for a period of two hours starting at the time of release in excess of 0.5 Rem to the whole body or 1.5 Rem to the thyroid or (2) a dose to any PUR-1 SAR 13-8 Rev 0,6/08

person occupying a restricted area during the length of time required to evacuate the restricted area in excess of 5 Rem to the whole body or 30 Rem to the thyroid.

This radiation exposure approaches the limits established in the Technical specifications, Sec 3.5.f for a singly encapsulated experiment. This experiment corresponds to the irradiation of 1.1 gm of U-235 in the mid-plane of the isotope irradiation tube located in position F6.

For the radiation calculations outside of the reactor building it was assumed that all fission products released in the reactor building would leak out within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Since the reactor room does not have any windows and only a few doors and emergency procedures calls for turning off the air exhaust system, the assumption about the leak rate is considered "to be reasonable. Another conservative as'sumption was made in that no radioactive decay and hence no decrease in the source strength was taken into account while calculation the dose rates outside the reactor building. The radionuclide concentration at a distance of 100 m from the release point was calculated using the atmospheric dispersion faction recommended in Ref.. 1; x

Q' mt CJI'CJz Where X is the concentration of radioactive material (Cilm\ Q' =source rate (Cils), u =

average windspeed (m/s), cry = lateral plume spread (m) and crz =vertical plume spread (m).

In the above expression for the atmospheric dispersion factor no credit was taken for so-called building wake effects and horizontal plume meandering, both of which help in spreading the radioactive plume. Also, no credit was taken for the fact that air from the basement area is' exhausted at a minimum height of 50 feet. Using an average wind speed of 1 mls and Pasquill type F atmospheric conditions the dispersion factor at 100 m is calculated to be 1.78 x 10-4 s/m 3 . Actually, the average wind speed is about 3.4 m/s. Therefore, the results of this analysis are conservative at least by a facto~ of 3).

Table 13-3:Oose Rates at 100 meters.

DR- DR-gamma DR-beta thyroid Isotope Imrem/hr\ mrem/hr) (rem/hr) 1-131 1.95E-03 9.06E-04 6.9BE-03 1-132 1.7BE-02 2.91 E-03 6.49E-05 1-133 5.01 E-03 3.BBE-03 2.51 E-03 1-134 2.2BE-02 4.67E-03 1.71 E-05 1-135 1.65E-02 2.50E-03 5.41 E-04 Kr-83m 1.21 E-05 4.20E-05 Kr-85m 1.52E-03 1.96E-03

PUR-1 SAR Kr-85 " 4.23E-06 Kr-87 2.17E-02 13-9 3.90E-04 1.45E-02 Rev 0,6/08

Kr-88 Kr-89 Xe-131m Xe-133m Xe-133 4.30E-06 7.36E-03 5.02E-02 3.64E-02 4.01 E-06 2.45E-05 3.4BE-04 1.45E-04 1.30E-03 5.54E-03 Xe-135m 5.91 E-03 1.19E-03 Xe-135 1.02E-02 1.17E-03 Xe-137 6.31 E-03 5.03E-02 Xe-138 4.B5E-02 3.0BE-02 2.10E-01 1.65E-01 1.01 E-02 Calculated dose rates at 100 m for an experiment power of 1 Ware shown in Table 7.6.

If it is assumed that an individual is located at this point for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following the fission product release from a postulated experiment failure then his/her resulting radiation does to the whole body would be .51 mrem and to the thyroid .02 rem. These doses are only fractions (about 1%) of those which are referred to in 10 CFR 100 in conjunction with the determination of an exclusion area.

It is concluded that experiments using fissile material can be irradiated at the PUR-I within the power limits.

13.2.2 Insertion of Maximum Allowed Excess Reactivity The analyses of this transient utilizes the reactor physics and reactivity coefficients determined by MCNP5 as described in Chapter 4, and thermal-hydraulic parameters determined by NATCON as described in Chapter 4, and the PARET/ANL 2 code. For this accident, the initial power before the transient was 10 kW.

The original PARET code has been adapted by the Reduced Enrichment for Research and Test Reactors (RERTR) Program to provide transient and thermal-hydraulics analysis for research and test reactors with both plate and pin-type' fuel assemblies.

The PARET/ANL version of the code has been subjected to extensive comparisons with the SPERT I and SPERT II (light and heavy water) experiments. These comparisons were quite favorable for a wide range of transients up to and including melting of the clad. Revisions of the code include new and more appropriate heat transfer, departure from nucleate boiling (ONB) and flow instability correlations, improved edits, reactor trips, control insertion model, a decay heat power model, and a loss of flow model.

\

The rapid insertion of the maximum worth of reactivity excess for PUR-1 (0.6% Llk/k) as specified by the Technical Specifications was evaluated. An assumption was made that the period trip (7s) failed, and scram was initiated on the power trip at 12 kW. Since the assumed measurement uncertainty on core power is 50%, a core power trip setting of 18 kW was utilized in the accident calculation. A delay of 0.1 seconds from the sending of scram signal to beginning of control rod motion is assumed. Fuel/coolant channels that are representative of the hottest fuel plates (as identified in Chapter 4) were PUR-1 SAR 13-10 Rev 0,6/08

modeled in the PARET analyses .

Results of this transient are summarized in Table 13-4: Peak power and clad temperature for insertions of 0.6% LlkJk with scram .. The reactor power increases fro m 1 kW to the tri p setting of 1.8 kW in less than 2.5 seconds. There is a negligible increase in the clad temperature as a result of this hypothetical accident. The safety limit is never in danger of being reached .

Table 13-4: Peak power and clad temperature for insertions of 0.6% IIk/k with scram.

T la mU °el Time of p.

~V/I Peak al at Peak SCRAM Reactivitv Inserted (kWI Power (s) t=O Power Maximum YES 0. 6% h.klk slep 10 40.3 0.187 49.94 49.73 49 .94 YES 0.6% 6.klk over 10 5 10 18.4 7.83 49.94 42 .30 49 .98 These results demonstrate th e ability of the LSSS to protect the safety limit of fu el temperatures not to exceed 530"C. The maximum temperatures achieved in the fu el are well below temperature of incipient boiling as well. Therefore PUR-1 can maintain the fuel integrity during this accident scenario .

Power and Clad Temperature Profile for 0.6% Ak/k Step Insertion with Scram

PUR-1 SAR

~ POWE RMW

==

- TCladC I Figure 13-1 : Power and Clad Temperatures for O.6%.llklk step insertion with scram .

13- 11 Re v 0, 6/08

Power and Clad Temperature Profile for 0.6% 6.k/k Slow Insertion with Scram 0.020 '"~~~~~~~~~~~~~~~~~~~~~~~~, 60 0.018 50 0.016 0 .014 40 "m

~ 0.0 12 *

E.

g

... 0.0 10 L-~ 30 ~

.~ 0,008 0.006 20 "a

0.004 10 0,002 0.000 "------~~~~~~~--~~_~---~--' 0 o 5 10 15 20 25 30 35 40 l- POWE R Time (5) r.::1W - Tclad C Figure 13-2:Power and Clad Temperatures for 0.6%1Ik1k step insertion with scram.

13. 2.2.1 . Insertion Without Scram While it is an extremely unlikely event that all reactor protective systems will fail , the analysis of the step and slow insertion of the maximum excess reactivity without scram was analyzed. In both cases it is assumed that both shim-safety rods are completely removed from the core, and are not available for insertion to shut the reactor down . For this accident, the initial power before the transient was 10 kW .

The reactivity coefficients determined in Chapter 4 act in this case to keep the reactor power from increasing without bounds. The power in both cases l step and slow insertion) levels off at about 2.38 MW , with the maximum clad temperature reaching 133.08°C. This temperature is still well below the safety limit of 530°C, so there is no danger of failure of the clad . It should also be noted that emergency boration is available to the operators, and this wou ld act to shut down the reactor in the unlikely case of failure of the reactor protection system. The peak power and clad temperatures are summarized in Table 13-5, and the power and clad temperature history for the

accident are shown in Figure 13-3 and Figure 13-4 PUR-1 SAR 13-12 Rev 0, 6/08

Table 13-5: Peak power and clad temperature for trip and no-trip insertions of 0.6%

Llk/k.

Time of T ladmn °C)

P,

~wi Peak al alPeak SCRAM Reactivity Inserted IkW) Power (5) 1=0 Power Maximum NO 0.6% h.kJk step 10 2388 672 49.94 133.08 133.08 NO 0 .6% ilklk over 10 s 10 2388 680 49.94 133.08 133.08 Power and Clad Temperature Profile for 0.6% Llk/k Step Insertion w/o Scram 3.00 140 2.50 120 100 .-.

2,00 U 0>

BO

!< *

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0-

" 1.50 E

~

0 60 *

~

1.00 40 "U 0.50 20 0.00 0 0 50 100 150 200 250 300 350 400 450 500 L J Time (s)

I- POWERMW - Tda<@]

Figure 13-3: Power and Clad Temperatures for O.6%D.k/k step insertion without scram .

PUR-1 SAR 13-13 Rev 0 , 6/08

Po.......er and Clad Temperature Profile for 0.6% l\k/k Slow Insertion w/o Scram 1

3.00 r~;::::======================::r 140 2.50 I 120 100 _

2.00 u

§" g>

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~ Q.

~ 1.50 ]

60 ~

Q.

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1.00 u 40 0.50 20 0.00 J.L. _ _ _ _ _ __ ~-------------------' 0 o 100 200 300 400 500 600 Time (s)

F - POWE RMW - Tciad C I Figure 13-4: Power and Clad Temperatures for O.6%.6.k/k slow insertion without scram .

J 13.3 Summary and Conclusions Evaluation of the accident scenarios detailed in this chapter lead to the concl usions that the PUR-1 reactor can operate safely and effectively at 10 kW co ntin uously. with the reactor limiting safety system settings of 110% (11 kW) and 120% (12 kW ) adequately protecting the reactor and its systems. Even in the very unlikely event of a complete failure of the reactor protection system and insertion of the maximum allowable reactivity, fuel integrity is maintained and there is little or no risk to the public.

PUR-1 SAR 13-14 Rev 0, 6/08

13.4 References 1 Lewis, E.E., "Nuclear Power Reactor Safety", John Wiley and Sons, Inc, (1977).

W. L. Woodruff and R. S. Smith, "A Users Guide for the ANL Version of the PARET Code, PARET/ANL (2001 Rev.)," Argonne National Laboratory (ANL),

ANLIRERTRITM-16, March 2001 .

PUR-1 SAR 13-15 Rev 0,6/08

14 TECHNICAL SPECIFICATIONS 14.1 Summary Description of the Document The NRC requires each applicant for and holder of a license to operate a non-power reactor to develop technical specifications that state the limits, operating conditions and other requirements for facility operation to protect the environment and health and safety of the facility staff and the public in accordance with 10 CFR 50.36. The technical specifications are typically derived from the facility descriptions and safety considerations contained in the SAR and represent a comprehensive envelope of safe operation.

The technical specifications for PUR-t were first generated in 1978 to satisfy these requirements. They were made before the publication of the ANSI Standard 15.1, but have been modified over time to meet the evolving format recommendations of that standard.

The technical specifications of the PUR-1 are maintained as a separate document from this safety analysis report.

14.2 Administrative Control of Technical Specifications The Committee on Reactor Operations (CORD) shall review any proposed changes to the Technical Specifications or licenses.

PUR-1 SAR 14-1 Rev 0,6/08

15 FINANCIAL QUALIFICATIONS 15.1 Financial Ability to Operate a Non-Power Reactor Since the facility is an existing one, the capital costs are low. The operational costs are minimal while the benefits in the education process of nuclear engineering students, radiation health scientists and technicians are great, both to the individual people and to the national, interests. No reasonable alternatives exist to the wide versatility of research/training reactors such as the PUR-1 in contributing to the education and scientific knowledge.

Purdue University, as a large research institution, has the capability to continue safe operation of PUR-1 for the foreseeable future ..

15.2 Financial Ability to Decommission the Facility Purdue University will provide a letter of intent to show it will meet the financial obligations of decommissioning the facility .

PUR-1 SAR 15-2 Rev 0, 6/08

APPENDIX 1: PUR-1 DRAWINGS

PUR-1 SAR Appendix 1-1 . Rev 0,6/08

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APPENDIX 2: NATCON INFORMATION A 2.1 Hot Channel Factors in the NATCON Code Version 1.0 The NATCON code version 1.0 [Ref. ANLIRERTRITM-12]* uses three hot channel factors (FQ, FW, FH). Using the source code and documentation, the factor FH used in NATCON is found to be the same as the factor FNUSL T used by E. E. F-eldman. Table 1 shows the tolerances and uncertainties included il") each of the six hot channel factors used by E. E. Feldman. The correspondence between the NATCON hot channel factors and E. E. Feldman's six hot channel factors is as follows.

Feldman's Hot Channel Factor NCATCON Input Variable System-wide Factors:

FFLOW . a factor to account for the uncertainty in total reactor flow FW (approximately)

FPOWER a factor to account for the uncertainty in total reactor FQ power FNUSLT a factor to account for the uncertainty in Nu number FH correla*tion

Local Factors:

FBULK a hot channel factor for local bulk coolant temperature rise FBULK (new input) f FFILM* a hot channel factor for local temperature rise across the FFILM (new input) coolant film FFLUX a hot channel factor for local heat flux from cladding FFLUX (new input) surface Hot Channel Factors in the NATCON Code Version 2.0 Sections 2.1 and 2.2 develop, for laminar natural convection, two thermal-hydraulic relationships that are used in section 2.3 to obtain formulas for the hot channel factors from user-supplied manufacturing. tolerances and measurement. uncertainties. The results of section 2.3 are summarized here for convenience .. The first three are local/random hot channel factors, and the last three are system-wide. An example of the use of these hot channel factors is given in section 4, with NATCON running instructions in section 3, and the new input description in section 5 .

PUR-1 SAR Appendix 2-1 Rev 0,6/08

FBULK is higher (conservative) if the temperature depen~ence of water viscosity is ignored.

FQ =1 + U7 FW =1 + Us FH =1 + Ug where U, =Fractional uncertainty in neutronics calculation of power in a plate U2 =Fractional uncertainty in U-235 mass per plate = /).m / M

U3 U4 U5

=Fractional uncertainty in local (at an axial position) fuel meat thickness

=Fractional uncertainty in U-235 local (at an axial position) homogeneity

= Fractional uncertainty in coolant channel thickness = (tnc - t hc ) / tnc U6 =Fractional uncertainty in flow distribution among channels U7 =Fractional uncertainty in reactor power measurement Us =Fractional uncertainty in flow due to uncertainty in friction factor Ug = Fractional uncertainty in convective heat transfer coefficient, or in the Nu number correlation' M = Nominal mass of U-235 per plate, gram

/).m =Tolerance allowed in U-235 mass per plate, gram The code obtains, for an input nominal reactor power CPWR, a thermal-hydraulic solution using the three systematic hot channel factors FW, FQ and FH. If the user-input reactor power is zero, then the code itself chooses the nominal power from a series of power levels (10 kW, 100 kW, 200 kW, and so on increasing in steps of 100 kW). This thermal-hydraulic calculation is done for a hot plate power of CPWR*FQ*(Radial power peaking factor RPEAK)/(Total number of fuel plates in standard and control assemblies). Also, the frictional resistance to flow is multiplied by FW 2 , and the convective heat transfer coefficient found for laminar flow in a rectangular channel is PUR-1 SAR Appendix 2-2 Rev 0,6/08

divided by FH. The random hot channel factors FBULK, FFILM and FFLUX are not used in this solution.

Having obtained the above solution, the random hot channel factors FBULK, FFILM and FFLUX are applied to the temperatures obtained, using the following equations. The temperatures calculated with all six hot channel factors are printed after the above solution. The onset of nucleate boiling ratio, ONBR, is computed using the temperatures with all six hot channel factors applied (using the equation below). If the user-input nominal power is zero, then the last nominal power for which the code prints a solution is that at which the ONBR is 1.0.

T;,6hcf =To + (T; - To)*FBULK Twall,;,6hcf =T;,6hcf + (T wall,; - T;)*FFILM T max,;,6hcf =Twall,;,6hcf + (T max,; - T wall,;)*FFLUX ON BR =

(r .-Yo)

~_I--,I/CJ,--,.I_---,-""7'"

(T.",all.i.6/iC.t - To )

where

=Bulk water temperature at the coolant channel inlet, i.e., the pool temperature, °C To T; = Bulk water temperature in node i of the channel with only systematic hot channel factors applied, °C I

T wall,; = Cladding surface temperature in node i with only systematic hot channel factors applied,oC Tmax,; = Fuel meat centerline temperature in node i with only systematic hot channel factors applied,oC .

T;,6hCf =Bulk water temperature in node i of the channel with all six hot channel factors, °C Twall,;,6hcf, =Cladding surface temperature in node i with all six hot channel factors, °C T max,;,6hcf = Fuel meat centerline temperature in node i with all six hot channel factors, C T;ncp,; = Incipient boiling temperature in node i with only systematic hot channel factors applied, C

Flow Rate in a Coolant Channel versus Power of a Fuel Plate NATCON is a laminar natural circulation code. The flow rate is calculated in the code by balancing the buoyancy pressure force to the laminar friction pressure drop. Following this concept, an analytical relationship is developed here (with some approximation) for the coolant flow rate in a single coolant channel in terms of the power generated in a

fuel plate and the channel geometrical dimensions. The analytical relationship is needed for obtaining hot channel factors.

PUR-1 SAR Appendix 2-3 Rev 0,6/08

The hot channe'l factor FW used in the code to account for the uncertainty in coolant flow rate is actually applied to the laminar friction factor in the code, that is, the laminar friction factor is multiplied by FW2. It is not applied directly to the flow rate. The relationship developed here explains how this technique works.

PI ,T 1 at channel outlet L = Channel height containing hot coolant (hotter than pool), m P = Power in a single fuel plate or the two half plates, W W=Upward flow rate in a single channel, kg/s Po ' To at channel inlet

Schematic of what the code analyses, that is, a single rectangular coolant channel heated by a half of a fuel plate on each side (right and left sides).

