ML20214M904
| ML20214M904 | |
| Person / Time | |
|---|---|
| Site: | 05000134 |
| Issue date: | 01/30/1987 |
| From: | Freese K, Matos J ARGONNE NATIONAL LABORATORY |
| To: | |
| Shared Package | |
| ML20214M890 | List: |
| References | |
| NUDOCS 8706010463 | |
| Download: ML20214M904 (102) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ - _ _ _ ~ DRAFT ANL CALCULATED RESULTS FOR THE WPI HEU AND REFERENCE LEU CORES I l l J. E. t1atos and K. E. Freese Argonne National Laboratory Argonne, IL 60439 l l l 1 hDR DO Sob 34 P PDR
Table 1. Summary of Core Physics Parameters for WPI HEU HEU LEU Reference Calculated Calculated Type Fuel UAl Alloy UAl Alloy UAlx -Al Enrichment, % 93.4 93.4 19.75 Number of Fuel Plates 10 10 18 Meat Thickness, em 0.991 0.991 0.762 i Cladding Thickness, um 0.762 0.762 0.381 235 U per Element, g 136.3 136.3 167.0 U Density, g/cm3 0.39 0.39 1.78 l Prompt' Neutron Lifetime, ys 82.0 69.0 61.1 Delayed Neutron Fraction - 0.0077 0.0077
~
Temperature Coefficient, '
% 6k/k per *C -0.005 -0.0172 -0.0163 Void Coefficient, % 6k/k per percent void -0.2 -0.223 -0.250 Excess Reactivity, % 6k/k all blades out 0.23-0.26 0.30 0.06 Total Worth of 3 Safety Blades, % 6k/k 14-18 .11.6 11.6 Worth of Regulating Blade,
% 6k/k 0.7 0.7 0.7 j Limiting Step Reactivity to
! Initiate Clad Helting, $ - 1.93a 2.63a 1.99 D - a Melting point of 6061 Al cladding is 582*C. b Melting point of " pure" (1100 or 2S) Al cladding is 660*C. 4 l
)
1
. l 1 l Energy Group Boundaries and Average Fluxes ) for the HEU and Reference LEU Cores The calculations were performed using a detailed three-dimensional model 1 of the reactor with ten group cross sections. The sideplates and fuel-bearing zones of each fuel element were modeled separately, along with explicit representations of the grid box, control blades, control blade guides, thermal column, and water gap between the grid box and the thermal column container. Energy Group Boundaries Group Upper Lower A Enerav Enerav 1 10.0 MeV 0.639 MeV 2 0.639 MeV 9.119 kev 3 9.119 kev 5.531 kev 4 5.531 kev 1.855 eV 5 1.855 eV 1.166 eV 6 1.166 eV 0.625 eV
, 7 0.625 eV 0.417 eV 8 0.417 eV 0.146 eV 9 0.146 eV 0.057 eV 10 0.057 eV 2.53 x 10-4 eV Note that there are four energy groups below 0.625 eV.
Core Average Fluxes (nv x 10 10) Energy Boundary ljEQ LEQ
< 0.625 eV 8.91 7.19
/
2 0.625 eV 18.0 18.4 10.639 MeV 6.73 6.86 As you can see from the energy group boundaries, we did not compute fluxes greater than 1 MeV in these calculations. The upper energy group that we have (Group 1) provides fluxes 10.639 MeV.
Isothermal Reactivity Feedback Coefficients HEU Core and LEU Reference Core Change of Water Temperature Only Temp. of Change in Core Core Waters, Excess Reactivity,
'C % Ak/k E M 20 0.0 0.0 38 -0.3086 - 0.2917 50 - 0.5155 - 0.4869 75 - 0.9491 -0.8957 100 - 1.3866 -1.3073
- Water density held constant at 0.998 g/cm3 (20 *C)
Change of Water Density Only (Vold Coefficient) Density of Equivalent Temp. Change in Core i Core Water *, of Core Water, Excess Reactivity, g/cm3 *C % Ak/k E M 0.998 20 0.0 0.0 O.993 38 - 0.1006 - 0.1209 0.988 50 - 0.2024 - 0.2430 0.975 75 - 0.4732 - 0.5656 0.958 100 - 0.8416 - 0.9994 0.900 100 - 2.2325 - 2.6041 0.800 100 - 5.1889 - 5.8769
- Water temperature held constant at 20 *C (0.998 g/cm3) l I
,- , _ _ . _ _ _ _ _ . _ _ . _ . _ _ ,-, ..-. r --_ - . . _ _ . , , ,. ._ , __ _ _ . , . _ _, _ . , _ _ . _ , - .
Isothermal Reactivity Feedback Coefficients HEU Core and LEU Reference Core l I Change of Fuel Temperature Only (Doppler Coefficient) l Temp. of Change in Core Fuel Meat *, Excess Reactivity,
*C % Ak/k E.U. LEl.
20 0.0 0.0 38 - 0.0007 - 0.0327 50 - 0.0010 - 0.0541 75 - 0.0015 - 0.0977 100 - 0.0018 - 0.1399 200 - 0.0026 -0.2980 l
- Water density held constant at 0.998 g/cm3 (20 *C).
Water temperature held constant at 20 *C (0.998 g/cm3).
SUMMARY
OF Basic Kinetics Parameters and Isothermal Reactivity Feedback Coefficients Parameter gg Lg.Q Prompt Neutron Generation Time, ps 69.0 61.2 Effective Delayed Neutron Fraction, % 0.7678 0.7697 Water Temperature Only - 0.1724 -0.1627 ( Ak/k/AT) x 10-3 /*C Water Density Only (38 - 50 *C) - 0.0848 - 0.1018 Fuel Temperature Only - 0.0003 - 0.0178 Vold Coefficient, (Ak/k)/ % Vold (0-10%) - 0.0022 - 0.0026
Sensitivity Calculations for WP1 LEU Core Reference Core: 24 Fuel Elements 167 g % per Element ' 10 ppm Nat. Boron Equivalent in 6061 A1. Flux Trap in Grid Position F3 ppm Excess Grams D5U Boron Reactivity, Der Element Eauiv. I Ak/k Sensitivity to g 55U per Element 166 10 - 0.111 167 10 0.013 Reference Core l 168 10 0.135 1 1 Sensitivity to ppm Nat. Boron Equivalent in 6061 Al ' 167 10 0.013 Reference Core 167 20 - 0.363 IG7 30 - 0.738 Sensitivity to Flux Trap Grid po'sition i 167 10 0.013 Fuel in 83 Flux Trap in F3 I 167 10 0.371 Fuel in F3 I Flux Trap in B3
\
l l 1 l l
Sensitivity Calculations for WPI LEU Core Reference Core Containing Element with Removeable Plates in Grid Position F3 Number of Excess Plates in g Z350 in Reactivity, Position F3 Position F3 % ak/k 0 0.00 0.013 Reference Core 3 27.84 0.263 i 6 55.68 0.450 9 83.52 0.618 12 111.36 0.767 15 139.20 0.897 18 167.00 1.006' l
- Excess reactivity for a 25 element core.
Combined with data in previous table for the 24 element core with fuel in position F3 and flux trap in position B3, the reactivity worth of adding 18 fuel plates into position B3 would be (1.006 - 0.371) = 0.635'% Ak/k.
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Reference Core Excess Reactivity for Various Blade Configuratfons Core Reactivity Worth Blade Configuration Excess Reactivity, of Blades,
% auk % A Uk Height
- o'f Height
- of 3 Safety Regulating Reference Blades Higdi HEU Core LEU Core HEU Core LEU Core 30' 30' O.303 0.064 0.0 0.0
- l' 30' - 11.28 - 11.5 0 - 11.58 - 11.56
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- 11.69 30' - l' - 0.026 - 0.498 0.563 0.562 30' - 2.5" -
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0.666 30' - 4' - 0.394 - 0.631 0.697 0.695 Reactivity Profile of Regulating Blade for the LEU Reference Core with the Three Safety Blades Withdrawn Core Worth of Height
- of Excess ~ Regulating Regulating Reactivity, Blade Blade % Ak/k % Ak/k
- 4' - 0.631 0.695
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- l' - 0.498 0.562 0* - 0.496 0.560 11.75' - 0.227 0.291 23.5" + 0.060 0.004 30' + 0.064 0.0
- Relative to bottom of fuel meat.
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i Safety Analysis Report for the Worcester Polytechnic Institute Reactor for Presentation to The United States k;p
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Nuclear Regulatory Comission September 25, 1979 i
I TABLE.0F CONTENTS Page 1 SECTION 1 - GENERAL INFORMATION AND REQUESTED ACTION '
SUMMARY
9 SECTION 2.- , Introduction 9 2.1 2.2 General Description of Reactor Facility 9 10 2.3 Reactor Data 11 2.4 Experimental Facilities 12 SECTION 3 - REACTOR FACILITY 12 31 Location - i 3 1.1 Site Description f 3 1.2 Population Distribution ; 3.1.3 Climatology ! 3 1.4 Geology. Hydrology and Drainage 3.1.5 Seismology 15 32 Building 3.2.1 Building Layout 3.2.2 Ventilation 17 SECTION 4 - REACTOR FACILITY DESCRIPTION 17 4.1 Reactor Description 4.1.0 General 4.1.1 Fuel Elements
'4.1.2 Control Elements
; g- 4.1 3 Regulating Element
' (;, 4.1.4 Drive Mechanisms 4.1.5 Grid Box 4.1.6 Suspension System 4.1.7 Locating Plate 4.1.8 Ion Chamber Support Assembly 4.1.9 Startup Counter Assembly 4.1.10 Pool 4.2 Experimental Facilities 20 4.2.1 Thermal Column i
4.2.2 Beam Port ' 4.2.3 Spare Fuel Elements 22 l 4.3 Control and Instrumentation , 4.3 1 General 4.3 2 Control Center Cubicle 4.3 3 Control Power Distribution System 4.3 4 Auxiliary Control Power Supply 4.3.5 Instrumentation Power Supply 4.3.6 Flux Level Safety Channels 4.3.7 Log N-Period Channel 4.3.8 Startup Channel 439 Area Radiation Monitors 4.3.10 Safety Blade Control 1 4.3.11 Regulating Blade Drive System I 4.3 12 Scram Circuits 4.3.13 Alarm and Trouble Monitor System '
'4 4.4 Neutron Source 26 4.5 Shielding 27 I
TABLE OF CONTENTS (continued) Page 4 28 1', SECTION 5 - REACTOR OPERATING CHARACTERISTICS 28 5.1 Introduction 28 5.2 Nuclear Characteristics 5.2.1 Critical Mass and Loading 5.2.2 Neutron and Gamma Flux 5.2.3 Control System
- 5. 2. 4 Bur n-up 5.2.5 Temperature and Void Coefficient 5.2.6 Neutron Lifettne 5.2.7 Alteration of Core Geometry 31 SECTION 6 - REACTOR OPERATING PROCEDURES ,
6.1 Reactor Management 31 6.2 Health Physics and Safety 31 6.2.1 Access Requirements 6.2.2 Health Physics and Safety 6.3 Operating Standards 32 6.3 1 Normal Operation 6.3 2 Core Alterations 6.3.3 Reactor Refueling 6.3.4 Approval of Experiments 6.3.5 ' Operation of Experiments 6.3 6 Limitation of Reactivity Insertion 6.4 ' Waste Disposal 36 37 6.5 Emergency Plan (E-39 SECTION 7 - SAFEGUARD EVALUATION 39 7.1 General 7.2 Accidents of Mechanical Type 39 7.2.1 Power Failure 7.2.2 Fuel Element Failure 7.2.3 Binding of Control Blades 7.2.4 Loss of Coolant 41 7.3 Accidents of Operating Type 7.3.1 Startup Accident t 7.32 Refueling Accident 7.3 3 Mishandling of Demineralizer Resin l
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7.4 Accidents of Experimental Type 42 7.4.1 Flooding Beam Port 7.4.2 Maximum Credible Accidents ; 7.4.3 Dropping Fuel Element on Full Core 7.4.4 Collapse of In-Core Experiment ! i APPENDIX A - TECHNICAL SPECIFICATIONS i APPENDIX B - REACTOR DATA - Figures 1 through 22 APPENDIX C - LIST OF TYPICAL EXPERIMENTS I $0} APPENDIX D - FISCAL STATEMENT APPENDIX E - RADIATION, HEALTH AND SAFEGUARDS COMMITTEE s
- RADIATION REGULATIONS Memorandum 1 - 60 l
APPENDIX F - REQUALIFICATION PROGRAM 1-27-82
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SECTION 2
SUMMARY
2.1 Introduction The Worcester Polytechnic Institute Pool Training Reactor (Figure 1) is a 10 kw (heat) light water cooled and moderated reactor designed ~ and built by the General Electric Company as a modification of' their standard open pool reactor. This type of pool reactor is similar to the Bulk Shielding Reactor at ORNL and the Geneva Aquarium except it is operated at low power and has additional safeguard features. Primary requirements for safe and flexible operation of the reactor as a student training aid have been met by the use of open pool design, low excess reactivity, and large negative temperature and void coefficients. The reactor is located in an existing building on the Worcester Polytechnic Institute campus about one mile from the center of Worcester, Mass. The primary use of the reactor is as an integral part of the Worcester Polytechnic Institute ele'ctive program in Nuclear Science and Nuclear Engineering. The reactor provides undergraduate and graduate students, under close supervision of qualified personnel, with reactor operating experience and experimental experience in the fields of reactor engineering, metallurgy, chemistry and physics. A list of representative experiments is included in Appendix C. 2.2 General Description of Reactor Facility The Institute Training Reactor, Figure 1, is a light water cooled and moderated heterogeneous reactor, fueled with approximately 4.0 kg of 20% enriched U-235, The core is located in the_ center of a pool of demineralized water 8 feet square by 15 feet , deep. There are 10 feet of water above the core. Experimental facilities converge toward the core and af ford opportunity for performing a number of different experiments at the same time. A typical core configuration is based on 24 fuel elements in rectangular array as shown in Figure 13. A 1 curie Pu-Be source occupies ! one module adjacent to the active core. Three safety blades and a manually 1 actuated regulating blade control the reactivity. The blades move vertically in two shrouds, extending the length of the core, The. core is moderated, reflected, and cooled by light water which is f circulated by natural convection. The thermal column side of the core is also reflected by graphite. Core elements are contained in a grid box enclosed on four sides to confine the flow of cooling water to the channels between and surrounding the elements. The grid box and contents, as well I as the blade drive mechanisms, are supported by a suspension frame from a reactor bridge. The cold, clean core with control _ blades removed has less than half percent excess reactivity. The safety blades, because of their location and large surface area, have a total shutdown worth of approximately 12% Ak eff. i
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l 2.3 Summary of Calculated Reactor Data (Furnished by Argonne National Laboratory i ,. . Tabulated below are the significant design' parameters which are used in this reactor. Reactor Materials: 1 Fuel Uranium aluminum alloy, 19.75% enriched Moderator High purity light water
- Reflector High purity light water and graphite
- Coolant High purity light water l Control Boral and stainless steel Structural material Aluminum Shield Water and aluminum lined
, concrete Structural Dimensions: Pool 8 x 8 by 15 ft. deep Core (active portion) 15 x 15 by 24 inches high Grid box 9 x 6 array of 3 inch modules Beam port One, 6 inch diameter
- Thermal column One, 40 x 40 inches in l cross-section Strategic Materials:
Fissionable material 4.2 Kg U-235-Burn-up Approximately 1% U-235 Fuel life Limited by factors other than burn-up l Thermal Characteristics: (Calculated) Operating power 10 kw (maximum)' Temperature, water 130 deg. F (maximum) Power peaking factor 2.17 Maximum hot channel factor 1. 51
, Maximum heat flux 400 Btu /hr.-sq. ft. l Specific power (clean, cold) 2.5 watt /gm U-235 '
Maximum gamma heat in core , 11 watt / liter Nuclear Characteristics: (Calculated) l Average thermal flux nv Average fast flux 7.2x1(0 25 x 10 nv 4 Maximum" operating excess reactivity 0.5% Akeff.
