ML20247K855
| ML20247K855 | |
| Person / Time | |
|---|---|
| Site: | 05000199 |
| Issue date: | 04/30/1989 |
| From: | Berlin R, Hu J MANHATTAN COLLEGE, RIVERDALE, NY |
| To: | |
| Shared Package | |
| ML20247K826 | List: |
| References | |
| NUDOCS 8906020040 | |
| Download: ML20247K855 (140) | |
Text
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L SAFETY ANALYSIS REPORT F
CONVERSION OF MANHATTAN COLLEGE L ZERO POWER REACTOR (MCZPR)
TO i
. LOW ENRICHED URANIUM (LEU) FUEL !
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Dr. Robert Berlin Dr. Jih Perng Hu 1
Submitted to:
United States Nuclear Regulatory Commission in Accordance with Requirements of 10 CFR 50.64 April,1989 8906020040 890519 PDR ADOCK 05000199 p PDC
TABLE OF CONTENTS Section Pape 1.0 ' INTRODUCTION AND GENERAL DESCRIPTION OF THE PROJECT . . . . I 1.1 Fuel Conversion Process . . . . . . . . . . . ............... .. ...... l' l.2 T h e M CZ P R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 1
' 2.0 SITE AND FACILITY DESCRIPTION . . ......... .................. 3 2.1 Site Parameters . . . ........ ........ ... ........ .. ...... 3-2.2 - Reactor Facility and Systems . . . . . . . . . . . . . . .. . ......... 5 2.2.1 - Facility Description . . . .. . ......... . ..... .. . 5 2.2.2 Reactor................ ... ... ............ ...... 11 2.2.3 Reactor Coolant System and Connected Systems . . . .. . ..... 13 2.2.4 Waste Management . . . . . . . ........ ...... .. .. ... 16
- j. 2.3 Current Reactor and Core Parameters and Limits . . . . ... .. 16
3.0 DESCRIPTION
OF THE CONVERSION PROCESS . . . ... . ... . . . 17 3.1 New Fuel Element Description .. .. . . ..... ........... . 17 1
3.2 LEU Core Parameters ... ...... ............. . .. .. . 17 l 3.3 Description of Fuel Removal and Replacement . . .... .. . 19 3.3.1 Steps in Removal and Replacement Processes . . . ... ...... 19 3.3.2 Equipment and Instrumentation Requirements . . ... . .. ... 21 3.3.3 Health Physics Report .. .. . .. ... ... . 22 3.4 Conversion Process Schedule . . . . .. ...... .... . 22
' 4.0 ACCIDENT AN ALYSIS . . . . . . . . . . . . . . .. .. . . . .... . . .. .. 23 4.1 Introduction .. ....... ...... .... . ... ... .. ..... 23 4.2 Operational Accidents and Conditions .. . .. .. .. .. 23 4.3 ' Handling Accidents . . . . . .. . .. . .. . . 23 4.4 Effect of Natural Phenomena .... . .... ... . . . .24 4.5 Transportation Accidents . . . ... ... . .... .... .... . 24 5.0 CONDUCT OF OPER ATIONS . . .... . ... . .. . .. 25 5.1 MCZPR . . . . . . . . ... ... ......... .... .. .. .. ... 25 5.2 Emergency Planning . . . .. .. .......... .. .. . 25 5.3 Radiation Protection During Fuel Conversion . . . ... .. ... 25 1
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1 Figure # Title Pape f 2-1 Plan of Manhattan College Campus . ... . ..... . .. ..... .... .. 4 2-2 Area Surrounding the Leo Engineering Building . . . . . . .. .. 6 2-3 First Floor of the Leo Engineering Building . . ... .. ... . .. 8 2-4 Second Floor of the Leo Engineering Building .. ... . .... . . 9 2-5 First Floor of the ZPR Room . . . . .. ... ... . . . ....... . . 10 2-6 Second Floor of the ZPR Room ..... .. . .... . . . . . ..... 12 2-7 Demineralizing System . . . . . . . . . . . . . .. .... .......... .. ..... 14 2-8 Reactor Tank Cleaning System . . . .. .... . , . . . .. .. . 15 3-1 HEU AND LEU Cores ... .. . . .... . . .. .... . . .. . . . 18 3-2 Removal and Replacement of HEU and LEU Cores . . . . . . 20 5-1 Oragnizational Structure of Manhattan College . . . .. . .. . .. 26 1
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i Table # Title page 2.1 Estimated Occupancy of Buildings Shown on Mby . . . . . .... ..... 7 Design Data of HEU arid LEU Cores . . . . . . . . . ..... . .. . .. . 17 3.1 19 3.2 Parameters of HEU and LEU Cores . . . .. .. .. . . . . . ...
Removal and Rep.lacement Sequence . . . . . . .. ...... 21 3.3 ..... . .
Investigational Levels of Exposure . . . . . .. ..... .... . . 28 5.1 ....
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1.0 INTRODUCTION
AND GENER AL DESCRIPTION OF TIIE PROJECT 1.1 Fuel Conversion Process The Nuclear Regulatory Commission (NRC), through an amendment to 10 CFR Part 50 regulations,is requiring that research and test reactor (non-power reactor)licensecs replace their Highly Enriched Uranium (HEU) fuel with Low Enriched Uranium (LEU) fuel to " reduce the risk of theft or diversion of HEU fuel used in non-power reactors and the consequence to public health, safety, and the environment from such thef t or diversion." As a consequence of the t.mendment, Manhattan College will be replacing the HEU fuct in the Manhattan College Zero Power Reactor (MCZPR) with LEU fuel.
The design specifications and drawings for the new fuel e.lcments have been prepared by EG & G Idaho,Inc. and approved by Manhattan College; analysis of the operation of the reactor und projected core parameters have been developed by the RERTR staff at Argone Material Laboratory (ANL); and Babcock and Wilcox of Lynchburg, Virginia is fabricating the fuel elements. This Safety Analysis Report (SAR) assesses the implications on safety of the changes in core parameters, and cf the steps in the conversion process from the HEU to LEU fuel.
1.2 The MCZPR The Manhattan College Zero Power Reactor (MCZPR) is lccated on the campus of Manhattan College in Riverdale, a residential section Gf the Borough of the Bronx in New Y0rk City. The reactor is situat:d in a Nuclear Engineering Facility within the Leo Engineering Building at 3825 Coricar Avenue The Nuclear Engineering Facility, and particularly the Zero Power Reactor, are however designed for isolation from the rest of the engineering building to provide additional security and safety.
The Nuclear Engineering Facility contains the Zero Power Reactor (a critical reactor),
a graphite moderated suberitical reactor, and a light water moderated suberitical reactor. The physicallayout includes a separate room containing the top of the critical reactor vessel and the control console, a basement containing the bottom of the critical reactor vessel and auxiliary equipment, a separate room containing tbc two suberitical reactors, a separate counting room and a separate classroom.
The critical reactor is a low power, pool type reactor, designed for a maximum power level of 0.1 watt by AMF Atomics of Greenwich, Connecticut. Prior to installation at Manhattan College in 1964, the reactor core had been used in 1961 by AMF Atomics in PTR (pressurized tube reactor) low critical research experiments at the IRL (In-dustrial Reactor Laboratory) reactor site in Plainsboro, New Jersey.
The reactor is a heterogeneous pool type reactor, light water moderated and fueled with 92 percent enriched uranium. It is intended solely for teaching and training.
Consequently the maximum power level of 100 milliwatts authorized first by the Atomic Energy Commission and later by the Nuclear Regulatory Commission is quite adequate to serve these purposes. It also guarantees a high degree of safety. The radiation intensity at the reactor deck level directly above the core is only 1 mR/hr when operated at the maximum allowed power.
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o There are two control rods in the reactor: one cadmium-stainless steel shim rod and one stainless steel regulating rod. The two detecting instruments in the reactor are a BF 3neutron detector and an uncompensated ion chamber. Two Geiger-Mueller counters are used for area radiation monitoring. The control console is located near the reactor.
vessel and contains all the necessary control switches, lights, and instrumentation required to operate the reactor efficiently and safely throughout its designed power range, i
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, 2.0 SITF AND FACILITY DESCRIPTION 2.1 Site Parameters Manhattan College is situated along Manhattan College Parkway on the heights above Van Cortlandt Park, in the Riverdale section of the Bronx, New York City, just a few blocks south of the Yonkers City line.
The Zero Fvwer Reactor is located on the first floor of the Leo Engineering Building (two blocks from the main campus) on Corlecr Avenue. This building has been owned and occupied by Manhattan College since 1963. The Zero Power Reactor has been in the same building and location since its installation at Manhattan College. The location of the building and its relationship to its surrounding is indicated la Figure 2.1.
The Leo Engineering Building provides classrooms, laboratories, library, and computer facilities for an estimated 1800 students at any one time.
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Thirteen of the trenty acres on which the majority of College buildings are constructed are situated over a rock formation known as the Fordham Schist which extends from this site south into the northern Manhattan Boroagh under the Harlem River. The Leo Engineering Building in which the reactor is located is constructed on alluvial fill which extends to a depth of some one hundred and fifty feet.
Hydroloev No wells have been drilled on the Manhattan College campus nor in its vicinity. The water supply of the College is part of the New York City water system supplied principally through Croton Reservoir in Westchester County.
Surface runoff water is collected in concrete-lined storm drains which empty into the New York City sewage disposal system. This drainage system has been adequate to prevent any flooding of the campus by heavy rains. It is conceivable that the Leo Engineering Building in which the reactor is located, constructed as it is on filled creek bottom land could have a drainage problem in a very severe rain storm. However, this has not occurred over the twenty-five years in which the building has been maintained by Manhattan College. Such an occurrence would not create a radiation hazard.
Seismology New York City has been seismically inactive in the geologic past and it is extremely improbable that an carthquake will occur here in the near future. The reactor site should be free from ay carthquake hazard. However, the Leo Engineering Building (which is structurally independent of any other building) does conform to the building code of the City of New York and should withstand even a severe shock. In the event of a rupture of the reactor vessel, loss of water will reduce moderation and scram the reactor. The water in the vessel would then be contained within an independent overflow area in the basement of the reactor facility.
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f Meteorology New York City generally has a warmer climate in winter and a cooler climate in summer than most other cities located in the same latitude. The average annual precipitation is forty-two inches. . Northwest winds prevail from November to April.
From May to October south or southwest winds are dominant. Winds of gale force are seldom experienced. Occasionally during the months of August and September tropical storms of hurricane force are felt in the City. At these times high winds of almost fifty miles per hour are experienced. Such conditions do not present a danger to the structural integrity of the Leo Engineering Building or the Zero Power Reactor facility.
Since the facility produces no gaseous effluent, the frequent temperature inversions in the area present no radiological hazard to the public from the facilities.
Demograohv and Land Use The demographic patterns and land use in the vicinity of the reactor building, including the Manhattan College campus buildings and local residences and businesses, have chang::d little since initiation of reactor operations in 1964.
Figure 2-2 is a map of the area surrounding the Leo Engineering Building, showing individual residence structurcs, businesses, and other land use designations. The accompanying Tabic 2.1 provides data on occupanc.y for the structures delineated on the map.
2.2 Reactor Fncilitv nnd Systen13, 2.2.1 Facility Description The MCZPR laboratory is located at the southea::t corner of the Leo Engineering Building of Manhattan College. The floor plans for the first and second floors of the building are shown in figures 2-3 and 2-4 respectively, with the ZPR area shaded.
j The only access from the first floor of the Leo Engineering Building is shown through door D1 (Figure 2-3) which is kept locked and bolted from inside at all times. Door D2 on the first floor leads to a staircase to room 221 from which access to the ZPR room is through door D4 (see Figure 2-4).
I Access to room 221 from the second floor of the Leo Engineering Building is through door D3. Access to the control console in the ZPR Room on the second floor is through !
door D4 from Room 221. Door D4 is visible to the operator at the console. The access !
doors D2 and D3 to Room 221 are provided with Fox Police locks.
I Room 107 on the first floor (Figure 2-3) is used as a Counting Room and Room 108 as a Briefing Room. Room 221 on the second floor contains a graphite moderated suberitical reactor and a water moderated suberitical reactor.
The reactor tank made of aluminum sits on a concrete slab on the first floor. The tank is held in place by five aluminum brackets welded to the side of the tank near the bottom. The brackets are bolted to the concrete floor. ,
1 Several concrete piers were added to the first floor of the structure to strengthen it (Figure 2-5). There are concrete walls on three sides of the room. The base of the fourth wall consists of a concrete curb l'0" high sufficient to permit the room to contain the entire contents of the tank. The remainder of the fourth wall consists of a metal partition to separate the reaMor room from the ventilation equipment room
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Table 2,1 Estimated Occupancy of Buildings Shown on Map Legend _ Type of Building Estimated Occupancy Al 2 story detached 1 family 5-6 1 family 7 A2 1 story A3 2 story brick semi-detached 1 family 6-7 2 family 10 B1 2 story brick B2 2 story frame 2 famil.y 8 i Converted dwelling 2 family 8 B3 C2 Walkup apartment 3-6 family 15-25 I Walkup converted dwelling 15 C5 D1 Elevator apartment 300 D2 6 story apartment 320 G4 Gas station with Work Shop 5-10 each i
G8 Garage with Show Room 10-20 K9 Proposed Manhattan College Research 300-500 and Learning Center Unclassified Miscellaneous Buildings 10-50 S9 5
U4 Utility Substation V Va cant Lot -
None Manhattan College Main Campus 7.000-3000 None Leo Engineering Building 1000-2000 None Farrell Hall 10-25 None Paulian Labs 10-30 None Gaelic Park 100-500 None Interborough Rapid Transit (Car Barn) 20-40 7
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for the Leo Engineering Building. The door D1 on this wall serves as access for bringing heavy items into the room. This door is kept locked and bolted from inside the reactor area.
The south wall of the room is an outside wall. This wall is reinforced and protected by a sloping slab of concrete 6'0" wide. A demineralizing tank sits on the floor next to the tank. Pipes to and from the domineralizer are suspended by supports from the ceiling. The height of this first floor room is 7'4-1/2".
The reactor tank which is 8 ft. high extends upwards through the ceiling of the first floor. The reactor vessel is surmounted by a platform 2'2-1/2" above the top of the tank and 2'4-1/2" above the floor. The reactor vessel and the edge of the platform are protected with chain fences. The room is 16'4" high. A window, Wl,4'3" 'x 4'3" is located on the south wall 8'0" above the floor and protected by wirc mesh and is ,
secured w.ith lock and key.
The control console is 5'3" x 2'0" and 6'1" high. It is located 3'0" from the west wall of the ZPR Room. Plans of the first and second floors of the ZPR facility are shown in Figures 2-5 and 2-6.
Conventional fire protection is available throughout the Leo Engineering Building.
In addition, carbon dioxide fire extinguishers are available on both floors of the };
Reactor Laboratory.
The ZPR Laboratory containr, a forced circulating ventilation system consisting of a blower ano ast,ociated duct work which are designed so as not to return air from the laboratory back into the ventilation system of the Leo Engineering Building. A separate blower, controlled by a switch located on the west wall of the ZPR Room, returns air into the atmosphere.
2.2.2 Reactor l
The Manhattan College Zero Power Reactor (MCZPR) is a heterogeneous, pool-type i using solid enriched uranium fuel and is and is moderated by light water. The principal components of the reactor are the reactor vessel and its associated equipment, the control s;' stem, and the demineralized system. The maximum power allowed by the Nuclear Regulatory Commission is 100 milliwatts. !
Reactor Vessel The reactor vessel consists of a large aluminum drum cignt feet high and ten feet in '
diameter. The drum wall is one quarter of an inch th:ck. The reactor core is centrally located at the bottom of the vessel and consists of a grid plate and stand upon which the fuel elements are mounted. 1 The existing HEU fuel is 92 percent enriched U-235. There are fifteen full fuel elements and one partial fuel element is made up of six concentric cylinders of fuel and is protected by a thick aluminum shield.
i Support for the neutron detectors, control rod drive mechanisms and other control system hardware is provided by the reactor platform above the pool. Neutron and gamma-ray detectors are suspended in the pool while drive mechanisms are located 'I on a mounting plate on the platform. ll i
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Control System The control console is located near the reactor vessel and contains all the necessary control switches, lights, and instrumentation required to efficiently and safely operate the reactor throughout its designed power range.
There are two control rods, a cadium-stainless steel shim rod and a stainless steel regulating rod. By operating the control console switches, the operator can drive the neutron-absorbing control rods either into or out of the reactor core, as required to control the reactor power level.
Measured neutron levels from the neutron detectors are amplified and displayed on the control console instruments, thus permitting the operator to monitor reactor performance. Several of these instruments can shut down the reactor automatically by providing a signal that will cause the control rods to be driven or dropped into the core. These controls insure that the reactor will always operate safely.