The above schematic shows what the code analyses, that is, a single rectangular coolant channel heated by a half of a fuel plate on each side (right and left sides). See Fig. 1 for details. The buoyancy pressure force is caused by the decrease in water density due to heating in the channel. The temperature dependence of water density can be written as P (T) = Po - Po P (T - To) (1 )

where T1 = Bulk water temperature at channel outlet, C

!1T = T 1- To = Temperature rise in channel from inlet to outlet, C 3

Po = Water density at channel inlet, i.e., the water density in the pool, kg/m P = Volumetric expansion coefficient of water, per C 3

P = Average coolant density in the channel, kg/m PUR-1 SAR ' Appendix 2-4 Rev 0,6/08

L g

= Channel height that contains hotter coolant (hotter than pool), m. It is the sum of heat generating length of fuel plate, non-heat generating fuel plate length at top, and the assembly duct length above the top of fuel plate

=Acceleration due to gravity, 9.8 m/s2 The buoyancy pressure force is given by Buoyancyi1p =(Po - p)g L (2)

The average coolant density p is given by (3)

Buoyancy i1 p =0.5 Po /3 i1T gL (4)

The coolanttemperature rise i1T can be written in terms of the input power P generated in a fuel plate, as shown by Eq. (5) below, and then the buoyancyi1p of Eq. (4) can be written in terms of the input power P, as shown by Eq. (6).

i1T =PI (W Cp ) (5)

Buoyancy i1 p =

Po/3gLP 2WCp Ignoring the minor losses at channel inlet and outlet, the laminar frictional pressure drop (6) in the channel is written below as Eq. (9) after using the laminar friction factor given by Eq. (7), and after replacing the coolant velocity by mass flow rate using Eq. (8). The parameter C in Eq. (7) is a constant for a given channel cross section, but it depends upon the channel cross section aspect ratio width/thickness, and varies from 57 for aspect ratio 1.0 (square channel) to 96 for an infinite aspect ratio (infinitely wide channel).

L=C/Re (7)

W= pAV (8)

Frictional i1 p = PfLc V2 = C j.lLc W (9) 2D 2pAD2 where

PUR-1 SAR f = Moody friCtion factor for laminar flow in the channel Appendix 2-5 Rev 0,6/08

Re A

0

= Reynolds number in the channel = p VOl Ji

= Flow area of the channel cross.section, m 2

=Equivalent hydraulic diameter of the channel cross section, m' Lc =Total coolant channel length causing frictional pressure drop, m.

V =Coolant velocity averaged over the channel cross section, m/s W =Coolant mass flow rate in the channel,. kg/s Ji = Average coolant dynamic viscosity in the channel~ N-s/m 2

. Ji (T) =Temperature-dependent dynamic viscosity of water, N-s/m 2 Jio =Ji (To) = Coolant dynamic viscosity at the channel inlet temperature To For the PUR-1 reactor, the temperature dependence of the dynamic viscosity of water over the temperature range 27 °e S; T S; 50 0 e can be approximated as follows.

Ji(T)= Ji(To)(1+T-To)-a (10)

where a = 0.12, To = 27 °e, Ji ( To) =O.875x1 0 -3 N-s/m 2 The average coolant dynamic viscosity ~ used in Eq. (9) can be set equal to the viscosity at the average coolant temperature (To + 0.5~ T) in the channel. Putting this temperature in Eq. (10), the average viscosity ~ is found to be Ji = Ji (To) (1+0.5~T)-a . (11 )

Equation (11 ) indicates that the average viscosity ~ can be set equal to Ji ( To) if ~ T is just a few °e (this is the case for the PUR-1 reactor at the operating power of 1 kW). If

~Tis greater than a few °e, i.e., 1 << O.5~T (this is the case for the PUR-1 reactor at an ONB power of about 100 kW), then Eq. (11) simplifies to the following.

if ~T>> 2 °C (12a) if ~T<< 2 °C (12b)

Substituting Eq. (12a) into Eq. (9), the frictional ~p becomes PUR-1 SAR Appendix 2-6 Rev 0,6/08

(13)

Equating the frictional L1 p of Eq. (13) to the buoyal)cy L1 p of Eq. (6) to find the steady-state coolant flow rate W in the channel, one obtains Eq. (14) below. Equation (14) can, be rewritten as Eq. (15).

(14) 2+a . Po PAD2 f3 gLpl+a, W - (15) 2 a C f.1o LC C pl +a .

Equation (15) relates the fuel plate power to the channel flow rate in natural circulation.

It is used to find the dependence of the flow rate on the. parameter C in the laminar friction factor (at constant power). All parameters in this equation are constant (p is also practically constant) except the parameter C in the laminar friction factor. Based on Eq.(15), the relationship between the flow rate Wand the parameter C is given by Eq.

(16) below. .

I Woc ( -1 )2+a (16)

C Equation (16) shows that the friction factor parameter C is multiplied by a factor (FW)2, 2

the coolant flow rate W will pe reduced by the factor (FW)2+a . This has been ,verified by actually running the NATCON code for the PUR-1 reactor. Since 0 is small (0 = 0.12 for the PUR-1 reactor), 2/(2+0) is nearly 1.0, and the flow rate Wis reduced approximately by the factor FW.

Bulk Coolant Temperature Rise versus Power of a Fuel Plate Equation (5) expresses, for laminar natural circulation, the bulk coolant temperature rise in terms of fuel plate power, coolant flow rate and specific heat Puttin'g the value of flqw rate obtained in Eq. (15) into Eq. (5), the bulk coolant temperature rise is given by Eq.

(17) below, purely in terms of power and the geometrical dimensions of the channel.

The right hand side of Eq. (17) is rearranged into two factors in Eq. (18), such that the second factor is sensitive to power and channel geometrical dimensions that usually have manufacturing tolerances and measurement uncertainties, and the first factor is insensitive to power and channel geometrical dimensions.

PUR-1 SAR Appendix 2-7 Rev 0,6/08

(17)

(18)

The nominal flow area and hydraulic diameter of a rectangular coolant channel are given by A =tne w nc (19)

(20)

(21 )

where te =Channel thickness (spacing between fuel plates), m tne = Nominal channel thickness (spacing between fuel plates), m the =Minimum channel thickness in hot ch~nnel (spacing 'between fuel plates), m We =Channel width, assumed not to change from its nominal value, m Pw =Wetted perimeter of the nominal channel, m Pne =Power generated in a fuel plate, without applying manufacturing tolerances, W Phe =Power generated in a fuel plate, after applying manufacturing tolerances, W Because the channel thickness te is much smaller than the channel width We in most experimental reactors, Eq. (21) reduces to (22)

Using the channel area and hydraulic diameter given by Eqs. (19) and (22) into Eq.

(18), the bulk coolant temperature rise can be written in terms of power, channel thickness, and channel width. This is the desired relationship for use in finding hot channel factors..

(23) . "

PUR-1 SAR Appendix 2-8 Rev 0,6/08

Formulas for Hot Cha'nnel Factors For use in the NATCON version 2.0, six hot channel factors (three global/systemic and three local/random) are obtained from 9 manufacturing tolerances and measurement uncertainties U1, U2, .... Ug that are defined below. These are fractional uncertainties rather than percent. Of these nine uncertainties, those affecting a particular hot channel factor are indicated in Table 1. The systemic hot channel factors are given by Eqs. (24) through (26), and the random hot channel factors are given by Eqs. (27) through (29). A utility Fortran computer program NATCON_HCF and a Microsoft spreadsheet NATCON.HotChanFactors.xls have also been developed to compute the hot channel factors using these formulas.

FQ =1 + U7 (24)

FW =1 + Us (25)

FH =1 + Ug (26)

The ratio of the power generated in hot plate to its nominal power, caused by the uncertainties in neutronics-computed power and in U-235 mass per plate, can be written as (27)

The ratio of bulk coolant temperature rise in hot channel to the temperature rise in the nominal channel, caused by the uncertainties in neutronics-computed power, U-235 mass per plate, and channel thickness, is obtained from Eq. (23). Only the quantity in the second parentheses is important here because the quantity in the first parentheses is insensitive to these uncertainties.

1 3 3 tlT. (P J2+a

_h_c = --.!!.£.

(t

...1!£.

J2+a

= (1 + U 1) 2+a (1 + U 2 ) 2+a

_I. _I (

1 J2+a*,

(28) tlTnc p"c thc 1- U5 The uncertainty in flow distribution is assumed to reduce the channel flow to (1- U6) times the flow without this uncertainty, and therefore the bulk coolant temperature rise is increased by the factor (1 + U6). This uncertainty in bulk coolant temperature rise is statistically combined with that given by Eq. (28) to obtain the following formula for the hot channel factor FBULK for input to the NATCON version 2.0.

_I _I 1 3 }2 2+a ?

FBULK=1+ (I+U,)""(I+U')"(I_U,] -I +U6- (29)

{

The temperature drop across coolant film on the cladding surface at an axial location is PUR-1 SAR Appendix 2-9 Rev 0,6/08

given by Eq. (30). Here the heat flux q" (W/m2) on the cladding surface is replaced by tf

q"'/2 in terms of the volumetric power density q"'(W/m 3) in the fuel meat.

q" tl q'"

I1T =-=-- (30) film h 2h The convective heat transfer coefficient h (W/m2-C) is given by Eq. (31). Here the laminar Nusselt number Nu is independent of flow rate, and varies only slowly with the aspect ratio (width/thickness) of coolant channel. The main variation of the heat transfer coefficient with channel thickness is due to the denominator of Eq. (31). The numerator of Eq; (31) is considered to be constant.

h = N"Kcool = N"K cool (31 )

D 2 tc Using Eq. (31) for the heat transfer coefficient, the temperature drop across coolant film can be written as Eq. (32).

q '" t. t I1T - f c (32) film- N K

" cool Equation (32) states that I1Tfil11l is directly proportional to the fuel meat thickness (having uncertainty U3), the channel thickness (having uncertainty U5), and the power density in meat. The uncertainty in power density is caused by three uncertainties, that is, U1, U2 and U4. Statistically combining these five uncertainties gives the following formula for the hot channel factor FFILM for input to the NATCON version 2.0.

(33)

The uncertainty in the heat flux at the cladding surface is included in the hot channel factor FFILM given by Eq. (33). A hot channel factor FFLUX for the heat flux alone can be found from Eq. (34) for heat flux in terms of the power density q'" in the fuel meat and the thickness of the meat. The fractional uncertainty in heat flux is the sum of fractional uncertainties in power density and meat thickness, as given by Eq. (35).

q,,= _ q"'t.

_f (34) 2 Sq" = Sq'" + Stlile! (35) q" q'" t filel In Eq. (35), the uncertainty in power density is caused by three uncertainties, that is, U1, U2 and U4. The uncertainty in the meat thickness is given by U3. Statistically combining these four uncertainties gives the following formula for the hot channel factor FFLUX for

input to .the NATCON version 2.0 .

PUR-1 SAR Appendix 2-10 Rev 0, 6/08

The uncertainty in the temperature drop ~ Tmetal from fuel meat centerline to cladding surface is not important in the case of the PUR-1 reactor because ~ Tmetal is very small compared to ~Tfilm. For example, ~T metal is 0.05 °C and ~Tfilm is 34.5 °C at 100 kW (36) without any hot channel factors.

Table 1. Uncertainties Included in the Six Hot Channel Factors Used in NATCON Version 2.0 (X implies that an uncertainty affects a hot channel factor)

Uncertainty Fraction FQ FW FH . FBULK FFILM FFLUX Local or random uncertainties 1 Neutronics calculation of X X X power in ;:t plate, U1 2 U-235 mass per plate, U2 X X X 3 Local fuel meat thickness, U3 X X 4 U-235 axial homogeneity, U4 X X 5 Coolant channel thickness, X X U5 6 Flow distribution among X X channels, U6 System-wide uncertainties 7 Reactor power measurement X

uncertainty, U7

8. Flow uncertainty due to uncertainty in friction factor, X Us 9 Heat transfer coefficient uncertainty X

due to uncertainty in Nu numb~r correlation, Ug The following information was presented as answers to Requests for Additional Information (RAls) during the conversion process. This information is not presented in the SAR chapters on thermal hydraulics. .

PUR-1 SAR Appendix 2-11 Rev 0,6/08

Question 28

28. Appendix 1. From the information in Appendix 1 it is not clear how insignificant are the channel inlet and outlet losses when compared to the wall shear. Please clarify.

Response

The information in Appendix 1 was used only to obtain hot channel factors for input to a more detailed thermal-hydraulic calculation using the NATCON code [Ref. 8 of the conversion proposal]. Therefore, Appendix 1 is.a simplified modeling of what is calculated in detail in NATCON, and it is used only for the purpose of obtaining closed-form equations from which hot channel factors could be found. Appendix 1 does not include the minor losses. The minor losses calculated by NATCON are reported below, and found to be about 16% of the total frictional pressure drop in the HEU core, and 14% of the total frictional pressure drop in the LEU core (see Table 027-1).

The pressure drop due to inlet and outlet losses were calculated (by the NATCON code) using loss coefficients of 0.5 and 1.0 respectively. The pressure drop due to wall shear along the channel length is found by summing the pressure drop for each axial mesh which is calqulated using temperature-dependent coolant viscosity and density for the axial mesh (14 mesh intervals were used over the channel length in all calculations).

The pressure drops are calculated by NATCON 'assuming fully developed laminar flow in a rectangular cross-section channel, and then multiplied by a factor FW2 (FW squared) where FW is an input which may be used to account for the increased pressure drop due to hydrodynamically developing laminar flow. In the calculations presented in the conversion proposal, FW was always set to 1.0, and thus the increased pressure drop due to developing laminar flow was not included. It is included in the calculations presented here (Table 027-1). The me~hod used is described below.

For the most limiting fuel plate in Table 4-27 of the conversion proposal for each core (HEU and LEU), a comparison of the pressure drops due to inlet plus outlet loss and wall shear, with and without the effect of developing laminar flow, are tabulated in Table 027-1.

NATCON calculates the Darcy-Weisbach friction factor f = C/Re for laminar flow, using a built-in table of the parameter C for different aspect ratios of the rectangular channel cross section (values of parameter C are given in the response to Ouestion number 29).

An apparent value of the parameter C averaged over the channel length, called Capp ,

was calculated using Eq. (576) of Shah and London [Ref. 2 listed at the end of all responses] to account for the increased pressure drop due to hydrodynamically developing laminar flow in the channel. The ratio Capp/C was found to be 1.1105 for the 207 mil HEU channel, 1.0985 for the 197 mil LEU channel. Since the NATCON code multiplies the fully developed friction factor by FW2 as mentioned above, the input FW equals 1.054 and 1.048 for the HEU and LEU channels respectively. NATCON calculations were done using these values of FW, and the pressure drops due to inlet

plus outlet loss and wall shear are compared in Table 027-1 (column,S for the HEU channel, and column F for the LEU channel).

PUR-1 SAR Appendix 2-12 Rev 0,6/08

Table 027-1 shows that the pressure drops due to wall shear and minor losses are 84%

and 16%, respectively, of the total pressure drop in the HEU channel at its ONB power; and the pressure drops due to wall shear and inlet plus outlet loss are 86% and 14%,

respectively, of the total pressure drop in .the LEU channel at its ONB power.

Question 29

29. Appendix 1. From the information in Appendix 1 it is not clear what is the I

functional. dependency of the laminar friction parameter C to the channel cross-section dimensions. Provide a reference for the evaluation of C.

Response

The following values (rows 1 and 2 of Table 029-1) of the parameter C for fully developed laminar flow in a channel of rectangular cross section versus the width-to-thickness aspect ratio (welte) of the .channel are used in the NATCON code that was used in the thermal-hydraulics calculations. The table starts from the square cross section (aspect ratio = 1.0) and goes to the infinite value of the aspect ratio (parallel plates). In order to find the parameter C for the aspect ratio of the PUR-1 reactor, the NATCON code simply interpolates between the tabulated values. The original author of the code obtained these values from an old Reference [E. R. G. Eckert and T. F. Irvine, Heat Transfer Laboratory, University of Minnesota (1957)] but thes.e values are also given in a textbook 'by Frank Incropera [Ref. 3]. These values are obtained from the closed-form analytical solution for the fully developed laminar velocity distribution in a rectangular channel summarize,d by R. K. Shah and A. L.' London [Ref. 2]. Equation (341) in [Ref. 2] is a fitted equation to easily find the parameter C. It should be noted that the aspect ratio used in [Ref. 2] is channel thickness-to-width ratio (the reciprocal of that used in NATCON and shown below in Table 029-1), and the friction factor in [Ref.

2] should be multiplied by 4 to get the Oarcy-Weisbach friction factor that is used in NATCON and tabulated below.

Table Q29-1. Friction Parameter C Used in the NATCON Code Welt:, 1.0 2.0 3.0 4.0 5.0 6.3 8.0 11.0 15.0 18.0 100.0 C in 58.0 6~.0 69.0 72.5 77.0 80.0 83.0 85.0 88.0 89.0 96.0 NATCON C.in 57.0 62.0 '69.0 73.0 82.0 96.0 Ref. 3 C in 56.9 62.2 68.4 72.9 76.3 79.5 82.4 85.6 88.1 89.3 94.7 Ref. 2

PUR-1 SAR Appendix 2-13 Rev 0, 6/08

Question 30

30. Appendix 1. From the information in Appendix 1 in both the calculation of the channel flow and the calculation of the bulk coolant temperature rise the ratio of the coolant kinematic viscosity to density (!-lIp) was assumed to be insensitive to temperature. Please demonstrate the validity of this assumption.

Response

The information in Appendix 1 was used only to obtain hot channel factors for input to a more detailed thermal-hydraulic calculation using the NATCON code [Ref. 8 of the conversion proposal]. NATCON does account for the temperature dependence of coolant viscosity and density in the calculation of the channel flow and the calculation of the bulk coolant temperature rise. Therefore, Appendix 1 is a simplifiep modeling of what is calculated in detail in NATCON, for the purpose of obtaining closed-form equations from which hot channel factors could be found.