; Critical mass 4.0 Temperature coefficient -0.0163% Akeff. per i degree C i Void coef ficient .25% Akef f. per i 1% void ! Prompt neutron lifetime 61.2 pseconds
Control: Safety Elements Number 3 vertical blades Dimensions 10.5 inches wide by 40.5 inches long by 0.375 inches thick Material Boral Reactivity control 3.5% each, minimum Total worth, 3 blades 12 Ak Maximum withdrawal rate 7.5 inc$fmin., one blade at a time Regulating Element Number 1 vertical blade Dimensions 10.65 inches wide by 40.5 inches long by 0.125 inches thick Material Stainless steel Reactivity control 0.7% Ak Maximum withdrawal rate 3.8 inch 7bfn/k/ Standard Fuel Type Flat plate Number of elements 24 for minimum critical loading Fuel alloy Uranium-alumnimum Clad material Aluminum Fuel enrichment U-235 19.75% enriched i Number of plates per element 18 Plate thickness 1.52 mm (0.060 in.) Clad thickness 0.381 mm (0.015 in.) Cooling System: Coolant Pool water Type cooling Natural convection Temperature 130 deg.F (maximum) ' Purification Recirculating gemineralizer i Purity required 1 ppm - 5 x 10 Ohm-cm 2.4 Experimental Facilities The experimental facilities provided with the reactor are a thermal column, a 6 inch beam tube and 2 fuel elements comparable to above elements but with 16 removable plates.
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SECTION 3 REACTOR FACILITY 3.1 Location 3.1.1 Site Description The neighborhood in which Worcester Polytechnic Institute is situated is completely residential in character. The Campus is bounded on its north side by a pond and public park, on its west side by a public park, and on the remaining two sides by residences. The reactor is located in Washburn Laboratory on the eastern side of the Worcester Polytechnic Institute Campus. As shown in the campus map, Figure 2, Washburn Laboratory is adjacent to Boynton Hall, Stratton Hall, Gordon Library, Salisbury Laboratories, the Project Center and the Power Plant. Boynton Hall houses the administrative offices while Stratton Hall is occupied by the Mathematics Department. The Campus Security Police and the Power Plant are located in the lower levels of Stratton. Salisbury Laboratories houses the Humanities Department, Life Sciences, Biomedical Engineering, Management Engineering and the Department of Social Science and Policy Studies. Student workshops are located in the Project Center. The Power Laboratory is a complete central power station which once supplied both heat and electricity for the campus but which now supplies only heating steam. Also located on the East Campus are the Atwater Kent Laboratories, used by the Department of Electrical Engineering, and Kaven Hall, occupied by the Civil Engineering Department. There are no private dwellings on West Street from Institute Road to Salisbury Street. The nearest private residence. to the reactor is approximately 500 feet away. The Institute buildings on the West Campus nearest Washburn Laboratory are Higgins Laboratory which houses the Mechanical Engineering Department, and Olin Hall of Physics. The nearest building used as a dormitory, Sanford Riley Hall, is over 500 feet away. 3.1.2 Population Distribution Worcester Polytechnic Institute is located in a residential section about one mile north of the civic center of Worcester, Massachusetts. Worcester is an important industrial center with a population of 171,566 (1975). Approximately 2500 students are enrolled at Worcester Polytechnic Institute. There are about 180 faculty members and about 280 staff workers.
3.1.3 Climatology Narrative Climatological Summary (supplied by U.S. Dept. of Commerce)
" Worcester Municipal Airport is located on the crest of a hill, 1,000 feet above mean sea level and about 500 feet above and 3-1/2 miles northwest of the City proper. It is surrounded by ridges and valleys with many of the latter containing reservoirs.
However, of the ridges, only two of them are higher, one 400 feet higher and 2-1/2 miles northwest, and the other 1,000 feet higher and 15 miles north. The proximity to the Atlantic Ocean, Long Island Sound, and the Berkshire Hills plays an important part in determining the weather and, hence, the climate of Worcester. Rapid weather changes occur when storms move up the east coast after developing through wave action off the Carolina Coast. In the majority of these cases, the waves pass to the south and east, resulting in northeast and easterly winds with rain or snow and fog. Storms developing in the Texas-0klahoma area normally pass up the St. Lawrence River Valley, and, while much depends upon the movement and scope of these systems, usually deposit little precipitation over the area, however, they do brir,g an influx of warm air into the region. Wintertime cold wage snaps incidental to Canadian High Pressure Areas following cold front passages are quite f requent, but a tempering of the temperatures usually occurs before the full impact of the high reaches the county. Summertime thunderstorms develop over the hills to the west, with a majority moving towards the northeast. From the use of radar, we find many break up just before reaching Worcester, or pass either north or south of the City proper. Airport site temperatures are moderate, as the normal mean for the warmest month, July, is 70.1. Though winters are reasonably cold, prolonged periods of severe cold weather are extremely rare. The three coldest months, December through February, together have an average normal of over 25 deg. The coldest i temperature since 1949 was -19 deg on January 15,1957, while i the warmest was 99 deg, on September 2,1953. A review of I Worcester Cooperative records since 1901 shows an alltime high j temperature of 102 deg. on July 4, 1911, and an alltime low 1 temperature of -24 deg, on February 16, 1943. The average last f reezing temperature in spring at the airport is April 26. The average first freezing temperature in fall is October 15. In this regard about two-thirds of both spring and fall dates f all within 10 days of the average. Airport temperatures produce an average season of 172 days without freezing. In comparison, the l Worcester Cooperative records at Winter Hill show the average l latest date in spring as May 7 and the average first freezing date in fall as October 3, resulting in a season of 149 days without freezing terperatures, which is 23 days less than the airport site. Some low-lying valleys are more frost prone than either Winter Hill or the airport. Local topography greatly affects this climatic factor.
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Precipitation is usually plentiful and well distributed throughout the year. Monthly normals range from slightly over 3 inches in February, the driest month, to over 4 inches in August and November. The snowfall for all Worcester sites since 1901 averages slightly less than 60 inches. Due partly to several unusually heavy March storms in recent years, the airport location now averages considerably higher." The Commonwealth of Massachusetts has an annual average of 5.7 tornadoes per 10,000 square miles or about 4 per year, according to the U.S. Department of Commerce, National Oceanic and Atmospheric Administration.. Ninety seven (97) people were killed in Massachusetts by tornadoes during the period 1950-1978 but of these 90 were killed in the June 9,1953 tornado that struck Worcester. Four (4) people were killed in the August 28, 1973 tornado that struck West Southbridge, .in the extreme western part of the state. Most of the tornadoes reported in Massachusetts are of the " mini-tornado" type with winds up to 125 mph, a width of 100 feet, and which touch ground only for a few minutes. 3.1.4 Geology, Hydrology and Drainage The reactor site is on Boynton Hill, a drumlin, composed predominantly of glacial till, which was formed during the Wisconsin state of the Pleistocene epoch. The upper three to five feet of the hill is composed of Paxton loam, followed ty bedrock by a very compact sand to sandy silt. Bedrock in the area is Oakdale quartzite and Worcester Phyllite. Borings on the hill indicate that the ground water table is at least 25 feet below the reactor site. In the rainy season, the development of a temporary water table as high as 10 feet below the surface may occur. Drainage from the reactor site follows along the basement floor of Washburn Shops into the city sewer system on .Boynton Street (see Figure 2) and f rom there to the Worcester Sewage Disposal Works near the Worcester-Millbury line. Water from there flows into the Blackstone, an industrial river. 3.1.5 Seismology In a paper entitled "The Seismicity of Massachusetts" by Joseph A. Fischer and Fred L. Fox (Economic Geology in Massachusetts, 0. C. Forguhor, Ed., 1967, publ., Univ. Mass. Graduate School ) Table 1 lists all earthquakes known or estimated to have intensities of V or greater on the modified Mercalli scale from 1627 to 1963. Worcester is not on the list and the western most longitude on the list is 71.2 degrees. The authors state, "The seismic history of Massachusetts indicates that all earthquakes of any consequence have occurred in the bordering slope east of the central upland (east of 71.5 degrees W Longitude). Very minor shocks have been felt in the scattered areas and their noting is probably related to population density."
i Further in the report they say, "Since faulting is evident over nearly all of the State, it is known that great tectonic events have af fected its entire area. Therefore, no area of the State can be judged immune from However, the likelihood seismic activity on a historical basis. of the occurrences of damaging shocks in that portion of the State west of 71.5 degrees W Longitude appears slight." Worcester Polytechnic Institute is located approximately 15 miles west of 71.5 degrees W Longitude. . 3.2 Building i 3.2.1 Building Layout
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Classrooms and laboratories assigned to the Materials Engineering Section of the Mechanical Engineering Department are located in the basement and on the first, second and third floors adjacent to the reactor facility. As shown in Figure 3, the reactor base is a concrete slab located in the basement. Access to the thermal column and beam port is on this level. The basement reactor room is partitioned off from the rest of the building and access is restricted by locks on all doors to the reactor room. The first floor plan is shown in Figure 4 A separate office and reception area is provided at the west end of the reactor compartment. Access to the reactor room is restricted by partitions and locked doors. An open space, protected by a railing, and a stairwell adjacent to the reactor permit visual inspection of the beam port shield and thermal column door from the operating level floor. 3.2.2 Ventilation The reactor compartment ventilating system isIt of conventional operates design at over 3000 and provides in excess of two changes per hour. Intake and cfm with a discharge duct velocity of over 800 feet per minute. Air enters exhaust are filtered through commercial fiberglass air filters.the basement and through -several duct openings along the north wall at first floor levels and exhausts through openings also located in the north wal l . The exhaust duct outlet is located on the south wall of the 50 reactor feet compartment at a point about 18 inches above the roof and about above the ground. As discussed in the Environmental Impact Appraisal experience and calculations show that gaseous wastes will not be generated Samples in any hazardous quantities in a low power reactor of this type. irradiated in the reactor for isotope study will be small in quantity review and of proposed at relatively low radiation levels. A careful i irradiation will be made by the Institute's Safeguards Committee to insure I that a spill of the most dangerous volatile isotope to be used would be J quickly and safely diluted and discharged by the ventilating system at a l
negligible concentration. For this reason, the system is not equipped with dampers, and emergency procedure in case of a spill wili be to continue to operate the ventilating system. l SECTION 4 REACTOR FACILITY DESCRIPTION 4.1 Reactor Description 4.1.0 General The reactor assembly comprises the core, control system, instrumentation system and supporting structure. A description of the major core components is presented in the following sections. 4.1.1 Fuel Elements Each fuel element consists of equi-spaced flat uranium-aluminum alloy fuel plates held vertically between two aluminum side plates. Fuel plates are of the sandwich construction similar to those of the ETR and the General Electric 3 MW Open Pool Reactor. Fuel meat is a uranium-aluminum alloy, 20% enriched. It is clad in aluminum by the picture frame technique. The active length of the fuel plates is 24 inches and the overall dimensions of each element including end boxes are 3 inches square by nearly 40 inches long. The end boxes position the. fuel elements in the grid and provide handles for fuel positioning. The elements may be inverted or rotated 180 degrees to achieve more efficient utilization of fuel. There are twenty five (25) elements, each of which has 18 plates 25 inches long by 2.79 inches wide. The plates are each 1.52 mm (0.060 in) thick including 0.381 mm (0.015 in) of aluminum cladding on each side. A space of 3 mm (0.12 in) between fuel plates provides a passage for the flow of cooling watfS5 by natural convection. Each element contains less than 170 grams of U per element. Fixed plate elements are designated F1 through F26 and two removable plate elements, similar in dimensions and loading to the fixed plate elements, are designated R27 and R28. 4.1.2 Control Elements Reactor control. for startup and shutdown is accomplished ' by three blade-type control elements (Figure 15) with a total shutdown worth of approximately 12% Ak . The poison section, of boron carbide and aluminum, approximately'd.380 inches thick, is sandwiched between aluminum side plates. It is 40.5 inches long, 25 inches providing active control of the core. The remaining 15.5 inches connect the poison section to the drive tube. Each safety blade is guided throughout its travel by a shroud shown in Figure 14 The shroud consists of two thin aluminum plates 38 inches high, separated by aluminum spacers to provide a 1/8-inch water annulus around the blade. The shroud is latched to the sides of the grid box, and can be removed, if necessary, by use of the grapple hook. Small flow holes at the bottom of the shroud minimize the effect of viscous damping on the scram line. arrangement absorbs scram shock during the last 5 inches of travel. 4.1.3 Regulating Element The regulating blade, Figure 16, is designed to hold the reactor sub-critical with all three safety blades withdrawn. This assures that the safety blades are in a position to insert maximum shutdown control if scramed during startup or operation. Criticality is attained by withdrawal of the regulating blade. The blade is a stainless-steel sheet, about 11 inches wide and 40 inches long, supported and guided in the same manner as the safety blade, described in Section 4.1.2 It compensates for small changes of reactivity during nonnal reactor operation and is actuated by a manually conrolled drive. The drive shaft is pinned to the upper drive mechanism in contrast to the electromagnetic coupling used for the safety blades. 4.1.4 Drive Mechanisms Three coarse control blades are each driven by an electromechanical drive which positions, holds and scrams its respective blade. The drives are mounted on the locating plate above the reactor pool . They are magnetically coupled to the blades. When the magnets are de-energized, the blades are gravity scrammed to shut down the reactor. The drive mechanism includes a motor, worm-gear reducer, slip-clutch, ball bearing screw assembly, limit switches, scram magnet and housing. The mechanism operates through a stroke of up to 32 inches at normal speed of less than 7.5 inches per minute. Limit switches at the ends of the stroke open the motor circuit and also cause an indication on the control panel. A third limit switch is incorporated within the scram magnet to provide remote indication when the drive shaft is engaged. After fl ux decays in the magnet, which takes less than 100 l milliseconds, the blades will fall into the core through their first 24 l inches of travel in about 500 milliseconds. An indicator on the control
- panel provides continuous indication of the rod drive magnet position, and I a red light is energized if a blade is disengaged from the magnet.