Aren Monitors One area monitor (gamma-1) is located at the level of the reactor deck and a second area monitor (gamma-2) is located at side of the reactor vessel about the height of the reactor core. When operated at the maximum allowed power, gamma-1 reads imR/hr and gamma-2 reads 2mR/hr.
2.2.3 Reactor Coolant System and Connected Systems The water in the reactor has an enormous heat capacity relative to the power rating (0.1 W). No recirculating cooling system is, therefore, provided. The total heat capacity of the pool is about 65 MJ/oC.
Water lost due to evaporation is replenished with city water. The water from the city system is passed through an electrically controlled check valve, a demineralized, a flow meter and a short flexible hose over the top of the reactor tank. The check valve insures no back flow of water from the reactor in case of pressure loss in the water supply system. The flexible hose is removed from the reactor tank when it is not in use. The water level in the tank is maintained at about 7 fect.
Figure 2-7 shows a schematic diagram of the demineralized system. The only tank wall penetration is a 3/4 inch aluminum coupling located 2" from the tank bottom, with a 3/4 inch short nipple and a 3/4 inch aluminum gate valve.
The pump is connected to a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> clock which is used to activate the pump switch.
The system is run almost continuously except for daily rest periods. The flow starts at the bottom of the tank and passes through the pump, heat exchanger, demineralized and into the tank through a goose neck over the edge of the tank. Valves are provided to alter the direction of flow through the demineralizing column. The steam-to-water heat exchanger was installed to study the temperature coefficient of reactivity.
The reactor tank bottom is cleaned several times a year using a pool vacuum. The vacuum head is connected to the inlet of the pump through a flexible hose. The demineralized column is bypassed (Figure 2-8) and the water returns to the tank through a gooseneck pipe at the end of which a polyester filter bag is attached. The sediments collected in the filter bag and the used bags are retained for testing by the Health Physicist and appropriate disposal. The vacuum head and flexible hose are disconnected and stored when the system is not in use.
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2.2.4 Waste Mn nacernent The reactor does not produce any solid radioactive waste either in the form of spent fuel or as a radioactive by-product. The reactor does not generate any radioactive water. When the demineralizing resin is replaced, a small amount of water (about 3 gallons)is extracted along with the resin. This water is stored in marked containers for testing by the Health Physicist and proper disposal.
The demineralizing resin is replaced about two or three times a year. The used resin is kept in marked containers for testing by the Health Physicist and appropriate disposal. No radioactivity has ever been found in the resin.
2.3 Current Reactor and Core Parameter and Limits The design characteristics of the MCZPR are described in section 5.0 of revision 4 of the Technical Specifications For The Manhattan College Zero Power Reactor, dated March 15,1985 which are incorporated as Attachment I to this document. A proposed updated version (Revision 5) of the Technical Specifications, dated February 20,1989, incorporating only required revisions in current parameters is included as Attachment II. The Technical Specifications provide among other information the following parameters: l Section 5.0 ' Reactor, core, control, and coolant design features.
2.0 Reactor safety limits and limiting safety system settings 3.0 Limiting conditions for operation Section 3.1 of this document provides the design parameters for the LEU core, and Figure 3-1 shows the fuel element configuration in the core. !
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3.0 DESCRIPTION
OF THE CONVERSION PROCESS 3.1 New F.u_cl Element Description The original Al-1100 claded, U Al alloy-type Highly-Enriched Uranium fuct (HEU 92%)in the Manhattan College Zero Power Reactor (MCZPR) will be replaced by the Al-606) claded, U3Si2-Al Low-Enriched Uranium Fuel (LEU 19.7 5 +0.2%). While all the LEU fuel elements have essentially the same dimensions as the HEU fuel elements, the Uranium-235 loadings in the 15 full fuel elements and 1 partial fuel element will be changed. In order to increase the reactivity worth of the regulating rod, the arrangement of these 16 fuel elements will be changed into a more symmetric geometry by having the full fuel element moved from position 46 to position 14. The LEU core will use the same control rods (1 shim rod and I regulating rod) and manual-controlled emergency shutdown rod that are currently used in the HEU core. The design data and core configurations are provided in Table 3.1 and Figure 3-1 respectively.
TABLE 3.1 Design Data of HEU and LEU Cores Design Data HEt1 Core LEU Core No. of Standard Fuel Elements 15 15 No. of Partial Fuel Elements 1 1 (Ring 2)
Fuel Type U-Al alloy U3Si2-Al Enrichment, % 92.0 19.7 5 +0.2 Uranium Density, g/cc 0.7 4.8 No. of Fuel Rings per Element 6 6 No. of Fuel Plates per Ring 3 3 U-235 per Standard Fuel Element, g 200 235 U-235 per Partial Fuel Element, g 24.0 27.4 Fuel Meat Thickness, mm 0.51 0.51 Cladding Thickness, mm 0.38 0.38 Cladding Material 1100 At 6061 Al Natural Boron Impurity Equivalent in Cladding & 10 20 Structural Al, ppm 3.2 LEU Core Parameters The following discussion of core parameters is based upon the design and safety analyses prepared by the Argonne National Laboratory (RERTR Program)in " Analyses For Conversion of the Manhattan College Zero Power Reactor From HEU to LEU Fuel" which is included in accompaniment to this SAR.
Due to the substantially increased amount of U-238 in each of the new LEU fuel elements, an increase of U-235 in all 16 fuel clements to compensate for neutron loss is necessary. In order to keep essentially the same safety margins as in the original ,
HEU core, most of the parameters in the LEU core have been recalculated. Based on l the simulation data provided by the Argonne National Laboratory, only those parameters which could change as a result of HEU/ LEU core conversion, and only 17
Figure 31HEU and LEU Cores l HEU CORE p Shlm Rod soure. (
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those which would appear in the proposed Technical Specifications for the LEU core (Rev.6), are listed in table 3.2. The technical specifications for the LEU Core (Revision
- 6) are included as Attachment 111 to this Report.
TABLE 3.2 Parameters of HEU and LEU cores Core Page No. in Reactor Parameters HEU LEU Tech Spec (rev. 6)
Excess Reactivity, % k/k 0.32-0.40 1.1 +0.4 3-1 (with -1.0% k/k Bias to LEU Core)
Worth of Reg. Rod, % k/k -0.9 -1.3 32 (with +0.3% k/k Bias to LEU' Core)
Shutdown Margin, % k/k -0.5 -0.6 3-1 (with Shim Rod Stuck Out)
Worth of Shim Rod, % k/k -2.5 -3.4 3-2 3.3 Description of Fuel Removal and Replacement 3.3.1 Stens in Removal and Replacement Processes During the process of HEU/ LEU core conversion, each HEU fuel element will be removed from the core and lowered into the fuel container (fuel cask) supplied by the EG&G Co. A sufficient number of containers will be obtained such that all 16 fuel elements (15 full and 1 partial fuel elements) can be sequentially removed from the core at one time, and then shipped to the DOE repository site. Immediately after the completion of HEU fuel removal, the new LEU fuel will be installed into the core, revising the procedure for HEU fuel element removal. In order to avoid an abrupt change of reactivity in the reactor core and to prevent the fuel elements from obstructing each other, all the outer (circumferential) HEU fuel elements will be removed prior to that of the central elements and a reverse process of installing the LEU fuel elements will be carried out from the center of the reactor core. The detailed removal and replacement sequence are schematically shown in Table 3.3 and Figure 3-2.
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l HEU Fuel LEU Fuel No. of Console Fuel Element No. Removal Order Insertion Order Meter Readings 25 (partial) 1 16 1st l
l 46 2 replaced by # 14 I 14 cmpty 15 13 3 14 2nd 12 4 13 22 5 12 32 6 11 3rd 43 7 10 54 8 9 55 9 8 4th 23 10 7 24 11 6 35 12 5 5th 45 13 4 44 14 3 6th 33 15 2 34 16 1 7th During the conversion process, continual readings on the neutron-flux and Gamma-radiation instrumentation will be taken. These readings will be recorded in the MCZPR logbook and will be used as reference for experiments (see Sec. 3.3.2) to be conducted after the core conversion is accomplished.
3.3.2 Eauinment nnd Instrumentation Requirements The tool currently available for HEU fuel element removal can also be used for the new LEU fuel element insertion,since all the new LEU fuel elements will be fabricated with essentially the same dimensions as those of the HEU fuel.
The handling and shipping equipment for the fuel containers will be provided by the EG&G Co. During the HEU/ LEU conversion process, the fuel containers (casks) will be shipped in-and-out from the first floor of the reactor room via the only accessible metallic door (29"W x 52"H). Three portable detcetors (I neutron monitor and 2 Gamma-ray detectors) will be used to count the variations in neutron flux and Gamma-radiation before and af ter fuel element replacement from the core and containers. Experiments to measure:
(1) Approach to Criticality (2) Reactor Period and Reactivity (3) Void Coefficient Measurement, and (4) Gamma-ray Spectrum in the Vicinity of the Core 21
will be conducted after core conversion to validate the calculated parameters and to assure the safety of subsequent reactor operation. For accuracy, a multi-channel analyzer along with a photo-multiplier tube will be used for Gamma-ray energy estimation (experiment [4]). An empty hold-down rod with about 118.25 cc air inside (considered as the " void" will be prepared to replace the regular lucite hold-down rod for the void effect testing (experiment [3]). The reactor console, which includes 3 strip chart recorders and other necessary control systems, will be used to conduct experiments (1) and (2), respectively.
3.3.3 Health Physics Supnort In addition to the routine radiation protection program conducted at the MCZPR,in accordance with the requirements of the Technical Specifications and the Radiation i
i Safety Manual, additional Health Physics support will be provided during the core conversion period. Dr. Stanley Malsky, the designated Health Physicist will be present at all times during the insertion of the new LEU fuel elements and the subsequent validation of core safety parameters. Dr. Malsky will monitor each phase of the process to assure personal and facility safety is being maintained and that radiation
!cvels are maintained ALARA. Detailed documentation will be made of the steps taken and instrument readings obtained. (See Section 5.3 for additional details).
3.4 Conversion Process Schedule It is anticipa'cd that the HEU fuel element removal, insertion of the LEU fuel elements, and follow-up experiments (see section 3.3.2) will be completed in one working session.
The HEU fuel elements will be loaded directly into the container provided as they are removed and will be available for shipment at that time. The shipment schedule will be established by DOE (EG&EG).
22 l -
4.0 ACCIDENT ANALYSIS 4.1 Introduction This section discusses the likelihood and consequences of postulated unanticipated variations in the conversion process, and of accidents occurring during and af ter the new LEU fuel elements are inserted, resulting from human error, equipment mal-function, or external natural phenomena.
It is to be emphasized that the MCZPR is inherently safe during normal operations, upset conditions, or postulated accident conditions. The primary safety features are the excess reactivity characteristics of the core, the redundant methods available to make the core suberitical, and the extremely low operating power level.
The core conversion will be conducted under the supervision of the Chief Reactor Supervisor. Safety-related activitics will be monitored by the Health Physicist and the Radiation Safety Officer. Routine operation of the reactor will not be initiated until it is validated that all fuel elements are in place, the control rods and instru-mentation are operating in accordance with the specifications, and all the calculated core parameters have been verified through repeated experiments.
4.2 Qocrational Accidents and Conditions Operational accidents during core conversion and subsequent testing, which may be caused by carcicssness, negligence,or mistakes of the reactor operator and/or supervisor not completely following routine and well-established operational procedures, are extremely unlikely in the MCZPR, since both neutron flux and Gamma-exposure will be monitored at all times during core conversion and reactor operation. Instrumentation monitoring will be performed by the Reactor Operator, Reactor Supervisor, Radiation Safety Officer, and Health Physicist to assure immediate and appropriate response to instrument readings. In addition, interlocks prevent inadvertent reactivity addition, and a scram system initiates rapid and automatic shutdown when safety margins are reached. Based on operational records during the last 25 years (since 1964), important parameters such as temperature coefficient and void coefficient, which may signif-icantly affect core reactivity, have always stayed within safety ranges under normal operational conditions. A forced circulating ventilation system is available to constrain the pool water temperature from fluctuation. A manual emergency shutdown rod will also be prepared for the unlikely case of a power failure during reactor operation with both control rods withdrawn (I shim rod and 1 regulating rod). The water level of the reactor tank will be maintained at 7 feet high providing a significant safety margin. Because of the extremely low power rating (0.1 watt) in the MCZPR, no measurable descent of pool water occurs as a result of normal reactor operations; however, a make-up water supply has beca equipped to replenish water loss due to evaporation. Any leaks from the reactor would lower the water level which in turn would activate the " Low-Water Scram" circuit shutting down the reactor.
4.3 Handling Accidents The MCZPR is primarily used for teaching and training purposes. As previously noted, the reactor operates at an extremely low power level of 0.1 watt, thus fuel element replacement after core conversion will not be necessary because burn-up will be minimal.
23
d During the core conversion, the only possible scenarios This could that occur could result in damage fuel. A fuel are handling accidents involving dropping f the reactor of a fuel element. ;
maximum drop of the fuel element from the second to the first floor o cicment) room (about 8-foot high)in terms of the fuel weight (about 1/8" 5.5 thickkilograms alu- per will yield approximately 37 Kilojoules potential energyi to the bendouter the outer minum cylinder. However, such a small amount of energy will h ne ther esults supporting cylinder nor crack the inner concentric fuel plates, It is also notedbased on t e r of the previously performed impact test on a dummy fuel i element. s (about 5-foot that during the core conversion the fuel element shipping conta neri in l nt high) will be placed directly underneath the fuel handling fuel is dropped into the container which would l'urther assure the element.
will be substantially reduced by the buoyancy force of the tan no damage to the fuel element. i Once the fuel elements are in place in the containers li and they arc closed, the conta idents.
will protect the fuel from impact resulting from any onsite hand ng acc 4.4 Fffect of Naturni Phenomena Any postulated natural chenomena (e.g. seismicf fects event, more severe weather, e occur during or af ter the core conversion process would not result in any e severe than discussed in sections 4.2 and 4.3 above.
4.5 Trnnsoortntion Accidents Tra tsportation of the fuel element casks to the i i designated their integrity repository will be th responsibility of the DOE (EG&G). The casks are designed to ma nta i n under any postulated severe accident that an accident conditions occurs that breachesthat acould shipping occurcaskduring transportal In the highly unlikely event ld i
l d and exposes the fuel element no impact,in terms of cle !
since burn-up of fuel has been negligible. f rsonnel.
would not result in any appreciable release of fission products or exposure o pe +
24
-- ~ _ ~ ~ - " ~ ~ - - ~ - - - - _ __ - - - - , _
M i
l 5.0 CONDUCT OF OPER ATIONS 5.1 MCZPR Organization The operation of the Manhattan College Zero Power Reactor (MCZPR)is supervised by the Manhattan College Reactor Operations Committee. The members of the Reactor Operations Committee report to the Reactor Administrator. The administrative reporting line from the Chairman of the Mechanical Engineering Department proceeds to the Dean of the School of Engineering, the Provost, the Executive Vice President, the President of Manhattan College, and the Board of Trustees of the Manhattan College Corporation. The Corporation owns the facility and has final legal and financial responsibility for the operation of the facility. Figure 5-1 shows the Organizational Structure of the Reactor Operation.
The Chief Reactor Supervisor, Radiation Safety Officer, and Health Physicist both report to and arc part of the Reactor Operations Committee. They have collateral responsibility with the Reactor Operations Committee for the review and evaluation of all proposed operations and procedures and audit of ongoing operations in order to assure that the reactor facility is operated in a safe and competent manner. That responsibility will apply to the core conversion process.
The Reactor Operations Committee meets semiannually and at other times if deemed necessary by the Reactor Administrator. The Committee will meet prior to initiation of core conversion operations to review the final plans for the operations.
The dutics of the Reactor Administrator, Reactor Operations Committee, Radiation Safety Officer, Reactor Supervisor, Chief Reactor Supervisor, and Reactor Operators are described in the Technical Specifications which are appended to the SAR. The Health Physicist performs all radiation surveys and samples under the supervision of the Chief Reactor Supervisor and other tests as requested by the Reactor Administrator or members of the Reactor Operations Committee. He also is responsible for monitoring records of exposure on film badges, maintenance of a log book on radiation tests and exposure records, and for a review of the reactor log book. The details of the core conversion process and subsequent validation of core parameters will be recorded in the reactor log book, and all exposure records for this operation will be maintained in the Health Physics log book and reviewed as required.
5.2 Emergency Planning An approved emergency plan for the facility exists and is appended to the SAR. The plan follows the guidelines stated in Appendix E to 10 CFR Part 50, U.S.N.R.C.
Regulatory Guide 2.6 (Revision 1, March,1983), and ANS 15.16- 1982. All participants in the core conversion process are familiar with steps to be taken should specific abnormal conditions arise.