As suggested in the question, water viscosity is temperature-dependent, i.e., it decreases with rising temperature. Appendix 1 was revised to account for the effect of temperature dependence of viscosity on hot channel factors, and the revised Appendix 1 is enclosed herewith. The temperature dependence of the dynamic viscosity of water over the temperature range 27°C :5 T :5 50°C (adequate for the PUR-1 reactor) can be written as follows .

(A1 )

where 0=0.12 To = 27°C = Pool temperature of PUR-1 J..l ( To) = 0.875x1 0 -3 N-s/m 2 J..l (T) = Temperature-dependent dynamic viscosity of water, N-s/m 2 As shown in the revised Appendix 1, the revised relationship between the flow rate W in a channel and the friction parameter C is given by Eq. (A2). The revised formula for hot channel factor FBULK for bulk coolant temperature rise is given by Eq. (A3).

I WOC -1 )2+a (A2)

(C (A3)

The exponent on the right hand side of Eq. (A2) changed from 0.5 (in the conversion PUR-1 SAR Appendix 2-14 Rev 0, 6/08

proposal ignoring temperature dependence of 1-1) to the revised value 1/2.12 = 0.4717 .

There exponents in Eq. (A3) for FBULK also changed, e.g., from 3/2 to 3/2.12 = 1.415.

As a result of this revision, the hot channel factor FBULK decreased from 1.312 (in the conversion proposal) to 1.301 for the most limiting fuel plate 262 in the HEU core.

Similarly, FBULK decreased from 1.321 (in the conversion proposal) to 1.308 for the most limiting fuel plate 1348 in the LEU core. The effect of ignoring the temperature dependence of viscosity is conservative.

NATCON calculations were done with these revised values of FBULK along with a value of FW > 1.0 to account for the increased friction due to developing laminar flow (in response to Question number 33). The results are shown in Table Q27-1 (column C for the HEU core, and column G for the LEU core).

As a consequence of the two effects (i.e., increased friction due to developing laminar flow and the temperature dependence of viscosity) on hot channel factors FW and FBULK, the ONB power of the HEU core changes from 76.3 kW (reported in the conversion proposal) to 75.9 kW, and the ONB power of the LEU core changes from 96.1 kW (reported in the conversion proposal) to 95.8 kW. The effect is small for the PUR-1 reactor.

Question 32

32. Appendix 1. Equation (30) has two terms and the conversion proposal states that the .expression within the parenthesis on the right hand side of the equation varies slowly compared to the heat flux tfuel q"'/2. Demonstrate the validity of the statement with reference to the PUR-1 fuel plate.

Response

Equation (30)' of Appendix 1 is for finding a hot channel factor for the temperature drop from the meat mid-plane to cladding surface (~T metal). This temperature drop is very small compared to the temperature drop from the cladding surface to bulk coolant

, (~Tfilm). For example, in the PUR-1 HEU fuel plate 262 without hot channel factors,

~ T metal is 0.07 °C and ~ Tfilm is 46.98 °C (at meat mid-height) at a high power of 100 kW.

Similarly, in the PUR-1 LEU fuel plate 1348 without hot channel factors, ~T metal is 0.05

°C and ~ Tfilm is 34.5 °C at a power of 100 kW. Therefore, the hot channel factor for

~ T metal is not important for PUR-1. The important hot channel factor is the factor FFILM for ~Tfilm. In the case of PUR-1, ~Tfilm is the bigger component (bigger than the bulk coolant temperature rise) in the total temperature rise from the inlet temperature to the cladding surface temperature at the axial level experiencing the onset of nucleate boiling. The hot channel factor FFILM found by Eq. (29) of Appendix 1 in the conversion proposal remains unchanged. It depends on the uncertainties in qtfuel and channel thickness (as shown in Eq. 28), but not on the uncertainty in [tfuel/(4Kfuel) + tclad/Kclad].

In short, PUR-1 is not limited by the fuel peak temperature, but by the onset of nucleate boiling, and the uncertainty in [tfuel/(4Kfuel) + tclad/Kclad] is not important for PUR-1. We believe that the hot channel factor FFILM has been determined accurately.

PUR-1 SAR Appendix 2-15 Rev 0,6/08

Question 33

33. Section 4.7.2. According to Appendix 1 the systematic uncertainty in flow rate is accounted for by applying the hot channel factor Fw to the laminar friction factor C.

Explain the reason for the value of the flow friction factor Fw being unity in Tables 4-25 and 4-26.

Response

As suggested in the question, a value of FW (hot channel factor for flow) greater than 1.0 should be used to account for the increased frictional pressure drop due to the hydrodynamically developing laminar flow in the entrance region of the coolant channel, otherwise the code (NATCON) accounts only for the fully developed frictional pressure drop. This has been done now and the results are presented in Table 027-1. Since each coolant channel creates its own buoyancy to drive its own coolant flow, there is no uncertainty due to redistribution of a total reactor flow rate. The loss coefficients of 0.5 and 1.0 at channel inlet and outlet are used in the calculations. To account for the reduction in flow rate due to the hydrodynamically developing laminar flow in the

. channel, the values of FW were calculated for the most limiting channels in the HEU and LEU cores as follows.

=

NATCON calculates the Oarcy-Weisbach frictio.n factor f C/Re using a built-in table of the parameter C for different aspect ratios of the rectangular channel cross section (values of param~ter C are given in the answer to Ouestion number 29). These values of parameter C are for the fully developed laminar flow in a rectangular cross-section channel. An apparent value of the parameter C averaged over the channel length, called Capp , was calculated using Eq. (576) of Shah and London [Ref. 2] to account for the increased pressure drop due to hydrodynamically developing laminar flow in the channel. The ratio Capp/C was found to be 1.1105 for the 207 mil HEU channel, and 1.0985 for the 197 mil LEU channel. Since the NATCON code multiplies the fully developed frictional factor by FW 2 , the input FW equals 1.054 and 1.048 for the HEU and LEU channels respectively. The flow reduction factor is input factor FW or more accurately FW 2/(2+a) = FW09434 (noting that a = 0.12 for the PUR-1 reactor as mentioned in the revised Appendix 1 enclosed herewith).

The results of using these values of FW in NATCON calculations (excluding the effect of temperature dependence of IJ on hot channel factors) are shown in Table 027-1. The ONB power of the HEU core changes to 75.8 kW from 76.3 kW reported in the conversion proposal. The ONB power of the LEU core changes to 95.7 kW from 96.1 kW reported in the conversion proposal. ' .

The channel flow indeed gets reduced by the factor FWO.9434 as expected. For the HEU plate 262, the flow reduces from 0.02083 kg/s to 0.01989 kg/s (see Table 027-1) when the input hot channel factor FW is changed from 1.0 to 1.054. The expected reduced flow should be 0.02083/(1.054)0.9434 = 0.01982 kg/s which is close to the NATCON-

calculated value of 0.01989 kg/so For the LEU plate 1348, the flow reduces from 0.01912 kg/s to 0.01834 kg/s (see Table 027-1) when the input FW is changed from 1.0 PUR-1 SAR Appendix 2-16 Rev 0, 6/08

to 1.048. The expected reduced flow should be 0.01912/(1.048)°*9434 which is close to the NATCON-calculated value of 0.01834 kg/so Question 36

= 0.01829 kg/s

36. Table 4-28. Define the parameter "margin to incipient boiling."

Response

The margin to incipient boiling shown in Table 4-28 was calculated at the nominal operating power of PUR-1 (Le., 1 kW), and it is the smallest value of the temperature difference (TONS - Tw) over the coolant channel length in the hottest channel where Tw is cladding surface temperature with all hot channel factors applied, and TONS is the local onset-of-nucleate-boiling temperature. This basically gives an idea of how far below the onset of nucleate boiling condition the reactor is operating. This definition can be written as an equation as follows:

where T(z) = Bulk coolant temperature at axial position z in the channel heated by the plate power of PopFr FaiN and applying the global hot channel factors for flow and Nusselt number of Fw and Fh Twall(Z) = Cladding surface temperature at axial position z in the channel heated by a plate power of PopFr FaiN and applying the global hot channel factors for flow and Nusselt number of Fw and Fh q"(z) =Heat flux at position z for the plate power of PopFr FaiN and applying the global hot channel factors for flow and Nusselt number of Fw and Fh p(z) =Absolute pressure in the channel at axial position z Tincp(p(z), q"(z)Fnux) = Onset of nucleateboiiing temperature at absolute pressure p(z) and heat flux q"(z)Fnux

=Operating power of the reactor (e.g., 1 kW for PUR-1)

N =Number of fuel plates in the core (e.g., 190 for PUR-1 LEU core)

To =Coolant temperature at the channel inlet Fr = RPEAK = Radial power factor of the plate cooled by the channel Fw =Hot channel factor for flow in the channel Fa =Hot channel factor for reactor power

PUR-1 SAR

=Hot channel factor for Nusselt number Rev 0,6/08

= FFILM = Hot channel factor for temperature drop across the coolant film on cladding surface '

=FFLUX =Hot channel factor for heat flux

=FBULK =Hot channel factor for b_ulk coolant temperature rise in the channel

/

PUR-1 SAR Appendix 2-18 Rev 0,6/08

Document ID: SPC-382 Revision ID: 1 Effective Date: 01/24/07 Specification Specification for Purdue University Standard and Control Fuel Elements -

Assembled for the Purdue University Reactor The INL is a U,S, Department of Energy National Laboratory operated by Battelle Energy Alliance, Idaho National laboratory

PUR-1 SAR Appendix 3-2 Rev 0, 06/08

INTENTIONALLY BLANK

PUR-1 SAR Appendix"3-3 Rev 0, 06/08

Form*H2.C"J (Re\". 09, Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: i ofii PURDUE UNIVERSITY REACTOR Document Project File No. Revision

1. Identifier: SPC-382 2. (optional): 3. No.: I Specification for Purdue University Standard and Control Fuel Elements - Assembled for the

4. Document

Title:

Purdue University Reactor Comments

5. All review and aQQroval signatures for this s~ecification are located on DAR Number 511249.

"*.';jj:X: .'~-"r,':*':,~~fL ,*;;,':;~\t1:*:

'. _':C' .:~~t_* ' .. .. - ....... < :.

.>:, . SIGNATuRES ";';;:

6. 7. 8.

Type or Printed Name Signature Signature Code I Date Organization!

Discipline See DAR Number 506184.

See DAR Number 511249 .

Document Control Release ~ :£.

9. Signature: . LL//, -' .;Cl-J

.... I Date: j. J..L/~ 0 7

", .': ,.::j,Wt1E:;';':';')~~~:k:,."c""*:RECO S MANAGEMENT' )~~:i~*i,~ ]~1~1~1'::* . ' ' ,.

10. Is this a Construction Specification'? Yes 0 No ~ II. NCR Related? Yes 0 No ~

Does document contain sensitive, unclassified infonnation? 0 Yes ~ No If Yes, what

12. category:

13. Can document be externally distributed? Yes ~ No D Area Index

14. Code: Area Ty"}Je SSC ill Review annually. Cutoff when Record superseded, obsolete or Unifonn File Disposition Retention cancelled. Destroy 75 years

15. Code: 0250 16. Authority: ENVI-b-4-a Period: after cutoff.

17. For QA Records Classification Only: Lifetime 0, Nonpermanent 0, Permanent 0 Item or activity to which the QA Records apply:

Periodic Review Frequency: N/A 0, 5 years [8), or

18. Other ,

j PUR-1 SAR Appendix 3-4 Rev 0, 06/08

Form 412.09 (Re\'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: ii of ii PURDUE UNIVERSITY REACTOR

INTENTION ALL Y BLANK

PUR-1 SAR Appendix 3-5 Rev 0, 06/08

FOl"m412.09 (Re\,. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: I of 57 PURDUE UNIVERSITY REACTOR Purdue University Reactor ISpecification I I DAR Number: 511249 I REVISION LOG Rev. Date Affected Pages Revision Description 0 05/31/06 All New Document.

1 01/24/07 All Revised to add Program Anneal requirements and update Drawing Titles

PUR-1 SAR Appendix 3-6 Rev 0, 06/08

Form -t 12.09 (Re\', (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-3S2 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 2ofS7 PURDUE UNIVERSITY REACTOR CONTENTS I.

SUMMARY

....................................................................................................................................... 5 1.1 General .................................................................................................................................. 5

2. APPLlCABLE CODES, PROCEDURES, AND REFERENCES ..................................................... 5 2.1 Standards, Specifications, Drawings and Attachments ......................................................... 5 2.1.1 Specifications and Standards ............................................................................... 5 2.1.2 Drawings (lNL) ................................................................................................... S

3. TECHNICAL REQUIREMENTS ...................................................................................................... 9 3.1 Production Qualification ....................................................................................................... 9 3.1.1 Fuel Plate Qualification: ...................................................................................... 9 3.1.2 Fuel Element Qualification: .............................................................................. 10

3.1.3 Requalification

................ :................................................................................ 10 3.1.4 Operator Qualification: .................................................................................... : II 3.2 Materials .............................................................................................................................. 12 3.2.1 Fuel Bearing Plates ........................................................................................... 12 3.2.2 Aluminum Weld Filler Metal: ........................................................................... 12 3.2.3 Dummy (Non-Fueled) Plate: ............................................................................. 12 3.2.4 Material Requirements ...................................................................................... 13 3.3 Mechanical Requirelnents ................................................................................................... 13 3.3.1 Fuel Plate Requirements ................................................................................... 13 3.3.2 Non-fueled (dummy) plates: ............................................................................. 15 3.3.3 Fuel Element Requirements .............................................................................. 15 3.4 Physical Properties .............................................................................................................. 15 3.4.1 Fuel Plate Requirements ................................................................................... IS 3.5 Surface Condition ................................................................................................................ 17 3.5.1 Surface Defects ................................................................................................. 17

3.5.2 Cleanliness

....................................................................................................... 17

3.5.3 Contmnination

.................................................................................................. IS 3.6 Marking ................................................................................................................................ IS 3.6.1 Fuel Plate Identification: ................................................................................... IS 3.6.2 Fuel Assembly Identification: ........................................................................... IS 3.6.3 Dummy Element Identification: ........................................................................ 19 3.7 Storage ................................................................................................................................. 19

- 3.S PUR-1 SAR Fuel Element Surface Treatment ......................................................................................... 19 Appendix 3-7 Rev 0,06/08

Fonn412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 30f57 PURDUE UNIVERSITY REACTOR 3.8.1 ........................................................................................................................... 19 3.8.2 ........................................................................................................................... 20 3.8.3 ........................................................................................................................... 20 3.9 Graphite Reflectors and Graphite Radiation Baskets .......................................................... 20

3.9.1 Material

............................................................................................................ 20

3.9.2 Asselnbly

.......................................................................................................... 20

3.9.3 Welding

............................................................................................................ 20

3.9.4 Identification

.................................................................................................... 20 3.9.5 Dimensional Inspection: .................................................................................... 21 3.9.6 Surface Finish and Defects: ............................................................................... 21

3.9.7 Storage

............................................................................................................. 21

4. QUALITY ASSURANCE ................................................................................................................ 21 4.1 Materials .............................................................................................................................. 22 4.2 Core Density ........................................................................................................................ 22 4.3 Fuel Loading ........................................................................................................................ 24 4.4 Fuel Honlogeneity ............................................................................................................... 24 4.5 Core Configuration .............................................................................................................. 26 4.6 Bond Integrity ...................................................................................................................... 26 4.6.1 Blister Anneal: .................................................................................................. 26 4.6.2 Ultrasonic Scanning: ......................................................................................... 27 4.6.3 Metallographic Examination ............................................................................. 27 4.7 Internal Defects ................................................................................................................... 27 4.8 Surface Finish and Defects .................................................................................................. 28 4.9 Clad-Core-Clad Dimensions ............................................................................................... 28 4.1 0 Cleanliness ............................................. :............................................................................. 29 4.1 1 Contamination ..................................................................................................................... 29 4.12 Dilnensional ......................................................................................................................... 29 4.12.1 Final Dimensional Inspection, ........................................................................... 29 4.13 Reactor Components and Spare Fuel Element Parts ........................................................... 29

5. PACKAGING AND SHIPPING ...................................................................................................... 30

6. NOTES ............................................................................................................................................. 30 PUR-1 SAR Appendix 3-8 Rev 0, 06/08

Fonn 412.0Y (Re\,. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I /24/07 Page: 4 of 57 PURDUE UNIVERSITY REACTOR 6.1 Definitions ........................................................................................................................... 30 6.2 Purchaser Tests .................................................................................................................... 33 6.3 Sublnittals ............................................................................................................................ 33

6.3.1 Preproduction

................................................................................................... 33

6.3.2 Pre-repair

.......................................................................................................... 34 6.3.3 Manufacturing Schedule: .................................................................................. 34 6.3.4 Delivery Submittals: .......................................................................................... 34 6.3.5 Fuel Plate Radiographs: .................................................................................... 35 6.3.6 Core Compact Data Sheets: ............................................................................... 35 6.3.7 RepOIi of Production by Unapproved Process: ................................................. 35 APPENDIX A ............................................................................................................................................. 37 APPENDIX B ............................................................................................................................................ 39 FIGURES

1. Purdue University Fuel Plate Sampling Procedures For Destructive Tests ............................. 36

PUR-1 SAR Appendix 3-9 Rev 0, 06/08

Fonn412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I124/07 Page: 50f57 PURDUE UNIVERSITY REACTOR

1.

SUMMARY

1.1 General This spec(fication (see def.) defines the materials, components, testing, inspection, certain processes, quality control (see def.) requirements and acceptance criteria for the fabrication of standard and controlfilel elements (see def.) and fuel element containers for the Purdue University Reactor at Purdue University at West Lafayette, Indiana.

2. APPLICABLE CODES, PROCEDURES, AND REFERENCES 2.1 Standards, Specifications, Drawings and Attachments The applicable portions of the following documents as defined herein, fonn a part of this specification. Where there is a conflict between the documents cited and the latest revision thereof, the supplier (see def.) shall notify the purchaser (see def.) of the conflict and use the latest revision in effect at the signing of the contract, unless otherwise directed by the purchaser.