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Control of the regulating blade is provided by a manually controlled motor driven drive mechanism. This drive is directly coupled to the regulating blade and operates at a speed of less than 4 inches per minute. The total stroke is less than 32 inches. Positioning accuracy is approximately 1/20 of 1%. The drive is equipped for position indication at the control console. A control shaft joins the regulating blade to its drive mechanism. The shaft is aligned by polyethylene sleeve bearings similar to those used for the control blade shafts, but without a dashpot. Radial clearances between shaft and bearings is 1/32 inch. The guide tube is attached to the suspension frame.
4.1.5 Grid Box 1 The core elements are supported and enclosed on four sides by the grid box. The grid box is approximately 28 inches long, 21 inches wide, and 36 inches high. The bottom is an aluminum grid plate with a 9-by-6 array of square holes, spaced to conform with the basic 3-inch element module. The sides of the grid box direct the convection current of the cooling water through the core. Four corner posts attached to the lower end of the suspension frame support the grid box. All parts, except for mechanical fasteners, are made of aluminum. The grid contains 52 spaces. Twenty-four of these are normally used for fuel elements. One is used for the start-up source and one for an irradiation device. The remaining spaces are blocked off with removable plugs for the possibility of future use for reflector and irradiation work. This arrangement will prevent the inadvertent placing of extra fuel in the core or improper location of fuel in the core. The initial grid arrangement, Figure 13, is based on a 9-by-6 array of modules each 3 inches square. The core is subdivided lengthwise by two 1-inch-wide shrouds containing the control blades. The fuel elements partially form a square near the center of the grid box in the " standard" loading as shown. The neutron source occupies one module adjacent to the active core. 4.1.6 Suspension System The core is suspended from an all-aluminum frame, Figure 14, which extends f rom the grid box to a height about one foot above the pool surface. The hollow corner posts of the suspension frame serve as guides for three ion chambers. A reactor bridge (mounted over the pool) supports the core suspension frame. The all-steel, prefabricated bridge was bolted together in the field and aligned with shims. 4.1.7 Locating Plate A locating plate, made of 1/2-inch steel, spans the upper end of the suspension f rame. It is bolted to the bridge and aligns the four control blade drive mechanisms with the core. The four mechanisms work through individual clearance holes, the base flange of each mechanism being secured to the locating plate. The plate and mechanisms are not removable as a unit, to prevent accidental withdrawal of the control elements. 4.1.8 Ion Chamber Support Assembly Three of the four corner posts of the suspension frame contain ion chambers for use in the flux monitoring channels of the control and instrumentation system. As indicated in Figure 14 the three compensated ion chambers are connected to adaptors and suspended from the corner posts. A clamp arrangement permist the assembly to be raised or lowered and then secured at the desired elevation.
4~1.9 Startup Counter Assembly The proportional counter assembly consists of the Boron 10 detector, ' enclosed in a watertight container, and the adaptor to which the container is joined. The detector is electrically insulated from the container. The entire assembly is suspended from the suspension bridge by a cable. An electrical cable connects the proportional counter with the preamplifier on the bridge. Three cables connect the preamplifier to the control center cubicle. The startup counter is located directly across the grid box from l the regulating blade and is contained in a guide tube and shield. A guide pin in the adaptor is free to slide in a vertical direction within the guide tube slot enabling the counter assembly to be raised or lowered as required. 4.1.10 Pool The Training Reactor is located in the center of a pool of demineralized water 8. feet square by 15 feet deep. The surf ace of the water is 11 feet above the core. Drains set in the concrete shield are located on two sides of the pool and the top surface of the shield is sloped to the drains. Overflow lines tie into these drains. The drains discharge to a steel drum which may be utilized as a hold-up tank. Impurities collecting in the pool water are removed by circulating the water through the pool cleanup demineralizer. This demineralizer is a mixed bed type. The resin in the unit may be regenerated if it is not excessively radioactive. Spent radioactive resin will be disposed of in compliance with 10CFR20. The pump used for circulating pool water to the cleanup demineralizer is a close-coupled centrifugal pump with a capacity of 10 gpm when pumping against the system head. The pool make-up water system in addition to- the demineralizer { includes, on an optional use basis, a motor operated valve, controiled by a float-actuated switch in the pool, a manual shutof f and throttle valve, and a check valve. The pool liner is fabricated of 1/4" thick aluminum and is watertight and has a coating on the side toward the concrete. 4.2 Experimental Facilities l All proposed new experiments must be reviewed by the Worcester Polytechnic Institute's Radiological Safety Officer who functions as a ; member of and in cooperation with the Institute's Radiation, Health and l Safeguards Committee (1) to insure that accidents causing changes in l composition and geometry of the experiments will not cause positive step changes or ramps in reactivity that might place the reactor on unsafe periods. No experiments with moving components shall be irradiated with the reactor unless the worth of the moving component is less than $0.25 (0.163% Ak/k)(2) to assure mechanical integrity, chemical compatibility and adequate protection against any other potential hazard. i 4.2.1 Thermal Column i The thermal column is a graphite-filled, horizontal penetration through the biological shield, to provide neutrons in the thermal energy range (about 0.025 ev) for irradiation of experiments. The column consists
; of an aluminum case, 40 inches square in cross-section, and about 6 feet i long, and is filled with graphite. Personnel in the building are protected against radiation frcm the column by a dense, concrete door which closes the column at the biological shield. The door is mounted on wheels which run on rails set into the concrete floor. The door can be moved The door is provided with a lock and is perpendicular to the shield face.
visible from the reactor operating level. Experimental access will be , granted only to authorized personnel. l 4.2.2 Beam Port The beam port is a source of neutrons primarily outside the thermal energy range. It is an air-filled aluminum tube extending from the reactor core face and terminating in a flange at the biological shield face. A shutter and removable shield plugs protect personnel in the building against radiation from the port. The plugs are aluminum casings, filled with concrete and containing spi ral conduits for Nssage of instrument leads. Provision is'made for ventilating the port and collecting and draining away any seepage that may accumulate between the port tube and the surrounding concrete. , The beam port plug and shutter each has a lock and is visible from the reactor operating level. Experimental access will be granted only to authorized personnel. 4 4.2.3 Spare Fuel Elements
! The removable plate spare elements are similar to the regular fuel elements except that the individual fuel plates and the top end boxes are removable. A lock arrangement in the pool underwater fuel storage facility will permit securing the spare elements or the elements that have been
- replaced. A locked storage cabinet outside the pool will be used for the j extra fuel when it is " cold." Individual plates from this type of element i may be stored separate f rom the element in the locked spare fuel storage
! cabinet. These spare elements will be used normally in critical approach tests and other low power experiments. If a single plate should ever need to be stored while " hot" an underwater rack will be provided. f The spare fuel elements which do not have removable plates will normally be stored in a locked storage cabinet outside the pool. In the event that the element is used in the reactor, a cable and lock device will permit securing the element in the underwater rack while it " cools." ! Two fuel storage racks are located on opposite sides of the reactor pool. Each rack is designed to contain not more than 18 fuel elements. When the reactor contains a critical mass, all additional fuel elements shall be locked in place except as authorized by the licensed senior ! i
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i i k operator in charge. I A fuel element shall not be stored outside of the reactor pool unless it produces radiatilon dose levels of less than 100 mR/hr at the storage container surface. Storage containers of fuel shall elements with fuel surf ace be locked closed when radiation dose levels greater than 100 mR/hr I
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1 unattended. handling situations, the Worcester In all of the above fuel i Safeguards Committee Polytechnic Institute Radiation, Health and regulations .concerning personnel monitoring and the use of survey probes i will be followed. Monitoring equipment conforming to Section 70.24, 10 CFR 70 is in operation near the spare element storage area. l Standard operating procedure requires the verification and recording of the number of fuel elements in the reactor core prior to eacn start-up. j All manipulations of fuel elements shall be under the direct supervision of i a licensed- reactor operator and shall conform to the provisions of Section j 6. - 4.3 Control and Instrumentation 4.3.1 General
- The normal operation of the reactor is manually initiated by the reactor operator. The instrumentation system provides the operator with all necessary information required for safe operation. In the event of a malfunction or accidental excursion, automatic safety controls will initiate a scram and shut down the reactor.
The control and instrumentation system consists of the nuclear instrumentation which measures neutron flux at the core and supplies the signals required for start-up, power operation, alarm and scram circuits. The system is supplemented by area radiation monitors in the reactor building, but outside the pool, as shown in Figure 21. 4.3.2 Control Center Cubicle l d The control center cubicle, a sheet metal cabinet, isA 8-1/2 feet long, front panel and 6-1/2 feet high, and 2 feet deep from front to back. rack type mounting provides space for reactor controls and nuclear instrumentation. Secured only by screws in the front panels, individual ! units may be separately removed. Control switches, push-buttons, indicator occupy a center panel recorders, and position indicators lights, An annunciator, alarm horn, log immediately above the operating-desk top. CR, period, and scram indicators and two power indicators and their associated multipliers, linear power channel selector, four switches and the necessary fuses to safeguard these units, fill out the upper panel. Two side panels contain amplifiers, a power supply unit, circuit Relays breakers, are area radiation monitors , a scaler, and micromicroammeters. e located in the rear of the cubicle. A _.
4.3.3 Control Power Distribution System Power for operation of the reactor equipment is supplied from a 115 volt, 60-cycle, ungrounded single phase source. Power for the control blade drives and the safety magnets is controlled from the master switch. The master switch is key-locked in the "0FF" position. The switch also provides an "0N" for normal operation, and a " TEST" position in which the control drives may be exercised without energizing the scram magnets and withdrawing the blades. Whenever the master switch is first turned "0N" from " TEST" or "0FF" an interlock causes the alarm horn to sound for 7 seconds to warn personnel of impending startup. A time delay mechanism prevents the control drive motors from withdrawing the blades during the 7 second delay period. 4.3.4 Auxiliary Control Power Supply The 115 volt, 60 cycle power f rom the 30 ampere auxiliary power circuit breaker is fed directly to an auxiliary control group which includes the cubicle blowers. 4.3.5 Instrumentation Power Supply Control and instrumentation power is supplied from the 115 volt, 60 cycle line through a 30 ampere circuit breaker, CB3. Two branch circuits, protected by separate fuses, supply power to the amplifiers, meters, recorders, and associated precision components of the reactor instrumentation system. The annunciator, scram relays, and alarm horn are energized by another branch circuit, while the fourth furnishes power to the drive motors of the regulating and control blades and to their switches and position indicators. 4.3.6 Flux Level Safety Channels i
- Two safety channels ar level over the entire flux range from 10 f available to monitor power watts to full power at the compensated ion chamber location and provide scram action. Each safety channel contains one micromicroammeter and the trip actuator. The signals are combined with the signals from the log N amplifier and the count rate amplifier in a trip mixer. The trip mixer scrams the reactor on high flux or fast period. The scram limit of flux level is preset at a value corresponding to less than 15 kw. A two-position channel-select switch permits recording of the output of either stable micromicroammeter on the linear power recorder on the control center cubicle.
4.3.7 Log N-Period Channel This channel monitors power level of the reactor over the range from 0.1 watt to full power. The system integrates the current in a gamma-compensated ionization chamber, amplifies this signal by means of an amplifier having a logarithmic characteristic covering a 6.3-decade range, and dif ferentiates the log N signal to give the reactor period. High voltage for the ionization chamber is obtained from a power supply in the equipment. One recorder permanently records log N, and is equipped with a high fl ux scram adjustable contact. The contact is set at a value
corresponding to less than 15 kw. 4.3.8 Startup Channel The startup channel will measure neutron flux in the " shutdown" condition, with the source present, up to a flux corresponding to 1.0 watts. A Boron 10 proportional counter serves as sensing element. Amplified counter output is monitored by a scaler and a log count rate meter on the control center cubicle. The scaler gives a direct reading of the number of pulses, or counts, occurring during a preset time interval. The scaler is intended primarily for the lowest flux levels. An audible indication of flux level is provided. During full power (10 kilowatt) operation, the proportional counter is fully withdrawn to its uppermost position (40 inches above the startup location) to prevent radiation damage to the counter. 4.3.9 Area Radiation Monitors Three radiation monitors are located as follows: one at the bridge, one near tne thermal column door, and one near the beam port plug. They give audible and visual indication when radiation levels exceed any preset level in the range of 0.1 to 100 mR/hr. The area monitors are connected to American District Telegraph ( ADT) and therefore provide continuous around-the-clock protection. 4.3.10 Safety Blade Control The three safety blades are normally controlled by two switches: one selects the blade to be moved; the other, a pistol-grip switch with a spring return , has positions " RAISE," "0FF," and " LOWER." It actuates the selected safety blade. Only one blade can be raised at a time. Indicator lights on the reactor control cubicle show when a safety blade has reached either limit of its travel. The safety blade normal control is overridden by an automatic or In case of a scram signal, the safety blades will scram from
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manual scram. any position during withdrawal or rundown. Each of the three control-blade drives is adjusted to operate at a rated speed less than 7.5 inches per minute in either direction, depending on which way the switch is held. Due to the relatively slow speed of the control-blade drives, the spring-return feature (to "0FF") permits close adjustment of blade position. Control-blade drive motors are interlocked against withdrawal of blades under the following conditions:
- 1. During a 7-second delay period subsequent to reactor startup.
- 2. With the log CRM count below 50/sec.
- 3. If the regulating blade is withdrawn from its lowest position.