5.3 Radiation Protection Durine Fuel Conversion Manhattan College is committed to conducting all reactor operations and associated maintenance and experimental activitics in a manner that a:sures the protection of all individuals, both on-site and in the surrour. ding environs. Radiation protection is carried out in a manner that is consistent with the applicable rules and regulations of the Nuclear Regulatory Commission and the State and City of New York, and with the specific conditions defined in our Special Material and Facility Operating Licenses.
The radiation protection program is based on the following premises:
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-- The College has a moral obligation to maintain personal health and safety with respect to all radiological ordinances which can never be compromised.
The Manhattan College Zero Power Reactor (MCZPR) is inherently safe because of its low power operation and built in design features
-- There is a need to inculcate reasonable and proper Health Physics procedures by requiring students to know and follow these regulations
-- Personnel safety must be the first consideration at all times, and no requirements will be allowed to override safety considerations An ALARA (As Low As Reasonably Achievabic) program will be conducted in l accordance with federal (NRC) guidelines The objectives of the Manhattan College radiation protection program are accomplished through the implementation of the components of a health physics program which will be in effect during the core conversion process. The primary purposes of this program are to assure the radiological safety of all College personnel, and to make certain that all sources of radiation are handled in accordance with Federal, State, and City regulations.
The MCZPR Radiation Safety Manual guides the activitics of faculty members and students using the Reactor, the supporting radiation facilities, and radioactive materials in the Nuclear Engineering Facility.
The procedural and radiation safety aspects of the Zero Power Reactor and isotope program are administered by the Reactor Operations Committee (ROC). The ROC, which is chaired by the Reactor Administrator, receives and evaluates all proposed operations and procedures in order to insure that the reactor facility is operated in a safe and competent manner. The Radiation Safety Officer (RSO)is responsible for enforcement of the rules, regulations, and operating procedures which conform with the NRC regulations (i.e.10 CFR Part 20) and the license conditions. The ROC anc i RSO will perform these functions for the fuel conversion process.
The services of the consulting Health Physicist will be employed to provide advice to the ROC and RSO in insuring radiation safety.
The Administration of Manhattan College and the Manhattan College Nuclear Engineering Facility are committed to a program for keeping exposures (individual and collective) as low as reasonably achievable (ALARA). In accord with this commitment, an administrative organization for radiation safety exists and is implemented by written policy, procedures and instructions which foster the ALARA concept within the lastitution. The organization includes the Reactor Operations Committee (ROC) and Radiation Safety Officer (RSO).
As a componcnt of the ALARA program, doses to individuals are maintained as far below the limits as is reasonably achievable, and the sum of the doses received by all exposed individuals is also maintained at the lowest practicable level.
On-site workers during the core conversion process will be instructed in the ALARA concept and its relationship to his working procedures and work conditions.
The on-site worker will be required to know what recourses are available if he feels that ALARA is not being promoted on the job.
27
The Manhattan College Nuclear Engineering Facility has established Investigational Levels for Occupational external radiation exposure which, when exceeded, will initiate review or investigation by the Reactor Operations Committee and/or the Radiation Safety Officer. The investigational Levels that we have adopted are listed in Table 5.1 below. These levels apply to the exposure of individuals.
TABLE 5.1 Investigational Levels
- (mrems per calendar quarter)
LEVEL I LEVEL II Whole body; head and trunk; active blood- 125 375 1.
forming organs; lens of eyes; or gonads Hands and forearms; feet and ankles 1875 5625 l 2.
Skin of whole body 750 2250 3.
Quarterly Exnosure of Individuals to Less Than Investigational Level 1 Except when deemed appropriate by the RSO, no further action will be taken in those cases where an individual's exposure is less than Table 5.1 values for Investigational Level 1.
Personnel Exoosures Ecual to or Greater Then Investigational Level L but Less Than Investigational Level 11 The RSO will review the exposure of each individual whose quarterly exposures equal or exceed Investigative LevelI. If the exposure does not equal or exceed Investigational Level II, no action related specifically to the exposure is required unless deemed appropriate by the ROC. The Committee will, however, consider cach such exposure in comparison with those of others performing similar tasks as an index of ALARA program quality and will record the review in the Committee minutes.
Exoosure Ecual to or Greater Than in >csticational Level 11 The RSO will Investigate in a timely manner the causc(s) of all personnel exposures equaling or exceeding Investigational Level 11 and if warranted, take action. A report of the investigation, actions taken,if any, and a copy of the individual's Form NRC-5 or its equivalent will be presented to the ROC at the first ROC meeting following completion of the investigation. Committec minutes will be sent to the administration for review.
Re-establishment of An Individual Occupational Worker's Investientional LevtL11 Ahpqye That Listed in Table 5.1 In cases where a worker's or a group of worker's exposures need to exceed Investigative Level .II, a new, Iligher Investigative Level 11 may be established on the basis that it is consistent with good ALARA practices for that individual or group. Justification for a new Investigative Level 11 will be documented.
28
~ _ _ - - _ __ _ _ _ _ - _ _ . _ _ _
The Reactor Operations Committee will review the justification for, and will approve, all revisions of Investigative Levels II. In such cases when the exposure equals or exceeds the newly established Investigative Level II, those actions listed above will . ,
be followed.
A During routine Reactor and Laboratory operations (including core conversion),
external personnel monitoring is accomplished by the use of film badges. Reactor operations personnel normally working in the reactor rooms wear beta-gamma-neutron film badges; visitors are required to wear film badges that record gamma radiation. '.
Badges are processed by a commercial vendor, and records of personnel exposures are maintained in the reactor area. -
In the event that any significant internal exposure is suspected to have occurred, bioassays of urine samples may be required. If analysis of the sampic verifics overexposure, follow-up medical examinations including blood analysis may be per-formed. An investigation of the cause of over exposure will be conducted, the situation remedied, and verified by the ROC.
All personnel performing the core conversion process will be monitored and area contamination surveys made to assure that no dispersal of radioactive material has occurred. Administrative controls (special work permits) are required for outside contractor personnel working in these areas.
Survey meters, located in the ZPR room, are used for area monitoring, and for monitoring of hands and clothing of individuals, suspected of picking up some contamination during the course of experimental activitics.
e In addition to the program described above other radiation protection measures are routinely performed at the Manhattan College Nuclear Facility. These include:
- All personnel using the Facility are instructed in radiation protection prior to participating in any of the activitics at the Facility. All students are required to familiarize themselves with the provisions of the Radiation Safety Manual.
Each area, room, or enclosure in which radioactive materials exist is posted with the specified radiation sign, as defined by Appendix B of 10CFR 20 or NBS69.
- No beverages, smoking foodstuff, or application of cosmetics is permitted in radiation zones.
- Protective clothing (e.g. gloves) and tools (e.g. tongs) are employed to avoid contact between radioactive material and the skin.
- A contamination control program exists involving the use of designated container and labels, controlled and labelled storage, cleanliness and rnaintainence standards for laboratory surfaces, and procedures for cicanup of spills, and for decon-tamination of equipment and structural surfaces.
1 ATTAC1E!ENT I TECHNICAL SPECIFICATIONS FOR THE
~
MANHATTAN COLLEGE ZERO POWER REACTOR FACILITY LICENSE R-94 l
MARCH 15, 1985
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Rev. 4 REVISION RECORD Revision No. Comment 0 Submittsd to U. S. N. R. C. on August 26, 1983 as part of the Safety Analysis Report for the Manhattan College Zero Power Reactor (MCZPR).
t 1 Complete revision submitted on January 12, 1984.
2 Complete revision submitted on November 15, 1984.
3 Revisions as noted on pages:
16-1, 16-2, 16-10, 16-11, 16-13, 16-16, 10-23, 16-25, and 16-28, on December 19, 1984.
I '
4 Page numbering changed from 16-1 to 1-1, etc.
Cover sheet revised from section 16 to Appendix A per USNRC request.
Revisions as noted on paJes i, ii, iii, and 1-1, 2-1, 2-2, 3-1, 3-2, 3-3, 3-4, 3-7, 4-1, 4-2, 4-4, 4-8, 4-11, 5-1, 5-2, 5-3, 5-4, 5-5, 6-1, 6-6, 6-8, 6-9, 6-11, 6-13, and 6-14 on March 15, 1985.
-i-
Rev. 4 TABLE OF CONTENTS _
1 Contents Page No.
Title Pac ........ .......................... ..... ..... ... i Revisien Record ... .. ..... .. . ..... ....... ... ii Table of Contents . .. . ... . ... .... ..... .. iii 1.0 DEFINITION . . ...... .. . . . . .... .. . ... 1-1
, 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS .. . 2-1 3.0 LIMITING CONDITIONS FOR OPERATION ..... . . . 3-1 4.0 SURVEILLANCE REQUIREMENTS ... ......... .... .... 4-1 5.0 DESIGN FEATURES . ...... . . .. .. ..... . . 5-1 6.0 ADMINISTRATIVE CONTROLS .. .. . .. .... . . . 6-1 Table 3-1, Safety System .... . . ...... .... . . 3-4 Figure 6-1, Table of Organization . ....... .. .... . 6-2 1
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Rev. 4 1.0 DEFINITIONS .
The terms Safety Limit, Limiting Safety System Settings, and Limiting Condition for Operation are as defined in paragraph 50.36 of 10 CFR Part 50.
ALARA - ALARA is a concept introduced by the Nuclear Regulatory Commission (NRC) to all reactor facilities. The basis of ALARA is that all exposure to radiation should be kept "as low as reasonably achievable" (ALARA).
channel - A channel is the combination of sensor, line, amplifer, and output devices which are connected for the purpose of measuring the value of a parameter.
channel calibration - A channel calibration is an adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a channel test.
channel check - A channel check is a qualitative verification of acceptable performance by observation of channel behavior. This verification, where possible, shall include comparison of the channel with other independent channels or systems measuring the saiue variable.
channel test - A channel test is the introduction of a signal into the channel for verification that it is operable.
/ control rod - Plates fabricated with neutron absorbing material used to establish neutron flux changes and to compensate for routine reactivity losses. This includes safety-type and regulating rods.
care - The portion of the reactor volume which includes the fuel elements, the source, and the control rods.
delayed neutron fraction - When converting between absolute- and dollar-value reactivity units, a beta of 0.00645 is used.
drop time - The elapsed time between reaching the complete removal setpoint and the full insertion of a safety-type rod. (It must be less than 1.0 second).
cycess reactivity - Excess reactivity is that amount of reactivity that would exist if all control rods (shim, regulating) were moved to the maximum reactive condition from the point where the reactor is exactb oritical (K 1).
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1 exteriment - Any object, other than a fuel element or handling tool, ;
which is inserted into the volume formed by projecting the rigid plate vertically to the tank pool water surface is to be regarded as e an experiment in the core.
measured value - The measured value is the value of a parameter as it '
appears on the output of a channel.
mayable eX9erinent - An experiment where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor pool water.
operable - Operable means a component or system is capable of .
performing its intended function.
l operating - Operating means a co ponent or system is performing its intended function.
reactivity limils - The reactivity limits are those limits imposed on reactor core excess reactivity. For the MCZPR the reactivity limits are 0.44% A k/k (0.68$) at 110.6*F.
reactivity worth of an exoeriment - The reactivity worth of an experiment is Lne maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration, reactor oonrating - The reactor is operating whenever it is not secured or shutdown.
, reactor coerat.or (RO) - An individual who is licensed to manipulate ,
the controls of a reactor.
react.or safety systems Reactor safety systems are those systems, including their associated input channels, which are designed to initiate automacic reactor protection or to provide information for initiation of manual protective action.
- reactor secured - A reactor is secured when:
- 1) It contains insufficient fissile material or moderator present in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or
- 2) A combination of all of the following:
a) All neutron absorbing control rods are fully inserted and other safety devices are in shutdown position, as required by the Technical Specifications, and b) The console key switch is in the off position and the key is removed from the lock, and c) No work is in progress involving core fuel, core structure.,
installed control rods, or control rod devices unless they are physically decoupled from the control rods, and d) No experiments are in or near the reactor.
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reactor shu_tdnan - The reactor is shutdown if it is suboritical by at least one dollar in the reference core condition and the reactivity worth of all experiments is between 60*F and 80*F.
l reculating rod - A low-worth control rod used primarily to maintain j an intended power level. Its position may be varied by operator i I
action.
research reactor - A research reactor is defined as a device designed j to support a self-sustaining neutron chain reaction for research J development, educational training, or laboratory purposes, and which d may have provisions for the production of radioisotopes.
reverse trio - The electronic setting within the console instrumentation which will initiate a reverse system and will drive in an electromagnet when certain specified limits are exceeded.
safetv-tvoe rod - A rod that can be rapidly inserted by cutting off the holding current in its electromagnetic clutch. This applies to both control rods.
scram tirae - The time for the control rods (shim, regulating) acting under the force of gravity to change the reactor from a critical to a suberitical condition. This will be equal to or less than the drop time.
scram trio - The electronic setting within the console instrumentation which will activate scram circuits when certain specified limits are exceeded.
senior reactor coerator - An individual who is licensed to direct the activities of a Reactor Operator (RO) and to manipulate the controls of a reactor.
shall. should and nav - The word "shall" is used to denote a requirement; the word "should" to denote a recommendation; and the word"may" to denote permission, neither a requirement nor a
, recommendation.
shutdown nnrain - Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made suberitical by means of the control and safety systems starting from any permissible operating condition although the most reactive rod is in its most reactive position, and that the reactor will remain suboritical without further operator action.
unscheduled shutdoEn - An unscheduled shutdown is defined as any unplenned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions which could adversely affect safe operation, not including shutdowns which occur during testing or checkout operations.
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o Rev. 4 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits 2.1.1 Applicability J
This specification applies to the inelting teinperature of the fuel cladding.
2.1.2 Objective To assure that the integrity of the fuel is maintained.
2.1.3 Specifications The safety limit shall be on the temperature of the fuel element cladding, which shall be less than 1220*F.
2.1.4 Bases The inelting ternperature of the alurninum used as cladding on the fuel eleinents is 1220'F. Therefore, in order to maintain fuel element integrity, the cladding temperature
' must not exceed 1220*F. As reported in Appendix C of the Safety Analysis Report for the MCZPR, the maxiinurn core temperature that can ever be reached is only 221*F and reaches this level only during the Maxirnum Hypothetical Accident. The specification, therefore, provides assurance on the integrity of the fuel within the cladding.
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2.2 Limiting Safety System Settings (LSSS) 2.2.1 Applicability l This specification applies to the setpoints of safety channels which monitor reactor power level.
2.2.2 Objective To assure that automatic trip action is initiated and that the operator is warned to take protective action against exceeding a safety limit.
2.2.3 Specifications The limiting safety system setting shall be en reactor maximum power level not exceeding 0.125 watt, or 125% of full power.
2.2.4 Bases Since there is no forced circulation cooling, the reactor core is cooled by the water surrounding the reactor core.
Therefore, the only paramcLer which could be used as a limit for the fuel cladding temperature is the reactor power. The analysis in Appendix C of the Safety Analysis Report shows that even for the Maximum Hypothetical Accident (a reactor power excursion of 147 kilowatts), the maximum core temperature reaches only 221*F (The tank water temperature would rise less than 10*F). This temperature is much lower than the temperature (1220*F) at which cladding damage could occur. Therefore, a large safety margin exists between the limiting safety system setpoint and the fuel safety limit.
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Rev. 4 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 3.1.1 Applicability These specifications apply to the parameters which describe the reactivity condition of the core.
3.1.2 Obj ective To ensure that the reactor cannot achieve prompt criticality and that it can be safely shutdown under any condition.
3.1.3 Specifications The reactor shall not be made critical unless the following conditions exist:
A. The total core excess reactivity with or without the movable experiments of section 3.8.3 shall not exceed 0.44% Ak/k (0.68$) at 110.6*F.
B. The minimum shutdown margin provided by control rods shall not be less than 0.46% Ak/k (0.72$) at 110.6*F. l C. Any change in the experimental apparatus shall be approved by the Reactor Operations Committee.
3.1.4 Bases Specification A is based upon the experimentally determined value for excess reactivity 0.44% ak/k (0.68$) at a reactor pool water temperature of 110.6*F.
i Specification B is based upon the negative worth of the regulating rod: that is, the control rod with the smaller negative worth.
Specification C limits the changes in the experimental apparatus to those approved by the committee charged with review and approval of experiments.
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l Rev. 4 3.2 Reactor Control and Safety System 3.2.1 Applicability These specifications apply to the reactor safety system and safety-related instrumentation.
3.2.2 -Objective To specify the lowest acceptable level of performance or the -
minimum number of acceptable conponents for the reactor safety system and safety-related instrumentation.
3.2.3 Specifications The reactor shall nct be made critical unless the following conditions exist:
A. The reactor safety system shall be operable in accordance with Table 3-1.
B. There shall be two safety-type control rods:
A regulating rod with a negative worth of 0.30% Ak/k (1.40$)
and a shim rod with a negative worth of 2.50% Aly'k (3.88$).