2.1.1 Specifications and Standards National Codes and Standards ASTM E 1742-00 Standard Practice for Radiograph Examination ASTM E 1417-99 Standard Practice for Liquid Penetrant Examination MIL-C-45662 Calibration System Requirements RDT F6-2T Welding of Reactor Core Components, Sections 1,2,3 and 6 American Society for Testing and Materials (ASTM)

ASTM B 209-00 Standard Specification for Aluminum and Aluminum-Alloy Sheet and Plate ASTM B 210-04 Aluminum and Aluminum-Alloy Drawn Seamless Tubes ASTM B 211-00 Standard Specification for Aluminum and Aluminum-Alloy Bar, Rod and Wire ASTM B 214-99 Standard Test Method for Sieve Analysis of Granular Metal Powders

PUR-1 SAR ASTM B 221-00 Standard Specification for Aluminum Appendix 3-10 Rev 0, 06/08

Form -112'<)9 (Re\'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 60f57 PURDUE UNIVERSITY REACTOR and Aluminum-Alloy Extruded Bars, Rods, Wires, Profiles and Tubes ASTM B 241-02 Aluminum and Aluminum-Alloy Seamless Pipe and Seamless Extruded Tube ASTM E 8-00 Methods of Tension Testing of Metallic Materials ASTM E 29-93a (1999) Recommended Practice for Indicating Which Places of Figures are to be Considered Significant in Specified Limiting Values ASTM E 2016-99 Standard Specification for Industrial Woven Wire Cloth American Welding Society (AWS)

A WS A5.1 0-1995 Aluminum and Aluminum Alloy Welding Rod and Bare Electrodes American National Standards Institute (ANSI)

ANSI B46.1-1994 Surface Texture ANSI YI4.5-1994 Dimensioning and Tolerancing for Engineering Drawings American Society of Mechanical Engineers (ASME)

ASME Section V - 2001, Boiler and Pressure Vessel Code without addendum Section V ASME Section IX - 2001 Boiler and Pressure Vessel Code Section IX ASME NQA-I-1997 Quality Assurance Requirements for Nuclear Facility Applications

PUR-1 SAR Appendix 3-11 Rev 0,06/08

form 412.09 (Re\', Ol))

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 70f57 PURDUE UNIVERSITY REACTOR Idaho National Laboratory (INL)

TRTR-ll Specification for Low Enriched Uranium Metal in Test Reactor Fuel TRTR-14 Specification for Reactor Grade Low Enriched Uranium Silicide Fuel Powder IN-F-4-TRA Specification for Aluminum Powder for Matrix Material in Test Reactor Fuel STD 7022A Cleanliness Acceptance Levels for Nuclear or Non-Nuclear Service Components American Society for Nondestructive Test (ASNT)

SNT -TC-l A (1996 or later) American Society For Nondestructive Testing (ASNT) Recommended Practice

PUR-1 SAR Appendix 3-12 Rev 0, 06/08

Form 412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FORPURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 8 of 57 PURDUE UNIVERSITY REACTOR 2.1.2 Drawings (lNL) 635454 Purdue University Test Research and Training Reactor Graphite Reflector Assembly and Source Drive Assembly 635455 Purdue University Test Research and Training Reactor Standard Fuel, Partial,

& Dummy Element Assemblies 635456 Purdue University Test Research and Training Reactor Control Fuel Element Assembly and Dummy Control Fuel

. Element Assembly 635457 Purdue University Test Research and Training Reactor Fission Chamber Fuel Element Assembly 635458 Purdue University Test Research and Training Reactor Standard Fuel Container Assembly 635459 Purdue University Test Research and Training Reactor Control Fuel Container Assembly 635460 Purdue University Test Research and Training Reactor Irradiation Facility Assembly 635461 Purdue University Test Research and Training Reactor Capsule Holder and Capsule Insert Assemblies and Details 635462 Purdue University Test Research and Training Reactor Graphite Container Assembly, and Source Drive Container Assembly 635463 Purdue University Test Research and Training Reactor Fuel Plate Assembly and Dummy Fuel Plate Detail 635464 Purdue University Test Research and

,Training Reactor Container Tube Assembly and Details

PUR-1 SAR 635465 Purdue University Test Research and Appendix 3-13 Rev 0,06/08

Form 412.09 (Re\'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I124/07 Page: 90f57 PURDUE UNIVERSITY REACTOR Training Reactor Graphite Block Detail And Graphite Capsule Holder Detail 635466 Purdue University Test Research and Training Reactor Miscellaneous Details 635467 Purdue University Test Research and Training Reactor Source Drive Nozzle Assembly and Source Drive Top 635468 Purdue University Test Research and Training Reactor Nozzle Preliminary Machined and Fission Chamber Top

3. TECHNICAL REQUIREMENTS 3.1 Production Qualification The supplier is required to qualify the processes or portions of the process or be exempt fi-om same by written approval of the purchaser. 1n qualification (see def.), only materials that comply with this specification shall be used .

Qualification processes, equipment, and operator qualification/training programs shall be identical to those used during production (see def.). To qualify, the supplier must demonstrate, to the satisfaction ofthe purchaser, that the process is capable of producing a product, which satisfies all the requirements of the specifications. Assembly of production fuel elements shall not be initiated until:

(1) all required data, to assure compliance with the qualification requirements, has been submitted to the purchaser; (2) data and records required by Section 6.3 have been submitted; and (3) written approval of qualification has been received by the supplier from the purchaser.

3.].1 Fuel Plate Qualification:

Fuel plate (see def) qualification shall be satisfied by supplier production of a minimum of two consecutively produced plate lots (see def), in lot quantities of24 plates (see def). The plates shall be made using low enriched uranium in the fonn of Silicide (see def) powder, which have a yield of no less than 65% acceptable fuel plates meeting all applicable requirements of this specification. The supplier may combine the results of two consecutive lots into a production run in determining the 65% yield requirement provided that there have been no changes in the manufacturing (see def) procedure (see def) between lots which would require requalification (see def) in accordance with Section 3.] .3 .

PUR-1 SAR Appendix 3-14 Rev 0, 06/08

Form 412.09 (Re\,. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0]/24/07 Page: ] 0 of 57 PURDUE UNIVERSITY REACTOR In the event that fuel plate qualification has been perfonned by the supplier meeting all the requirements of this specification during the past twelve months, and qualified operators are performing the fabrication, fuel plate qualification requirements listed above will be waived.

Fuel plates made in development (see def.) (prior to and including qualification runs that fail to meet the 65% yield requirements) will not be used in fabricating production fuel elements without prior approval of the purchaser.

3.1.2 Fuel Element Qualification:

The supplier shall fabricate 1 dummy standardfilel element assembly (see def.), which shall meet the requirements of this specification.

3.1.3 Requalification

The supplier shall notify the purchaser of any proposed process change.

A changed process may not be used in production until the supplier has met all the requirements of Section 3.1.3, submits the results and data of the requalification effort, and receives written approval from the purchaser.

Requalification for any fucI plate attribute to the requirements of the specification will be required when the processes, materials, fuel loadings, equipment or equipment operators (welding and rolling) which have been previously qualified are changed, unless the supplier can demonstrate to the satisfaction of the purchaser by engineering explanation or proof test that such changes will have no detrimental effect on the product.

Requalification for compacting, pack (see def.) assembly, and rolling mill operators can be less than qualification basis, since the procedure has already been established. Candidate operators who are not qualified for compacting operations, pack assembly operations, and hot/cold rolling mill operations must demonstrate their abilities in perfom1ing the individual operations they are assigned.

An operator must qualify by processing two lots of fuel plates with minimum lot size of 24, for the operation he is assigned to qualify, before perfom1ing any production operation independently. Each lot of fuel plates shall be processed through final inspection, with a minimum yield rate of 90% acceptable fuel plates required for the operator to be

PUR-1 SAR termed qualified .

Appendix 3-15 Rev 0,06/08

Form ~ 12.09 (Rt'\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: ]

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0]/24/07 Page: ]] of 57 PURDUE UNIVERSITY REACTOR NOTE: Failure of an operator to qual!fjl, because offilel plate deviations, must be based on deviations related to the operation being qualified.

The purchaser on a case-by-case basis will determine the quantities and sizes ofrequal[ficatiol1filel plates selected to be destructively examined.

3.1.4 Operator Qualification:

Operator qualification will be accomplished via an approved supplier intell1al qualification program for the following operations:

A. Arc melting B. Compacting C. Pack assembly D. Hot rolling E. Cold rolling

F.

3.1.4.1 Final machining.

In addition to the operations specified above, the supplier shall also show evidence of the training and competency of those individuals who perform any of the following fuel element fabrication and inspection activities:

A. Powder sieving, weighing, and testing B. Compact weighing, visual and dimensional inspection C. Fuel plate/element and component cleaning D. Fuel plate annealing operations E. Dimensional inspection of plates, elements, and subcomponents F. Metallographic sample preparation and inspection G. Visual inspection of plates, elements, and subcomponents H. Void volume inspection

PUR-1 SAR

1. Fluoroscope inspection of fuel plates Appendix 3-16 Rev 0, 06/08

Form 412.09 (Re\,. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY ldentifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I124/07 Page: 12 of 57 PURDUE UNIVERSITY REACTOR

1. Radiography and inspection of fuel plate radiographs K. Ultrasonic testing and interpretation.

The individuals performing these operations shall have specific requirements imposed on them that will demonstrate their knowledge and ability to perfoml their respective assignments. Documented evidence ofthe training of these individuals shall be maintained and shall be made available to the purchaser upon request.

3.2 Materials The material requirements for the components comprising the fuel element are as specified on Drawings per Section 2.1.2 and requirements of this section.

3.2.1 Fuel Bearing Plates 3.2.1.1 Fuel Cores: Thefllel cores (see def.) of the fuel plates shall be uranium silicide powder dispersed in aluminum alloy powder which meet the requirements ofIN-F-4-TRA and TRTR-14, per Section 2.1.1 of this specification.

3.2.1.2 Frames and Covers: Aluminum for the frames and cover plates shall conform to ASTM B209, Alloy 6061-0. The aluminum plate stock used for frame and cover plates shall be certified by the supplier to contain less than 30 PPM boron, 80 PPM cadmium, and 80 PPM lithium.

The subcontractor shall furnish certified physical properties and chemical analyses of ingots or plates of the 6061 materials to INL.

3.2.2 Aluminum Weld Filler Metal:

All aluminum weld filler metal shall be ER4043 as required by Specification A WS A5.10-1995.

3.2.3 Dummy (Non-Fueled) Plate:

Dummy (l1011:/ileled) plates (see def.) shall be fabricated from aluminum Type 6061-0, that meets the requirements of Section 3.2.1.2 .

PUR-1 SAR Appendix 3-17 Rev 0, 06/08

Form 41 ~.09 (Rev, (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 13 of 57 PURDUE UNIVERSITY REACTOR 3.2.4 Material Requirements All material used or contained in the product shall comply with all the requirements of this specification and Drawings per Section 2.1.2 unless exempted by written document by the purchaser.

3.3 Mechanical Requirements 3.3.1 Fuel Plate Requirements 3.3.1.1 Fabrication: The supplier shall furnish the details of his fuel plate rolling schedule and component cleaning process to the purchaser for approval prior to use in production per 6.3.1.

Compacting details shall include silicide - aluminum compacting pressure and compacting press dwell time.

After hot rolling, each fuel plate shall be blister annealed per Section 4.6.1 and then cold rolled to final thickness at room temperature. After cold rolling operation, the fuel plates shall be subjected to program annealing. The rolling schedule shall contain, at a minimum, the following:

A. Nominal plate reduction B. Minimum number of hot roll passes C. Nominal inter-pass reduction and target thickness D. Hot rolling furnace temperature E. Preheat time for all hot roll passes F. Final hot roll plate thickness G. Type and frequency of roll lubricant utilized H. Nominal cold roll reduction.

1. Final cold roll thickness.

Fuel plate cladding (see def.) thickness required by Section 3.3.1.4 and fuel core homogeneity requirements of

. Section 4.4 are independent requirements that must be met.

PUR-1 SAR Appendix 3-18 Rev 0, 06/08

Fonn 412.()9 (Re\', (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 14 of 57 PURDUE UNIVERSITY REACTOR 3.3.1.2 Core Configuration: No fuel particles are allowed within the fuel free zones located at the ends of the plates as shown on Drawing 635463.

The nominally unfueled area of each fuel plate as defined by Drawing 635463 may contain random fuel paIiicles defined as flaking and limited in size, location, and spacing per this Section, as determined by Section 4.5.

The presence of fuel particles detected between the maximum fuel core outline and fuel plate edges and ends is allowed provided they do not violate the following restrictions:

One or more fuel particles, which fit in a rectangle whose area is not more than 4x I 0-4 in 2 is acceptable AND The fuel particle(s) are no closer than 0.080 in. to any other particle edge to edge AND No particle is closer to the plate edge or end than the major dimension of the particle.

Stray fuel particles (see def.) that violate the above requirements may be removed from fuel plate edges by filing, provided the following:

The filed out area is no deeper into the edge of the plate than 0.050 in., no longer than 0.250 in.

AND Each filed area is at least 1.0 in. apart Filing of fuel plate ends, for the removal of stray particles, is not allowed, unless previously approved by the purchaser.

3.3.1.3 Internal Defects and Bond Integrity: Metallurgical bond, as determined by Section 4.6 is required at interface areas of the finished fuel plates, specifically fuel core-to-c1ad and

PUR-1 SAR c1ad-to-frame. The presence of grain growth across the fuel Appendix 3-19 Rev 0,06/08

Form 412.1)9 (Rc\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY ldentifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 15 of 57 PURDUE UNIVERSITY REACTOR matrix-cladding interface and across the aluminum frame-cladding interface of at least 50% is required. Fuel core defects in excess of 0.06 in. in any dimension as detemlined by Section 4.7 are not allowed.

3.3.1.4 Cladding Thickness: During production, all plates will be subjected to UT min-clad inspection. The standard will be calibrated at the nominal 0.008-inch scan depth. The gage will then be adjusted to a 0.010 inch scanning depth and the fuel plates will be scanned at 0.010 inch. Fuel plate UT traces, which display min-clad indications at the 0.01 O-inch depth, shall be visually compared with the 0.008-inch Standard trace. Fuel plates for which the UT reports show a comparable density of indications, or worse, than the indications displayed on the standard UT report are unacceptable. Fuel plates, which fail the 0.01 O-inch UT scan, shall be rescanned at 0.008 inch. Only fuel plates which are acceptable when rescanned at 0.008 inch shall be submitted to the Purchaser and User for evaluation .

3.3.2 Non-fueled (dummy) plates:

The supplier shall uSe a cold rolling method to obtain plate thickness.

Non-fueled (dummy) plates shall be subjected to program anneal.

3.3.3 Fuel Element Requirements 3.3.3.1 Welding: All welding shall be performed using procedures and welding personnel qualified in accordance with ASME Section IX or the criteria defined in Appendix B. Quality acceptance of production welds shall be in accordance with Appendix B, Section 5.

3.4 Physical Properties Fuel plates shall have a core of U3Si 2 and aluminum and completed fuel plates and fuel elements shall have fuel loadings per Sections 3.4.1.2, and 3.4.1.5.

3.4.1 Fuel Plate Requirements 3.4.1.1 Fuel Core: The fuel core shall consist of 19.75 --0.2 weight

% enriched uranium silicide powder dispersed in aluminum powder. The uranium silicide powder shall be -100, +325 U.S. standard mesh particles. However, a blend may PUR-1 SAR Appendix 3-20 Rev 0, 06/08

Form 411.09 (Rev. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 1/24/07 Page: 16 of 57 PURDUE UNIVERSITY REACTOR contain up to 35 weight percent of -325 U.S. standard mesh particles. Any powder particles greater than 100 mesh particles shall be reground such that they will go thru the 100 mesh sieve. The fuel core shall be fabricated according to standard powder-metallurgical and roll-bonding techniques. The supplier shall provide to the purchaser, a written procedure for pack assembly and the initial rolling step which describes the method used to prevent excessive oxidation that causes non-bond of fuel core to the cladding.

3.4.1.2 Fuel Loading: By using the approved supplier's method of assigning U-235 content, per a detailed description as to the weighing procedure by which the supplier proposes to assign fu~Each fuel plate shall contain~. The weight of each core shall be measured and recorded to within 0.01 gram U-235 based upon weight of the final compact and chemical and isotopic analysis of the constituents.

3.4.1.3 3.4.1.4 Fuel Homogeneity: Fuel homogeneity requirements are located in section 4.4.

Void Volume: In the qualification process, all fuel plates shall be inspected for void volume using the method described in Section 4.2. The percent voids i~ the fuel cores of all fuel plates shall be detem1ined by the inspection procedure developed by the supplier. The percent voids in the fuel cores shall be at least 3.0% and not more than 11.0%.

3.4.1.5 Fuel Element Requirements 3.4.1.6 for each fuel element

. Each Control Fuel

. Control limits for the method used to measure this weight are established at the 95% confidence level for a significant population of measurements of a particular standard. The U-235 enrichment shall be 19.75 --0.2 weight % of total uranium per specification TRTR-ll.

PUR-1 SAR Appendix 3-21 Rev 0, 06/08

Form 412.0~) (Re\', OY)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I /24/07 Page: 17 of 57 PURDUE UNIVERSITY REACTOR 3.5 Surface Condition Fuel plates and completed fuel elements must comply with the surface condition requirements of Section 3.5.1,3.5.2, and 3.5.3 and drawings of Section 2.1.2, per ANSI B46.1. Sanding, or any other finishing procedure that will smear the aluminum surface, will not be allowed on fuel plates unless approved by the purchaser.

3.5.1 Surface Defects 3.5.1.1 Compliance with surface finish and defect requirements shall be established by 100% visual inspection of all fuel plates and elements. The surface of the finished fuel plates shaIl be smooth and free of gouges, scratches, pits, or removal of metal in excess of 0.005 inch in depth. Dents in the fuel plate shall not exceed 0.0 I 2 inch in depth or 0.25 inch in diameter. If there is evidence of dogboning in the plates, surface defects in the dogbone (see def.) area shall not exceed 0.003 inch in depth. No degradation ofthe fuel

3.5.1.2 plates beyond these limits shall be permitted .

Fuel Plates shall be free of stringiness, scabs, or cracks.

Surface finish shall be as required by Drawing 635463.