4 If the flux rises above a preset level before the safety blades are completely withdrawn.
- 5. If the reactor is in the scram condition.
The first interlock gives personnel in the reactor building a 7-second warning of impending startup. The second interlock prevents flux level from increasing without reliable indication in the startup channel . The third interlock prevents criticality being achieved while raising a control blade. The fourth interlock gives a measure of protectionA against position erroneous loading to an abnormally high excess of reactivity. transmitter is provided for each control-blade drive to show drive position in the core relative to the fully inserted position. Digital indication to an accuracy of 0.02 inch is furnished by number wheels on a mechanical counter which is chain driven from the ball-bearing screw of the drive. The indication is transmitted electrically through a segmented commutator in the counter to the " WITHDRAWAL IN INCHES" indicator on the control center cubicle. Operation of the safety blade drive mechanism may be tested under the following conditions:
- 1. Master control switch (on reactor control cubicle) in " test" position;
- 2. Safety blade magnet de-energized (i.e., safety blades in core and reactor shutdown); and
- 3. Scram relay de-energized.
4.3.11 Regulating Blade Drive System The regulating blade drive is adjusted to operate at a rated speed of less than 4 inches per minute in either direction depending on which way the switch is held. Jog switches and a time delay relay are provided for close adjustment of the regulating blade. Actuating a jog switch energizes the control drive circuit. After a time delay of from 0.2 to 0.25 seconds, the time delay relay opens the circuit. Thi_s permits short movements in 0.01 inch increments of the regulating blade in either direction. 4.3.12 Scram Circuits The scram circuits initiate either relay (" slow") or electronic (" fast") scram. Relay scram is accomplished by de-energizing the scram relay under one of the following conditions: l 1. Fast period (adjustable between 3 and 30 seconds)
- 2. Manual scram
- 3. Hi'gh voltage failure in control center cubicle 4 High neutron flux
- 5. Either micromicroammeter switch in the position equivalent to the 10 to 100 kw range or above.
Electronic scram is caused by a flux level exceeding 1.5 times the normal preselected value in any safety channel, or by a reactor period less than 5 seconds. Electronic scram is effected by cutting off the current in each scram magnet so that the weight of the control blade will cause it to fall. 4.3.13 Alarm and Trouble Monitor System The annunciator and alarm horn on the control center cubicle informs the operator of abnormal conditions developing in 'the reactor system. Green lights indicate normal operation. Trouble in a particular phase of the system is indicated by lighting of the appropriate red light and the sounding of the alarm horn. The horn is silenced by pressing the " ANNUNCIATOR ACKNOWLEDGE" button which also turns the green light off. The red light remains on until the fault is cleared and the green light again appears. Abnormal conditions within the following areas will actuate a safety ala rm:
- a. Scram
- b. Neutron flux
- c. Period
- d. Water level
- e. Control blade disengaged
- f. Area radiation
- g. Ar 41 release A scram alarm will accompany each relay or push button (operator actuated) scram. The annunciator scram alarm is energized through an extra set of contacts on one of the scram relays.
4.4 Neutron Source A one curie plutonium-beryTlium source utilizing 16 gms of plutonium is contained in a steel hermetically sealed cylindrical capsule 1 inch in diameter by 1.6 inches long. The capsule is normally kept inside a small plastic bottle weighed down with paraffin and aluminum scrap and fitted with a lifting ring to which a strong nylon fishing line is attached. The source is normally positioned at the midplane of a modified radiation ! basket. During the instrument check-out preceding each start-up the source l bottle is swung close to each compensated ion chamber to verify proper ! operation and scram point setting. l The source will normally remain in the reactor pool at all times. l 1 When it is desired to remove the source from the pool for any substantial time period, the operational procedure will be reviewed by the radiological safety officer and the source will then be stored in the manner described below. The source will be leak tested at intervals no longer than 30 days during any storage outside the pool. The leak test will consist of wiping the surface of the source thoroughly with a piece of filter paper. The paper will be counted, using windowless flow proportional counting to count-alphas only. If an alpha count in excess of 10 counts per minute is t r obtained , the NRC and the source material supplier will be notified at once, and the source will be kept in its storage container unti.1 procedures for handling the situation are approved. The source will also be leak tested when it is removed from the pool for storage if it has not been previously tested within 30 days. In any case a leak test will be performed at least twice a year. In the event that the source needs to be stored outside the pool, a paraffin lined steel drum similar to that used as the original shipping container will be used. The container will be kept under lock and key and will be posted with radiation hazard signs. 4.5 Shielding The reactor 'is shielded for 10 kw operation, such that the maximum I radiation levels 1 meter above the pool surface and at the surface of the concrete shield, when the beam port and thermal column are closed, is less than 50 mR/hr. 1 In order to achieve the required shielding, the reactor is kept under at least 10 feet of water. The shield around the core consists of 3 feet of water surrounded by about 5 feet of concrete. I 1 1 1 i i 27-
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SECTION 5 REACTOR OPERATING CHARACTERISTICS 5.1 Introduction This section summarizes reactor operating characteristics as calculated by the General Electric Company in support of the initial reactor designs and by Argonne National Laboratory in support of the fuel change to low enriched uranium (LEU). . The G.E. calculations were " checked, where possible against data from the Bulk Shielding Facility, the Geneva Swimming Pool Reactor, and critical experiments at ORNL" and were confirmed by " low-power tests and actual operation of the Spanish Pool Reactor," modified where necessary to reflect the increase from the original I kw to 10 kw in 1967. 5.2 Nuclear Characteristics 5.2.1 Critical Mass and Loading The critical mass obtained from ANL diffusion calculations is 4.0 kg of U-235 in 20% enriched fuel fo" the 5-by-5 element core surrounded by water. The initial core loading :onsisted of 3.50 Pg of U-235 in highly enriched fuel with an excess reactivity of 0.23% Ak/k. 5.2.2 Neutron and Gamma Flux Average neutron fluxes for the reactor core at 10 kw are as follows: 10 thermal (<.625 ev): 7.2 x 10 nv 10 epi-thermal (>.625 ev): 25 x 10 nv 10 fast (>.639 Mev): 6.9 x 10 nv The distribution of the midplane fluxes in the core is shown in figure 6. The gamma heating in the core has a peak value of about 11 watt / liter. Figure 7 shows the distribution .of the midplane thermal, epithermal, and fast fluxes in the thermal column. Dose rates at 10 kw in the thermal column are as follows: gamma dose at shield face: 3 rem /hr gamma dose at outside of thermal column door: <10 mrem /hr h maximum neutron dose at outside of thermal column door from streaming:
<10 mrem /hr 5.2.3 Control System The three safety blades, Figure 13, cut the core into three sections.
One blade will be worth approximateiy 3.5% Akeff. The blades shadow each other, so their combined control effectiveness is more than three times the individual effectiveness of one blade. The total control-worth of the three blades lies around 12%. The control blades span almost the full width of the core. As a result, they keep the reactor subcritical even if a loading error were made and more fuel included than planned. To go critical with the entire 6 x 9 grid filled with fuel elements would require at least 300 gm of U-235 per element, about twice the loading of the fuel elements intended to be used. 5.2.4 Burn-up The burn-up is estimated to be approximately 1% U-235. The fuel life will be limited by factors other than burn-up. 5.2. 5 Temperature and Void Coefficient Negative temperature and void coef ficients aid stable operation of the reactor. The coef ficients have been calculated by perturbation techniques, and measured on the Spanish Open-Pool Reactor, which is similar in nuclear design to the reactor under consideration. The calculations used a reference temperature of 20 degrees C: results are tabulated as follows: Temperature Void Calculated Coefficient Coefficient Coefficients Ak Ak eff eff
-4 O (10-3 per % void)
(10 per H a) Normal core loading: Core average (blades withdrawn) -1.7 -2.5 Fuel section alone -2.2 -2.8 b) Central fuel element removed: -3 Core average -0.6 Fuel section alone -0.7 -4 Measured Coefficients (Spanish OPR) Two measurements Core heated with reflector -0.5 -2, -4 l I The effect of water temperature change on reactivity is shown in ! figure 8. With the water density held constant, the temperature at which the maximum loaded excess reactivity would be offset is about 50 C. Figure t 9 shows the effect of water density change on reactivity of the core. In an excursion, the water outside the core and in any control blade gaps would not be heated to the same temperature and density as the water flowing through the fuel elements. Therefore, the values of the reactivity coefficients shown in Section 2.3 are reasonably conservative. > l r The ef fect of fuel temperature change on reactivity due to Doppler broadening is shown in figure 10. Reactivity effects of. fuel temperature, water density, and water temperature are all additive so that any credible postulated excursion would be easily offset before temperatures . could approach that of fuel damage. 5.2.6 Neutron Lifetime ^ The neutron lifetime has been calculated as 6.1 x 10-5 seconds. This value is comparable to the measured value of 6.5 x 10-5 on the Borax ~ I reactor. 5.2.7 Alteration of Core Geometry l Alterations of core geometry will affect the available reactivity. If 4 the reactor is loaded to criticality with a square array of 25 fuel elements, arranged symmetricaly with respect to .the control blades, the }' changes in reactivity are estimated to be as follows:
- 1. An extra fuel element is added plus 1% Ak eff
- 2. A center fuel element is removed minus 4.6% Ak eff Adding an extra fuel element will not make the reactor critical provided that at least one safety blade is in the core.
1 I 4 i - .
SECTION 6 REACTOR OPERATING PROCEDURES 6.1 Reactor Management A single individual, the Director of the Nuclear Reactor Facilities, will have full responsibility over the operation of the reactor facility. The Director will report to the Dean of Faculty and will be responsible to the Radiation, Health and Safeguards Comittee for the proper carrying out of all pertinent local and NRC regulations and for proper maintenance of such records and operating practices as the Safeguards Committee may deem necessary for the safe operation of the facility. 6.2 Health Physics and Safety A Radiation, Health and Safeguards Committee (RHSC) has been established by the President of the Institute to develop policies and procedures relative to health hazards from radiation consonant with the Code of Federal Regulations. It is the function of this committee to establish maximum dose limits, survey and monitoring and handling procedures, access regulations and procedures, and emergency regulations and procedures. The committee will review isotope applications and experimental procedures, and keep records. 6.2.1 Access Requirements for the WPI Reactor Facility
- 1. Entrance to and primary exit-from the reactor room will be through the main entrance on the first floor of the reactor facility.
Personnel entering through this door must pass through a reception area where records of persons entering or leaving the facility may be recorded and where personnel dosimeters and film badges can be distributed when necessary by the authorized person in charge.
"No Admittance" signs will be posted on all other entrances to the reactor facility.
Keys to the facility doors will be limited in number and signed out under the supervision of the Facility Director. A key will be readily available for campus police or the night watchman, but will be placed in an alarm box covered by the American District Telegraph service.
- 2. Emergency exits from the reactor room will be locked to the outside at all times when not in use but can easily be opened from the inside for emergency exit.
- 3. All personnel entering the reactor facility will be required to
- conform to the Personnel Monitoring Requirements set forth in the Worcester Polytechnic Institute Radiation, Health, and Safeguards Comittee (RHSC) Regulations.
- 4. Areas containing various critical mrem levels will be posted and personnel entering the facility will be instructed on their meanings. Criteria for-establishing these levels will be set by the RHSC and will comply with pertinent sections of the Code of Federal Regulations. .
Application of these criteria will be the responsibility of the reactor facility director. Various areas of the facility may allow differing maximum doses during regular operatien. 6.2.2 Health Physics and Safety
- 1. Personnel Monitoring
- a. All persons entering a radiation posted area must wear film badges and/or pencil-type dosimeters, depending on the classification of the area. Permanent records will be kept for each person using such a device.
- b. Periodic surveys shall be made throughout the reactor facility during critical operations, using a. portable survey _ instrument for gamma detection and with particular attention paid to the beam port shield, thennal column face, demineralizer, and hold-up tank.
- c. The reactor facility ventilating system filters shall be checked periodically and whenever removed or changed. A general environmental check will be made from time to time to provide background records over a long period of time.
6.3 Operating Standards The basic premise of all proposed operating standards is the safety of the reactor, its operating personnel, and the ininediate surroundings. The limitations described below will be imposed upon operation of the reactor. The reactor operation will be supervised ~ by a Nuclear Regulating licensed operator at all times in- conformity with the Commission regulations expressed in Title 10, Code of Federal Regulations, Parts 50, 55 and elsewhere. The reactor will be normally operated on a less than eight-hour day, five-day week. Nonnally . formal laboratory course work will occupy several days per week with thesis, experiment, demonstration, repair and modification work occupying the remaining work week. 6.3.1 Normal Operation Before each day's operation a complete check-out of the reactor safety system shall be made and results recorded in the reactor log. f The check-out shall include testing each detection channel by use of the Pu-Be neutron source, verification of a scram setting.on each detector ( I using the source neutrons, a scram of each safety blade from at least 4" ! j out, check of the manual scram, and a visual check of the beam port, thermal column, and core. The fuel loading shall be recorded. Drive a h
mechanisms, signal lights and position indicator lights shall be checked. If any vital component does not function properly, the reactor will not be allowed to go critical until the condition is corrected. Power for the operating controls is supplied through the master control switch on the reactor control cubicle. For safety during servicing and refueling, as well as to prevent unauthorized operation, this switch can be locked in the "off" position by a key-lock integral with the operating handle. The key to this switch will always be in the possession of approved operating personnel. When the operator unlocks the master control switch and turns it to "on," a startup warning horn is sounded for 8 seconds. The startup warning also sounds when returning the switch to "on" from the " test" position. Af ter the 8-second warning, the operator prepares for operation by resetting the scram relay if it is not already reset. This relay may be reset if no condition exists which would cause a scram. The safety magnets are then engaged if necessary, by driving the drive mechanisms down, using the switch on the reactor control cubicle. (Normally the magnets and armatures are left in contact when the console is secured after operation.) When the maanets engage, a limit switch is actuated which prevents further downward motion but does not interfere with upward motion. Interlocks prevent upward motion as described in section 4.3.10. An indicating light for each control magnet shows when the magnet is engaged. Before attaining criticality, all three safety blades are withdrawn, with the regulating blade fully inserted. If the power level exceeds a preset limit during withdrawal of the safety blades, an interlock will prevent further withdrawal . Criticality is attained by partial withdrawal of the regulating blade. The schedule of operation is then continued as follows:
- 1. The scaler, count rate meter, and period meter are observed during withdrawal of the safety blades.