C. The drop time for either safety rod shall not exceed 1.0 l second; measurements of rod drop times shall be made once semi-annually. [
D. The reactivity insertion rate _ for a single rod shall not exceed 0.10% A k/k (0.154$) per second.
3.2.4 Bases Specification A provides assurance that the reactor safety system which may be needed to shut down the reactor is operable. Each feature of the system is described in Table 3-1.
- A scram system is provided that causes interruption of the magnet current to the electromagnets, should a scram trip be exceeded. The control rods then fall into the reactor core under the force of gravity. This system provides a conservative response to an instrumentation system failure, electric power failure, low water level, high neutron flux, and high gama activity.
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Rev. 4
- A reverse system is provided that lowers the control rods into the reactor should a reverse trip be exceeded. This system provides a conservative response to the high/ low power reading from each channel's recorder, and to the power being off on each recorder.
- A bypess system is pro'.<ided that causes the elimination of reverse circuit in the linear channel or the gamma channel.
This system provides a special controllable function for the reactor operation during initial fuel loading and during J-reactor startup.
Specification B provides assurance that reactor can be operated safely at the critical state because these negative worths make it possible to shutdown the reactor rapidly.
Specification C provides assurance that both reactor safety rods can be fully inserted into the core to decrease the power level within 1.0 second.
Specification D assures a safe rate of power change during startup and during power ascensions.
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Rev. 4 Table 3-1 Safety System The following circuits shall be functioning whenever fuel is in the reactor and power is available to the control devices.
- 1. Screm Circuits A scram system shall be provided that shall cause interruption of l the magnet current to the electromagnets supporting the control rods whenever a scram trip is exceeded. Power to the magnets shall be available when the " reactor on" switch is on and there are no scram trip signals. A scram trip shall be provided for each of the conditions below, with the trip setting as specified.
- a. High neutron flux - Count Rate Channel electronic trip set for 125% or less of Full Power (Full Power shall be equal to 0.1 watt).
- b. High neutron Flux - Linear Channel electronic trip set for 125% or less of Full Power.
- c. High gamma activity - high level signal from either of the two Gamma Channels electronically set for 10 mR/hr or less.
- d. Hanual scram - operates upon actuation of the manual scram button on the console.
- e. Low water level - operates when the tank water level drops one foot below the tank full position. Tank full position is defined as seven feet above the bottom of the reactor vessel.
- f. Reactor key switch off - operates when the " REACTOR ON" switch is turned to the off position,
- g. Power failure - operates whenever the power supply.to the console or to the nuclear instrumentation fails.
- 2. Reverse Circuits A reverse system shall be provided that shall cause both control l rod devices to drive the control rods into the reactor whenever a reverse trip is exceeded. The reverse action shall override any rod selection made by an operator and shall persist as long as a reverse trip is exceeded. Reverse trip shall be provided for
' each condition as below with the trip setting as specified:
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Table 3-1 Safety Systent. (continued)
- a. Count Rate Channel reverse trip shall occur for any of the following conditions:
(1) Count Rate recorder off.
(2) Count Rate recorder down scale - shall occur when the recorder indicates less than 2 counts per second.
(3) Count Rate recorder up scale - shall occur when the recorder indicates greater than 50,000 counts per second.'
- b. Linear Channel reverse trip shall occur for any of the following conditions:
(1) Linear recorder off.
(2) Linear recorder down scale - shall occur when the linear recorder is less than 5% of full scale.
(3) Linear recorder up scale - shall occur when the linear recorder is greater than 95% of full scale.
- c. Ganuna Channel reverse trip shall occur for any of the following conditions:
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(1) Gamma recorder off.
1 l (2) Ganuna recorder down scale - shall occur when the recorder indicates less than 0.2 mR/hr.
(It should be noted that the minimum reading on the recorder is 0.1 mWhr while the minimum reading on the instruments on the Gamma Channels for Area Radiation Monitoring is 0.01 mR/hr).
l (3) Ganuna recorder up scale - shall occur when the recorder indicates greater than 95 mR/hr.
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- d. Any scram c ondition shall cause the control rod devices to drive in the electromagnets.
, e. Manual run-in trip shall occur upon actuation of the "Run-In" switch on the control console.
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4 Table 3-1 Safety System. (continued)
- 3. Bypm s in Safety Systems The only bypasses in the scram or reverse circuits shall be those described below. The bypasses shall be key operated switches located on the console.
- a. A bypass to eliminate a reverse as a consequence of the genutta recorder being down scale utay be utilized during startup until the ganuna recorder reads on scale.
- b. A bypass to eliminate a reverse as a consequence of the linear recorder being down scale may be utilized in the initial fuel loading while conducting experiments to determine suboritical multiplication.
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Rev. 4 3.3 Coolant Condition I
Because of the low value of the maximum steady power level (0.1 watt), no recirculating cooling equipment or systems are required other than the pool of water maintained in the reactor tank. The pool water is basically used as the moderator to slow down the fast neutrons; however, due to its large heat capacity, the pool water can also be used to dissipate heat from the reactor core.
3.3.1 Applicability This specification applies to the coolant condition of the pool water in the reactor tank.
3.3.2 Objective This specification insures that all the heat being generated in the reactor core can be dissipated by the pool water.
3.3.3 Specifications A, The maximum heat capacity of the pool water shall be 65 MJ/*C, with a minimum water level requirement of seven feet as measured from the tank bottom.The minimum heat capacity of the pool water shall be 56 MJ/*C, with a water level of six feet.
B. The lower limit of water resistivity based on the requirements :
for the corrosion protection system shall be set at 0.1 Megohm
-cm. There shall be no upper limit for the water resistivity.
3.3.4 Bases Specification A provides assurances that the pool water, due to its large heat capacity relative to the low steady power level (0.1 watt), can completely remove and dissipate heat in the reactor tank.
Specification B provides the minimum resistivity of water in the reactor tank. If the water resistivity is less than this level of 0.1 megohm-cm, the resin bed in the deionizer is i
replaced. The resistivity of the water has little or no l bearing on the corrosive effects experienced or possible other i than to offer some guarantee that catalyzing anions are
! excluded from the system. In the absence of the other l protective mechanisms that are in place, corrosion will occur l
regardless of the resistivity of the water. The supplemental l protective device selected for use is a dynamic cathodic protection system operating at 10 volts and an expected current density of 25 to 50 milliamps. The lower limit of 0.1 Megohm-cm has been selected to allow this system to provide an electron saturated barrier layer over the entire internal surface of the tank thus preventing the development of any galvanic cells as a result of discrete areas becoming anodic.
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3.4 Confinement or Containment No operations requiring confinement or containment are performed with the MCZPR.
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3.5 Ventilation System 3.5.1 Applicability This specification applies to the ventilation system within the MCZPR room.
3.5.2 Objective To ensure that the air in the reactor room is always clean and free of dust. t 3.5.3 Specifications A. The MCZPR Laboratory shall contain a forced circulating ventilation system consisting of a blower and associated duct work. There shall be no connection between this system
- and any other part of the building.
B. A switch shall be provided in the reactor room to turn the ventilating system on and off.
3.5.4 Bases Specification A shows that the ventilation system is independent of the building.
Specification B provides assurance that the ventilation system is controllable.
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3.6 Emergency Power No emergency power is supplied to the MCZPR. In the event of power failure while the reactor is operating, a scram trip shall occur.
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3.7 Radiation Monitoring Systems 3.7.1 Applicability These specifications apply to the radiation monitoring systems
/ and to the limits on the radiation detection level of each channel.
3.7.2 Objective To specify the minimum number of acceptable components or the lowest acceptable level of performance for the radiation monitoring systems.
3.7.3 Specification A. Two radiation monitoring channels shall be provided to measure ganuna intensity. These channels shall also be used to monitor reactor operation and shall be used in the reactor safety system as described in Table 3-1. Each channel consists of a Ganuna Detector and Ganuna Indicator Unit. A conunon strip chart recorder shall be provided with a selector switch for recording the output of either channel.
B. Each Geruna Detector shall be a sealed unit containing a Geiger-Mueller tube, transistorir.ed count rate amplifier, and check source. The output from the Detector shall be logarithmic with respect to the radiation level. The check source shall be exposed to the Detector by a solenoid which is actuated by a push button on the control console.
C. One of the Detectors (Gamma 1) shall be located on the reactor platform directly over the core area while the other Detector (Ganuna 2) shall be mounted on the side of the reactor tank.
D. The Gamma Indicator shall contain the power supply for the system, the alarm reset check source control, and the output connector for the Detector. Also contained on the front of the Indicator ia a logarithmic meter relay for indication and alarm of the camma level. The alarm shall be set to give audible annunciation whenever the radiation level exceeds G ruWhr for Ganuna 1 and 10 mVhr for Garuna 2.
E. The range of both detectors shall be from 0.01 to 100 mR/hr.
The systein shall be designed so that if the radiation intensity is greater than 100 mR/hr, the detector shall indicate full scale.
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3.7.4 . Bases Specification A provides the functions and components of two radiation atonitoring channels end provides assurance that these two channels can also be used as safety-related channels.
Specification B provides the composition of each detector and provides assurance that each detector can function well by using the check source.
Specification C provides assurance that radiation in both radial and axial directions can be detected.
Specification D provides assurance that the safety alarm system will be activated when radiation exceeds the allowable limit.
Specification E provides the rar.ge and design configuration of the monitoring system.
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3.8 Experiments -
t 3.B.1 Applicability u
These specifications apply to the experiments installed in the reactor.
3.8.2 Objective
~ To prevent dataage to the reactor and release of radioactive material'in the event of experiment failure, and to avoid exceeding any safety limit.
3 8.3 Specifications Limitations on experiments and material. irradiations in the reactor shall meet the following conditions:
A. No experiment. shall be installed in the reactor in such a location that any part of the apparatus will touch or in any way interfere with the action of'the control rods.
B. No experiment shall be installed in the reactor that can shadow the nuclear instruments, thereby giving erroneous or unreliable information to the reactor operator.
C. No experiment which has explosive properties shall be irradiated.
D. Experiments containing materials whose release to the water could result in a violent chemical reaction (e.g. Sodium) or-would result in chemical or corrosive attack to the reactor components (e.g. Mercury) shall not be irradiated.
E. Experiments containing materials whose release could result in overexposure of personnel to gaseous or particulate radioactivity shall not be irradiated.
F. Each experiment, other than laboratory exercises defined in
. Item G shall receive the specific approval of the Reactor Operations Committee. In addition, all operations leading to the production of more than one millicurie of any radioisotope outside of fuel elements shall receive the approval of the Reactor Operations Committee.
G. Laboratory exercises - The following laboratory exercises, while not requiring the presence of a Reactor Supervisor, require the p2esence of a Reactor Operator. Laboratory exercises involving ase of experiments (other than experiments used in the following exercises) shall be conducted in the presence of a Reactor Supervisor.
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- 1. Startup and Operation of MCZPR: Approach to Criticality
- 2. Critical Mass Determination
- 3. Reactor Period and Reactivity
- 4. Void Coefficient Measurement
- 5. Flux Distribution in the MCZPR
- 6. Determination of Buckling
- 7. Meastirement of Diffusion Length and Age
- 8. Temperature Coefficient of Reactivity
- 9. Ganuna Ray Energy Spectrum in the Vicinity of the Reactor Core
- 11. A maximum of seven aluminum covered indiurn foils with a e combined maximum negative reactivity of -1.113 x 10%k/k may be used in a laboratory exercises (Flux Distribution in the MCZPR). A void with a maximum negative worth of -10.4x10%k/k may also be used in a laboratory exercises (Void Coefficient Measurement).
I. A record of each material irradiation shall be included in the reactor log. The record shall include at least the following data:
Material irradiated Position in core Reactor power Irradiation time, tiine in, time out Dose rate on contact at time of removal Supervisor's signature 3.8.4 Bases Specifications A, B, C, D, E, F, and I are based on I requireinents stated in the Standard for the Development of Technical Specifications for Research Reactors, ANSI /ANS -
15.1 - 1982.
Specification G lists all the laboratory exercises that have been performed with the MCZPR for teaching and training j purposes. There are no limiting conditions for these exercises except item 5 - Flux Distribution in the MCZPR.
With regard to Specification H, for the laboratory exercise involving the irradiation of the seven indium foils, the reactor is always scrammed before these foils .are removed.
Also, in the laboratory exercise for determination of the void coefficient, both control rods are fully inserted before the rod is fully removed.
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3.9 Facility Specific LCO No limiting conditions for operations (LCO) unique to the
- facility other than those listed above are necessary.
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i Rev. 4 4.0 SURVEILLANCE REQUIREMENTS Surveillance tests, except those specifically required for safety when the reactor is shutdown, may be deferred during reactor shutdown: however, they must be completed prior to i reactor startup. j j
l 4.1 Reactor Core Parameters 1 4.1.1 Applicability These specifications apply to the surveillance activities required for reactor core parameters.
4.1.2 Objectives These requirements give the frequency and type of testing to assure that the reactor core parameters conform to the Specifications of section 3.
- 4.1.3 Specifications A. Total core excess reactivity shall be checked immediately after any change in the core configuration. All core configuration changes shall have prior approval by the Reactor Operations Committee and the Nuclear Regulatory Commission.
B. The shutdown margin shall be checked at least semi
-annually by comparing the two control rod positions dut ng the laboratory exercises of Approach to Criticality (see 3.8.3, G. 1).
C. Any change in the experimental apparatus shall be approved and documented by the Reactor Operations Committee.
4.1.4 Bases Specification A to preclude operating the reactor without adequate shutdown capability provides assurance on the core excess reactivity.
Visual confirmation and the experimental measurements of Specifications A, B, and C are sufficient to provide assurance that the reactor core parameters are as specified in 3.1.3 A, B, and C.
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Rev. 4 4.2 Reactor Control and Safety System 4.'2.1 Applicability These specifications apply to the surveillance activities required for the reactor control and safety system, 4.2.2 Objective LTo specify the frequency and type of testing or calibration to assure that the reactor control and safety system conforms to the specification of section 3 of these Specifications.
~4.2.3 Specifications' A. It is assumed that the worth of the control rods remains unchanged . However, the relative worths of the control rods shall be' checked annually by comparing the criticality positions of the control rods with the criticality positions of previous years.
B. The rod withdrawal speeds shall be measured at least twice a year, and shall be such that the maximum speed is no greater than 12 inches per minute in the MCZPR.
C. The operability of the control rods and driving mechanisa shall be tested daily when the reactor is operating.
D. An operability test, including trip action, of each safety channel listed in Table 3-1 that provides a scram function shall be completed prior to each reactor startup following a period when the reactor has been secured for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or at least weekly during continuous operating periods.
l E. A calibration of the channels listed in Table 3-1 shall be performed at least annually and whenever any maintenance on a channel which may affect its performance is completed.
4.2.4 Bases Specification A assures that relative worths of the control rods shall be checked annually in order to maintain the required shutdown margin. This specification also provides means for determining the relative worth of experiments inserted into the reactor core.
The rod withdrawal and insertion time measurement intervals required in Specification B verify the limits in Specification 3.2.3 D and are appropriate to detect abnormal performance.
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Specification C verifies the operability requirements in Specification 3.2.3 C during each day of operation, In Specif.ication D each channel capable of generating a scram signal is tested during the pre-critical procedure, prior to startup, so that the conditions of Specification 3.2.3 A are satisfied, i
Specification E requires calibration of safety and safety-related channels at an interval which is appropriate and justified by prior experience at this facility.
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Rev. 4 4.3 Coolant Condition This specification applies to the surveillance of the pool water condition.
4.3.1 Applicability This specification applies to the surveillance activities required for the reactor coolant condition.
4.3.2 Objective These requirements a cify the frequency and type of testing to assure that the coolant condition conforms to the specifications of section 3.3.
4.3.3 Specifications A. The water level in the reactor pool shall be maintained at seven feet from the tank bottom and shall be checked and recorded every day that the reactor is operated.
B. The lower limit of water resistivity shall be checked and recorded every day that the reactor is operated. A pool water sampling analysis shall be performed semi-annually by the Health Physicist. The result of the sampling analysis shall be reviewed by the Reactor Operations Committee.
4.3.4 Bases Specification A provides assurance on the amount of pool water in the reactor tank.
Specification B provides assurance on the water resistivity of the pool water. Test results on the reactor pool water showing no indication of radioactivity provide evidence that there has been no breach in the fuel cladding and no corrosion of the cladding by the pool water.
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4.4 Confinement or Containment There .is no confinement or containment system.
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4,5 Ventilation Systems 4.5.1 Applicability This specification applies to the surveillance activities required for the reactor ventilation system.
4.-5,2 Objective To specify the frequency and type of testing to assure that the ventilation system conforms to the specifications of section 3.5 of these Specifications.
4.5.3 Specification The blower and switch of ventilation system shall undergo testing for normal operation at least once per year.
4.5.4 Bases This specification requires that the blower and switch of ventilation system be tested to ver$fy that they can be operated when needed. The testing interval is adequate to verify operability based on experience at this facility.