Compliance with requirements of this section shall be accomplished by visual inspection of all fuel plates and fuel elements.

3.5.1.3 Defects on fuel plate edges or ends are pemlissible provided they are evaluated and acceptable to the requirements of Paragraph 3.3.1.2.

3.5.1.4 Compliance with surface finish and defect requirements shall be established by 100% visual inspection of all fuel element containers. Fuel element containers shall be free of surface defects such as pits, dents, or scratches in excess of 0.010 inch in depth and 0.12 inch in diameter or equivalent area.

3.5.2 Cleanliness

The suppliers fabrication, assembly, and storage areas used for the production of Purdue University fuel elements and/or components shall confonn to the requirements of "controlled work area" (see def.) as defined in Paragraph 1.3.6 ofINL Standard 7022A. Cleanliness shall be PUR-1 SAR Appendix 3-22 Rev 0, 06/08

Fonn-' 12.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Jdentifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 18 of 57 PURDUE UNIVERSITY REACTOR in compliance with INL Standard 7022A, Paragraphs 1.1, 1.2.3, 3.1, 3.2-b, d, i, 3.3 - d, e, 4.1.3, 4.2, and 4.3. Freon shall not be used to clean fuel elements or components.

As determined by Section 4.10 of this specification, there shall be no foreign materials on the finished fuel plates or surfaces of the finished fuel elements. All oil, metal chips, tumings, dusts, abrasives and spatter, scale, and other particles shall be removed from the fuel surfaces by procedures which assure that the minimum cladding thickness has not been violated. All components shall be cleaned by a method approved by the purchaser.

3.5.3 Contamination

The surfaces of each fuel plate shall be counted or smeared and counted for alpha-beta-gamma contamination. The alpha count shall be less than five dpm per 100 cm 2 , and the beta-gamma count shall be less than 200 dpm per 100 cm 2

Each fuel element shall be smeared and counted for radioactive contamination. The alpha count shall be less than five dpm per 100 cm 2 ,

and the beta-gamma count shall be less than 200 dpm per 100 cnl.

3.6 Marking NOTE: AI/file! plates, file! assemblies, andfile! e!ement containers will be marked per this section.

3.6.1 Fuel Plate Identification:

Each finished fuel plate shall be identified, as shown on Drawing 635463, by a combination of numbers and/or letters that will maintain positive identification relative to the complete traceability to the supplier fabrication history, including the basic material lots, heat or metal, manufacturing cycle, and quality control phases. The identification number shall be stamped, etched or vibro-peened at the location specified by Drawing 635463. The depth of the identification characters shall not exceed 0.0 lOin.

3.6.2 Fuel Assembly Identification:

Each fuel assembly shall have an identifying number such as 07-XX (07 signifying year of fabrication). The number shall be placed on the

PUR-1 SAR container assembly as shown on Drawings 635455, 635456 and 635457 .

Appendix 3-23 Rev 0, 06/08

Fonn412.(19 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 19 of 57 PURDUE UNIVERSITY REACTOR The identification shall be stamped or entered by a method approved by the purchaser, with two inch block characters not in excess of 0.010 inches in depth. Standard assemblies should be labeled: E2, F2, G2, H2, F3, H3, E4, F4, G4, H4, F5, H5. Control assemblies should be labeled: E3, G3, and E5. The fission chamber assembly should be labeled as G5. The source assembly shall be labeled as C3. The spare Standard Assemblies should be labeled: SP-I, SP-2, SP-3. The spare Control Assembly should be labeled as SPA.

3.6.3 Dummy Element Identification:

The dummy standard fuel element assembly shall have the identifying number DUM-I. The number shall be placed on the container assembly as shown on Drawing 635455. The identification shall be stamped or entered by a method approved by the purchaser, with two inch block characters not in excess of 0.0 10 inches in depth.

3.7 Storage

All fuel plates, fuel assemblies, and fuel element containers that have received final cleaning per Section 3.5.2 shall be protected in clean polyethylene containers or other containers approved by the purchaser while (1) awaiting final assembly, (2) being transferred into or being maintained in storage, or (3) being prepared for packaging or shipment. Any material exposed to contamination shall be reinspected to the requirements of Section 3.5.

3.8 Fuel Element Surface Treatment I f boehmite treatment is required during fuel element fabrication, the following shaH apply. After fuel elements are assembled and inspected they shall be subjected to an environment that will cause an evenly distributed boehmite layer of 0.00006 to 0.0003 in. thickness (averaged over the surface using eddy current instrumentation) to form on all surfaces of the entire assembly .. The treatment process shall be perfonned under controlled conditions, which shall require the supplier to maintain a record of the thermal history of the autoclave. The records shall include heat charts of recorded time and temperature. Documented evidence of the controls placed on the autoclave shall be maintained by the supplier.

3.8.1 After the boehmite process has been qualified, one fuel element from nd every 2 autoclave run shall be inspected following a procedure approved by the Purchaser.

PUR-1 SAR Appendix 3-24 Rev 0, 06/08

Form 412.<)9 (ReY. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 20 of 57 PURDUE UNIVERSITY REACTOR 3.8.2 Each fuel element shall have a corresponding aluminum plate coupon, made from fuel plate end crops, placed near the fuel element during the boehmite formation process. The aluminum plate coupon shall be subjected to the same environment as the fuel elements and each coupon measured for boehmite thickness.

3.8.3 Fuel elements and aluminum plate coupons subjected to the boehmite formation process must be carefully handled to preclude scratches, dents, and gouges that would cause removal of boehmite.

3.9 Graphite Reflectors and Graphite Radiation Baskets Graphite reflector assemblies (see def.) and irradiation facili()l assemblies (see def.) shall be fabricated as per requirements contained in this section and in drawings 635454, 635460, 635461, and 635465.

3.9.1 Material

All materials used shall comply with all the requirements of this

3.9.2 specification and applicable drawings .

Assembly:

The assembly of the graphite reflector assemblies and irradiation facility assemblies shall be as shown on the applicable drawings.

3.9.3 Welding

All welding shall be perfonned using procedures and welding personnel qualified in accordance with ASME Section IX or the criteria defined in Appendix B. Quality acceptance of production welds shall be in accordance with Appendix B, Section 5.

3.9.4 Identification

The graphite reflector assemblies shall have identifying numbers such as GR-X placed on the side of the assembly as shown drawing 635454.

The graphite reflector shall be labeled as follows:DI, D2, D3, D4, D5, EI, FI, GI, HI, II, 12, 13, 14, and 15. The irradiation facility assemblies shall have identifying numbers such as IF-X placed on the side of the assembly as shown on drawing 635460. The irradiation facility assemblies shall be labeled as follows: D6, E6, F6, G6, H6, and ]6. The identification shall be stamped or entered by a method approved by the

PUR-1 SAR purchaser, with two inch block not in excess of 0.0 I 0 inches in depth .

Appendix 3-25 Rev 0, 06/08

Form 412.()~ (Re\'o (l1J)

Idaho National Laboratory SPECIFICA TION FOR PURDUE UNIVERSITY ldentifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 21 of 57 PURDUE UNIVERSITY REACTOR 3.9.5 Dimensional Inspection:

Verification of all external dimensions of the graphite reflector assemblies and irradiation facility assemblies shall be by 100%

inspection, in accordance with drawings 635454 and 635460. All dimensions of this specification shall apply at a temperature of 75°F +/-

5°.

3.9.6 Surface Finish and Defects:

The graphite reflector assemblies and irradiation facility assemblies shall be free of surface defects such as pits,dents, scratches in excess of 0.010 inch deep and 0.12 inch diameter or equivalent area.

3.9.7 Storage

All graphite reflector assemblies and irradiation facility assemblies shall have received final cleaning and shall be protected in clean polyethylene containers or other containers approved by the purchaser while (a) being

4. QUALITY ASSURANCE transferred into storage, (b) being maintained in storage, or (c) being prepared for shipment or packaging.

The supplier shall document, implement, and maintain a quality program in compliance with ASME NQA-I-1997.

The supplier shall permit the purchaser to conduct pre-award and continuing evaluation of the Quality Program. .

PersOlmel perfonning NDE examinations, specifically radiographic, ultrasonic, liquid penetrant, and visual shall be certified to American Society for Nondestructive Testing (ASNT) Number SNT -TC-IA and certification documentation shall be made available to the purchaser.

Unless otherwise specified, the supplier shall be responsible for the perforn1ance of all tests and inspections required prior to submission to the purchaser of any fuel element for acceptance. Provided, however, that the perfornlance of such tests and inspections is in addition to, and does not limit, the right of the purchaser to conduct such other tests and inspections as the purchaser deems necessary to assure that all fuel elements are in confonnance with all requirements of this specification. Except as otherwise specified, the supplier may use for inspection purposes either his own or any commercial laboratory acceptable to the purchaser. Records of all tests and examinations shall be kept complete PUR-1 SAR Appendix 3-26 Rev 0, 06/08

Fnnn 412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Jdentifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 22 of 57 PURDUE UNIVERSITY REACTOR and available to the purchaser. All test and measuring equipment shall be calibrated per the requirements of Standard MIL-C-45662.

The following applies to specified limits for requirements on core density per Section 3.4.1.1 and 4.2 and all dimensional requirements of this specification. For purposes of determining confonnance with this specification an observed value or a calculated value shall be rounded off to the nearest unit in the last right hand place of figures used in expressing the limiting value in accordance with the rounding-off method of the Recommended Practices for Designating Significant Places in Specified Limiting Values (ASTM E29).

The supplier shall prepare for his use and the purchaser's approval an integrated manufacturing and inspection test plan. The plan shall include all manufacturing operations, equipment and tooling used, inspection requirements and gaging used, and mandatory hold points established by the purchaser.

Any materials or fuel element components that are fabricated using equipment, personnel, or processes that are not in accordance with approvals as previously granted by the purchaser are subject to rejection (see def.). A report of any such incident must be

submitted in accordance with Section 6.3.7 .

Fuel clement inspection for shipment or rejection will be made by the on-site purchaser's representative at the supplier's plant. Final fuel element acceptance will be made by the purchaser at the User's facility.

4.1 Materials Compliance with the material requirements of Section 3.2 shall be established by supplier certification. A "Certification of Chemical Analysis" or a certified mill test report shall be supplied to the purchaser for each lot of material used in the fabrication of fuel elements. This certificate shall give the results of the chemical analysis for the material. All fuel element materials shall be traceable.

4.2 Core Density The density of the fuel cores required in Section 3.4.1.3 shall be determined by the Archimedes principle. During qualification of the fuel plate core void density required by Section 3.4.l.3 shall be detennined on all qualification fuel plates submitted. After the particular plate typ~ has been qualified, 100% inspection for void density is not required for production lots of fuel plates. For production lots, three randomly selected fuel plates from each lot shall be inspected for void volume density. Should anyone of these plates be discrepant, the entire lot must then be inspected for void volume density. Ifvoid density discrepancies appear regularly in the process, the purchaser may request 100% inspection.

PUR-1 SAR Appendix 3-27 Rev 0,06/08

Form 412.(l9 (RC'v. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 23 of 57 PURDUE UNIVERSITY REACTOR The actual core volume shall be calculated by the following formula where:

weight units are in grams and volumes in cubic centimeters.

where:

Vc immersion volume of fuel 'plate core Vp volume of fuel plate PAL density of aluminum used for fuel plate cladding 2.715 gms/cc Wp weight of plate We deburred weight of fuel plate core compact

PUR-1 SAR Appendix 3-28 Rev 0, 06/08

1\lf111 412.09 (Rev_ (9)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 24 of 57 PURDUE UNIVERSITY REACTOR The theoretical core volume shall be calculated by the following fonnulas:

Vel = (WU 3~~2) + (WAI) pU3::m PAl where:

calculated theoretical core volume weight ofU3Sh powder in core weight of aluminum matrix powder in core density ofU3Si2 powder as measured PAl density of aluminum powder used for core matrix

2. 7lO gms/cc The void percent in the core shall be calculated using the following formula:

where:

V% percent voids in the fuel plate core 4.3 Fuel Loading Verification of the fuel loading as specified in Section 3.4.1.2 shall be in conformance to the supplier's procedure required in Section 6.3.1.

In order to determine compliance with the fuel density requirements of Section 4.4, the U-235 loading of the fuel plate, as detemlined in accordance with the procedures of Section 6.3.1, will be divided by the core volume (V c) as calculated by the method described in the second paragraph of Section 4.2.

4.4 Fuel Homogeneity Fuel core homogeneity requirements shall be complied with by a one-piece radiograph of all fuel plates from each fuel plate lot and evaluation of the radiograph by calibrated densitometer measurements. Purchaser approved density standards may be used by the supplier. Fuel plates and density standards shall be PUR-1 SAR Appendix 3-29 Rev 0, 06/08

Form 412.09 (Re\'. ()~)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24107 Page: 25 of 57 PURDUE UNIVERSITY REACTOR exposed simultaneously. Fuel plate density variations shall be determined by comparison of fuel plate areas to corresponding areas of the standard.

All fuel plates shall be inspected for homogeneity. Homogeneity of the fuel plate core shall be determined by radiograph film density measurements with a densitometer having a 0.080 inch aperture.

When determining fuel core density from plate radiographs, the brighter the image on the radiograph, the more dense is the uranium and the lower the number indicated on the densitometer. The darker the image on the radiograph, the less dense is the uranium and the larger the number indicated on the densitometer. A

+30% fuel core density and a +20% fuel core density is indicated by the densitometer readings in the suspect area being 30% or 20% lower than the average densitometer readings for all core locations. A -30% or a -20% fuel core density is indicated by the densitometer readings in the suspect area being 30% or 20% higher than the average densitometer readings for all fuel core locations.

Anyone-half inch diameter or greater spot in the plate fuel core area, other than the dogbone area shall not be less in fuel density than -20% of the average fuel

density for all fuel core locations. To determine the low density of a one-half inch diameter area, the film is maneuvered under the densitometer in the low-density area until the highest number possible is obtained on the densitometer. This number is recorded. Then four readings are taken one-fourth inch from this spot and symmetrical around it. The average of these five readings is compared to the average densitometer readings for all fuel core locations.

If density standards are used, the average densitometer readings of all fuel core locations will be replaced by the nominal density standard and comparisons will be detennined between the suspect spot on the radiograph and the -30% and

-20% standards. For the +30% and +20% homogeneity overload inspection, compare the nominal density standard to the suspect area. In this case densitometer units from nominal of the fuel plate represent the following percentages: -0.15 = +30%; -0.10 = +20%. Fuel plates exceeding these limits are discrepant.

For rectangular shaped, suspected discrepant areas that are evaluated to the one-half inch criteria, orient the four symmetrical readings such that worst case readings will be taken.

Between the minimum and maximum pennissible fuel core length boundary, fuel underload condition shall not be evaluated.

Any indication of un-alloyed uranium as determined by radiography shall be cause for rejection.

PUR-1 SAR Appendix 3-30 Rev 0, 06/08

Form 412.()~ (ReY. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 26 of 57 PURDUE UNIVERSITY REACTOR Any 0.080 inch diameter spot in the fuel plate dogbone area (area within one inch of each fuel core end) shall not be greater in fuel density than +30% of the average fuel density for all core locations. Anyone-half inch 'diameter area in the dogbone area shall not be less in fuel density than -30% of the average fuel density for all fuel core locations. The actual dogbone shall not be more than one-half inch in the longitudinal direction.

Other than the dogbone areas near ends of fuel core, anyone-half inch diameter area shall not be greater in fuel density than +20% of the average fuel density for all fuel core locations. To detennine the high density of a one-half inch diameter area, the film is maneuvered under the densitometer in the high-density area until the lowest number possible is obtained on the densitometer. This number is recorded. Then four readings are taken one-fourth inch from this spot and symmetrically around it. The average of these five readings is compared to the average densitometer readings for all fuel core locations.

Unless otherwise specified, purchaser approval of all radiographs is required prior to assembly of fuel plates into elements.

4.5 Core Configuration Each finish-cut flat fuel plate shall be radiographed in accordance with Appendix A and evaluated for compliance with Section 3.3.1.2.

Visual radiograph inspections will be perfonned without magnification on a light table having a light intensity of 450 to 600 ft-candles at the table surface and the area darkened to give. a light range of 5 to 15 ft -candles 18 in. above the light table with radiograph filh1 in place on the table.

4.6 Bond Integrity 4.6.1 Blister Anneal:

After the fuel plate has been hot rolled, it shall be heated to 900°F+/-13°F, held at that temperature for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , -15 minutes, +30 minutes, removed from furnace, and allowed to air cool.

Any blisters, in the fuel core region larger than a 0.060 in. diameter or any blister in the frame region of the fuel plate larger than 0.120 in. diameter shall result in rejection ofthe associated fuel plate. A maximum oftwo blisters less than 0.060 in. diameter is allowed in the fuel core area, provided they are more than 0.250 in. apart. A maximum of two blisters in any of the four sides of the picturef;-ame (see def.)(a maximum of eight) region smaller than 0.120 in. can be tolerated providing that no blister is PUR-1 SAR Appendix 3-31 Rev 0, 06/08

form 412.09 (Rev. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 27 of 57 PURDUE UNIVERSITY REACTOR any closer to the plate edge or end or to another blister than the major dimension of the blister and no blister is closer to the plate edge or end than 0.050 inch. When there is question as to size or location of the blisters, the acceptance or rejection of the plate shall be detennined in the ultrasonic inspection of Section 4.6.2.

4.6.2 Ultrasonic Scanning:

The finished fuel plate area shall be ultrasonically inspected in compliance with ASME Boiler and Pressure Vessel Code,Section V, Article 5, Paragraphs T-110, T -510, T -520, T -521, T -522-a, b, c, e, g, i, j, k, 1, 0, T -523, T -523-1, and T -534. Any indication of discontinuity in the fuel core region equivalent to that indicated by a 0.060 in. diameter standard or any indication of a discontinuity in the frame region of the fuel plate equivalent to that indicated by a 0.120 in. diameter standard shall result in rejection of the associated fuel plate. Acceptance criteria for number of blisters revealed by ultrasonic scanning are per Section 4.6.1. Any discontinuities, inside the fuel plate, other than blisters and for which acceptance criterion is not already stated, shall be

4.6.3 described by the supplier and evaluated by the purchaser.