- 2. After the safety blades are withdrawn, withdrawal of the regulating blade is started. At that time the log count rate meter is approaching the upper limit set point and the B10 detector is raised to a preset location that reduces the count rate by a factor of about 100 but still permits an adequately on scale LCRM reading. Observations and regulating blade adjustments are then resumed to increase power toward the desired level.
- 3. When the safety channels and the Log N Channel are providing neutron flux level indications the Boron 10 counter is withdrawn to its upper position to reduce radiation damage. Fl ux indications are available from the safety channels beginning at power levels of about 0.02 watts and continuously thereafter up to full power of 10 kw. The Log N Channel Chart recorder comes on scale at about 0.1 watts.
[ l A log book will be kept of day to day operations containing in part the following infonnation:
- 1. N&me of operator and experimental personnel
- 2. Name of senior operator in charge
- 3. Date, time, and power level of operation
- 4. Non-routine experimental infonnation 6.3.2 Core Alterations (
All pool training reactor core alterations must be made only after careful planning and personnel orientation. Calculations of the required and known- effects of each course of action will be made prior to any core al teration. The following general rules apply to any course of action which demands core alteration:
- 1. Changes to be made will be evaluated in advance based on calculations or experimental data.
- 2. Before making any core alterations, two control blades must be cocked approximately 15 inches to provide the required fast safety scram during core alterations. (If it becomes necessary to remove one or more of the control rods, for maintenance purposes, all fuel must first be removed from the core.)
- 3. Before reloading the core with fuel, the neutron source will be positioned in the holder and control and regulating blades will be placed in the core with at least two blades cocked at-the-ready.
Only when the operator is sure that the blades and source have been placed in the core will he insert the fuel elements.
- 4. When the neutron source is in the core, a reading will be indicated on the startup channel . As fuel is added to the core, this reading will increase according to the calculations and records made prior to actual reactor refueling.
- 5. Carefully written records will be kept of the contents of the core, including source position, blade position, fuel-element position, and the location of all other core components and inserts.
- 6. a) When more than one grid position is involved in a loading i change, the core shall be unloaded to less than one-half the
' estimated critical mass and all experiments shall be removed.
Multiplication infonnation with all rods fully withdrawn shall be incorporated in a reciprocal multiplication curve and a new value of critical mass extrapolated. The fuel mass in each loading step shall be not more than one half the difference between loaded and extrapolated critical fuel mass until such j difference is less than a single standard element. Rods shall not be more than 50% withdrawn during the actual loading of fuel into the core. b) Af ter the completion of the core loading, tests shall be performed to ascertain that the excess reactivity limits set forth in these specifications are not exceeded. For a core geometry which has been previously loaded and for which an excess reactivity measurement has previously been made, criticality checks shall be made in loading the last 3 elements in lieu of the loading step requirements of Paragraph 6(a). Whenever a change of core configuration involving a single grid ! position is being made, two safety blades shall be cocked at the half withdrawn position and the third. shall be fully inserted during the fuel transfer. For a previously untried configuration the removable plate element shall be used first in the new position with only two plates present. Thereafter not more than two plates shall be loaded in any step and core excess reactivity measurements shall be made after each step to insure that the total excess reactivity of'the core after each fuel insertion will be below the maximum pennissible value of 0.5% a k/k. 6.3.3 Reactor Refueling The reactor is refueled manually, using either a hook grapple or a fuel element ' grapple. Each fuel element may be carried to one of the storage racks located in the pool. The water surface is at least 10 feet above the top of any fuel element positioned in either the core grid box or a fuel rack. During transfer the water coverage varies as the element is manipulated, but after the element clears the grid box it is usually under about 8 feet of water. Radiation levels at the pool surface during the entire transfer are negligible. The geometry of the storage racks is such that poisoning is not required. The fuel elements can be loaded with the locating plate in place, and the control drives connected. This is desirable from the safety standpoint, since it prevents the control blades being raised by hand. 6.3.4 Approval of Experiments Experimental procedures are reviewed periodically by the Radiation, Health and Safeguards Comittee. When high radiation levels or potentially_ dangerous experiments are envisioned, the details of the experiment are submitted to the Radiation, Health and Safeguards Committee for approval. If, in the Committee's opinion, the experiment is not safe, final approval of the experiment will be withheld by the Facility Director until adequate safeguard provision has been made by redesign of the experiment of 5-inclusion of required safety circuits. i I- ? h
6.3.5 Operation of Experiments Experimental facilities will be kept clean to minimize the amount of foreign matter likely to become activated and escape into the building. The interior of the thermal column will be checked periodically for corrosion. 6.3.6 Limitation of Reactivity Insertion The reactor is designed to be inherently safe by holding the excess reactivity generated in any accident within a safe limit extrapolated from the Borax tests. (1, 2) To assure that this design provision is not violated, experiments to actTvi"ty nse"rt ubWSccfNnt conditNns\ $!ack #"1 ohng theNm port, or dropping a fuel element on top of a fully-loaded core.
- 1. Dietrich, J. R. " Experimental Determination of the Self-Regulation and Safety of Operating Water-Moderated Reactors," U.N. Geneva Conference, 1956, Paper 481.
- 2. Luckow, W. K., and Widdoes, L. C., " Predicting Reactor Excursions b Extrapolating Borax Data," Nucleonics, Vol . 14, No. 1, Jan.
1 56 The maximum reactivity that could be inserted in the Borax I reactor without' damage to the core was detemined to be approximately 4.7% AKeff
- with the moderator at saturated connditions. Conservative extrapolation to the subcooled moderator conditions and plate design for this reactor yields a lower limiting value of 1.29% Ak eff.
The nuclear excursion produced by addition of 4.7% Ak in Borax led to nothing more than expulsion of the moderator from the*coYe, without release of fission products contained in the core. The limitation to 0.6% A keff. therefore includes a safety factor in excess of two. 6.4 Waste Disposal The only solid waste anticipated in connection with the operation of this reactor will be spent resin from the pool cleanup demineralizer. If the resin is radioactive, it will be flushed into a shielded cask and removed from the site for burial or other suitable disposition by an approved vendor. It is not anticipated that gaseous wastes will be generated in any finite qtlantities. The ventilating system provided for the facility is described in Section 3.2.2. In the unlikely event that appreciable gaseous fission products were released, remote monitoring equipment at the pool surface would sound an alarm. Detailed procedures will be developed for disposing of radioactive isotopes either purchased or created by irradiating samples in the reactor. These procedures will conform to the Code .of Federal Regulations, Part 20. 6.5 Emergency Plan _ 6.5.1 Planning Basis The WPI reactor continues to be a conservatively operated teachingDuring the
- f. reactor operated intermittently to accomodate academic schedules.
first twenty years of core life the total power produced totaled less than 8000 kilowatt hours. Thus, the core fission fragment inventory is actually No mueb less than the postulated values used in Section the 7facility of thisand report. the WPI exph.olve or pressurized samples are irradiated at By Product Material License limits the amount of radioactive material that may be separated from the core to less than 100 mei for isotopes with half lives of less than 40 minutes and to less than 1 mei for isotopes with half lives greater than 20 hours. Thus, the consequences of contamination of personnel or the facility during any conceivable accident Section are minimized and 7 details this consideration is reflected in the Emergency Plan. postulated accidents and their probable severity. Regulations and emergency procedures are kept posted conspicuously at all Washburn Laboratory exits and in each classroom area. Copies of emergen:- regulations will be provided each instructor regularly using the building. Meetings have been held with officials.from the Worcester Police and Fire Departments and both departments are kept informed of any substantial changes in the reactor facility that might in any way affect their response in an emergency. 6.5.2 Personnel Emergency In the event of an accident or medical emergency involving a person at s the reactor facility, such as a fall, heart attack, etc., the campus police N, will be notified immediately. 6.5.3 Emergency Alert In the event that an emergency alert is received, the operation of the reactor will be terminated in a timely manner. The Facility Director or ranking individual at the facility will order the facility evacuated if this appears necessary in his judgment. In case of security problems of any kind, the WPI Security Plan will take effect. The WPI Campus Police will be promptly notified of an emergency alert of any kind. 6.5.4 Reactor Emergency and Organizational Control of Emergencies An emergency evacuation alarm may be initiated by automatic radiation detection systems or by a reactor staf f member manually setting off the alarms via either switching off a fail safe radiation station at the console or by If the tripping a low water alarm device adjacent to the reactor bridge. reactor is in operation the operator will initiate a scram in a timely fashion. t Evacuation of the Washburn Laboratory shall follow immediately upon the continual sounding of the reacIor horn. A continual sounding is defined as a steady or intermittent loud buzzing sound of duration longer than 30 seconds. Horn extensions activated by r:he radiation monitors are placed at key loca-tions throughout Washburn Latdr% tory and are equipped with strobe lights. Evacuation drills are held twice a year and observed by members of the RHSC. y A record of the drill effectiveness is kept for the RHSC files. Upon an evacuation signal other than a drill:
. 1 I. A 24 hour per dry monitoring sarvice will diepetch polics end security personnel to the scene with instructions to prevent unauthorized persons from approaching or entering the Washburn Laboratory. They will also contact by telephone one of the 9
following officials:
- 1) Reactor Facility Director
- 2) Chairman Radiation, Health and Safeguards Committee (RHSC)
- 3) Any other licensed WPI reactor operator or RHSC member whose name is posted on the emergency list.
It shall be the function of the first of the above indi-viduals on the scene to perform the following tasks: a) Verify that the building is under guard, b) Procure a portable radiation monitor from its storage place in the Security Police office. (This monitor will be kept readily accessible in the Security Office and will be given a battery and operational check at least once a month.) Persons on the Emer-gency Call list will be briefed on the location and operation of this monitor when they are placed on the list, c) Check individuals who have reported personal contam-ination and instruct them on decontamination pro-cedures. Check buildings as may be judged appropriate, d) Ascertain the radiation level at the reactor facility building and entrance, continuing into the reactor site as the radiation level termits. e) Order a return to work for areas separate from the facility if it is ascertained that no radiation hazard exists outside the immediate confines of the facility. f) Order a return to normal work in the factaity if no radiation hazard exists or proceed with clean-up procedures if a spill has occurred and radiation levels permit decontamination to proceed. g) Isolate the reactor and/or building area with police assistance if it is ascertained that a radiation hazard exists, check reactor walls and doors for leakage and seal if necessary, post the area, cnd notify the U.S. Nuclear Regulatory Commission. II. Upon an evacuation signal all individuals in Washburn Laboratory, excepting the reactor operator and Reactor Facility Director, will follow the general evacuation procedure. Faculty or staff members in responsible charge of specific activities in the building are charged with the responsibility of evacuating their own students or transients in the area and assisting in the proper carrying out of the emergency regulations. Revision 1 (19 Jan. 1981) in
The general evacuation procedure is as follows: r
- 1) Leave the building immediately.
- 2) Assemble in the front door area of Gordon Library.
Persons who have reason to believe they have been contam-inated; i.e. , were close to a spill, etc. , will report to the Security Office. I l 4 3) Personnel shall remain in the assembly area until they receive further instructions from the officials listed above. i The reactor operator or Facility Director before evacuating will: , i
- 1) Open the master power switch to the reactor.
- 2) Turn room ventilation on in case of a spill and off in case of a fire.
- 3) Leave the reactor area, closing the automatically locking door behind him.
The exceptions to the above regulations are as follows: 1 1 1) In the case of a deliberate f alse alarm (drill) the Reactor Director, or his appointee, may, at his discretion, notify the monitoring agency of the precise time and duration of (4,3 the alarm and excuse their participation.
- 2) If the evacuation signal is caused by a false alarm of known origin, the laboratory instructor in charge at the !
time may notify ADT and security of the situation and excuse their participation. 6.5.5 Facility Emergency Organization j The reactor facility personnel will promptly report any accident or emergency to the Campus Police. Their written procedures include the following document: I l ws { Revision 1 (19 Jan. 1981) i 38a
Personal Injury or Sudden Illness Person taking emergency call should caution caller not to move victim unless absolutely necessary. Make victim comfort-able and assure that help is on the way. Campus Police Office will respond immediately. Assess the situation and render such immediate first aid as required. As soon as possible, report first assessment to the Campus Police Office. If the victim has collapsed and is unconscious and a heart attack is suspected, call for the Worcester Police Department ambulance. Administer cardiac pulmonary resuscita-tion (CPR) as first aid treatment. If personal injuries appear severe, call for the Worcestcr Police Department ambulance and stay with victim. Make the victim comfortable until help arrives. Render first aid cs indicated. Severely injured or suspected heart attack victims should be taken to City Hospital. If the personal injuries are minor the Campus Police Officer should transport the victim to the WPI Infirmary for first aid treatment. If the nurse on duty recommends hospi-tal treatment, the Campus Police Officer should transport the victim to Hahnemann Hospital Emergency Room. If the injuries appear to require treatment beyond the capabilities of the Infirmary (example: apparently broken arm, cuts requiring stitches, etc.) take victim directly to Hahnemann Hospital. QG ~ In all events, when a victim is being transported to a hospital, the Campus Police Officer in charge will notify the hospital emergency room that the person is on the way, alerting them as much as possible as to the nature of the injuries or condition of the victim. Your Campus Police Department is composed of a chief, uni-formed officers, dispatchers, a secretary, and student assis-tants. All uniformed personnel hold sworn appointments as special police officers. All officers are required to undergo and complete successfully basic police training. All officers have undergone basic training in first aid, including cardio-pulmonary resuscitation (CPR) . Campus Police may make arrests. The Campus Police Office is located in Stratton Hall and can be reached from the outside by dialing 753-1411 and asking for extensions 270, 433 or 463. The Campus Police operate "around the clock," seven days a week, 52 weeks a year. Mobile campus patrol equipped with radio cocaunication (two-way UHF) links mobile units to the Campus Police Office. Revision 1 (19 Jan. 1981) 38b
The main functions of Campus Police are protection of life and property, and assistance in case of illness, accidents, or p; special messages. Additional information can be found in the 1980-81 Student Handbook. 6.5.6 Coordination with Participating Government Agencies Contact has been made with the Massachusetts State Police and with Massachusetts Department of Public Health Radiation Control personnel, and arrangements are complete for message authentication. Our first contact in case.of emergency will be the State Police and via police channels we vill then be in touch with the Director of Disaster Preparedness, the Massachusetts Civil Defense authorities and the Nuclear Incident Advisory Team (NIAT) . 6.5.7 Assessment Actions As shown in Section 7, the consequences of contamination of personnel, the facility or the public during any conceivable accident are minimized by the limitations of the WPI Licenses. Survey equipment is kept in storage at the WPI Security Office and checked periodically for operability and calibration accuracy. Specific details on postulated releases are contained in Section 7. 6.5.8 Corrective Actions The nature and simplicity of the WPI facility is such that routine (fy review of emergency actions are adequate. The facility has a sprinkler system and is equipped with wall mounted fire extinguishers, a wall mounted first aid kit and an eye wash station. Emergency contact with security can be initiated in a variety of ways from a number of locations throughout the facility. Per-sonnel are instructed to contact Security immediately via appropriate means in the event of any kind of accident to personnel or if any potentially hazardous condition arises. 6.5.9 Contamination control Measures Signs posted throughout Washburn Laboratory instruct the occupants on proper procedures in case of an evacuation alarm. Persons having any reason to believe they might be contaminated report directly to Security where they can be monitored and either decontaminated, if the incident is minor, or transported to Worcester City Hospital if their condition warrants this action. Worcester City Hospital has confirmed that they have procedures in place to admit and treat persons contaminated with radioactive materials. If in fact there is any evidence of facility contamination, routine use of the facility will be suspended until the Campus Radiological Safety Officer has certified that the contamination has been adequately removed. Revision 1 (19 Jan. 1981) 38c
, c 65.10 Emergency Personnel Exposure In the event that injured persons need evacuation, emergency workers may accept up to 25 rem of exposure in the rescue operation. As shown in Section 7 this would permit a minimum entry time of 2 minutes if the reactor had been operating at full power continuously for one year and if there were maximum release of the fission product inventory. Typically the reactor over the past several years has generated less than 17. of the maxi-mum kwhr represented by full power operation around the clock.