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4.0 Emergency Power This specification'does not apply to this facility since there
.is no emergency power supply.
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Rev. 4 4.7 Radiation Monitoring System 3 4.7.1 Applicability These specifications apply to the surveillance activities required for the radiation monitoring system.
I 4.7.2 Objective i To specify the frequency and type of testing to assure that the radiation monitoring system conforms to the specification of section 3.7 of these Specifications.
4.7.3 Spe,cifications These surveillance activities are required for safety.
A. A calibration of the two radiation monitoring channels shall be perforined at least annually and whenever any maintenance on a channel which may affect its performance is completed. This calibration shall be performed by cornparing the readings of these instrutnents with those on a portable beta-ganuna survey ine ter . The latter shall be calibrated annually by a recognized diagnostic laboratory.
B. An operability test, including source checks, of the radiation monitoring channels shall be perfortned at least quarterly (and recorded in the reactor checkout sheets).
C. Readings of the radiation levels of all instruinents shall be recorded hourly during operation with the reactor being critical.
D The environmental filtn badge and sinear surveys in and around the reactor enclosure shall be perforined at least twice a l year.
E. An ALARA program shall be established and inenitored by a Radiation Safety Officer (RS0).
4.7.4 Bases Based on experience at this facility and average usage pattern of the reacter, Specifications A-E are odequate to verify that the operations conform to the Specifications of 3.7.3. The usage pattern shall be subject to review by the Reactor Operations Conunittee.
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4.8 Experiments 4.8.1 Applicability These specifications apply to the surveillance activities required for experiments installed in the reactor.
4.8.2 Objective l To specify the frequency and type of testing to assure that the experiments conform to the specifications of section 3.8 of these Specifications.
4.8.3 Specifications A. The identification and location of all installed experiments shall be recorded prior to each reactor startup.
B. Other specific surveillance activities shall be established during the review and approval process specified in section G.O.
4.8.4 Bases Specification A requires that the reactor operator verify that the installed experiments are approved.
Specification B recognizes that detailed surveil]ance requirements will vary among experiments, and that the Reactor Operations Committee specifies the appropriate type and frequency of surveillance.
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.4.9 Facility Specific Surveillance No Facility Specific Limiting Conditions for Operations are -
provided in section 3.9.
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Rev. 4 4.10 Frequency of Testing 4.10.1 Applicability This specification . applies to all surveillance requirements of Section 4 of these Technical Specifications.
4.10.2 Objective The objective of this specification is to establish inaximurn time intervals for surveillance periods. It is intended that this specification provide operational flexibility and not reduce surveillance frequency.
4.10.3 Specifications 4.10.3.1 Tiine intervals used elsewhere in these specifications shall be defined as follows:
A. Biennially - Interval not to exceed 30 rnonths.
B. Annually - Interval not to exceed 15 months.
C. Semi-annually - Interval not to exceed 32 weeks.
D. Quarterly - Interval not to exceed 18 weeks.
E. Monthly - Interval not to exceed 6 weeks.
F. Weekly - Interval not to exceed 10 days.
G. Daily - Must be done prior to the first startup of the calendar day following a shutdown greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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t Rev. 4 5.0 DES 1GN FEATURES 5.1 Site and Facility Description 5.1.1 The Manhattan College Zero Power Reactor (HCZPR) shall be located in the Leo Engineering Building of Manhattan College at 3825 Corlear Avenue, Bronx, New York.
5.1.2 A total of fif teen full fuel eleinents and one partial fuel eleinent shall be peraianently niounted upon the reactor grid plate, which shall be located at. the center of the reactor tank bottoin. Due to the extreinely low power rating of the MCZPR (0.1 watt),
periodic fuel replaceinent is not necessary. Ilence, no built-in provision is inade for the storage of spent fuel.
5.1.3 The control systein shall consist of two safety-type control rods, that is, a shirn rod and a regulating rod. The control r<.x1s shall be attached to Lheir associated drive niechanisnis by an electroinagnet and shall fall by gravity to the least reactive position upon a decrease of inagnet current following screin action.
5.1.4 The MCZPR Laboratory shall be provided with a forced circulating ventilation systein. This ventilation systeni shall be independent of the Leo Engineering Building and shall be controlled by a switch located on the west wall of the MCZPR roont.
5.1.5 The MCZPR shall be designed to operate at a full power of 0.1 watt. There shall be no fission product release or gaseous effluent under such low power rating.
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5.2 Reactor Coolant Condition Because of the extremely 1cw power rating of the reactor (0.1 watt), there is no recirculating coolant or coolant systent '
in the reactor other than the' reactor cool water. Water lost due to evaporation shall be replenished with New York City water. The water front the city system shall pass through a demineralized at a flow rate of two to three gallons per hour.
Under the reference core condition, the water level in the reactor tank shall be niaintained at seven feet as measured from the tank bottom.
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l l 5.2 Reactor Coolant Condition 1
' Because of the extreinely low power rating of the reactor (0.1 watt), there is no recirculating coolant or coolant systent in the reactor other than the' reactor pool water. Water lost
- due to evaporation shall be replenished with New York City i water. The water front the city systeni shall pass through a deinineralizer at a flow rate of two to three gallons per hour.
- Under the reference core condition, the water level in the reactor tank shall be tuaintained at seven feet as ineasured front the tank bottotti.
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Rev. 4 5.3 Reactor Core, Fuel, Control Rods, and Startup Source 5.3.1 Reactor Core A rigid plate stand is welded to the bottom of the reactor tank. Bolted to the grid plate stand is a grid plate. Fuel eleinent hold-down rods are passed axie.lly through the center of the fuel eleinents to hold the latter rigidly in position.
These hold-down rods, each with total length of 35 inches, are threaded into the grid plate. The shaft of these hold-down !
rods are made partly of aluininuin and partly of lucite. The l lucite portion, which consists of a solid rod one inch in l.
diameter, is 24 inches long. The lower portion of the hold-down rod is made out of a aluminuin tubing having a wall thickness of 1/8 inch and total length of 5-1/2 inches. The bottom 1-1/2 inches is threaded and secures the hold-down rod to the grid plate. The bro top of the hold-down rod, which 1 extends over the top of the tuel element is also made of aluminum with thickness of 3/8 inch. The aluininum portions of the hold-down rod are securely fastened to the lucite by aluininuin pins and epoxy cement.
5.3.2 Reactor Fuel The fuel portion of the elements consists of six concentric cylinders formed by inechanically joining and posit.ioning eighteen curved fuel plates within grooves of three spacer webs. The cylinder fuel plate consists of 0.02 inch-thick U-Al alloy of 92% enriched uraniura, clad on both sides with 0.015 inch of aluminum, making the total plate thickness 0.05 inch. The nominal U-235 content of each full fuel eleinent is 200 grains. The inner diameter of the innerrnost cylinder is about 1.25 inches and the spacing between adjacent cylinders (water channel width) is 0.118 inch. A inaxiinum of fifteen full fuel elements plus one partial fuel element (containing a nominal 24 grams of U-235), is used in the facility. The partial fuel element has one cylinder.
5-3 a _
Rev. 4 5.3.3 Reactor Control Rods A.' Control Rods - The reactor is controlled by two Y-shaped control rods which pass in the clearance between adjacent fuel elements. One control rod (the shim rod) is constructed so that the blades are fortned by sandwiching a 1/16 inch sheet of cadmium between 1/16 inch layers of stainless steel. The other control roJ is an all stainless steel regulating rod.
Either one of these control rods is capable of preventing the reactor from becoming critical.
B. Control Rod Drive Mechanisms - Each control rod drive system-is a cantilever drive with a design drive speed no greater than 12 inches per minute.
5.3 A Resctor Startup Source The startup source is a Pu-Be source encapsulated in. tantalum.
The source strength is one curie (approxiinately 1,000,000 neutrons per second).
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5.4 Fissionable Material Storage Fuel elements are permanently stored on the reactor grid plate l with the exception that three fuel plates are permanently stored in a ' locked steel container fastened to the floor of the first floor of the HCZPR Laboratory.
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'- Rev. 4 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization 6.1.1 Structure The organization for the inanageinent of the reactor facility shall be structured as shown in Figure 6-1.
Levels of authority indicated divide responsibility as follows:
Level 1: Responsible for the facility license and site administration.
Level 2: Responsible for the Reactor facility operation and Inanageinent.
Level 3: Responsible for daily operations.
The Reactor Operations Conimittee shall be appointed by the Reactor Administrator and shall be responsible to hiin for the review and evaluation of all proposed operations and procedures in order to insure that the reactor facility shall be operated in a safe and collipetellt InannCr.
6.1.2 Responsibility Individuals at the various inanagement levels shown in Figure 6-1, in addition to having responsibility for the policies and operation of the facility, shall be responsible for safe-guarding the public and facility personnel from undue. radiation exposures and for adhering to all requirernents of the Operating License and the Technical Specifications.
In all instances, responsibilities of one level inay be assumed by designated alternates, or by higher levels, conditional upon appropriate qualifications.
The detailed description of -duties of each individual in Level 2 and Lesel 3 are as follows:
A. The Ecactor Mministralot shall provide final policy l decisions on all phases of reactor operation and on regulations for the facility es a whole. He will be advised in all inatters concerning the safe operation of the reactor by the Reactor Operations Cominittee. i 6-1
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MANHATTAN COLLEGE CORPORATION BROAD OF TRUSTEES i Level 1
{ERFAIDEt[P OF THE_COLLEGEl l I
l EXECUTIVE VICE_EBES.IDEtlIl
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{EBRYD3}
))EAN _OF_ TilFuSCHOOL_DE..EttGINEERINGl REACTOR ADMINISTRATOR ;
___ Lev _el_ 2 RADIATION SAFET'i _
REACTOR OPERATIONE OEELCER C0 tit [ITTEE blEALTH_EllYSICLSJ}
CHIEF I<EACf0R SUPERVISOP Ley.eL 3 l8EACIOR SUPERVISORS}
l REACTOR OPERATOP
. _JaYel - 4 OLERIC5[ STAFF,l MAINTENANCE i AIDLlSSISIAlfi'SI ,_EEESQUUEL i
Figure 6-1, Table of Orgm ization 6-2 1
The Reactor Administrator shall be responsible for the overall' adininistration and supervision of the reactor facility. He shall appoint qualified meinbers to the Reactor Operations Conunittee from tinie to time as necessary. He shall designated Reactor Supervisors, na;ne the Chief Reactor Supervisor, and appoint the Radiation Safety Officer. The Reactor Administrator shall approve and proinulgate all regulations, instructions and procedures governing the operation of the reactor facility. The Reactor Administrator shall be appointed by the Provost of Manhattan College.
B. 'The ReacInr_DneraftlonLCnilmujlen shall be responsible to the Reactor Administrator for the review and evaluation of all proposed operations and procedures in order to insure that the reactor facility shall be operated in a safe and coinpetent manner. Particular einphasis shall be placed on the examination of new and untried operations and procedures, and ;
the Conunittee shall take action on proposed new experiments.
The Conunittee shall review and evaluate all proposed changes in the reactor system. The Reactor Operations Cotantittee '
shall advise on and be available for advice and assistance on any probleins relative to the safe operation of the reactor facility.
C. The Radiatiomfaafaty_01Licer shall be responsible for the promulgation and enforcernent of rules, regulations end operating procedures which conform with the regulations set forth in 10 CFR, Part 20. The Radiation Safety Officer in conjunction with the Reactor Operations Coiumittee shall approve suggested procedures for the purchase, possession, storage, use, and disposition of all radioisotopes, consistent with general or specific licenses for use of by-product material issued to Fanhattan College. The Radiation Safety Officer in conjunction with the Reactor Operations Conunittee, shall be available for advice and assistance on problenis involving radiological safety arising ,
from the operation of the reactor facility. The Reactor i Operations Con >mittee shall evaluate and approve all proposed procedures leading to the production of radioisotopes with a half-life longer than one hour. All operations leeding to the production of inore than une millieur4.e of radioactivity, with any half-life. nost receivt prior approval of the Reactor Operations Cominittee. l S-3 1
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D. The Health Physicist shall be responsible for monitoring records of exposure on film badges, maintenance of a log on radiation tests and exposure records. He also shall review the reactor log. Periodic radiation surveys of the critical reactor laboratory, the suboritical laboratory and the counting room, and other areas where radioactive materials are being used, shall be
~ made by the Health Physicist under the direction of the Chief Reactor Supervisor. The Radiation Safety Officer shall be notified if any abnormal radiation problem is encountered.
Results of these surveys shall be recorded or filed in the log.
The Health Physicist shall also be responsible for proper disposal of samples and radioactive materials. The Health
' Physicist shall be appointed by the Reactor Administrator after consultation with the Reactor Operations Committee.
E. The Beactor supervisors shall be appointed by the Reactor
' Administrator. These individuals shall have general competence in reactor technology and associated fields. Each supervisor shall hold a Senior Operator's License issued by the Nuclear Regulatory Commission. The Reactor Supervisors shall be responsible to the Reactor Administrator, through the Chief Reactor Supervisor, for the preparation and submission of complete detailed proposed procedures, regulations and administrative rules to insure the maintenance, safe operation, proper and competent use, and security of the reactor equipment.
Appointment as a Reactor Supervisor shall in all cases be accompanied by appointment to the Reactor Operations Committee.
The Reactor Supervisors shall be responsible for the preparation end submission of operating schedules of the reactor facility, and shall insure that all activities and experiments involving the facility conform to both local and Commission regulations.
They shall establish, in coordination with the Reactor Operations Committee, procedures for activities to be performed with the reactor. They shall establish procedures and be responsible for the keeping of adequate, complete, and currently accurate records for the operation and maintenance of the facility.
A Reactor Supervisor shall be in charge of the facility and shall ,
witness the startup and intentional shutdown procedures. In addition, he shall be responsible for prompt execution of emergency procedures.
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1 F. The Chief RelatpI_ Supervisor shall hold a valid Senior Operator's License issued by the Conunission. He shall be responsible for the prornulgation and enforcement of administrative rules, regulations and operating procedures.
He shall inform the Reactor Operations Couunittee of any unusual operations proposed to be performed on the reactor, or any proposed changes in procedure. He shall not authorize the operation or proceed with the proposed changes until appropriate evaluation and approval has been inade by the Peactor Operations Cornmittee, and authorization given by the Reactor Adtninistrator. The Chief Reactor Supervisor shall have the authority to authorize any activities or procedures which have received prior approval of the Reactor Operations Comtnittee . He shall be directly responsible for enforcing operating procedures and insuring that the reactor facility is operating in a safe, coinpetent and authorized inanner at all tiines. In addition, he shall be directly responsible for the preparation, authentication and storage of all prescribed logs and operating records.
G. The Bean. tor 0oerators shall hold a valid Operator's License issued by the Cortunission. They ntust conform to the rules, instructions and procedures for the startup, operation, and shutdown of the reactor facilities. They inust also conform to the specifications of the Emergency Plan. Within the constraints of the adatinistrative and supervisory controls outlined above, a Reactor Operator shall De in charge of the control console at all titnes that the reactor is operating.
The Beactor Operator shall be required to inaintain cotuplete and accurate records of all reactor operations in the operational logs.
H. All other personnel using the facility shall be instructed in the hazards involved, and given a copy of the laboratory regulations concerning use of radioactive material. All personnel working in the vinicity of the reactor shall wear filin badges.
G.1.3 Staffing
- 1. The ininimuta staffing when the reactor is not secured shall ]
be: J 1
- a. A licensed Reactor Operator in the control room. i
- b. A 1icensed Senior Beactor Operator present in the Leo i
Engineering Building.
- c. A Health Physicists - qualified individual contactable by '
phone.
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- 2. Operating Personnel Requireinents
- a. The controls of the reactor shall be operated only (the reactor controls are to be regarded as operating if the " Reactor-On" switch is turned to "Oll" and fuel is present in the core tank) with the specific' authorization of a Reactor Supervisor. The Reactor Operator shall be responsible for obtaining the e authorizing signature of a Reactor Supervisor at the top of the checkout sheet. The Reactor Supervisor signing the authorization is the supervisor "in charge".
- b. Whenever the reactor controls are operated, a licensed Reactor Operator shall be present and in the inunediate vicinity of the console. An up-to-date list of licensed reactor operators l
shall be posted near the reactor console. A person is considered "present" if he is in the console rooin within view of the instruinents on the console, o A Reactor Supervisor shall be present in the Leo Engineering Build.ing at all tiines that the reactor controls are operated and shall be cognizant of the reactor operation at all tiines.
If the supervisor in charge of the operation ntust leave the building, the reactor controls inust either be turned off and locked or another supervisor inust accept responsibility. The React.or Operator shall be inforined of such a transfer of authority. A list of Reactor Supervisors shall be posted near the reactor console.
- 3. Personnel Requireinents for Fuel and Experituental Loading
- a. Any tuoveinent of fuel eleinents or of tuaterial into or out of the reactor core can be done only on specific written authorization of or in the presence of a Reactor Supervisor.