Metallographic Examination.

During qualification, One fuel plate per lot selected for qualification per Section 3.1.1 will be sectioned per Figure 1, polished and etched, and examined at 50x or above for bond and clad-core-clad dimensions per the requirements of Sections 3.3.1.3 and 3.3.1.4, and Drawing 635463, respectively.

If the fuel plate fails the metallographic examination for grain growth, voids, laminations, core cracking or separation, or foreign particles or materials, then randomly selected another plate in the lot for metallographic examination. If this plate fails the examination, reject the lot.

Fuel plates selected for destruction tests may be rejected fuel plates, providing the attribute to be tested for is not affected by the cause for rejection. Reject fuel plates so used must have purchaser approval before destruct tests are perfonned.

4.7 Internal Defects Any internal defect in excess of the requirement of Section 3.3.1.3 in the fuel

core, including voids, laminations, U3 Sh segregation, clumping, core cracking or PUR-1 SAR Appendix 3-32 Rev 0, 06/08

Form 412.{)1) (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 1/24/07 Page: 28 of 57 PURDUE UNIVERSITY REACTOR separation, or foreign particles or materials, which is identified by any measurement technique, including radiography per Section 4.4, ultrasonic scanning per Section 4.6.2, or metallography per Section 4.6.3, shall be cause for rejection of the fuel plate.

4.8 Surface Finish and Defects Compliance with requirements of Section 3.5 shall be established by visual inspection of all fuel plates and fuel elements. Out-of-specification defects shall be measured for size and depth and reported to the purchaser.

4.9 Clad-Core-Clad Dimensions Fuel Plate Qualification requirements of section 3.1.1 shall be established by ultrasonic techniques using the purchaser-supplied, min-clad inspection gage. All fuel plates will be subjected to ultrasonic min-clad inspection with the fuel core region scanned for each plate. Ultrasonic min-clad inspection shall be accomplished by calibration of the min-clad gage, using the Advanced Test Reactor (A TR) Standard (8E0777) scanned at the nonnal mode of 0.008 inches.

The min-clad gage will then be adjusted and the fuel plates will be scanned at a depth of 0.010 inches. Ultrasonic Test (UT) traces showing fuel at the 0.010 inch depth will be compared to the 0.008 inch standard to determine plate acceptability. If the density of indications from fuel plate exceeds the A TR standard density of indications, the plate is rejectable.

NOTE: The ATR standard is a small piece of an ATRfzlel plate that hasfllel particles near the slll.iace. It is used on the UT min-clad machine to indicate min-clad indications and compare the density of these indications to any indications notedfi'om afue! plate being inspected by UT.

During the fuel plate qualification process, compliance with the requirements of Section 3.3.1.4 shall be established by destructive analysis of one fuel plate per lot in accordance with Figure 1.

After fuel plate qualification, all production plates shall be min-clad ultrasonic inspected at a depth of 0.0 10 inches. Those plates discrepant at 0.010 inches shall be rescanned at 0.008 inches. Plates which are acceptable when re-scanned at 0.008 inches shall be submitted on Information/Change Request (Form 540.33) to the purchaser.

PUR-1 SAR Appendix 3-33 Rev 0, 06/08

Form 412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY ldentifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 29 of 57 PURDUE UNIVERSITY REACTOR 4.10 Cleanliness Fuel plate, fuel assembly, and fuel element container cleanliness requirements of Section 3.5.2 shall be established by visual inspection without magnification of all fuel plates, fuel assemblies, and fuel element containers.

4.11 Contamination The surfaces of each fuel plate and fuel assembly shall be counted or smeared and counted for alpha-beta-gamma contamination and meet the requirements of Section 3.5.3.

4.12 Dimensional It shall be the supplier's responsibility to assure that fabrication is perfol111ed in accordance with all dimensions delineated in the Drawings referenced in Section 2.1.2. Noncomplying design dimensions on fuel plates, fuel assemblies, and fuel element containers (actual measurements) shall be submitted to the purchaser for review and approval. Any discrepant component shall not be used in a fuel

element assembly unless approved .

The supplier is to certify to compliance with the design dimensional requirements delineated in the Drawings referenced in Section 2.1.2.

All dimensions of finished fuel plates, fuel assemblies and fuel element containers apply at 75°F+/-5°F.

4.12.1 Final Dimensional Inspection.

Dimensions required by this specification and drawings of Section 2.1.2 shall be inspected llsing a purchaser approved sample plan and recorded on an inspection sheet with "in specification" dimensions recorded by check mark, "OK," or actual measurements and "out of specification" dimensions recorded as actual measurements.

4.13 Reactor Components and Spare Fuel Element Parts Reactor components and spare fuel element parts not assembled into fuel element assemblies are required to be certified. The certification shall consist of material certification, fabrication verification, and supplier certificate of compliance to the specification and drawing requirements. The certification documents shall be submitted to the purchaser and user.

PUR-1 SAR Appendix 3-34 Rev 0, 06/08

Form 412.09 (Re\'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 30 of 57 PURDUE UNIVERSITY REACTOR

5. PACKAGING AND SHIPPING Packaging and shipping of the fuel elements shall be performed using a Purchaser approved procedure in compliance with this section.

The purchaser shall provide shipping containers to protect the fuel elements from damage during shipment and which confonn to the applicable requirements of the Departments of Energy and Transportation, and other regulatory agencies having jurisdiction of the shipment of radioactive materials. Re-useable shipping containers will be returned to the Supplier by the User at the Purchaser's expense.

The Supplier is responsible for loading the fuel elements into shipping containers in a sealed polyethylene sleeve in a cleaned dry condition and free of extraneous materials.

The Supplier shall take necessary precautions during packaging to prevent damage to the fuel elements during shipment. Each container shall be provided with a tamper-proof seal. Loading and shipping documents for the container shall be prepared in accordance with the applicable regulatory requirements.

The Supplier shall make arrangements for shipment to the User.

6. NOTES 6.1 Definitions For the purpose of this specification, the following terms are identified:

Batch. The amount of silicide powder mixture which is handled as a unit or traceable to a common step.

Blended. To mix or mingle constituents of a batch.

Cert[ficatio11. The action of detennining, verifying and attesting in writing (signed by a qualified party) to the qualifications of personnel and material.

Cladding. The aluminum covers bonded to the fuel core and the picture frame.

Control Fuel Element Assembly. An assembly consisting of the control fuel element container with eight fuel plates.

Controlled Work Area. A work area to which access of personnel, tools, and materials is limited and physically controlled. Temporary enclosures may be used where adjacent activities produce contamination which is detrimental to the job .

PUR-1 SAR Appendix 3-35 Rev 0, 06/08

Form 412.(19 (Rev. (9)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I /24/07 Page: 31 of 57 PURDUE UNIVERSITY REACTOR Development. A determination of processes, equipment, and parameters required to produce a product in compliance with this specification.

Dogbone Area. Thickening of the fuel core usually in the last 1/2 in. of the core, which may result in clad thinning in those areas.

Dummy Fuel Element Assembly. An assembly consisting of a fuel element container with unfueled simulated dummy fuel plates.

Dummy Fuel Plate. A non-fueled plate made entirely from the aluminum material specified in this document.

Edge Clad. The distance between the edge of the fuel core and the edge of the finished fuel plate, before any stray particles are removed, in the width direction as determined by radiography of a flat fuel plate.

Failure. A condition where the fabrication process appears to be out of control or a breakdown or damage to equipment creates excessive costs andlor schedule

. delays .

Fuel Compact. A quantity of uranium silicide powder and aluminum powder, cold compacted by pressing into a solid block for assembly into packs for hot roll and cold roll into fuel plates. The compacts are encased in frames and cover plates to form the pack.

Fuel Assembly. An assembly of fuel plates and hardware components. This includes both the standard and control fuel elements.

Fuel Core. The uranium-bearing region of each Fuel Plate.

Fuel Plate. The Fuel Core complete with aluminum frame and cladding.

Graphite Reflector Assemblies. A component consisting of a graphite container assembly with a graphite block inside.

In-Process Controls. Inspections and tests made during production to ensure that the manufacturing processes, equipment, and personnel are producing a product meeting specified requirements.

Irradiation F acili~)I Assemblies. A component consisting of a round tube attached inside a graphite container assembly with graphite blocks filling the annulus between the tube and container. Inserted within the tube is the isotope capsule assemblies .

PUR-1 SAR Appendix 3-36 Rev 0, 06/08

Form 41 :!.09 (R~v. (9)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 32 of 57 PURDUE UNIVERSITY REACTOR Lot. A group of pieces handled as a unit or material traceable to a common processing step.

Mmn!facture(ing). All fabrication, assembly, test, inspection and quality control processes. Fabrication is a synonym for Manufacture.

Pack. The fuel compact, picture frame, and cover plates, assembled together for hot rolling.

Picture Frame. The window shaped aluminum frame, which holds the fuel compact.

Plates. See Fuel Plates.

Procedure. The detailed description of the series of processes during manufacture and inspection, which follow a regular definite order (not to be construed as an outline).

Production. That phase of the program, following Qualification, during which the product is in Manufacture.

Purchaser. Idaho National Laboratory (INL).

Qualification. A demonstration that the Manufacturing process, equipment and personnel can produce a Product in compliance with this Specification.

Quality Contra/. The sampling plans, inspections, tests and records required and used during Production to assure that the Product is in compliance with this Specification.

Rejection. Materials, paJ1s, components, or assembly products, which will not be accepted as fulfilling the contract requirements because of noncompliance with this Specification.

Requalification. A demonstration that a single or group of manufacturing processes, equipment and personnel can produce a product in compliance with this specification after the original qualification has been completed and becomes invalid.

Silicide. Uranium metal alloyed with silicon and fabricat'ed per the requirements of Specification TRTR-14. The word "fuel" is a synonym for Silicide.

Specification. All parts and appendixes to this document, its references, drawings, and standards, as may be modified from time to time by contractual document.

Standard Fuel Element Assembly. An assembly consisting of the fuel element container with fourteen (14) fuel plates.

PUR-1 SAR Appendix 3-37 Rev 0, 06/08

Form 412.09 (Rev. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 33 of 57 PURDUE UNIVERSITY REACTOR Stray Fuel Particles. Isolated fuel particles lying outside the maximum fuel core outline defined on Drawing 635463.

Supplier. The primary vendor selected by INL to manufacture the product.

User. Purdue University, at West Lafayette, Indiana.

6.2 Purchaser Tests None 6.3 Submittals The following data and records shall be supplied to the purchaser in the quantities stated. The purchaser's approval, prior to implementation, is required on those marked with an asterisk. All records and data shall be maintained by the supplier for the duration of the Purdue University fuel element contract.

The granting of approval by the purchaser of design, working drawings, specifications,requests, and other technical data submitted by the supplier under the provisions of the subcontract or specification shall not affect or relieve the supplier from such responsibility as the supplier has with respect to adequacy or correctness of the design, working drawings specifications, reports, and other technical data.

6.3.1 Preproduction

Documents requiring approval must be submitted prior to production use. The number of copies shall be as specified by the Vendor Data Schedule. These documents include:

A detailed description as to the weighing procedure by which the supplier proposes to assign Plate U-235 content as required in Section 4.3.

Included in the description must be sampling, analytical, and quality control procedures; a statement as to the established accuracy and precision of the assigned fuel plate and fuel element U-235 content; developmental and production data in support of the accuracy and precision estimate; and data which at the 95% confidence level, shows that the method used to assign U-235 values has a bias which is less than 0.2% relative

A detailed description as to the manner the supplier will use to verify the fuel Plate U-235 value as required by Section 4.3 PUR-1 SAR Appendix 3-38 Rev 0,06/08

Form -t 12.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effccti ve Date: 0 I124/07 Page: 34 of 57 PURDUE UNIVERSITY REACTOR

All fabrication, assembly, cleaning, surface treating, handling, and decontamination procedures (not to be construed as an outline)

All production test, inspection, and quality control procedures, including all nondestructive and destructive tests and all standards and sampling section drawings. All data from these tests, including but not limited to: radiographs, metallographic samples, ultrasonic testing traces, and qualification yield rates

All packaging, storage and shipping procedures

6.3.2 Pre-repair

All repair programs and procedures prior to use.

6.3.3 Manufacturing Schedule:

A schedule using a purchaser approved technique.

6.3.3.1 Reports .

I. Biweekly qualifications phase summary status report. The first such report shall be initiated I month after date of contract award.

2. Three (3) copies of a monthly report detailing program progress against a previously submitted schedule shall be supplied by the supplier to the purchaser. RepOIt type, fonnat and submittal schedule shall be as agreed upon between the purchaser and supplier.

6.3.4 Delivery Submittals:

Three copies (except as noted) of the following data and records shall be sent prior to or accompany the shipments. The supplier shall maintain copies of these records for at least 10 years and until the supplier has received written approval from the purchaser for disposition or disposal:

Certification of product compliance to the requirements of this specification to include any test data pertaining thereto Supplier's core compact data sheets, with individual fuel plate uranium composition data including:

Serial number with batch (sec def.) identification

PUR-1 SAR Appendix 3-39 Rev 0, 06/08

Form ~ 12.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY ldentifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 35 of 57 PURDUE UNIVERSITY REACTOR Uranium content Fuel plate core weight U-235 enrichment Total quantity U-235 content Core void density data lndividual fuel clement composition data, including:

Uranium content U-235 content Serial number of each plate in the element Radiation count from fuel plate and fuel element exterior as required by Section 3.5.3 and 4.11. The counting period, counter, background, efficiency, and type of counter used shall be reported List of all applicable waivers and deviations and related fuel plates or fuel clements If perfonned, documented evidence of the performance and test results of the boehmite fonnation from the fuel clement surface treatment per Section 3.8.

6.3.5 Fuel Plate Radiographs:

Fuel plate radiographs of all accepted fuel plates required by Sections 4.4 and 4.5 and Appendix A shall be sent to the user.

6.3.6 Core Compact Data Sheets:

Supplier's fuel core compact data sheets shall be supplied to the INL Quality Assurance Representative as they are generated.

6.3.7 Report of Production by Unapproved Process:

Whenever the supplier's previously submitted and approved process control limits are exceeded, or any material or fuel element components are fabricated using equipment, personnel, or processes which are not purchaser approved, the time, nature, description, corrective action to be taken, and proposed further corrective action shall be reported immediately by the supplier, with a written report to the purchaser to follow within 10 working days .

PUR-1 SAR Appendix 3-40 Rev 0,06/08

Form 412.(19 (Rev, (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Jdentifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: OJ/24/07 Page: 36 of 57 PURDUE UNIVERSITY REACTOR I. T Samples. Transverse to be taken equally space along Fuel Core length.

2. L Samples. Longitudinal to be taken at centerline and to include the Dogbone Area.

T1 T~ T,-,

~

T4

-~ -~ -~ -~

r - - - - - --- - - - r--- - -,

I I I

I

, + +-+-+ + , , I I

I I I

1, I I I I L --- --- - - --- r--- - ...J

-~ -~ -~ -~

Figure I. Purdue University Fuel Plate Sampling Procedures For Destructive Tests .

PUR-1 SAR Appendix 3-41 Rev 0,06/08

Form 412.09 (Rc\'o ()9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I124/07 Page: 37 of 57 PURDUE UNIVERSITY REACTOR APPENDIX A Requirements for Radiography of Purdue University Fuel Plates

1. Scope This specification provides requirements for radiography of Purdue University reactor fuel plates, acceptable film quality and film identification.

2. Requirements A procedure must be written to specify the details for achieving acceptable fuel plate radiographs. The procedure must include the requirements given in this specification.

2.1 Equipment Setup The voltage shall be 100 k. v .p. with a focal spot size of 5 mm maximum. The distance between the focal point and the plate shall be at least twice the length of the plate. The focal point shall be centered laterally and longitudinally over the plate or group of plates.

2.2 Film 2.2.1 The image outline shall be clear and sharp; the film shall be free of runs, streaks, scratches, blurs, and cassette defect that will affect the area covered by the fuel plates.

2.2.2 The film density of all points of the radiograph that correspond to the fuel plate border locations outside the plate core shall provide densitometer readings between 1.5 and 2.7. Film density as read over the nominal density standards shall provide densitometer readings between 0.9 and 1.5.

2.2.3 The film shall be extreme sensitivity, extra fine grain, high contrast, double emulsion, industrial x-ray type, (Kodak type M or equal) which is acceptable to the purchaser. Development of the film shall be in accordance with the manufacturer's recommendation.

2.2.4 Film Identification

PURc1 SAR Appendix A Appendix 3-42 Rev 0, 06/08

Form 412,(1) (Rev, ()t))

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I 124/07 Page: 38 of 57 PURDUE UNIVERSITY REACTOR 2.2.5 A system of identification of the film shall be provided by the supplier, which shall show as a minimum:

A. Plate lot number B. Plate type and serial number C. Orientation of density standard D. Density standard identification E. Date of radiography .

Appendix A PUR-1 SAR Appendix 3-43 Rev 0, 06/08

Form 412.0Y (Re\', (9)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I124/07 Page: 39 of 57 PURDUEUMVER~TYREACTOR APPENDIX B Welding Requirements and Qualification for Purdue University Fuel Elements

1. Scope The requirements for welding and for the evaluation of welds applicable to the Purdue University Fuel Element Container and components are established by this Appendix.

1.1 Application. This document defines requirements for the following:

1.1.1 Welding procedure qualification.

1.1.2 Performance qualification of welders, welding equipment, and special fixturing.

1.1.3 Information to be included in welding procedure specifications.

1.2 1.1.4 1.1.5 Application of qualified procedures to production welding.

Destructive testing and nondestructive examination for qualification and for production welding.

Special Limitations for Applicability. The requirements contained in this appendix are to some degree based on RDTF6-2T. Those requirements applicable to Manual, GTAW, single pass, welding of Plug Joint welds, Comer Joint welds, and Partial Penetration Butt Joint welds have been included in this appendix. The introduction of a new weld design or weld process requiring a change in these limited parameters would require an appropriate review of RDT F6-2T for requirements applicable to the new parameters.