Entries to the facility required for corrective action where the emer-gency does not involve personnel will be limited such that entry workers will not exceed 1.25 rem per calendar quarter. Any required transportation of injured personnel to the hospital will be via a Worcester Police Department Ambulance. 6.5.11 Organizational Preparedness As previously noted, two evacuation drills are annually scheduled for Washburn Laboratory by the RHSC. All licensed operators are required to review the emergency procedures periodically. A branch station of the Worcester City Fire Department is located a short distance from the campus, contact is made with officials there periodically, and they occasionally visit the facility. M Revision 1 (19 Jan. 1981) 38d
SECTION 7 SAFEGUARD EVALUATION 7.1 General The Worcester Polytechnic Institute Reactor is designed to insure minimum hazard in view of its anticipated use and location. The accidents considered are discussed below: 7.2 Accidents of Mechanical Type 7.2.1 Power Failure In the event of an electric power failure, the reactor will automatically scram. The pool water is more than adequate for safe dissipation of any decay heat. No damage to the reactor or its environs could result. 7.2.2 Fuel Element Failure Fuel element failure may be caused by excessive hydraulic forces or excessivefue} temperature. The heat flux in the reactor core is extremely low (1080 w/m mgximum) compared to the predicted burnout heat flux of at least 50,000 w/m . Failure due to hydraulic pressure unbalance or due to excessive temperatures during controlled operation is, therefore, not to be expected. Fuel element failure may happen during accident emergencies when the water in the core is suddenly expelled. Such an accident is discussed in Section 7.4.2. 7.2.3 Binding of Control Blades The control blades have been investigated with regard to:
- 1. entry time during scram operation;
- 2. distortion as a result of unequal heating; and 3 deflection caused by manufacturing clearances and unequal forces acting against the blades.
Measurements indicate that the time for a blade to drop through the first 24 inches of its travel is on the order of 500 milliseconds, as shown in Figure 11. However, a conservative drop time of 1 second (over 24' inches) was assumed for purposes of accident analyses (Figure 11) to show the relative insensitivity of reactor safety to this parameter. A 1 second blade drop time was found to limit hot spot fuel temperature to 320 degrees F, or less in all of the excursions examined in this report. This is well below the level at which fuel melting would occur (about 1220 degrees F).
Calculations indicate that distortion due to unequal heating is negligible. The clearances in the guide bearings permit the control blade to touch the shroud during f all. The calculateo contact pressure is at most two pounds. The points of contact between blade and shroud will be lubricated by a water film, so that the ef fect of f riction becomes negligible. This has been verified by test of a prototype control ekment through approximately 100 scrams with no apparent wear. In actual operation over a 19 year period annual inspections have failed to find any evidence of wear. On the basis of the above calculations and tests, the possibility of scramming the control blades is judged to be sure under all conditions. 7.2.4 Loss of Coolant The pool is specifically designed to preclude the possibility of drainage. It is of reinforced concrete construction to resist the most severe earthquake that might be reasonably expected in the area. Nevertheless, pool drainage must be considered because of the severity of the potential ha zard . First, the uncovered core creates a direct radiation hazard (less than 1 rem / min) to personnel on or near the reactor bridge, although people beyond the concrete pool walls would be shielded from direct radiation by the walls. Second, the problem of decay-heat removal during and after loss of coolant must be considered. If the cladding temperature rose to the melting point (near 1220 degrees F for aluminum), or if the cladding failed due to occluded gas pressure, or due to scme other cause, then radioactive fission products from a portion of the fuel in the core might be released to the building atmosphere. There is no danger of melting so long as the core heat can be conducted into water, that is, while the pool water surface is above the grid plate of the core. When the water no longer wets the grid plate, then heat must be removed by other means, namely, .by natural convection of air through and around the core, or by emergency measures. Calculations indicate that at least 5 kw of internal heat can be di ssi pated f rom the core by natural convection of air alone. The calculation is conservative in the following respect; it takes no account of heat conducted through the core supporting structure; it assumes a steady-state hot-spot fuel surface temperature of 1130 degrees F, with no change in heat distribution from the full-power distribution in the core; it neglects decay heat leakage by gamma radiation and by radiation from the end of the fuel elements. Even if the pool drained instantaneously, no fuel melting or dangerous release of radioactive fission products are anticipated, since the loss of water moderator would shut the reactor down and the rate of internal heat generation would rapidly decay to a small fraction of the calculated safe 5-kw limit. 7.3 Accidents of Operating Type 7.3.1 Startup Accident Analysis indicates that the fuel elements will not melt if a startup accident should occur. A pessimistic estimate of the energy released in a startup accident has been made under the assumption the log N-period channel fails to operate, and that scram does not take place until the power reached 1-1/2 times normal operating level. The maximum withdrawal speed of the safety blades is 7.5 in/ min. The time delay from generation of a scram signal to the instant when the safety blades are free to drop is less than 100 milliseconds. During scram, the blades are assumed to accelerate (for purposes of the accident evaluation) at rates corresponding to the 1-second drop curve in Figure 11. The energy release is about 40 i KJ, giving a peak typergture rise of less than 1 degree C. Fuel heat capacity is 2.35 x 10 J/m K. 4 The calculation assumes the neutron source to be in placegnd uses a conservative ratio of trip power to cource power, equal to 10 . source is absent from the core, this ratio may be of the order of 10 gthefor a clean core. source power" then grresponds to the spontaneous fission rate of U}bg ("0.3 fission /sec/kg , and a multiplication factor of 6, due to 17% shutdown reactivity. Startup under this condition is i normally prevented by an interlock which requires the B10 startup counter to read at least 5 counts /sec (Section 4.3.10). If, however, the interlock were inoperative, and the g des were withdrawn, the final period would be only about 10% shorter than in the case considered above, with unimportant effect upon the fuel temperature rise. 4 7.3.2 Refueling Accident Erroneous loading of a fuel element should be easy to avoid because of the small core size, good visibility of the core, and easily recognizable reference points. All fuel loading will be under the supervision of a senior licensed operator and will be in accord with the provisions of Section 6. It is not normally possible to go critical with the safety blades in the core. In the event a loading error had taken place, the reactor could conceivably go critical with one safety blade partially in the core. Since the position of the blades is indicated on the control console, the operator would know that an error had been made and could shut down the reactor before going to power. 7.3.3 Mishandling the Demineralizer Resin Following an accident such as cladding failure of a fuel element, the resin would become radioactive and the dose rate at the demineralizer may temporarily rise as high as 10 rem /hr. To avoid such peaks of radioactivity, replacement of the resin, would be scheduled when the dose rate had decayed to a low value. In normal facility operation it is expected that the resin in the demineralizer would be transferred rarely, only two resin changes have been required during the past nineteen years. 1 l
During transfer, the likelihood of spillage is high. Precautions will be taken, as in the handling of other radioactive materials, to protect operating personnel and to prevent contamination of the public sewer system. 7.4 Accidents of Experimental Type 7.4.1 Flooding Beam Port A leak which would flood the beam port involves less than a 0.1% addition of reactivity. This amount of excess reactivity (0.1%) is below prompt critical and the effects of the accident would be minimal.
- 1. " Reactor Physics," AEC Reactor Handbook, Vol . I (Pg 106)
- 2. Newson , H. W. , "The Control Problem in Piles Capable of Very Short Periods," Mon (1947) 7.4.2 Maximum Credible Accidents Consider the hypothetical case of four concurrent failures, viz.,
instantaneous addition of excess reactivity and three failures to scram. It is difficult to conceive of a mechanism by which even a single one of these failures could actually happen. The accident will presumably lead to " chugging" as steam is expelled f rom the core and replaced by water in successive power bursts. The available excess reactivity will diminish due to the effects of steam voids and core temperature rise, the buildup of fission-product poison in the fuel, and the deformation of fuel elements under the mechanical and thermal stresses induced by the bursts. Eventually, chugging will stop and the co re will reach an equilibrium power such that voids due to steam l production just balance the available excess reactivity. i Let us now take an extreme point of view and assume that meltdown is the only mechanism by which the core will shut down, taking no credit for the compensating effects of temperature, fission-product buildup, and core deformation. Physics calculations indicate that removal of a fuel element from the hottest region of the core, i.e., at the center, changes the reactivity by about 4.6%. Consistent with the assumption of 0.6% instantaneous reactivity insertion, one may then postulate that about 1/4 of a fuel element may melt. The fission-product buildup in that much fuel at the core center corresponds to about 1-1/2% of the total build-up in the Co re . The core is under 10 feet of subcooled water. In experiments conducted by General Electric in cooperation with the Pacific Gas and Electric Company, up to 80,000 lb/hr of superheated steam was injected from i i
- a horizontal 14-inch pipe into water at temperatures in the range of 100 to 150 degrees F. to determine the depth of water required to condense all the
, steam. This depth was found to be never more than 6 feet. It is concluded that 10 feet of water above the core is more than sufficient to insure that
- no fission products are carried to the water surface by way of steam bubbles generated in the nuclear excursion.
It is possible, though very unlikely, that some of the molten aluminum
; might react with water during the excursion. The heat evolved in the t
reaction of 30 pounds of aluminum would raise the pool temperature by about i 1 degree F. The fuel contains about 170 pounds of aluminum. Actually, no a more than' a small fraction of this amount could get into the molten ! condition necessary for rapid reaction, and it is clear that the effect upon the course of the accident is negligible. The postulated accident creates no mechanism, other than a violent reaction, by which solid i particles might be propelled through the water and into the atmosphere, j The aluminum-water reaction, if it takes place at all, would be too weak to j accomplish such expulsion. i The release of radioactive noble gases from the molten portion of the core provides the only means for contaminating the air above and around the reactor. It is possible that traces of solid fission products and halogens will be carried along with the bubbles of noble gas and dispersed in the
, air above the pool.
- The following calculation gives an order-of-magnitude estimate of this release, assuming that the " traces" constitute 1% of the solids and i halogens in the molten fuel, i
y 1. Total energy release rate by fission products in core after one j year of operation at 10 kw: 1.05 E 15 MeV/sec from ys { 1.03 E 15 MeV/sec from es 1
- 2. Distribution of fission products released to atmosphere:
1 Noble l Gases Halogens Solids Total
- Portion of total 10% 10 % 80% 100%
F.P. in core
, Portion of total F.P.
- in core contained in .015(10 )=.15% .015(10 ) =.15% .015(80)
- 1.2% 1. 5%
j melted 1/4 fuel element Portion of total F.P. in core released from 1( .15) =.15% .01(.15)=.0015% (.01)(1.2)=.012%0.16% j melted 1/4 fuel element 1 l MeV/sec from ys released 1. 79 E 12 2.86 E 10 1.11 E 11 1.93 E 12. i ] MeV/sec from ss released 9.91 E 11 9.30 E 9 1.36 E 11 1.14 E 12 J i l 1
With a compartment volume of over 30,000 cubic feet, the concentration of fission products inside would create an exposure rate of the order of 40 rem /hr, giving personnel adequate time to take measures for the protection of the facility and for their own safety. The dose rate outside the building does not constitute a significant hazard to the public. These figures are very conservative in that in actuality the reactor is at power less than 10% of the total hours in a year. Loss of Coolant with Damaged Fuel Suppose that a leak develops and the pool drains at a moderate rate, similar to the accident reviewed in Section 7.2.4, but made more serious by the fact that the cladding on one or more fuel elements is already damaged. As before, assume that the reactor had been running at full power up to the beginning of the accident, and that no emergency measures are taken to prevent pool drainage or to cool the core. To simplify a probably much more complex situation, suppose that a single plate of one element is completely stripped of cladding, and that 1 the fission products released are those which vaporize below 1000 degrees F., i.e., xenon, krypton, bromine, and iodine. Assume that the gaseous fision products are uniformly distributed throughout the core fuel and that all the atoms within recoil range (5 microns) of the surface are emitted. The release of fission products is calculated as follows: Volume fraction released to building: 1 element x 1 plate x zb elements 10 plates 2 x 5 microns x 1 mil = 0.4 x 10-4 39 mils 25.4 microns Total energy release rate by fission products released to building: 1.30 E 10 MeV/sec from ys 5.39 E 9 MeV/sec from ss With a building volume of over 30,000 cubic feet, the concentration of fission products inside would create an exposure rate of the order of 0.2 rem / hour, with no significant hazard to the public. Spill of a Radioisotope If a vial or an isotope solution being irradiated should disperse the isotope in the pool water, it would be removed by the pool cleanup demineralizer. The worst condition of this type is considered to be the spilling of a vial containing AuCl 3 in solution after it has been
i irradiated for five days in a flux of 10 11 nv. It is calculated that a 10 gram sample dispersed in the pool water would produce 140 mrem /hr at the pool surf ace and area monitoring equipment will sound an alarm. The room would be evacuated until such time as the activity had decayed to a safe level and/or the contaminanat had been removed by the pool cleanup demineralizer. In the event of the spill of a volatile highly radioactive isotope on i the floor, area monitoring equipment will sound an alarm. The room will be evacuated until such time as the activity has decayed- to a safe level and/or the contaminant gases have been removed by the ventilating system. j The worst condition of this type is considered to be the spilling of ten . I
- grams of methyl iodide, CH 3 I, after it had been irradiated for 4 hours with the reactor at full power. Assume that the I-128 formed is similar to I-131 except that the I-128 has a half-life of only 25 min. As a first j
- approximation , to determine the limiting case, assume that the air is l
still, that the entire solution evaporates in approximately 30 seconds into l an 8 cubic meter volume and that decay is a negligible factor. These are all conservative assumptions. Calculations indicate that the maximum permissible ingestion dose could be inhaled in the first 0.2 seconds. In a i short time the operation of the exhaust system would increase this time
- figure appreciably. If the vapor were uniformly. distributed throughout the i reactor compartment, for example, the dilution would be increased by a f actor of 100. With an exhaust duct velocity of over 800 feet per minute 2
and 3,000 cu. f t. per min, exhaust air, there will be four room changes per hour and the exhaust-gas dilution is such that the radiation hazard at the duct exit is negligible, i i 7.4.3 Dropping fuel element on Full Core This accident would increase reactivity by less than 0.5% 7.4.4 Collapse of in-Core Experiment - Each in-core experiment must be evaluated to assure that no more than 0.6% reactivity is added by the worst mode of failure. , it is possible, though unlikely, that the reactivity due to one of the above accidents could be inserted in less time than it takes the reactor to scram. Assuming that 0.6% reactivity were added instantaneously when the reactor is operating at normal power, the heat released in the first second j (no scram) would be about 5 kw-sec. At the same time, the fuel temperature would rise less than i degree F at the core center. Actually, scram of even a single one of the three safety blades will shut the reactor down in a fraction of a second. The flux scram level of 1.5 times normal corresponds to the prompt rise after the insertion of only
- 0.25% excess reactivity. The accident is thus judged to create no hazard
- to the facility, much less to personnel.