- b. At least two persons, one a licensed Reactor Operator, shall be in the ZPR laboratory when any fuel eleinents or any experitnent is inoved in 1.he reactor core. One person shall be at the reactor console. ;
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- c. Whenever the final fuel eleinent necessary for attaintnent of l criticality is transferred into the core, a Reactor Supervisor i shall be present. )i
- d. A Reactor Supervisor shall be present in the ZPR rocin during the loading of an experinent ir.to the core for the first titne, or its reinoval frotn the core. A Supervisor shall be prerent in j the 2.PR rootn or give his written aut horization for repstitiva i insertions of an experimant.. l 9-G 1
6.1 A Selection and Training of Personnel The selection, training, and requalification of operators personnel shall ineet or exceed the requirements of American National Standard for Selection and Training of Personnel for Research Beactors ANSI /ANS 15.4 - 1967, or its successor, and be in accordance with the Requalification Plan approved by the Nuclear Regulatory Commission.
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Rev. 4 O.2 Review and Audit The Reactor Operations Conmlittee shall perforin the independent w
review and audit o'f the safety aspects of reactor facility operations.
~0.2.1 Cotuposition and Qualifications
' The Reactor Operations Connaittee shall be coluposed of the Reactor Adntinist.rator and the ilealth Physicist, both ex of ficio, and at least three ot.her nieinbers having experti:e in reactor tecluiology. Conunittee ineinbers shall be appointed by the Reactor Adulinistrator.
6.2.2 Cinrter and Bules
- 1. The Reactor Operations Cotntuittee shall nieet at least seini-annually and niore frequently as circuinstances warrant, consistent with effective inonitoring of facility activities. Written records of its tueeting shall be kept.
- 2. The Reactor Operations Co:muittee inay appoint one or tuore qualified individuals to perforn t.he audit function.
6.2.3 Review Function The follwing itenis shall be reviewed:
- 1. Deterulination that proposed changes in equipinent, systents, t.est.s, experitnents, or procedures do not involve an unreviewed safety question.
- 2. All new procedures and niajar revisions thereto having safety significance and proposed changes in reactor facility equipinent, or systeins having safety significance.
- 3. All new experiinents or classes of experiinents that could affect, react.ivity or result in the release of radioactivity.
- 4. Proposed changes in the Technical Specifications or the Operat.ing License.
- 5. Reports of external (NRC, state and local authorities, and insurors) inspectors and auditors.
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Rev. 4 0.2.4 The Audit Function
'"he Audit Function shall include selective (but cornprehensive) examination of operating records, logs, and other docuinents. 1 j
The audit will be performed biennially by an outside '
individual or group' familiar with the research reactor operations. They chall submit a report to the Reactor Adininistrator and the Reactor Operations Conmiittee. The following itetus shall be audited:
- 1. Facility operations for conforttiance to the Technical Specifications and applicable Operating License canditions, at least once a year.
- 2. The retraining and requalification prograin for the operating staff.
- 3. The results of action taken to correct those deficiencies that inay occur in the reactor facility equipinent, systems, structures, or it.ethods of operation that affect reactor safety, at least once per calendar year.
- 4. The reactor facility - Einergency and Physical Security Plans and itupleinenting procedure at least once every other calendar year.
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6.3 Procedures .!
Written procedures shall be prepared, reviewed and approved prior to intiating any of the activities listed in this section. The procedures shall be reviewed by the Reactor Operations Conunittee and epproved by the Reactor Administrator.
The following activities, not already described in the Technical Specifications, may be included in a set of procedures.
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- 1. Startup, operation, and shutdown of the reactor.
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- 2. Fuel . loading, unloading and movement within the reactor.
- 3. Routine maintenance of inajor components of systetus that could have an effect on reactor safety.
'd. Surveillance tests and calibrations required by the Technical Specifications or those that may have an effect e on reactor safety.
- 5. c'ersonnel radiation protection consistent with applicable regulations.
- 6. Administrative controls for operations and maintenance and fo'r the conduct of irradiations and use of experiments that could affect reactor safety or core activity.
- 7. Iinplementation of the Emergency and Physical Security Plans.
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G.4 Experinients; Review and Approval Approved laboratory exercises shall be carried out in accordance with established and approved procedure.
- 1. All new exercises shall be reviewed by the Reactor Operations Committee and approved by the Reactor Administrator prior to initiatiori.
- 2. Substantive changes to previously approved experiitients shall be node only af Ler they are reviewed by the Reactor Operations Coimtlittee and approved by the Reactor Administrator.
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6.5 Required Actions 0.5.1 Action to be Taken in Case of Safety Liniit Violation
.1. The reactor shall be shutdown and reactor operations shall not be resuined until authorized by the Nuclear Regulatory Conunission (NRC). i
- 2. The safety liinit violation shall be protnptly reported to the Reactor Adininistrator or a designated alternate.
- 3. The safety limit violation shall be reported to Nuclear Regulatory Conunission.
- 4. A safety lituit violation report shall be prepared. The report, and any follow-up report shall be reviewed by the Reactor Operations Coininittee and shall be subtaitted to the Nuclear Regulatory Conunission when authorization is sought to resuine operation of the reactor. The report shall describe the following:
- a. Applicable circumstances leading to the violation including, when known, the cause and contributing factors.
- b. Effect of the violation upon reactor facility coniponents, systeins, or structures and on the health and safety of personnel and the public.
- c. Corrective action to be taken to prevent recurrence.
6.5.2 Action to te Taken in the Event of an Occurrence of the Type Identified in 6.6.2 - 1.b and 6.6 2 - 1.c.
- 1. Reactor conditions shall be returned to norinal or the reactor shall be shutdown. If it is necessary to shutdown the reactor to correct the occurrence, operations shall not be recuined unless authorized by the Reactor Adininistrator or a designated alternate.
- 2. Occurrence shall be reported to the Reactor Aininistrator or a designated alternate and to the Nuclear Regulatory ,
Conunission .
- 3. Occurrence shall be reviewed by the Reactor Operations Conunittee at its next scheduled nieeting.
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Rev. 4 6.6 Reports 6.6.1 Operating Reports Internal reports are kept as minutes of the semi-annually meetings of the Reactor Operations Conunittee.
A report summarizing facility operations will be prepared annually where the reporting period ends August 31. A copy of this report shall be submitted to the Nuclear Regulatory Comntission (NRC) Region I office by October 15 of each year, with a copy to the Director, Office of Nuclear Reactor Begulation, Nuclear Regulatory Commission. The report shall include the following:
- 1. A narrative summary of reactor operating experience.
- 2. A description of unscheduled shutdowns including where applicable, corrective action taken to preclude recurrence.
- 3. Tabulation of major preventive and corrective maintenance operations having safety significance.
- 4. Tabulation of major changes in the reactor facility and procedures, and tabulation of new tests or experiments, or both, that are significantly different from those perfortned previously and are not described in the Safety Analysis Report, including conclusions that no unreviewed safety questions were involved.
- 5. A cununarized result of any radiation surveys perforated by the facility personnel.
- 6. A suinmary of exposures received by facility personnel and visitors uhere such exposures are greater than 25% of that allowed or reconunended.
6.6.2 Special Reports
- 1. There shall be a report not later than the following I working day by telephone and confirrned in writing by telegraph or similar conveyance to Nuclear Regulatory Commission to be followed by a written report that describes the circumstances of the event within 14 days 1 of any of the following: {
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- a. Violation of Safety Limit (see 6.5.1).
- b. Helease of radioactivity from the site above allowed limits (see 6.5.2).
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- c. Any of the following (see 0.5.2):
- 1) Operation with actual safety system settings for required systems less conservative than the litniting safety systein settings specified in the Technical Specifications.
- 2) Operation in violation of litniting conditions for operation-established in the Technical Specifications unless prornpt reinedial action is taken.
- 3) A reactor safety systein coniponent inalfunction which renders or could render the systern incapable of performing its intended safety function unless the malfunction or condition is discovered during maintenance tests or periods of reactor shutdown.
- 4) An unanticipated or uncontrolled change in reactivity greater than the licensed excess reactivity, or one dollar, whichever is smaller.
- 5) Abnormal and significant degradation in reactor fuel, or eladding, or both which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both. l l
- 6) An observed inadequacy in the implementation of adtuinistrative or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reector operations. ,
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- 2. A written report within 30 days to the Nuclear Regulatory Conmiission concerning the following:
- a. Permanent changes in the organization involving the Reactor Adininistrator, Chief Beactor Supervisor, or Radiation Safety Officer. l
- b. Significant changes in the transient or accident analysis as described in the Safety Analysic Peport.
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6.7 Records-l l
6.7.1 Records to be retained for a Period of at Least Five Years or '
for the Life of the Component if Less than Five Years
- 1. Norinal reactor facility operation (but not including supporting docuanents such as checklists, log sheets, etc. ,
which shall be maintained for a period o!' at least one 1 year). !
'2. Principal maintenance operations.
- 3. Reportable occurrences.
- 4. Surveillance activities required by the Technical- 1 Specifications.
' 5. Reactor facility radiation and contamination surveys where required by applicable regulations.
- 6. Laboratory exercises perforined with the reactor.
- 7. Fuel inventories, receipts, and shipinents.
- 8. Approved changes in operating procedures.
- 9. Records of ineetings snd audit reports of the Reactor Operations Coimnittee.
G.7.2 Records.to be retained for at Least One Training Cycle Retraining and requalification of licensed operators:
Records of the anost recent complete cycle shall be snaintained at all times the individual is einployed.
G.7.3 Records to be Retained for the Lifetime of the Reactor Facility Applicable annunl reports, if they contain all of the required inforination, may be used as records in this section.
- 1. Gaseous and liquid radioactive effluents, if any, release e to the environs.
- 2. On-site environmental inonitoring surveys required by the Technical Specifications.
- 3. Radiation exposure for all personnel inonitored.
- 4. Drawings of the reactor facility.
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-l ATTACHMENT II TECHNICAL SPECIFICATIONS FOR THE MANHATTAN COLLEGE ZERO PO'iER REACTOR FACILITY LICENSE R-94 February 20, 1989 i
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REVISION 9ECOPD _ T' '
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d Comm-nt "
Submitted to U. S. N. R. C. on August 26, 1983 '
as part of the Safety Analysis Report for the
. Manhattan College Zero Power Reactor (MCZPR).
J Complete revision submitted on January 12, 1964.
Co:nplete revision submitted on November 15, 1984.
Revisions as noted on pages:
16-1, 16-2, 16-10, 16-11, 16-13, 16-16, 16-23, 1 16-25, and 16-26, on December 19, 1984.
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Pagel numbering' changed from 16-1 to 1-1, etc.
Cover sheet revised from section 16 to Appendix A
.per USNRC request.
Revisions as noted on pages i, ii, iii, and 1-1, 2-1, 2 -2, ' 3-1, 3-2, 3-3, 3-4, 3-7, 4-1, 4-2, 4-4, 4-8,14-11, 5-1, 5-2, 5-3, 5-4, 5-5, 6-1, -;
6-6, 6-8, 6-9, 6-11, 6-13, and 6-14 on March 15, l 1955.
'l Revisions as noted on pages i, ii, and 1-1,1-2, 3-1, and 3-2 on February 20, 1989.
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- h. 1. 0 - DEFINITIONS The. terms Safety-Limit, Limiting Safety System Settings, and Limiting Condition for Operation are as defined in paragraph 50.36 of 10 CFR
-Part'50.
' ALAFA -' 4.LABA is e ' concept introduced by the Nuclear Segulatory
"'cem.iezion 180) to all reacter facilities. The basic cf A1 ABA is the eil e::posure to rsdistion should be k+;t "as low as re?senabl/
aenie"able" (ALAFA).
L c'mnnil - A channel is the combination of sensor, line, explifer, 2nd I cutput devices which are connected-for the purpose of measuring the
-value of a parameter.
chann+1 calibration:- A channel calibration is an adjustmen; of the channel such that-its output corresponds uith acceptable accuracy to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip, and shall be deemed to include a channel test.
cJmnel check - 'A' channel check is a qualitative verification of acceptable performance by observation of chsnnel behavior. This verification, where pcssible, shall include comparison .of the channel with other independent channels or systems measuring the same variable.
cher.21 tart - A chonnel test is the introduction of a signal into the channel for verification that it is operable.
control _rnd - Plates fabricated with neutron absorbing material used to establish neutron flux changes and to ecmpensate for routine reactivity losses. This includes safety-type and regulating rode, core - The portion of the reactor volume which includes the fuel elements, the source, and the ccatrol rods.
delnved neutron fracticn - When converting between absolute- and dollar-value reactivity units, a beta of 0.0078 is used.
i t
dren tims - The elapsed time between rea'ching the complete removal setpoint and the full insertion of a saf'ety-type red. (It must be less than 1.0 second) l excerr resetivity - Excess reactivity is that amount of reactivity that would exist if all control rods (shim, regulating) were moved to the maximum reactive condition from the point where the reactor is exactly critical (K =1). I eff 1-1 t
1 Rev. 5 l experiment - Any object, other than a fuel ~ element or handling tool, J
which .is inserted into the volume formed by projecting the rigid j plate vertically to the tank pool water surface is to be regarded as i an experiment in the core. I i
.]
titeurad v,1ue - The measured value is the value of a parameter as it 1 appears on the output of a channel. l movabla areeriment - An experiment where it is intended that the entire experiment may be moved in or near the core or into and out -
of the reactor pool water. j i
co+r?bl+ - Operable means a component :,r system is capable of performing its intended functica.
crerating - Operating means a component or system is performing its
. intended function.
resetivity limite - The reactivity limits are those limits imposed on reactor core excess reactivity.
reactivity worth of an eroeriment - The reactivity worth of an experiment is the maximum absolute value of the reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment position or configuration.
raneter coerating - The reactor is operating whenever it is not secured or shutdown.
raseter coerator (RO) - An individual who is licensed to manipulate the controls of a reactor.
reenter r,fety evstems - Reactor safety systems are those systems, including their associated input channels, whion are designed to initiate automatic reactor protection or to provide information for initiation of manual protective action.
raseter -eeured - A reactor is secured when:
- 1) It contains insufficient fissile material or moderator present in the reactor, adjacent e::periments or control rods, to attain criticality under optimum available conditions of moderation and reflection, or
- 2) A combination of all of the following:
a) All neutron absorbing control rods are fully inserted and othe r safety devices are in shutdown position, as required by the Technical Specifications, and b) The console key switch is in the off position and the key is removed from the lock, and c) No work is in progress involving core fuel, core structure, installed control rods, or control rod devices unless they are physically decoupled from the control rods, and d) No experiments are in or near the reactor.
1-2
Rev. 5
?0 LIMITING CONDITIONS FOR C?ERATIO.N '
3.1 Reactor Core Parameters 3.1.1 Applicability These spe ificati:ns apsly to the psrsr.eters which describt ' he re ._vity condi icr. of the core.
3.1.2 Objective To ensure that the reacter ca.nnot achieve prompt criticality end that it can be safely shutdown under any condition.
3.1.3 Specifications The reactor shall not be made critical unless the following ecnditiens exist:
A. The total core excess reactivity with or without the movable experiments of sectica 3.8.3 shall not exceed 0.44% Ak/k at 110.6*F.
~B. The minimum shutdcwn margin provided by control rods shall not be less than 0.46% ak/k at 110.6"F.
C. Any change in the experimental apparatus shall be approved by the Reactor Operations Committee.
3.1.4 Bases Specification A is based upon the experimentally determined value for excess reactivity 0.44% Ak/k at a j reactor pool water temperature of 110.6'F. I I
Specification B is based upon the negative worth of the
)
{
regt.lating red: that is, the control rod with the smaller I nega'tive worth. l I
Specification C limits the changes in the experimental apparatus to those approved by the committee charged with review and approval of experiments.
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3.2 Beactor Control snd Safety System 3.2.1 Applicability.
These specifications apply tu the reacter safety system and safety-related :nstrumentatic"
.: . 2 . 2 Obj+o-1v+
To specif:, th+ lwe: t awes tat a le cel ;f performance or the mininum number of acceptable ccmponents for the reactor safety syst+m and safety-related instrumentation.
3.2.3 Specifications The reactor shall not te made critical unless the following conditions exist:
A. The reactor safety system shall be cpere.ble in accordance with Table.3-1.
B. There shall b? two safety-typ+ control rede:
A regulating rod with a negative worth of 0.90% Ak/k end a shim rod with a negative worth of 2.50% ak/k.
C. The drop time for either safety rod shall not exceed 1.0 second; measurements of rod drop times shall be made once semi-ennually D. The reactivity insertion rate for a single rod shall not I exceed 0.10% Ak/k per second.
1 3.2.4 Eases Specification A prevides assurance that the reactor safety system which may be needed to shut down the reactor is operable. Each feature of the system is deceribed in !