1.3 Definitions.

Arc Strike. Any localized melting, heat affected zones, or change in the contour of the surface of the finished weld or adjacent base metal resulting from an arc or heat generated by the passage of electrical energy between the weld or base metal and a current source; such as welding electrodes, electron beams, ground clamps, high frequency arc, etc.

Automatic Welding. Welding with equipment which performs the entire welding operation without constant observation and adjustment of controls by an operator.

The equipment mayor may not perform the loading and unloading of the work.

Appendix B PUR-1 SAR Appendix 3-44 Rev 0, 06/08

Form 412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I124/07 Page: 40 of 57 PURDUE UNIVERSITY REACTOR Face of Weld. The exposed surface of a weld on the side from which welding was done.

Face Rei1?lorCe171ent. Reinforcement of weld at the side ofthe joint from which welding was done.

Heat. A single homogeneous melt of metal or alIoy.

Joint Penetration. The minimum depth a groove or flange weld extends from its face into a joint, exclusive of reinforcement.

Machine Welding. Welding with equipment which perfonns the welding operations under the constant observation and control of an operator. The equipment mayor may not perform the loading and unloading of the work.

Position of Welding. The tenns related to positions of welding for joint types and welding processes and the position limits are defined in Section IX, ASME Boiler and Pressure Vessel Code.

Repair. The process of restoring a nonconforming item characteristic to an acceptable condition, although it does not confom1 to a specified requirement.

Rework. The process by which a nonconforming item is made to confom1 to specified requirements.

Root of a Joint. That p011ion of a joint to be welded where the members approach closest to each other. In cross section the root of the joint may be a point, a line or an area.

Root ofa Weld. The points, as shown in cross section, at which the back of the weld intersects the base metal surfaces.

Root Penetration. The depth a groove weld extends into the root of a joint measured on the centerline of the root cross section.

Root Rei1?iorcement. Reinforcement of weld at the side opposite that from which welding was done.

Root SlIIiace. The exposed surface of a weld on the side opposite that from which welding was done.

Size ofa Groove Weld. The joint penetration (depth of chamfering plus root penetration when specified) .

PUR-1 SAR Appendix B Appendix 3-45 Rev 0,06/08

Fonn 412.0Y (Rcv. OY)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: 1 ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 41 of 57 PURDUE UNIVERSITY REACTOR Undercut. A groove melted into the base metal adjacent to the toe or root of a weld and left unfilled by weld metal.

Undel:fill. A depression on the face of the weld or root surface extending below the surface of the adjacent base metal.

Welder and Welding Operator Peljor177ance Qualification. The tests to demonstrate a welder's or welding operator's ability to produce welds meeting prescribed standards.

Welder. One who is capable of perfomling a manual or semiautomatic welding operation (sometimes erroneously used to denote a welding machine).

Welding Operator. One who operates machine or automatic welding equipment.

Welding Procedure Qualification. The test to demonstrate that welds made by a specified procedure can meet prescribed standards.

Welding Procedure Spectfication. A written welding procedure which specifies the detailed methods and practices to be used in the production of a weldment and how they shall be carried out. A specification includes all elements of a procedure necessary to produce a satisfactory weldment. Examples of some of the elements included in a specification are: material used, preparation of base materials, preheat and postheat cleaning, assembly method and sequence, fixturing, heat treatments, joint welding procedures, preweld and postweld nondestructive examinations, repair, rework, etc.

Welding Procedure. The detailed methods and practices including all joint welding procedures.

2. Reference Document The following documents are a part of this appendix to the extent specified herein. The issue of a document in effect on the date of the invitation to bid, including any amendments also in effect on that date, shall apply unless otherwise specified. Where this appendix appears to conflict with the requirements of a reference document, such conflict shall be brought to the attention of the purchaser for resolution.

Appendix B PUR-1 SAR Appendix 3-46 Rev 0,06/08

Form 412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: . SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 1/24/07 Page: 42 of 57 PURDUE UNIVERSITY REACTOR 2.1* American Society for Testing and Materials (ASTM) Standards 2.1.1 ASTM E2, Preparation of Micrographs of Metals and Alloys 2.1.2 ASTM E3, Preparation of Metallographic Specimen 2.2 American Society of Mechanical Engineers (ASME) Codes 2.2.1 ASME Boiler and Pressure Vessel Code,Section IX, Welding Qualifications 2.3 American Welding Society (A WS) Standards 2.3.1 A WS A2.2, Nondestructive Testing Symbols 2.3.2 AWS A3.0, Tenns and Definitions

3. Weld Qualification Requirements 3.1 General Requirements

PUR-1 SAR Appendix B Appendix 3-47 Rev 0,06/08

Form 412.09 (Re\'o (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 43 of 57 PURDUE UNIVERSITY REACTOR 3.1.1 All welding procedures, welders, or welding machine operators shall be qualified in accordance with the provisions identified in this Appendix.

3.1.2 Weld Procedure and Performance Qualification Testing previously qualified to these requirements under other contracts may be used.

Existing records to support previously qualified procedures and personnel are subject to review by the purchaser.

3.1.3 Base materials and filler material shall comply with the requirements of the drawings.

3.1.4 Welding processes which satisfy the specified requirements and produce the quality required by this Appendix are permissible. Welding procedures which utilize fluxes and coatings shall not be used.

3.1.5 Fixtures

The capability of fixtures for aligning parts shall be demonstrated before welding of production parts is initiated. If chilI bars or blocks are used, the type of material and their location with respect to the joint shall be included in the procedure specification .

3.1.6 3.1.7 Position of Qualification Welds. All procedure and performance qualification test welds shall be made in the same positions as for production welds.

Special Conditions for Qualification Welds: All procedure and performance qualification test welds shall be made under conditions which simulate the actual production welding conditions. These conditions shall include space limitations, joint accessibility, degree of comfort due to heat, position and other handicaps or environmental factors which the welder or welding operator will endure during actual production welding.

3.1.8 Heat Treatment. Weld preheatand postheat treatments shalI not be used without prior approval by the purchaser.

3.1.9 Interpass Temperature. For multi-pass weld, the weld interpass temperature shall not be less than 60° F or greater than 350° F without prior approval by the purchaser.

3.1.1 0 Records. Records of welding, associated processing, and inspection shall be maintained for all welds. Complete records may consist of inspection fonns, routings, or reference to Operating Procedures or other documents. These records shalI include at least the following:

Appendix B PUR-1 SAR Appendix 3-48 Rev 0, 06/08

Fonn ~ 12.09 (Re\'. (9)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 44 of 57 PURDUE UNIVERSITY REACTOR

1. Base Material (Type, material specification, heat or lot number).

2. Filler Material (Type, material specification, heat or lot number).

3. Cleaning procedures.

4. Joint identification and weld maps when applicable.

5. Welding machine type and identification.

6. Welding procedure specification.

7. Welder or welding operator qualification.

8. Procedure and perfomlance qualification.

9. Current-voltage data for machine or automatic welding.

10. Date welds are made.

11. Inert gas mixture, when applicable .

12. Nondestructive examination procedure.

13. Nondestructive examination personnel identification.

14. Examinations and tests (nondestructive and destructive) and the results.

15. Photomacrographs and photomicrographs.

16. Meta110graphic specimens.

17. If applicable, rework and repair of welds.

18. Disposition of welds.

3.2 Welding Procedure Specification 3.2.1 The welding procedure specification shall meet the requirements of this Appendix, and shall be submitted to the purchaser for infomlation.

3.2.2 The welding procedure specification shall include all essential elements and details, as required by this section, to cover each joint to be welded by the supplier. Each joint sha1l be identified in the welding procedure specification. The specification shall include a joint design sketch for Appendix B PUR-1 SAR Appendix 3-49 Rev 0, 06/08

Form 412.09 (Re\'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 45 of 57 PURDUE UNIVERSITY REACTOR each joint welding procedure even if the specification references drawing numbers.

3.2.3 The following basic information and essential variables shall be included in sufficient detail to assure that compliance with the requirements of the specification can be verified:

1. Basic Information

a. Joint Design: (the joint geometry, fit-up, and other required dimensions of the welded joint) tolerances and material thickness.

b. Method of arc initiation

c. Electrode size (for gas tungsten arc welding)

d. Gas type and flow rate (shielding and backing gas)

e. Welding current range for manual welding

f. Whether tack welds or fixtures are used for assembly of the joint for welding

g. Method and frequency of cleaning

h. Number of weld layers and passes I. Whether stringer beads or weave beads are used

2. Essential Variables

a. General, All Welding Processes.

I. A change from a base material type or grade (materials of the same nominal chemical analysis and mechanical property range, even though a different product form) to any other base material type or grade. When joints are made between two different types or grades of base material, a procedure qualification shall be made for the applicable combinations of materials, even though procedure qualification tests have been made for each of the two base materials welded to itself.

Appendix B PUR-1 SAR Appendix 3-50 Rev 0, 06/08

FOfm412.09 (Re\'o (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 46 of 57 PURDUE UNIVERSITY REACTOR II. A change of filler metal type or classification to any other type of classification 111. A change in welding position.

IV. A change in vertical welding direction, i.e., from upward to downward or vice versa.

v. The addition or omission of integral backing (e.g.,

"butt-lap" type joint).

VI. The addition or omission of non fusing metal retainers.

VII. The addition or omission of filler metal to the joint.

Vl1I. Any change in the method by which filler is added, such as prep laced shims, prep laced wire, prep laced consumable inserts, wire feed, or prior weld metal surfacing ("buttering") of one or both joint faces .

IX. The addition or omission or any type of prep laced consumable inserts or joint surfacing.

x. A change in the shape or size of preplaced consumable inserts or joint surfacing.

XI. A change from multiple pass welds to single pass welds.

XII. The omission of inert gas backing during welding, except that requalification is not required where a qualified welding procedure is changed to omit the inert gas backing and then is used only for a single welded butt joint with a backing strip, or a fillet weld. For multiple pass welding, the omission of inert gas backing during welding until three layers or 3/16 of weld metal thickness has been deposited, whichever is greatest.

XlII. A change from one welding process to any other process or combination of welding processes.

b. Manual Welding, All Welding Processes .

Appendix B PUR-1 SAR Appendix 3-51 Rev 0, 06/08

Form 412.09 lRcv. (9)

Idaho National Laboratorv SPECIFICA TION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I124/07 Page: 47 ofS7 PURDUE UNIVERSITY REACTOR

1. An increase in the standard size of filler metal from that stated and qualified in the procedure specification.

II. A change in joint geometry which violates the tolerances given for the joint geometry elements listed below:

Bevel Angle: State in procedure specification.

Tolerance: Minus 5%.

Groove Angle: State in procedure specification.

Tolerance: Minus 5%

Alignment Tolerance: Assign value in procedure specification. Qualify procedure for single welded joints using maximum permissible misalignment in a portion of the joint.

c. Gas Tungsten Arc Process.

I.

II.

A change of electrode material type.

A change in arc starting methods.

III. A change from a single shielding gas to any other shielding gas or to a mixture of shielding gases or a change in specified composition of gas mixture.

IV. A decrease in shielded gas flow rate of more than ten percent.

PUR-1 SAR Appendix B Appendix 3-52 Rev 0, 06/08

Funn -111.09 (Re\'. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 48 of 57 PURDUE UNIVERSITY REACTOR 3.3 Welding Procedure Qualification 3.3.1 The welding procedure shall be qualified to the requirements of this section.

3.3.2 All welding used in qualifying a welding procedure shall be performed in accordance with a welding procedure specification.

3.3.3 Before any welding is performed on production components, the supplier shall qualify each proposed welding procedure by:

1. Recording all essential elements of the welding procedure in a welding procedure specification (see Section 3.2)

2. Verifying the welding procedure specification by welding test specimens representing each joint to be welded in production and performing nondestructive examination and destructive tests in accordance with the requirements of this Appendix.

3. Submitting to the purchaser, for infom1ation, the welding procedure specification and a certified copy of the detailed results obtained from the tests performed on the test welds. The metallographic sections required by this Appendix shall also be submitted to the purchaser.

PUR-1 SAR Appendix B Appendix 3-53 Rev 0,06/08

Fnnn 412.09 (Rev_ 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 49 of 57 PURDUE UNIVERSITY REACTOR 3.3.4 Essential Variables. The welding procedure shaH be set us as a new welding procedure specification and shaH be completely re-qualified when any of the changes listed in SeCtion 3.2.3.2 are made in the procedure.

3.3.5 Chart Recordings. Current-voltage-time charts shaH be used for each procedure qualification weld for automatic or machine welding.

Calibrated current and voltage indicating meters may be substituted for trace chaJ1 type equipment for manual welding. The current and voltage ranges shall be recorded for manual welding.

3.4 Welder Perfonnance Qualification 3.4.1 Performance qualification weld tests shall meet the requirements of this section, except that any welder used to qualify the welding procedure shall also be considered qualified and additional performance weld tests are not required.

3.4.2 General.

1. The perfonnance qualification tests are intended to detennine the ability of welders to make sound welds.

2. The perfomlance test may be tenninated at any stage of the testing procedure whenever it becomes apparent to the supervisor conducting the tests that the welder does not have skill required to produce satisfactory results. In this event, the welder may be re-tested at the discretion of the supplier in accordance with 3.4.3.

3. Each supplier shall maintain a record of the procedures, including the essential variables, under which welders are examined and the results of the examinations.

3.4.3 Qualification of Welders.

I. Each welder shall pass the tests prescribed for procedure qualification except that tensile tests are not required. The essential variables and the test results obtained by each welder shall be recorded in a Performance Qualification Test Report. Any welder who performs acceptable welding procedure qualification tests shall be considered qualified.

2. Renewal of Qualification. Requalification of a welder is required

PUR-1 SAR when:

Appendix B Appendix 3-54 Rev 0, 06/08

Form 412.09 (Rev. 09)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 01124/07 Page: 50 of 57 PURDUE UNIVERSITY REACTOR

a. 90 or more days have elapsed since he last produced acceptable welds using the specific welding process, or

b. He has not perfol111ed acceptable welds using the production welding procedure.

c. Any time there is a specific reason to question a welder's ability to make welds meeting the requirements of this Appendix, requalification shall be required. Only one test weld shall be required for renewal of qualification. If this test weld fails to meet all of the original requirements, then a complete performance requalification shall be required.

3.4.4 Chart Recordings. Curi'ent-voltage-time charts shall be used for each procedure and perfomlance qualification weld for automatic or machine welding. Calibrated cunent and voltage indicating meters may be substituted for trace chart type equipment for manual welding. The current and voltage ranges shall be recorded for manual welding.

3.5 Welding Machine Qualification 3.5.1 Performance qualification weld tests shall meet the requirements of this section, except that any welding machine used to qualify the welding procedure shall also be considered qualified and additional performance weld tests are not required.

1. The perfomlance qualification tests are intended to determine the ability of welding machines to make sound welds.

2. Any time there is a specific reason to question a welding machine's ability to make welds meeting the requirements of this Appendix, requalification shaH be required. Only one test weld shaH be required for renewal of qualification. If this test weld fails to meet all of the original requirements, then a complete performance requalification shall be required. Welding machines used for the manual welding of any successful procedure or welder qualification tests shall be considered qualified for manual welding of all core components covered in this Appendix.

PUR-1 SAR Appendix 8 Appendix 3-55 Rev 0,06/08

Fllnll ~ 12.09 (Re\'. OY)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision: J ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I 124/07 Page: 5 J of 57 PURDUE UNIVERSITY REACTOR 3.5.2 Welding machines used for the manual welding of any successful welder perfonnance qualification tests shall be considered qualified for manual welding of all components covered in this session.

3.6 Examination & Tests 3.6.1 Type of Test Required. The following tests shall be used for the qualification of welding procedures and / or welders as applicable:

I. Nondestructive examination by a liquid penetrant method.

2. Nondestructive examination by Visual to test for soundness and surface characteristics of the weld.

3. Destructive examination by sectioning for metallographic examination of weld joints and adjacent areas to test for fusion, weld geometry, weld reinforcement, and soundness of the weld.

4. When the purchaser has reason to believe that the quality of any weldment is doubtful, he may require additional inspection .

5. Nondestructive Examination and Tests

a. Visual. The test weld shall be examined visually prior to welding and after welding in accordance with Section 5.1

b. Liquid Penetrant. The test weld shall be examined after the final layer in accordance with Section 3.6.2.2 using a color contrast method.

c. Unless otherwise specified, inspection of procedure and performance qualification welds shall be perfomled in the final surface condition.

6. Destructive Examination. Each test weld shall be sectioned transversely to metallographically examine a minimum of:

a. Three section faces for welds on cylindrical components less than I ~ inch in diameter or for welds that are one to four inches long on non-cylindrical components.

b. Four sections faces for welds in cylindrical components that are greater than ll~ inch in diameter or for welds that are greater than four inches long on non-cylindrical components.

Appendix B PUR-1 SAR Appendix 3-56 Rev 0, 06/08

Fonn 412.09 (Re\', (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Jdcntificr: SPC-382 STANDARD AND CONTROL FUEL Rcvision:

ELEMENTS - ASSEMBLED FOR THE Effcctive Date: 01/24/07 Pagc: 52 of 57 PURDUE UNIVERSITY REACTOR

c. One section face for plug welds, arc spot welds, and welds that are less than one inch long on non-cylindrical components.

d. The cross section shall be polished and etched to provide clear definition of the structure in the fusion zone and the heat-affected zones.

e. For welds in (a) and (b) of this paragraph, one cross section shall be made through a weld start and a weld stop area and the remaining sections shall be made at random. For weld described in (c) of this paragraph, the cross section shall be made at the approximate centerline of the weld.

Examination of the welds shall be in accordance with Section 3.6.2.3.

3.6.2 Acceptance Criteria for Qualification Test Welds I. Visual Examination. Visual examination shall be in accordance with Section 5.1.

2. Liquid Penetrant Examination. Unless otherwise specified, final weld surfaces shall be examined using a color contrast method.

a. For welded joints in materials less than 1/8 inch thick the following relevant indications are unacceptable.