l 1. Barnes, R.S., et al, " Swelling and Inert Gas Diffusion in Irradiated Uranium," A/ CONF.15/P/81,1958 (Pg. 54) I
- 1 l
5 ,- i ! l
/,*
APPENDIX A TO LICENSE.NO. R-61 TECHNICAL SPECIFICATIONS FOR THE WORCESTER POLYTECHNIC INSTITUTE REACTOR DOCKET N0; 50-134 mi% e e K s
- g Is f
i TABLE OF CONTENTS Page 1.0 DEFINITIONS...................................................... 1 2.0 SAFETY LIMITS AND OPERATING RESTRICTIONS......................... 3 2.1 Safety Limits............................................... 3 i 2.2 General Operating Limitations............................... 4 j 2.3 Experiments................................................. 4 1 3.0 S URVEI L LANC E REQU I REMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.1 Frequency of Surveillance................................... 6 3.2 Action to be Taken.......................................... 6 i 3.3 Radiation 0etec. tion......................................... 6 4.0 SITE AND DESIGN FEATURES......................................... 8
- 4.1 Site........................................................ 8 I
(.. 4.2 Restricted Area and Exclusion Area.......................... 8 i 4.3 Reactor Building and Ventilation System..................... .8 i 4.4 Reactor Core..........................................:..... 8 i 1 4.5 Reactor Safety and Control Systems.......................... 8 5.0 ADMINISTRATIVE AND PROCEDURAL REQUIREMENTS 11 l 5.1 Facility Administrator...................................... 11 i 5.2 Radiation, Health, and Safeguards Committee................. 11 5.3 Radi ol ogi cal Safe ty 0 f fi ce r. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 , 5.4 Fire Protection............................................. 11 5.5 Procedures.................................................. 11 5.6 Operating Records........................................... 12
- 5. 7 Reports..................................................... 12
- 5.8 Annual Operating Reports.................................... 13 5.9 Fuel Storage................................................ 13
! 5.10 Initial Startup of Al tered Core Con figuration. . . . . . . . . . . . . . . 14 f, . . t l l 1 I t I i I I i f i
w I J The actual values of dimensions, measurements, and other numerical values may differ from values given in these specifications to the extent of normal con-struction and manufacturing tolerances, or normal accuracy of instrumentation.
- 1. 0 DEFIt41TI0t45 Cold, Clean, Critical Condition: Since xenon and samarium effects are neg-ligible for this reactor in its normal operations, the term cold, clean, cri-tical shall refer to the condition of the reactor core when it is at the normal ambient water temperature of 70 to 75'F and free of any experiments that could affect reactivity.
Critical Reactor Ooeration: Critical reactor operation shall refer to any situation unen more than 12 fuel elements are loaded in the core and any control blade is withdrawn more than 6 in: Exoeriment: An experiment shall mean any apparatus, device, or material installeo in the core or in external experimental facilities that is not a normal part of those facilities. Movable Exoeriment: ar manipulated wnileA the movable experiment is one that may be inserted, removed, reactor is critical.
.s., Ooerable: An instrument or channel is operable when the instrument or channel will be operational once it is energized.
Ooerational: An instrument or channel is operational when that instrument or enannel is installed, energized, and in all other respects performing the moni-toring and safety functions for which it was intended. Reactor Ooeration: Reactor operation shall be any condition wherein either the reactor key is inserted into the console lock or the reactor is not in a shut-down condition. Reactor Safety System: The reactor safety system is that combination of control channels ano associated circuitry that forms the automatic protective system for the reactor or provides information that requires manual protective action to be initiated. Reactor Scram: Reactor scram shall be the rapid insertion of the three control blades into the core by either of the following methods: (1) Relay (slow) scram: Reactor relay scram (slow scram) shall be instigated by the relay scram circuits which control current inputs for the trip amplifier. Interruption of this current shall de energize the scram magnets. (2) Electronic (fast) scram: Reactor electronic scram (fast scram) snall be c caused by the application of suf ficient negative bias in the trip amplifier to terminate current to the scram magnets. . l Worcester Tech Specs 1
Readily Available on Call: Readily available on call shall mean the licensed ( senior operator on cuty snail ensure that he/she is within a reasonable driving time (1/2 hr) from the reactor building. The licensed senior operator shall always keep the licensed operator informed of where he/she may be contacted. Reoortable Occurrence: A reportable occurrence is any of the following concitions: (1) a safety system setting less conservative than the limiting setting estab-lished in the Technical Specifications (2) operation in violation of a limiting condition for operation established in the Technical Specifications (3) a safety system component malfunction or other component or system mal-function that during operation could, or threatens to, make the safety system incapable of performing its intended safety functions ( 1) release of fission products from a failed fuel element (5) an uncontrolled or unplanned release of radioactive material from the restricted area of the facility (6) an uncontrolled or unplanned release of radioactive material that results in concentrations of radioactive materials within the restricted area in excess of the limits specified in Appendix 8, Table 1 of 10 CFR 20 (" (7) an uncontrolled or unanticipated change in reactivity in excess of 0.5% ak/k (8) conditions arising from natural or man-made events that affect or threaten to affect the safe operation of the facility (9) an observed inadequacy in the implementation of administrative or proce-dural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the facility Shutdown Condition: The reactor shall be deemed to be in the shutdown condition if no control or regulating blade is withdrawn from its fully inserted position or if there are less than 12 fuel elements loaded on the grid plate. Worcester Tech Specs 2 f
l 2.0 SAFETY LIMITS Att0 OPERATING RESTRICTI0tlS ~ 2.1 Safetv Limits i Cri ticality: The reactor shall be subtritical when the three control blades are at their fully withdrawn positions and the regulating blade is in its fully inserted position. 1 Macnet Release and Blade Oroo Times: The interval between the occurrence of cutof f voltage (scram) and the separation of each control blade from its magnet shall not exceeo 100 msec. Total time. of insertion of the first 24 in. of the control blades following initiation of a scram signal shall be less than 600 msec, including the magnet release time. Maximum Excess Reactivity: The maximum excess reactivity above cold, clean, critical shall be 0.5% ak/k. Radiation Alarms: Upon indication of radiation levels in excess of 50 mrems/hr (20 mrems/hr for fuel storage) area monitors shall actuate audible evacuation alarms in the reactor room and in the second and third floor areas above the
- reactor pool.
^ Radiation levels: The maximum radiation levels 1 m above the pool surface and
. .: at the surface of the concrete shield, when the beam port and thermal column are closed, shall be less than 50 mrems/hr.
Control Blade Withdrawal: The maximum withdrawal rate for a control blade shall , be 7.5 in./ min. The maximun reactivity addition rate through movement of the regulating blade shall be 0.06% Ak/k sec. Interlocks shall prevent simultaneous ' withdrawal of more than one control blade and shall prevent withdrawal of any control blade unless the regulating blade is fully inserted. Startuo Source Recuirement: Ouring reactor startup, a neutron source produc-ing at least 10' neutrons /sec shall be located adjacent to the fuel region. ! When readi,1gs on the log count rate meter are below 50 counts /sec, an interlock shall prevent withdrawal of any control blade, i Temoerature and Void Coefficients: The temperature and void coef ficients of reactivity snall be more negative than -2 x 10 5 ak/k F and -2 x 10 3 ak/k % void, respectively, at 80 F. Water Level: The minimum depth of water above the top of the end box of the
- core fuel elements in the reactor pool shall be 10 ft.
- Water Purity
- Corrective action shall be taken promptly if the following limits
! for the pool water are not met: (1) pH less than 8.0 and greater than 6.0 . (2) resistivity greater than 5 x 105 ohm-cm (3) pool water activity less than 10 5 pCi/mi 1 Worcester Tech Specs 3 t 4
l Water Temoerature: The maximum bulk water temperature of the reactor pool shall be 110"F and the minimum shall be 40 F. During critical operation of the g reactor, the corresponding temperature limits shall be 100 F and 60 F. 2.2 General Coeratino Limitations (1) Personnel Requirement - Reactor operation shall be permitted only when two or more persons are in the reactor facility, at least one of whom is a licensed operator at the controls. A senior operator shall be readily available on call during reactor critical operation. (2) System Integrity - The reactor shall not be operated when there are signi-ficant defects in fuel elements, control blades, or the reactor safety and control systems. (3) Abnormal Conditions - When abnormal operation of the reactor occurs, in-cluding its controls, safety systems, and auxiliary systems, action shall be taken immediately to scram the reactor and determine the cause of the abnormal . behavior. The operator at the controls shall have authority to scram the reactor whenever he/she believes that a question of adequate safety exists. 2.3 Exoeriments (1) The graphite. thermal cclumn and the beam port shall be vented to the facil-ity ventiTation exhaust system. n b (2) No experiments with moving components shall be irradiated with the reactor unless the reactivity worth of the moving component is less than 0.255. (3) Experiments shall be designed so that they do not significantly block natural circulation flow within the reactor core. (A) The total worth of all experiments with positive reactivity contributions shall be limited so that summed with the cold, clean, core excess reactiv-ity the total is not greater than 0.5% ak/k. (5) All samples or experiments shall be doubly encapsulated and ensured leak tight if release of the contained materials could cause corrosive attack to the facility or excessive contamination of the pool water. (6) No experiment shall be installed in such a manner that (a) it could significutly shadow the nuclear instrumentation system monitors (b) failure of the experiment could interfere with the insertion of a control blade f (c) failure of the experiment could damage the reactor (d) failure of the experiment could release excessive airborne-contamination , w Worceste Tech Specs 4 i
1 (7) No explosive or other materials that could combine violently shall be irradiated in the reactor or in external experiment facilities, in quanti-ties greater than the equivalent of 25 mg of TNT. In addition, the stress , that would be prodJced in the experiment container upon detonation of the explosive shall be calculated and/or experimentally determined to be less than the yield stress of the container. (8) If a container fails and releases material that could damage the reactor fuel or structure by corrosion or other means, physical inspection shall be performed to determine the consequences and need for corrective action. I a h O
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l t I l 4 1 Worcester Tech Specs 5 1
t 3.0 SURVEILLANCE REQUIREMENTS 3.1 Frecuency of Surveillance Daily: Before each day's critical operation (with the excepion of those experiments that require the reactor to be operated continuously for more than one full day), the two safety channels, the log-N period channel, and the con-sole annunciator system shall be checked and ensured to be operational. Quarterly: The area radiation monitoring systems and the pool water level switch shall be checked and ensured to be operational quarterly. Semiannually: At least semiannually, a reactor inspection shall be performed consisting of ' (1) The excess reactivity of the core above cold, clean, critical shall be measured. (2) The console instrumentatico shall be calibrated by a foil activation measurement of reactor power where applicable, or calibrated by other means, and checked for proper conditions. (3) Pool water pH shall be measured and conductivity and pH devices shall be ( caliorated. Annually: At least once each year, all fuel elements shall be removed from the storage racks. While the fuel elements are thus stored, the control blades shall be brought to the surface and visually inspected and the blade drives lubricated. Blade drop times and magnet release times shall be measured for each control blade, and a plot of blade drop times versus distance shall be obtained for each safety blade and compared with data of previous years. Abnormal deviation from previous data will be investigated and reviewed by the Radiation, Health, and Safeguards Committee. 3.2 Action to be Taken If maintenance or recalibration is required for any of the items, it snall be performed and the instrument snall be rechecked before reactor startup proceeds. 3.3 Radiation Detection Area Monitors: Area radiation sensors capable of detecting gamma radiation in the range of 0.1 to 100 mrems/hr shall be installed near the beam port, deminer-alizer, thermal column door, fuel storage area, and less than 1 m above the core pool surface. Upon indication of radiation levels in excess of 50 mrems/hr (20 mrems/hr for fuel storage) th.ese monitors shall actuate audible alarms in the reactor, room and in the second and third floor areas above the reactor pool. gA v Worcester Tech Specs 6
3 Portable area monitors capable of detecting gamma radiation in the range of 0.10 to 50 mRems/hr may temporarily replace fixed area monitors described ', above provided that the required alarms are operational. Portable Monitors: During reactor operation, operable portable survey instruments shall be readily available to the reactor operator for measur-ing beta-gamma exposure rates with ranges designated 1.0 mR/hr or less to over 50 R/hr, and fast plus thermal neutron dose rates from 0.04 to 1,000 mrems/hr. One or.more portable survey instruments for measuring beta-gamma exposure rates with a minimum range of 10 mR/hr to 50 R/hr will be kept available to the reactor staff in an external location (normally the security office) to facilitate obtaining radiation readings if a reactor radiation alarm should be activated. es b e
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4 ( h 1 i Worcester' Tech Specs 7 7/25/84
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7 4.0 SITE AND DESIGN FEATURES 4.1 Site The reactor and associated equipment is housed in the Washburn Laboratories located between West Street and Boynton Street on the campus of Worcester Polytechnic Institute in Worcester, Massachusetts. 4.2 Restricted Area and Exclusion Area The reactor room shall constitute a restricted area as defined in~ 10 CFR 20 and shall be controlled by partitions and normally locked doors. In addition, two small areas, one each on the second and third floors of Washburn Laboratories, directly above the reactor control drives, shall become restricted areas whenever the reactor is operating at power levels in excess in 1 kW and radiation levels in any of the rooms exceed 2 mrem /hr. The exclusion areas, as defined in 10 CFR 100, shall consist of the reactor room and the areas above the reactor. 4.3 Reactor Building and Ventilation System The reactor shall be housed in a closed room that is designed to restrict leakage. The ventilation system shall provide at least two changes of air per hour in the, reactor room whenever the reactor is operating. 4.4 Reactor Core b' Fuel Elements: Standard fuel elements shall be flat plate type consisting of uranium-aluminum alloy clad with aluminum. The width and depth of each fuel element shall be 3 in. x 3 in. Each element shall have an active length of 24 in. There shall be a maximum of 10 g of U-235 in each fuel plate and not more than 170 g of U-235 in any fuel element. Standard fuel elements have 18 fuel plates, each plate 1.52 mm thick with a clad thick-ness of 0.381 mm on each side. A maximum of 28 standard fuel elements may be installed in the core. Not more than two experimental fuel elements with sixteen removable fuel plates similar to standard fuel plates and fitted with removable top end boxes may be installed in the core. These elements may be used as part of the core assembly either as complete elements or as partial assemblies, I loaded with from 2 to a total of 18 fuel plates each. l 4.5 Reactor Safety and Control Systems The safety system shall be designed so that no single electrical fault that partially or completely disables the automatic scram function can, in any manner, impair or disable the manual scram function, and vice versa. The safety system shall be fail safe with respect to loss of voltage.