Table 3-1. I 1
- A scram system is provided that causes interruption of ithe ,
nagnet current to the electromagnets, should a scram tr'.ip be '
exceeded. The control rods then fall into the reactor ' core under the force of gravity. This system provides a l conservative response to an instrumentation system failpre, electric power failure,1cw water level, nigh neutron flux, and high gamma activity.
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ATTACHMENTIII TECHNICAL SPECIFICATIONS FOR THE MANHAH AN COLLEGE ZE50 POWEB REACTOF FACILITY LICEN3E P-34 MARCH 10. 1989 l
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PEVISION RECORD l
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.Pevision No. Comment 0 Submitted to U. S. N. R. C. on August 26, 1983 as part of the Safety Analysis Report for the Manhattan College Zero Power Reactor (MCZPR).
1 Complete revision submitted on January 12, 1984.
2 Complete revision submitted on November 15, 1984.
3 Revisions as noted on pages:
16-1, 16-2, 16-10, 16-11,'16-13, 16-16, 16-23, 16-25, and 16-28, on December 19, 1984.
4 Page numbering changed from 16-1 to 1-1, etc.
Cover sheet revised from section 16 to Appendiv. A per USNBC request.
Revisions as noted on pages i, ii, iii, and 1-1, 2-1, 2-2, 3-1, 3-2, 3-3, 3-4, 3-7, 4-1, 4-2, 4-4, 4-8, 4-11, 5-1, 5-2, 5-3, 5-4, 5-5, 6-1, 6-6, 6-8, 6-9, 6-11, 6-13, and 6-14 on March 15, 1985.
5 (HEU core) Revisions as noted on pages i, ii, and 1-1, 1-2, 3-1, and 3-2 on February 20, 1989.
6 (LEU core) Revisions as noted on pages i, ii, and 2-1, 2-2, 3-1, 3-2, 5-1 and 5-3 on March 10, 1989.
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Fev. 6 l
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits 2.1.1 Applicability l
This specification applies to the melting temperature of the fuel cladding.
2.1.2 Objective To assure that the integrity of the fuel is maintained.
2.1.3. Specifications The safety limit shall be on the temperature of the feel element cladding, which shall be less than 1030*F (582'C).
2.1.4 Bases The melting temperature of the aluminum used as cladding on the fuel elements is 1080*F (582*C). Therefore, in order to maintain fuel element integrity, the cladding temperature must not exceed 1080'F (582*C). As reported in Reference 1, the maximum cladding temperature that can ever be reached is only 239'F (115'C) and reaches this level only during the Maximum Hypothetical Accident. The specification, therefore, provides assurance on the integrity of the fuel element cladding.
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Rev. 6
-2.2 Limiting Safety System Settings (LSSS) 2.2.1 Applicability This specification applies to the setpoints of safety channels which monitor r+ actor power level.
2.2.2 Objective To assure that automatic trip action is initiated and that the operator is warned to take protective action against exceeding a safety limit.
2.2.3 Specifications The limiting safety system setting shall be on reactor maximum power level not exceeding 0.125 watt, or 125% of full power.
2.2.4 Bases Since there is no forced circulation cooling, the reactor core is cooled by the water surrounding the reactor core. Therefore, the only parameter which could be used as a limit for the fuel cladding temperature is the reactor power. The analysis in Reference 1 shows that even for the Maximum Hypothetical Accident (a reactor power excursion of 183 kilowatts), the maximum cladding temperature reaches only 239"F (115*C). This temperature is much lower than the temperature (1080*F, 582*C) at which cladding damage could occur. Therefore, a large safety margin exists between the safety system set point and the cladding safety limit.
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_ _ _ _ _ _ _ _ _ _ _ ___j
8ev. 6 3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Core Parameters 3.1.1 Applicability These s;-::ificatiens a;;1y t:i the ; ara:rsters which describe the reactivit-/ Ocr.dition 7:f the core.
3.1.2 Objective To ensure that the reactor cannot achieve prompt criticality and that it can be safely shutdown under any condition.
3.1.3 Specifications The reactor shall not be made critical unless the following conditions exist:
A. The total core excess reactivity with or withou; the movable experiments of section 3.8.3 shall not exceed 1 0.44% Ak/k at 110.6*F (43.7'C) . l B. The minimum shutdown margin provided by the control rods shall not be less than a calculated value of 0.46% Ak/k at 110. 6*F (43.7'C) . l C. Any change in the experimental apparatus shall be approved by the Reactor Operations Committee. I i
3.1.4 Bases Specification A is based on an estimated value of ;
0.44% Alvk for the excess reactivity at a reactor pool ;
water temperature of 110.6*F (43.7'C). j a
Specification B is based upon the negative worth of the !
regulating rod: that is, the control rod with the smaller negative worth.
Specification C limits the changes in the experimental apparatus to those approved by the committee enarged with review and approval of experiments.
3-1 i
l Rev. 6 3.2 Reactor _ Control and Safety System 3.2.1 Applicability These specifications apply to the reactor safety system and safety-related instrumentation.
?.2.2 !b :nivt To r;6cify tne lower. acceptable level of performance or the minimum number of acceptable components for the reactor safety system and safety-related instrumentation.
3.2.3 Specifications The reactor shall not be made critical unless the following conditions exist:
A. The reactor safety system shall be operable in accUrdance with Table 3-1.
B. There shall be two safety-type control rods:
A regulating rod with a calculated negative worth equal to or greater than 0.90% Ak/k and a shim red with a calculated negative worth equal to or greater than 2.50% Ak/k.
C. The drop time for either safety rod shall not exceed 1.0 second; measurements of rod drop times shall be made once semi-annually.
D. The reactivity insertion rate for a single rod shall not exceed 0.10% Ak/k per second.
3.2.4 Bases Specification A provides assurance that the reactor safety system which may be needed to shut do@n the reactor is operable. Each feature of the system is described in Table 3-1, i l
- A scram system is provided that causeb interruption of the magnet current to the electromagnets, should a scran trip be exceeded. The control rods then fall into the reactor core under the force of gravity. This sysgem provides a conservative response to an instrumentation system failure, electric power failure, low water level, high neutron flux, and high gamma activity. '
3-? ;
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'5. 0 DESIGN FEATUEES.
5.1 Site and Facility Description 5.1.1 The Manhattan College Zero Power Peactor (M ZPR) shall be located in the Leo Engineering Building of Manhattan College at 3825 Corlear Avenue, Brons:, New York.
$ 5.1.2 The full fuel elements and the partial fuel element shall be mounted on the reactor grid plate, which shall be located at the center of the reactor tank bottom.
Due to the extremely low power rating of the MCZPR (0.1 watt), periodic fuel replacement is not necessary. Hence no built-in provision is made for the storage of spent fuel.
5.1.3 The centrol system shall consist of two safety-type control rods, that is, a shim red and a regulating rod. The control rods shall be . attached to their associated drive mechanisms by an electromagnet and shall fall by gravity to the least reactive position upon a decrease of magnet current following scram action.
5.1.4 The MCZPR Laboratory shall be provided with a forced circulating ventilation system. This ventilation i system shall be independent of the Leo Engineering i Building and shall be controlled by a switch located on the west wall of the MCZPR room.
5.1.5 The MCZPR shall be designed to operate at a full power of 0.1 watt. There shall be no fission product release or gaseous effluent under such low power rating.
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5.3 Reactor Core, Fuel, Control Rods, and Startup Source l
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5.3.1 Reactor Core i A rigid plate stand is welded tc the bcttom of the reactor tank. Bolted to the grid plate stand ic a grid plate. Fuel element hold-down roas are passed axially through the center of the fuel elements to hold the latter rigidly in position.
These hold-down rods, each with total length of 35 inches, are threaded into the grid plate. The shaf t of those hold-down rods are made partly of aluminum and partly of lucite. The lucite portion, which consists of a solid rod one inch in d iameter, is 24 inches long. The lower portion of the hold-down rod is made out of a aluminum tubing having a wall thickness of 1/8 inch and total length of 5-1/2 inches. The bottom 1-1/2 inches is threaded and secures the hold-down rod to the grid plate. The broad top of the hold-down red, which extends over the top of the fue! clement is also made of aluminum with thickness of 3/8 inch. The aluminum portions of the hold-down red are securely fastened to the lucite by aluminum pins end epoxy cement.
5.3.2 Reactor Fuel The fuel portion of the elements consists of six concentric cylinders formed by mechanically joining and positioning eighteen eurved fuel plates within grooves of three spacer webs. The cylinder fuel plate consists of 0.02 inch-thick U3 Sin -Al fuel meat containing uranium enriched to 19.75 0.2% ,
in U-235 and cle.d on both sides with 0.015 inch of aluminum, !
making the total plate thickness 0.05 inch. The nominal U-235 content of each full fuel element is 235 grams. The inner l diameter of the innermost cylinder is about 1.25 inches and the spacing between adjacent cylinders (water channel width) is 0.116 inch. The partial fuel element can have up to three fueled cylinders with a nomina] U-235 content between 9.1 and 129.5gramsj l
5-3 4
REFERENCES F. Frecsc, " Analyses For Conversion of the Manhattan College Zero Power hEU to LEU Fuel", Argenne National Laboratory (RERTR), February I
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APPENDIX E ,
EMERGENCY PLAN: ST JOSEPH'S HOSPITAL AGREEMENT (St. Joseph's Agreement is Appendix 3A of Emergency Plan)
E-1
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THE' EMERGENCY PLAN ' !
FOR THE MANHATTAN COLLEGE ZERO POWER REACTOR Fe.cility Operating License: R-94 Ibcket No: 50-199
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I TABLE OF CONTENTS Page
- 1. O Introduction 1 6
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- 2. O Definitions !
Organization and Responsibilities 12
- 3. O Emergency Clas sification System '17
- 4. 0 Emergency Action Levels for 22
- 5. 0 Notification of Unused Events
- 6. 0 Emergency Planning Zone .23 Emergency Response 24
- 7. 0
- 8. O . Emergency Facilities and Equipment 27
- 9. 0 Recovery 29 10.0 Maintaining Emergency Preparedne ss 30
TABLE OF CONTENTS s Page_
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- 1. O Introduction 6 i
- 2. O Definitions 12
- 3. 0 Organization and Responsibilities 17 4.0 Emergency Classification System 22 f
- 5. 0 Emergency Action Levels for Notification of Unused Events 23 6.0 Emergency Planning Zone 24
- 7. 0 Emergency Response 27 I 8. 0 Emergency Facilities and Equipment 29
- 9. 0 Recovery 30 ;
10.0 Maintaining Emergency Preparedne ss 1
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LIST OF FIGURES Figure No. Page 1.1 Map of Area Surrounding the Leo Engineering Building 2 Location of Leo Engineering Building
- 1. 2 3 13 . Plan of First Floor of the Leo Engineering Building 4
- 1. 4 Plan of Second Floor of the Leo Engineering Building 5 2.1 Plan of First Floor of the Leo Engineering Building 7 showing the ZPR Room (Cross Hatched) 2.2 Plan of Second Floor of the Leo Engineering Building 8 showing the ZPR' Room (Cros s Hatched)
- 2. 3 Plan of First Floor of the Leo Engineering Building 9 showing the MC ZPR Facility (Single Hatched)
- 2. 4 Plan of Second Floor of the Leo Engineering Building 10 showing the MC ZPR Facility (Single Hatched)
- 2. 5 Plan of Second Floor of the Leo Engineering Building 11 showing Emergency Support Center (Cross Hatched) 3.1 MCZPR Normal Operating Organization 13
- 3. 2 MCZPR Emergency Organization Chart 14 7.1 Location of Emergency Organization Chart 25 8.1 Emergency Support Center (Shown Cross Hatched) 28 J
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_ - _ _ _ _ _ _ _ _ - _ - _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _- __ _ a
- 1. 0 INTRODUC TION _
This Emergency Plan applies to the Manhattan College Zero Power 1.1 Reactor (MCZPR). The reactor is owned and operated by the Manhattan College Corporation of Bronx, New York,10471. Under Facility Operating License R-94, Docket No. 50-199.
l
- 1. 2 The objectives o %J , plan are to designate responsibility among l the reactor personnel and establish guidelines of action in the event of an accident or incident at the reactor that may present undue risk to the health and safety of individuals, or re sult in damage to the property. The plan also identifies off-sit- support organizations that may be activated if required.
- 1. 3 The MCZPR is a U-233 fueled light water moderated open pool The type heterogeneous reactor with plate type fuel elements.
reactor tank is 8 ft. high and 10 ft. in diameter. The core consists of 15 full fuel elements and one partial fuel element. Each full element consists of six fuel plates containing 200 gm of uranium and the partial element contains 25 gm making a total of 3025 gm of uranium. The reactor is licensed to operate at a continuous maximum power of 0. I watt. Because of the very low power level, no recirculating cooling system is provided.
- 1. 4 The MCZPR is a research reactor. The major functions of the reactor are training of reactor operators and for experimentation as part of Nuclear Engineering courses offered at Manhattan College.
- 1. 5 The reactor is operated whenever needed for training or for class experiments. All experiments performed on the reactor The require Committee prior closely approval of a Reactor Operations Committee.
monitors all experiments performed on t.'.e reactor. Based on the operating history of the past three years , the reactor was made critical about 34 times a year with each critical operation laating an average of 16 minutes. The power levels were most often well below the license.d level of 0. I watt.
- 1. 6 The MC 7_.PR is located in the Bronx, New York. It is easily accessible from Interstate Highway 87 and the 'Honry Hudson Parkway by connecting roads. An area map is given in Figure 1.1. The reactor is housed in the Leo Engineering Building on Corlear Avenue between 238 Street and 240 Street. Figure 1. 2 shows the location of the Leo Engineering Building, The reactor facility occupies portions of the first and second floors of the Leo Engineering Building. The floor plans are shown in Figure s 1. 3 and 1. 4. Acces s to the facility is either through door D i on the first floor or through D4 on the second floor.
Access door D3 is kept locked and bolted from 4 1 side at all time s.
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-S -
_ 6 -
- 2. O Definitions MCZPR 2.1 MCZPR stands for the Manhattan College Zero Power Reactor located in the Leo Engineering Building of Manhattan College.
- 2. 2 ZPR Room or Reactor Room Consists of rooms on the first and second floors in the Leo Engineering Building in which the reactor is built. The area is shown cross hatched in Figures 2. I and 2,2.
2.3 MCZPR Facility or Reactor facility Consists of the ZPR room and rooms 107, 108 and 109 on the first floor and room 221 on the second floor which are shc.wn single hatched in Fi gure s 2. 3 and 2. 4.
- 2. 4 Emergency Support Center Consists of the platform area of room 221. This is shown in Figure 2. 5.
- 2. 5 Reactor Operations Committee (ROC)
The ROC consists of the Reactor Administrator who acts as the Chairman of the Committee, the Chief Reactor Supervisor, Health Physicist, Radiation Safety Officer and others who might be helpful in the operation of the reactor and appointed to the Committee by the. Reactor Administrator.
- 2. 6 NRC Nuclear Regulatory Commission.
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FIGURE 2 3 PLAN OF FIRST FLOCR OF THE LEO ENGINEERING BUILDING ~
SHOWING THE MCZPR FACILITY ( SINGLE HATCHED )
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12 -
- 3. 0 Organization and Re sponsibilitie s The Manhattan College Zero Power Reactor Emergency Plan is designed to provide a means of meeting the additional demands that would be encountered if an emergency situatim arises. To effect this goal an Emergency Organization drawn primarily from the norn.a1 operating personnel is identified. One single individual is assigned the responsibility to direct all emergency related activities.
3.1 Normal Organization Structure Figure 3. I shows the normal operating organization of the MCZPR.
- 3. 2 Emergency Organizatim Structure, Figure 3.2 shows the Emergency Organization.
This organizations formed from the normal operating organization so that a smooth transition is pos sible.
3, 2.1 Emergency Director The Reactor Administrator will be the Emergency Director. In his absence, the Chief React or Supervisor, Reactor Supervisor on duty or the Reactor Supervisor to arrive first on the scene will be the Acting Emergency Director, in that order.
Basic Function The Emergency Director is responsible for taking all action necessary to manage any reactor related emergency.
Primary Responsibilities The Emergency Director
- 1. Coordinates and directs activitie s of the MCZPR Emergency Organization.
- 2. Classifies and declares an emergency when needed.
l
- 3. Assures notification of College, local and federal agencies as delineated in the procedure s required. The actual notifica-tion of outside agencies will be made, if time allows, only J
by the Reactor Administrator. 1
- 4. Issuing instructions to the Emergency Organization and assuring that appropriate action is ta' en. i
- 5. Insures health physics activities on campus. i
]
- 6. Declares termination of emergency. 1
- 7. Authorize s re-entry into the ZPR Room after an emergency. ]
}
- 8. Authorize s volunteer emergency workers to incur radiation exposure in excess of 10CFR20 limits.