I. Any cracks.

II. Linear indications.

III. Indications with dimensions exceeding 1/64 inch.

IV. Rounded indication separated by li4 inch or less edge-to-edge.

v. Five or more rounded indications in any six square inches of weld surface with the major dimension of this area not to exceed six inches with the area being taken in the most unfavorable location relative to the indication being evaluated.

b. For all welds in materials 118 inch thick or greater, the following relevant indications are unacceptable. (Only Appendix B PUR-1 SAR Appendix 3-57 Rev 0,06/08

Furm 412.(1) (Re\', ()9)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 0 I 124/07 Page: 53 of 57 PURDUE UNIVERSITY REACTOR those indications with major dimensions greater than 1/64 inch are considered relevant for item (iii).)

I. Any cracks.

II. Any linear indications.

III. Rounded indications with dimensions exceeding 10 percent of the nominal weld thickness or 1/8 inch, whichever is smaller. Rounded indications separated by 1I 16 or less edge-to-edge shall be evaluated as a single indication.

IV. Four or more rounded indications in a line separated by 1116 inch or less edge-to-edge.

v. Six or more indications in any six square inches of weld surface with the major dimension of this area not to exceed six inches with the area taken in the most unfavorable location relative to the indications being evaluated.

VI. Aligned indications in which the average of the center-to-center distance between anyone indication and the two adjacent indications in a straight line is less than 3/16 inch.

3. Metallographic Examination Metallographic examinations shall be performed on qualification test welds at not less than SOX on test welds as required in this Section in accordance with ASTM E.2.

Any cross section which is shown by metallographic examination to contain any of the following relevant defects shall be cause for rejection of the test welds.

a. Any cracks.

b. Incomplete fusion, or insufficient joint or root penetration.

c. Any tungsten inclusions, slag inclusions, or porosity having a maximum dimension greater than 20 percent of the weld thickness or 1132 inch, whichever is smaller.

d. More than four tungsten inclusions or pores which have a maximum dimension less than in (c) above .

Appendix B PUR-1 SAR Appendix 3-58 Rev 0,06/08

Form 412.0~ ~Re\'. O~)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE* Effective Date: 0 1/24/07 Page: 54 of 57 PURDUE UNIVERSITY REACTOR

e. Any deviation from specified weld geometry or weld reinforcement.

3.6.3 Test Welds.

1. Procedure and 1 or welder performance qualification shall be made on test welds which duplicate the production weld joint type and which simulate the conditions to be used in production with respect to orientation, the essential variables listed in Section 3.2.3.2, and the dimensions of the pm1s to be joined to the extent that they affect heat requirements, relative motions, and distortions. All welding used in qualifying a welding procedure and lor welder performance shall be performed in accordance with the procedure specification.

2. For manual welding, two consecutive test welds shall be made when the weld joint is less than six inches in length. Only one test weld shall be required when the weld joint is 6 inches or greater in length .

3. All test welds shall be tested using the required tests listed in Section 3.6.1. To qualify the procedure specification used in making the test welds, each weld shall pass the required tests.

4. Repair of procedure or perfonnance qualification test weld(s) is prohibited.

4. Production Welding All production welding shall be accomplished using approved welding procedure specifications and qualified welders andlor welding operators.

5. Quality Acceptance of Production Welds 5.1 All completed production welds shall be visually examined in accordance with the following requirements:

5.1.1 General Visual Inspection Requirements. All visual examinations shall comply with the following:

1. Visual examination shall be made under direct daylight-type fluorescent lighting of at least 100 foot-candles at the work examination area .

PUR-1 SAR Appendix B Appendix 3-59 Rev 0,06/08

Form 412.(1) (Rev. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 55 of 57 PURDUE UNIVERSITY REACTOR

2. Visual examination shall be performed with the aid of a 5x (minimum) magnifying glass.

3. The inspection required by this standard shall not be performed by the welder who made the welds. However, if the welder is qualified in accordance with this standard, he may visually inspect his own welds prior to the inspections required by this standard.

4. Personnel performing visual inspection shall have 20-20 vision, natural or corrected, stereo acuity, and shall not be color-blind.

5.1.2 Visual Acceptance Criteria (except for porosity). Visual examination of weld joint preparations and welds shall be perfomled in accordance with the following requirements to verify conformance to the written welding procedure, the design requirements, and the requirements of this standard:

1. Prior to welding, the weld joint edges and adjacent surfaces shall be examined for:

a. Proper edge preparation, dimensions, and finish.

b. Alignment and fitup of the pieces being welded.

c. Verification of correct material by check of records.

d. Verification of the cleanliness requirements ..

2. After welding, the joint shall be examined in the final surface condition for:

a. Contour, reinforcement and surface finish of welds.

b. Degree of underfill, undercut, and overlap.

c. Arc strikes, weld spatter and impression marking.

d. Bum-through and fuse-through

3. Weld joints and surfaces which are shown by visual examination to have any of the following defects or areas of nonconformance are unacceptable:

a. Any nonconformance revealed by 5.1.2.1 .

Appendix B PUR-1 SAR Appendix 3-60 Rev 0, 06/08

Form 412.09 (Re\'. (9)

Idaho National Laboratory SPECIFICATION FOR PURDUE UNIVERSITY Identifier: SPC-382 ST ANDARD AND CONTROL FUEL Revision: I ELEMENTS - ASSEMBLED FOR THE Effective Date: 01/24/07 Page: 56 of 57 PURDUE UNIVERSITY REACTOR

b. Any zone of incomplete fusion.

c. Insufficient joint or root penetration.

d. Any undercutting, underfill, or bum through.

e. Any concavity on the face side of groove welds.

f. Any arc strikes, weld spatter, and impression marking.

g. Any visible inclusions, porosity, cracks, and unfilled craters.

4. Machined welds shall meet the drawing requirements.

5. All welds shall be free from surface markings resulting from mishandling, punching, scratching, etc., which exceed the specified surface requirements.

6. All welds shall be free of dross, or slag .

7. All welds shall be free of oxidation due to improper shielding and overheating which produce black or gray spalling or loose particles. Iridescent temper films and the dark metallic vapor deposits which may occur adjacent to the welds are acceptable.

These films and deposits shall be removed by approved cleaning procedures when accessible.

5.1.3 Visible unacceptable porosity is as follows:

1. Four or more pores with a major dimension of 0.048 inches or more randomly positioned.

2. A single pore with a major dimension of 0.064 inches or more.

3. Six or more pores with a major dimension of 0.016 inches or greater in one weld.

4. Four or more porosity with a major dimension of 0.0 16 inches or greater, in line separated by less than 0.063 inches from edge to edge.

5.2 Repair of a defective weld by welding shall be limited to two attempts.

Unacceptable defects shall be removed and re-examination made using liquid penetrant color contrast method to assure complete removal of the defect. If the Appendix B PUR-1 SAR Appendix 3-61 Rev 0, 06/08

FOnl1412.(l9 (Rev, (9)

Idaho National Laboratorv SPECIFICATION FOR PURDUE UNIVERSITY Jdentifier: SPC-382 STANDARD AND CONTROL FUEL Revision:

ELEMENTS - ASSEMBLED FOR THE Effective Date: 0]/24/07 Page: 57 of 57 PURDUE UNIVERSITY REACTOR removal of the defects results in reducing the thickness of the weld metal below the thickness of the base metal, the area shall be rewelded using a welding repair procedure which has been approved by the Purchaser. Whenever a defect is removed and subsequent repair by. welding is not required, the excavated area shall be blended into the surrounding surface to remove any sharp notches, crevices or comers. Completed repairs shall be visually re-examined per Section 5.1. Records shall be maintained on all repairs and shall include the following:

5.2.1 Location of joint.

5.2.2 Location of defect.

5.2.3 Description of defect, including type and size.

5.2.4 Reference to approved repair procedure.

5.2.5 Inspections before and after repair and the results thereof.

5.2.6 Identification of repair welders or welding operators .

PUR-1 SAR Appendix B Appendix 3-62 Rev 0,06/08

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OPERATOR REQUALIFICATION PROGRAM for the.-

PUR-1 REACTOR FAC.ILlTY This program is designed to comply with the:intent of 10 CFR 55, AppeAdix A, concerning the continued training. and requalification of operators for the PUR-1 reactor. It will be mandatory for all operators licensed-on the PUR-1 reactor to participate in the program.

The requalification program consists of the following parts:

A. INSTRUCTION A series of eight meetings will be held over a two year period, during which all topics listed below in part A.1.b will be covered .

. 1. Each meeting will consist of:

a. A review of reactor operations and modifications, if any.

b. A lecture of one or more of the following topics:

i. Theory and principles of operations.

II. General and specific plant operating characteristics.

iii. Plant instrumentation and control systems .

iv. Plant protection systems.

v. Engineered safety systems.

vi. Normal, abnormal, and emergency operating procedures.

vii. Radiation control and safety.

viii. Technical specifications.

ix. Applicable portions of Title 10, Chapter I, Code of Federal Regulations.

2. The lectures will be given by the reactor operators, senior operators, health physics personnel, faculty members of the School of Nuclear Engineering, or other qualified personnel as determined by the Laboratory Director.

B. PROGRAM EVALUATION Completion of the biennial operator requalification program will consist of a written examination and a demonstration of operator proficiency in reactor operation.

1. Written examination:

a. One of the senior operators will be exempt from taking the examination. This senior* operator will make up and administer the

examination to all other operators and senior operators. The senior operator may receive assistance for making up questions on the topics PUR-1 Operator Requalification Program Page 1 of 6 June 25, 2008

in part A.1.b from the instructor for each topic. The senior operator exemption will rotate through the entire senior operator roster.

b. The written examination for requalifying licensees will contain representative questions measuring. the knowledge, skills and abilities needed to perform licensed duties. These,will be identified from the licensed operator's duties performed, information in the Safety Analysis Report, operating procedures, facility license and amendments, License Events Reports, and"any other information requested from the facility licensee by the NRC.

The representative questions for the operator's examination will sample the following topics:

i. Fundamentals of reactor theory including"the fission process, neutron multiplication, source effects, control rod effects, criticality indications, reactivity coefficients, and poison effects.

II. General design features of the core, fuel assemblies, control rods, core instrumentation, and coolant flow.

iii. Mechanical components and design features of the reactor coolant system.

iv. Auxiliary systems that affect the facility.

v . Facility operating characteristics during steady state and transient conditions.

vi. Design, components, and functions of reactivity control mechanisms and instrumentation.

VII. Design, components, and functions of control and safety systems including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

viii. Components, capability, and functions of emergency systems.

ix. Shielding, isolation, and containment design features, including access limitations.

x. Administrative, normal, abnormal, and emergency operating procedures.

xi. Purpose and operation of radiation monitoring systems, including alarms and survey equipment xii. Radiological safety principles and procedures.

xiii. Procedures and equipment available for handling and disposal of radioactive materials.

c. Representative questions for the senior operators examination will sample the topics in the operators list and in addition will sample-the following list:

i. Conditions and limitations in the facility license.

PUR-1 Operator Requalification Program Page 2 of6 June 25, 2008

ii. Facility operating limitations in the technical specifications and their bases.

iii. Licensee procedures required to obtain authority for design and operating changes.

IV. Radiation hazards that may arise during normal and abnormal situations including maintenance activities and various contamination conditions.

v. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

vi. Procedures and limitations involved in initial core loaeing, alterations in core configuration, and determination of various internal and external effects on core reactivity.

vii. Fuel handling facilities and procedures.

d. Any person who scores less than 70%, overall, on the examination will be relieved from licensed duties and enrolled in an accelerated.

program until such time as they can satisfactorily pass an examination covering the material. The course content and duration will depend upon the individual's deficiencies.

2. Operator proficiency:

a. The exempt senior operator will also administer an operator proficiency examination to all other operators and senior operators.

b. The content of the operating test will be identified from duties of the licensed operator/senior operator and reference documents listed in Part B.1.b

c. The operations test for the requalifying licensee will demonstrate an understanding of and the ability to accomplish a representative sample of the following items:

i. Perform the prestartup procedures.

ii. Manipulate the console controls as required to operate the facility between shutdown and designed power levels.

iii. Identify annunciators and condition-indicating signals and perform appropriate remedial actions where appropriate.

iv. Identify the instrumentation systems and the significance of facility instrument readings.

v. Observe and safely control the operating behavior characteristics of the facility.

vi. Perform control manipulations required to obtain desired operating results during normal, abnormal, and emergency situations .

vii. Safely operate the facilities auxiliary and emergency systems.

PUR-1 Operator Requalification Program Page 30f6 June 25, 2008

viii. Demonstrate or describe the. use and function of the facilities radiation monitoring systems, including fixed radiation monitors and alarms, portable survey instruments, and personal monitoring equipment.

ix. Demonstrate knowledge of significant radiation hazards .and the ability to perform procedures to reduce elevated levels of radiation and to maintain exposures .as low as' reasonably achievable (ALARA).

x. Demonstrate.knowledge of the emergency plan, including, as' appropriate, the operator's or senior operator's responsibility to decide whether the plan should be executed and the duties assigned under the plan.

xi. Demonstrate the knowledge and ability, as appropriate to the assigned position, to assume the responsibilities associated with the safe operation of the facility.

xii. Demonstrate the ability to follow all procedures in a way that conforms to all license limits, amendments, and applicable conditions and specifications.

d. Any person who can not demonstrate proficient operation of the reactor will be relieved of his licensed duties until such time as proficient operation could be demonstrated. Proficient operation may

be established by performing a minimum of six hours of supervised reactor operations and demonstrating proficiency of section B.2.

C.' ON THE JOB TRAINING

1. Each licensed operator in the requalification program may at the option of the exempt senior operator, be required to make 8 reactor startups, shutdowns, or power level changes during the two year period covered by the program.

2. Each licensed operator at the facility will manipulate the plant controls, and each licensed senior operator will either manipulate the plant controls or direct the activities of individuals during plant control manipulations during the term of the operators/senior operator's license. Manipulations by operators/senior operators must consist of the following activities:

a.' Completed annually.

i. Plant shutdown.

ii. Significant power changes (>1 0%).

iii. Loss of coolant. (Not considered credible) iv. Loss of electrical power.

v. Loss of coolant flow. (Not considered credible)

b. Completed on a.two-year cycle.

i. Loss of protective system channel.

ii. Mispositioned control rod or rods.

PUR-1 Operator Requalification Program Page 4of6 June 25, 2008

iii.

iv.

v.

vi.

vii.

Inability to drive control rods .

Conditions requiring use of emergency boration Fuel cladding failure or high activity in reactor coolant.

Failure of servo system.

Reactor trip.

viii. Failure of nuclear instrumentation.

Note: When the control panel of the facility is used for training, the action taken or to be taken for the emergency or abnormal condition may be discussed; actual manipulation of the controls is not required.

10 CFR 55.59 (c) (4) (iv).

3. Each licensed operator at the facility will perform the function of the license held. An SRO is credited with performing the function any time the operator is on call, instructing classes, student, or student operator in training, inside the reactor room with the key on, or maintaining custody of the key. Additionally unstructured activities such as participation in facility-related design and safety review groups, Emergency Plan, emergency drill, Committee on Reactor Operations (CORO) participation, experimental activities, related technical presentations, performing security related functions, and performing maintenance and calibration activities contribute to training in all parts of the program except parts B.2, C, D and E. A statement to the file is sufficient to

document the training and/or time accounting .

D. LITERATURE REVIEW Each reactor operator and senior operator will annually review the contents of the operating manual, technical specifications, and the emergency procedures. A statement to this fact will be kept in the requalification file.

E. RECORDS Records will be maintained to document each instructor,' each topic discussed, each licensed operator's and senior operator's participation in the requalification program. The records will contain copies of each written exam, answer sheets, results of evaluation, and the biennial operator proficiency demonstration. Documentation of additional training and test required for individuals exhibiting deficiencies will also be included in the files. All records of the requalification program will be retained .by the as determined by the Laboratory Director until the licenses of the participants are renewed.

F. EXEMPTIONS During intervals when the licensed operations crew consists only of senior operators who are instructors for topics in part A.1.b., the requalification program will be modified to exempt those senior operators from parts A and B: 1. Parts B.2, C, D, and E will remain in effect.

PUR-1 Operator Requalification Program Page 5 of 6 June 25, 2008

When the licensed operations crew increases to include those who do not instruct in the program, the program will revert to its initial content Operators may place a statement into the file stating that they have done a literature review and/or instructed the topics 'in section A and B.1 in lieu of meetings',

and exams.

During intervals when the licensed operations crew consists of only.one" senior operator this operator will be exempt from parts A*andH, part C would be documented in the*console log book and as stated in C.3,**parts D'and E-will remain in effect In any of the above requalification activities exclusive of operations; additional methods may be used to accomplish the training requirement These may include mail, electronic classroom or other methods may be used for training, meetings, testing or other-required communication(s).

- PUR-1 Operator Requalification Program Page 6 ot6 June 25, 2008

ML111890201: Difference between revisions

Revision as of 19:25, 10 March 2020

Contents

Text

Response

Response

Response

Response

Response

Response

Title:

SUMMARY

3.1.3 Requalification

3.5.2 Cleanliness

3.5.3 Contmnination

3.9.1 Material

3.9.2 Asselnbly

3.9.3 Welding

3.9.4 Identification

3.9.7 Storage

6.3.1 Preproduction

6.3.2 Pre-repair

SUMMARY

3.1.3 Requalification

3.5.2 Cleanliness

3.5.3 Contamination

3.9.1 Material

3.9.3 Welding

3.9.4 Identification

3.9.7 Storage

6.3.1 Preproduction

6.3.2 Pre-repair

3.1.5 Fixtures

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ML111890201: Difference between revisions

Revision as of 19:25, 10 March 2020

Text

Response

Response

Response

Response

Response

Response

Title:

SUMMARY

3.1.3 Requalification

3.5.2 Cleanliness

3.5.3 Contmnination

3.9.1 Material

3.9.2 Asselnbly

3.9.3 Welding

3.9.4 Identification

3.9.7 Storage

6.3.1 Preproduction

6.3.2 Pre-repair

SUMMARY

3.1.3 Requalification

3.5.2 Cleanliness

3.5.3 Contamination

3.9.1 Material

3.9.3 Welding

3.9.4 Identification

3.9.7 Storage

6.3.1 Preproduction

6.3.2 Pre-repair

3.1.5 Fixtures

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