- 4. 5.1 Nuclear Instrumentation The channels of nucler instrumentation (listed below with their minimum operating ranges) shall during all reactor critical operations be opera-tional and shall be connected to the safety system, except as noted in Table 4.1.
Worcester Tech Specs 8
i l l (1) startup channel, background to 10-2% full power, i.e., background to 1W (2) log-N period channel; 2 x 10-3% to 150% full power; i.e. , 0.2 W to 15 kW (3) Linear safety channels 1 and 2; 2 x 10-3% to 150% full power; i.e., 0.2 W to 15 kW
- 4. 5. 2 Control Blades There shall be three control blades, each consisting of vertical aluminum-clad boral blades 10.5 in, wide x 40.5 in, long x 0.375 in. . thick.
The minimum reactivity worth of each blade shall be 3.5% ak/k. The total worth of the three blades shall be at least 11% Ak/k. 4.5.3 Regulating Blade There shall be one regulating blade consisting of a vertical stainless-steel blade 10.65 in, wide x 40.5 in, long x 0.125 in, thick. It shall
, have a calculated reactivity worth of less than 0.7% Ak/k.
- 4. 5.4 Blade Position Indicators i '
The blade position indicator on the console shall provide an indication of
,- the blade position to within + .02 in. Signal lights shall be provided for (49; each control blade drive and Tor the regulating blade to indicate the upper and lower limits of travel and, in the case of control blades, an armature engaged by a magnet.
1 l 4 r
.n
( i Worcester Tech Specs 9 l
2 4 i , Table 4.1 Safety system functions Set point < Condition
- Detector- range Action Comment Neutron count S-10 chamber in > 50 counts / Interlock prevents Nay be bypassed rate i set point startup channel sec safety blade with-
,only when k,ff drawal _< 0.9 Reactor period b
Comoensated ion Relay scram and i set point chamoer in log-N , -'> + 5 see alarm period channel - I Reactor period Compensated ion -> + 3 see Electronic scram / ; i set point chamber in log-N- alarm - period. channel i Reactor power 2 compensated < 150% full Electronic scram /
. > set point ion chamoers in power alarm
. level safety t channels 1 and 2
' 1. compensated ion ( 150% full Relay scram / alarm chamber in log-N icale reading period channel of log-N (
~
recorder - ! Reactor power Compensated ion < 115% full Alarm , > set point chamber in Icale reading linear level of linear
- channels 1 and level 2 recorder i Actuation of Contacts within Relay scram / alarm manual scram switch i
switch 4 Pool water level Level switch i 1 foot Building evacua- May be bypassed
-< set point orco in pool tion alarm provided a water level . licensed senior 3
operator is
.present in the facility
' "Except as notec aoove none of the conditions listed may be bypassed during critical oper-ation of the reactor. 'When any of the nuclear detection systems listed above are disabled
- or undergoing maintenance, the reactor must be maintained in the shutdown condition.
l , 1 1
.l l
{ Worcester Tech Specs 10 l
v 5.0 ADMINISTRATIVE AND PROCEDURAL. REQUIREMENTS 5.1 Facility Administrator-The Director of the Nuclear Reactor Facilities shall have full responsibility for operation of the reactor facility. The Director shall report to the Dean of Faculty and shall be responsible to the Radiation, Health, and Safeguards Committee for conformance to the facility license provisions and all local and NRC safety regulations. The Director also shall be responsible for proper main-tenance of such records and operating practices as the Committee may deem neces-sary for the safe operation of the facility. 5.2 Radiation. Health. and Safecuards Committee A Radiation, Health, and Safeguards Committee shall review and approve all pro-posed modifications affecting reactor safety, as well as general and specific types of experiments and procedures, including determination of unreviewed
~
safety questions pursuant to 10 CFR 50.59. This committee also shall conduct, at least quarterly, reviews of operations, equipment performance, records, and procedures. The Ccamittee shall establish written procedures regarding review methods, quorums, and subcommittees, and it shall maintain written records of its activites. The members of the Committee shall be appointed by the President or Vice President of Worcester Polytechnic Institute (WPI) and a majority shall be WPI faculty members. 5.3 Radiolocoical Safety Officer A Radiological Safety Officer shall be appointed to serve on the Radiation, Health, and Safeguards Ccamittee and to review and approve all proposed pro-cedures and experiments concerning radiological safety. The Radiological Safety Officer shall advise the Director of the Nuclear Reactor Facilities of rules, regulations, and procedures relating to radiologicial safety and shall routinely conduct radiation surveys. 5.4 Fire Protection The licensee shall provide heat or ionization-type smoke detectors, which will alarm when there is a fire in the reactor room. At least two such detectors shall be operable at all times. 5.5 Procedures Detailed written procedures shall be provided for all normal operations of the reactor, supporting facilities, maintenance operations, radiation protection, experiments, and emergency plans and operations. These procedures shall be approved implemented.by the Radiation, Health, and Safeguards Committee before they are (- Worcester Tech Specs 11
Temporary procedures that do not change the intent of the initial approval procedures may be authorized by two members of the facility staff at least one (_. of whom shall be a licensed senior operator. Such procedures shall be subse-quently reviewed by the Radiation, Health, and Safeguards Committee. 5.6 Ooeratino Records In addition to records required elesewhere in the license application, the following ecords shall be kept of . (1) reactor operation, including power levels and periods of operation at each power level (2) maximum radioactivity released or discharged into the air or water beyond the effective control of the licensee as measured at or before the point of such release or discharge (3) emergency shutdowns and inadvertent scrams, including reasons for emer-gency shutdowns (4) maintenance operations involving substitution or replacement of reactor equipment or components (5) experiments installed including description, reactivity worths, locations, exposure time., total irradiation, and any unusual events involved in their performance and in their handling n , (6) tests and measurements performed pursuant to the Technical Specifications (7) incore irradiations 5.7 Reoorts In addition to reports otherwise required under this license and applicable regulations (1) The licensee shall inform the Commission of any incident or condition relating to the operation of the reactor that prevented or could have pre-vented a nuclear system from performing its safety function as described , in the Technical Specifications. For each such occurrence, WPI shall promptly notify, by telephone or telegrapo, the Administrator of the appro-priate NRC Regional Compliance Office listed in Appendix 0 of 10 CFR 20 and shall submit within 10 days a report in writing to the Director, Division of Licensing (OL), with a copy to the Regional Office. (2) The licensee shall report to the Director, GL, in writing within 30 days, any observed occurrence of substantial variance disclosed by operation of the reactor frcm performance specifications contained in the Safety Analysis Report or the Technical Specifications.
)
(3) The licensee report to the Director, OL, in writing within 30 days, any occurrence of significant changes in transient or accident analysis as described in the SAR. Y. l Worcester Tech Specs 12 l
5.8 Annual Ooerating Reoorts (f A report covering the previous year shall be submitted to the Administrator of F the appropriate Regional Office not.later than March 31 of each year. It shall include (1) Ooerations Summary: a summary of operating experience having safety sig-nificance occurring during the reporting period, including (a) changes in facility design (b) performance characteristics (e.g. , equipment and fuel performance) (c) changes in operating procedures that relate to the safety of facility operations (d) any abnormal results of surveillance tests and inspections required by these Technical Specifications (e) a brief summary of those changes, tests, and experiments that required authorization from the Commission pursuant to 10 CFR 50.59(a) (f) changes in the plant operating staff serving in the positions of Reactor Facility Director, Health Physicist, or Radiation, Health, and Safety Committee members (2) Power Generation: the most current summary of thermal output of the facil-ity available together with a summary of the total thermal power generated over the life of the reactor (3) Shutdowns: a listing of unscheduled shutdowns which have occurred during
~
the reporting period, tabulated according to cause, and a brief discus-
,7sm sion of the actions taken to prevent recurrence (4) Maintenance: a discussion of corrective maintenance (excluding preventa-tive maintenance) performed during the reporting period on safety related systems and ccaponents (5) Chances. Tests, and Exceriments: a brief description and a summary of the safety evaluation for tnose changes, tests, and experiments that were car-ried out without prior Commission approval, pursuant to the requirements of 10 CFR 50.59(a)
(6) Radioactive Effluent Releases: a statement of the quantities of radio-active effluents released from the plant 5.9 Fuel Storace Two fuel storage racks are located on opposite sides of the reactor pool. Each rack shall be designed to contain not more than 18 fuel elements. When the reactor contains a critical mass, all additional fuel elements not 'in the core shall be locked in place except as authorized by the licensed senior operator in charge. A fuel element shall not be stored outside of the reactor pool unless it produces radiation dose levels of less than 100 mrems/hr at the storage container surface. Storage containers of fuel elements shall be locked closed when unattended.
.h Worcester Tech Specs 13
c All fuel element transfers to or from the reactor core shall be conducted by a staf f of not less than three persons, which shall include a licensed senior (c operator in charge and a licensea operator. Staff members will continuously monitor the operations using appropriate radiation monitoring and core nuclear instrumentation. 5.10 Initial Startuo of Altered Core Confiouration (1) During a critical experiment of a new (not previously used) core config-uration, subcritical multiplication plots shall be obtained from at least two instrumentation channels. (2) When a change of core configuration involving a single grid position is being made, two control blades shall be cocked at the half withdrawn posi-tion and the third shall be fully inserted during the fuel transfer. For a previously untried configuration, the removable plate element shall be used first in the new position with only two plates present. Thereafter not more than two plates shall be loaded in any step and core excess reactivity measurements shall be made after each step to ensure that the total excess reactivity after fuel insertion will be below the maximum permissible value of 0.5% ak/k. (3) When more than one grid position is involved in a loading change, the core shall be unloaded to less than one-half the estimated critical mass and all incore experiments shall be removed. Multiplication information with all blades fully withdrawn shall be incorporated in a reciprocal multipli-cation curve and a'new value of critical mass extrapolated. The fuel mass hq in each loading step shall not be more than one-half the difference between
. loaded and extrapolated critical fual mass until such difference is less than a single standard element. Blades shall not be more than 50% with-drawn during the actual loading of fuel into the core.
After the completion of the core loading, tests shall be performed to ascertain that the excess reactivity limits set forth in these specifica-tions are not exceeded. For a core geometry which hs been previously loaded and for which an excess reactivity mesurement has previously been made, criticality checks shall be made in loading the last 3 elements in lieu of the preceding loading step requirements of this paragraph. All standard fuel elements shall be unloaded before a control blade may be manually removed. Procedures for reloading the last half of the core shall prescribe the , loading of fuel nearest the grid box center first with fuel loading there-after proceeding outward. The reactor shall not be brought critical nor shall more than two control blades be fully withdrawn when any vacant grid box position is surrounded by four fuel elements. The control blades shall be positioned to intersect the completed core. e b. Worcester Tech specs 14 ..
/
APPENDIX B TO LICENSE NO. R 61_ REACTOR FACILITY DATA
- Figures 1 - 22 WORCESTER POLYTECHNIC INSTITUTE REACTOR } Docket No. 50-13h a
3. i 7 I l
TABLE OF CONTENTS ~
- 1. Reactor Picture
- 2. Campus Map
- 3. Basement Floor Plan
- 4. First Floor Plan
- 5. a. Basement Floor Plan
- b. First Floor Plan
- c. Second Floor Plan
- d. . Third Floor Plan
- 6. Midplane Neutron Fluxes at 10 kw
- 7. Thermal Column Neutron Fluxes at 10 kw
- 8. Temperature Coefficient
- 9. Void Ccef ficient
- 10. Doppler Coef ficient
- 11. Control Blade Drop Time
- 12. Decay Power after Infinite Irradiation
- 13. Typical Core Arrangement ~
- 14. Block Diagram of Safety Instrumentation
- 15. Assembly Control Blade
- 16. Assembly Regulating Blade
- 17. Beam Port' f 18. Pool Outline
- 19. Pool Reactor Drive
- 20. Pool Training Reactor Core
- 21. Block Diagram of Safety Instrumentation
- 22. Fuel Element
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WPI LEU CORE, MIDPLANE THERMAL COLUMN FLUXES FROM CORE FACE OF THERMAL COLUMN
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Temperature Coefficient of Reactivity Water Density Constant at 0.998 g/cc 0
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Figure 8
Void Coefficient of Reactivity Water Temperature Constant at 20C 0 p. N g 8 E 5
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' for Rod 3 BLOCK DIAGRAM OF SAFETY INSTRUMENTATION FIGURE 21
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