1 i
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Manhattan College President l Provost 1 l
Dean of Manhattan College Engineering Radiation Protection Council I
I Reactor Operations --
3 l
Committee L-- ,
Reactor ;
l Administration .
Reactor Reacter Health Radiation Physicist Safety Officer Chief Reactor Supervisor Reactor Supervisor Reactor Operators Figure 3 1 MCZ.PR ORGANIZ ATION CHART
Manhattan College Pre sident Provo st Dean of , _ , _ _ _ _ ,
Emergency Enginee ring News Center Off-Site .___ Emergency Support Dire cto r Organizations Health Radiation Physicist Safety Office r Emergency l O pe rations j: Supe rviso r Emergency Ope rato r s I
l Figure 3. 2 MCZPR Emergency Organization Chart 1
l l
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- 15 -
9, Prepares news releases to the public when requested by the Dean of Engineering.
- 10. Reports to the Dean of Engineering.
3.2.2 Emergency Operations Supervisor The Chief Reactor Supervisor will be the Emergency Operations Supe rvi so r. In his absence, the Emergency Director will appoint one of the reactor supervisors to act as the Emergency Operations Supe rvi so r.
Basic Furrtions The Emergency Operations Supervisor is responsible for the implementation of activitie s connected with safe shutdown and maintaining of conditions that would minimize the effect on health and safety of the public.
The Emergency Ope rations Supervisor
- 1. Reports to the Emergency Director
- 2. Implements reactor emergency pr ocedure s
- 3. Supervises reactor emergency personnelin work at the reactor connected with the emergency
- 4. Supervises recovery operationt 3.2.3 Emergency Operators Emergency Operators are drawn from the Reactor Operating personnel.
Basic Function The basic function of the Emergency Operators is to assist the Emergency Operations Supervisor in his activities connected with an emergency. They report to the Emergency Operations Supervisor.
3,2.4 Radiation Safety Officer The Reactor Radiation Safety Officer will continue to function as the Radiation Safety Officer during an emergency.
Basic Function The basic function of the Radiation Safety Cfficer during an emergency is to advise the Emergency Director on matters of radiation safety l during an emergency. He reports to the Emergency Director.
3.2.5 Health Physicist The MC ZPR. Health Physicist will, continue to function a s the health physicist during e.n emergency.
16 -
Basic Function The basic function of the Health Physicist during an emergency is to implement procedures that would minimize the radiological effects on the health and safety of the personnelinvolved in an emergency work and that of the public.
The Health Physicist
- 1. Reports to the Emergency Director 2 Develops pla n s and procedure s for checking for contamination 3.2.6 Emergency News Center .
The College Relations office of Manhattan College will serve as the Emergency News Center. Any news pertaining to an emergency will come to the Emergency News Center for the Dean of Engineering or his designate.
3.3 Off-Campus Service Support To assist Manhattan College Emergency Organization, outsiA agencies may be called to actionby the Emergene,y Directc r. l
- 3. 3.1 Medical Assistance St. Jo seph's Medical Cente r, Yonkers, New York has medical facilities needed to render immediate treatment to contaminated and non-contaminated injured personnel. Appendix 3A is a copy of the letter of understanding between Manhattan College and St. Jo s e ph's Medical Center.
3.3,2 Ambulance Service St. Joseph's Medical Center will provide ambulance service when needed.
3.3.3 Polic e The Manhattan College Security Department will provide routine i
security for the reactor.-If additional security is needed, the Blice Department of the City of New York will provide the needed extra s e cu rity.
3.3.4 Fire As sistance The Fire Department of the City of New York may be called for assistance if the need arises.
3.3.5 Nuclear Regnlatory Commis sio_n If the need arises assistance may be sought from the Nuclear Regulatory Commis sion, Region I office in King of Prussia, Pa.
ace w um om- enum 1
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- l[;, ;i :'\ JMedical venter 127 Souh Doodway, Yonkes, New YcA 10701 Clune 13, 1983 cal Engineering arkway
!471 onfirm St. Joseph's Medical Center's ability to diation hazards from your college due to its y our department along with the hospital's mmittee is updating and modifying both the internal ion Safety policies and procedures. Once this mpleted, I will forward a copy to your office for I has procedures and protocols designed to handle fgencies24hoursperday,sevendaysperweek.
19 and selected Radiologic technical personnel are Emergencies. The Emergency Room personnel also ifollow in case an accident occurs during evening In the meantime, I have supplied you some information
'ial in case an emergency occurs (attached).
$his letter you have additional questions or concerns, .
bopy of our final report will be sent to you when l i
l f sin cerely , ,
- IC- s Robert Kleinbauer '
. i Administrator-Radiology Services St. Joseph's Medical Center i
l cy File l .,
(al (09 Ai OARA 7F.n A t :-m LJ-- - /n d Ai n A R A A nn i I ________________-O
APPENDIX 3A - PAGE 2
\TTACHMENT I IUCLEAR MEDICINE DEPARTMENT:
HOURS OF OPERATION: MONDAY - FRIDAY 7:30 AM - 5:30 PM EMERGENCY COVERAGE : CALL ADMINISTRATOR ON CALL OR RADIOLOGIST ON CALL PHONE NUMBERS : (914) 965-6700 EXT. 687,688 PERSONNEL:
MS. JEANINE WYKA - NUCLEAR MED. TECHNICIAN MR. GENE TOLENTINO - NUCLEAR MED. TECHNICIAN MR. ROBERT KLEINBAUER - ADMINISTRATOR DR. SMILJAN PULJIC - DIRECTOR OF RADIOLOGY 4 CASE NUCLEAR MEDICINE DEPARTMENT IS CLOSED CALL RADIOLOGY DEPARTMENT CXT. 683) OR EMERGENCY ROOM (E XT . 4 71) , EXPLAIN SITUATION , LEAVE NAME ID PHONE NUMBER; THEY WILL CONTACT ALL APPROPRIATE STAFF.
l L - - - _ - _ -
- 17 ency Classification System stually, all possible emergencies at any research and test e have been classified into four groups. Appendix I of the ed Review Plan [1] (repeated on the next few pages) lists s
Notification of Unusual Events Ale rt Site Area Emergency General Emergency e MC ZPR licensed at 0. I watt and with no recirculating cooling ements, only the first class of emergency, viz; Notification sual Event is hypothesized as the most severe credible nt.
- ation of Unusual Events nergency class " Notification of Unusual Events" is hypothesized it at MCZPR if events are in progress or have occurred which
- e potential degradation of the safety of the reactor. Spe cific levels are detailed in Section 5. O, Emergency Action Levels tification of Unusuc 1 Events.
F. Bate s, B. K. Grime s and S. L. Ramo s r
andard Review Plan for the Review and Evaluation Emergency Plans for Research and Test Reactors S. Nuclear Regulatory Commis sion JREG-0849, May 1982
- 18 -
APPENDIX I EMERGENCY CLASSES AND EXAMPLE EMERGENCY ACTION LEVELS Emergency Class Example Action Levels Notification of 1. Actual or projected radiological effluents at the Unusual Events site boundary exceeding 10 MPC for unrestricted areas when avoraged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or 15 mrem whole body accumulated in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 1_/
2 Report or observation of severe natural phenomenon that are imminent or existing such as:
(1) earthquakes that could adversely affect the reactor safety systems, (2) high or low natural water sources that could adversely affect reactor safety systeme and: (3) tornado or hurricane winds that could strike the facility.
- 3. Threats to or breaches of security.
- 4. Fuel damage accident that could release radio-
, nuclides to confinement or containment.
- 5. Fire within the facility lasting more than 10 minute s.
-1/
It should be noted that the radiation dose levels of the emergency action levels established for the various emergency classes are slightly different from those specified for power reactors. However, in the judgment of the NRC staff, the radiation dose levels specified are adequate for the credible accidents associated with the operation of research and test reactors, and the specified action levels provide reasonable assurance that protective measures associated with the action levels specified can and will be l taken, provided appropriate emphasis is also given to developing emergency l action levels that relate directly to facility parameters (e. g. , pool water levels and area radiation monitors).
19 -
Emergency Class Example Action Levels Ale rt 1. Actual or projected radiological effluents at the site boundary exceeding 50 MPC for unrestricted ar when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or 75 mrem whole body accumulated in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l__/
- 2. Radiation levels at the site boundary of 20 mrem /hr for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> whole body or five times this level to l
the thyroid. ._/
- 3. Abnormal loss of water used for shielding or coolant to irradiated reactor fuel at a rate which either exhausts the initial bacle.ip system capacity or exceeds makeup capacity.
4 Loss or radioactive material control that causes radiatica dose rates or airborne radionuclides to increase ambient exposure levels by a factor of 1000 throughout the reactor building.
- 5. Fire that may affect any reactor safety system (s).
- 6. Other Imminent or existing hazards such as (1) mis s'les impacting on the reactor fe cility, (2) explosion that affects facility operation, and (3) uncontrolled release of toxic or flammable gase s into the facility environs.
- 7. Radiation dose rates in the reactor building requiring evacuation of all per uonnel (e. g. ,
100 mrem /hr for one hour throughout the reactor building.
1_ / It should be noted that the radiation dose levels of the emergency action levels established for the various emergency classes are slightly differen.t from those specified for power reactors, However, in the judgment of the NRC staff, the radiation dose levels specified are adequate for the credible accidents associated with the operation of resee rch and test reactors, and the specified action levels provide reasonable assurance that protective measures associated with the action levels specified can and will be taken, provided appropriate emphasis is also given to developing emergency action levels that relate directly to facility parameters (c. g. , pool water levels and area radiation monitors).
- 20
. Example Action Levels Emergency Clas s _ Actual or projected radiological effluents at the 1.
Site Area site boundary exceeding 250 MPC for unrestricted Emergency areas when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or 375 mrem accumulated in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, l__/
2.
Actual or projected radiation levels at the site boundary of 100 mrem /hr for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> whole body or five time s this level to the thyroid, l__/
3.
Abnormal continuing loss of reactor coolant, to fuel requiring coolant, at a rate greater than the capacity of the backup system (s).
4.
Imminent loss of physical control of the reactor.
5.
Several natural events being experienced.
Examples a) include: earthquake that is causing observable damage to the reactor safety equipment within the building.
b) high or low natural water levels that are affecting the operability of any reactor safety system; and c) tornado or hurricane winds that are damaging the reactor structure.
1/ It should be noted that the radiation dose levels of the emergency action levels established for the various emergency However,classes are slightly in the judgment different of the from those specified for power reactors.
NRC staff, the radiation dose levels specified are adequate for the and credible accidents associated with the operation of research and test reactors, the specified action levels provide reasonable assurance that protective measures associated with the action levels specified can and will be taken, provided appropriate emphasis is also given to developing emergency action levels that relate directly to facility parameter s (e. g. , pool water levels and area radiation monitors).
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Example Action Levels
- 1. Sustained actual or projected radiation levels at the site boundary of 500 mrem /hr.
- 2. Actual or projected doses radiation levels at the site boundary in the exposure pathway of 1 rem whole body or 5 rem thyroid.
- 3. Loss of reactor coolant that could lead ,,.
to fuel melt.
- 4. Loss of physical control of the reactor building or reactor control room and areas housing vital equipment.
Events that have caused or will cause massive 5.
facility or reactor system damage that could lead to fuel melt.
1
'ified for facilities with authorized power levels le ss lMW thermal and determined on a case by case basis l
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- ZZ -
- 5. 0 Emergency Action Levels for Notification of Unusual Events
- 1. Indication of fuel damage by increased concentration of radionuclides in the reactor water Loss of Confinement integrity j 2.
- 3. Threats to or breaches of security
- 4. Fire v/ithin the facility lasting more than 10 minutes
- 5. Report or observation of severe natural phenomenon that are imminent or are existing such as:
a) ea rthquake
'b) hurricane c) flood
- 6. Other hazards or events at the facility such as:
a) aircraft crash b) explo sion c) release of toxic or flammable gas
- 7. Injury to personnel (contaminated or not) that requires transportation to an off-campus medical facility It is recognized that the above items have very low probability of occurrence at the MC ZPR facility.
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- 6. 0 Emergency Planning Zone No radiological emergency that could result in off-site plume exposures exceeding 1 rem whole body or 5 rem thyroid is plausible at MC ZPR and hence no Emergency Planning Zone is identified for MCZPR.
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- 7. 0 Emergency Response j
Emergency classes of Alert, Site Area Emergency and General I Emergency are not postulated to occur at MCZPR. Hence, the emergency response is given only for the emergency class, Notification of Unusual Events.
7.1 Activation of Emergency Organization The Emergency Organization will be activated by the following steps:
- 1. If and when a MCZPR staff member becomes aware of or is informed of the possible existence of an emergency, he notifies the Reactor Administrator, the Chief Reactor Supervisor or a Reactor Supervisor, whoever he can reach in that order.
- 2. The first person who is entitled to act as the Emergency Director :
to reach the Reactor Facility verifies the existence of an emergency, -
declares an emergency and assumes the role of the Emergency Dire cto r.
- 3. A Reactor Supervisor who assumes the role of the Emergency l Director will relinquish that role to the Chief Reactor Supervisor when the latter reache s the facility. The Chief Reactor Supervisor will relinquish the role of the Emergency Director when the Reactor Administrator arrives at the facility.
- 4. The Emergency Organization is activated by the Emergency Dire cto r. The call-out roster of the members of the Emergency Organization with their telephone numbers (both home and office) shall be posted at the Control Console of the reactor and at the j
Emergency Support Center. The se locations are shown in Figure 7.1.
The roster will also contain telephone numbers of the off site -
agencies such as NRC, ambulance, hospital, police and fire department that may be activated if necessary.
Telephones are provided at the Reactor Control Console and at the Emergency Support Center. Telephones are also available in faculty and department offices and in the Dean's office.
- 7. 2 As se ssment Action The reactor operations staff are p rovided with film badges. Extra badge s are also available . These may be used to assess radiation doses, if any, to personnel during an emergency.
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@ EMERGENCY SUPPORT CENTER FIGURE '7,1 LOCATICIT OF EMERGENCY ORGANIZATION CHART
-25_
Calibrated portable radiation meters are available at the reactor facility. An assessment of radiation levels may be made using ;
these instruments. If deemed nece s sary, the health physicist will take wipes and air samples and assess the extsnt of the contamina-
- j. tion, if any.
- 7. 3 Notification
" Notification of an Unusual Event" shall be made no later than the next working day to NRC, Region 1, King Of Prus sia, Pa. by the Reactor !
l Administrator r. If this cannot be accomplished by the Reactor Administrator in the allotted time due his absence, then the Emergency Director who acts in that capacity on behalf of the Reactor Administrator will accomplish the notification process.
- 7. 4 Leaving the Facility Before the Termination Of An Emergency i If the Emergency Director has to leave the facility for any reason, he may do so only after appointing a qualified substitute to act as the {
Emergency Director. l I
- 7. 5 Termination of Eme rgency i When appropriate, the Emergency Director shall declare the termination of emergency. ,
- 7. 6 Emergency Log An Emergency Log will be maintained by the Emergency Director.
The log should contain the time date of the declaration of every ;
emergency, the name of the Emergency Director, action taken during '
an emergency, time and date of the declaration of the termination of the emergency. The Emergency Log should be kept at the Emergency Support Center.
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- 8. 0 Emergency Facilities and Equipment
- 8. I' Emergency Support Center The platform in Room 221 (shown double hatched in Figure 8.1) will serve as the Emergency Support Center.
- 8. 2 Emergency Equipment Portable Radiation Survey meters are available in the ZPR room.
l 8. 3 First Aid Facilities ,
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First aid supplies are available in cabinets at the Emergency Support Center.
- 8. 4 Communication, Equipment Telephones are provided at the Reactor Control Console and at the Emergency Support Center. Additional phones are available in the offices of the Dean of Engineering and faculty and department offices.
- 8. 5 Decontamination Equipment Decontamination equipment is provided at the Emergency Support C e nte r.
- 8. 6 Handling of Injured and/or Contaminated Personnel Arrangements hav neen made with a medical center (Appendix 3A) for the handling oi injured and/or contaminated personnel.
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- 9. 0
- 9. I Recovery Manager The Chief Reactor Supervisor, or in his absence, a Reactor Supervisor will act as the Recovery Manager.
The Recovery Manager will:
- 1. See that the facility is brought back to rormal for the normal operation of the reactor
~
- 2. Perform the normal checkout procedures of the reactor.
- 3. Enter in the Reactor Operations Log book that the reactor has been recovered after the emergency and that the reactor is ready for normal operation.
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Maintaining Emergency Preparedness 20.0 .
- 10. 1 Annual Review These emergency plans shall be reviewed by the Reactor Operations Committee at least once a year.
10.2 The Emergency Organization chart and the call out roster shall be corrected whenever necessary and the updated charts should be posted and distributed as needed.
10.3 Implementing procedures affected by any emergency plan j changes shall be revised, approved and distributed to authorized l recipients within30 days after the revised plans have been issued.
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