ML20077L327
| ML20077L327 | |
| Person / Time | |
|---|---|
| Site: | 05000131 |
| Issue date: | 07/31/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0988, NUREG-988, NUDOCS 8308080731 | |
| Download: ML20077L327 (70) | |
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NUREG-0988 Safety Evaluation Report related to the renewal of the operating license for the research reactor at the Omaha Veterans Administration Medical Center Docket No. 50-131 U.S. Nuclear Regulatory Commission Offico of Nuclear Reactor Re.gulation July 1983 pa arouy khf)
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NUREG-0988 Safety Evaluation Report related to the renewal of the operating license for the research reactor at the Omaha Veterans Administration Medical Center Docket No. 50-131 U.S. Nuclear Regulatory Commission l
Office of Nuclear Reactor Regulation July 1983 p*== n,
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l ABSTRACT This Safety Evaluation Report for the application filed by the Omaha Veterans Administration Medical Center (0VAMC) for a renewal of operating license number R-57 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission.
The facility is owned and operated by the Veterans Administration and is located in its Medical Center in the city of Omaha, Nebraska.
The staff concludes that TRIGA reactor facility can continue to be operated by OVAMC without endangering the health and safety of the public.
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OVAMC SER iii I
l
TABLE OF CONTENTS P_jgte ABSTRACT.............................
iii 1
INTRODUCTION.
1-1 1.1 Summary and Conclusions of Principal Safety Considerations...
1-1 1.2 Reactor Description..................
1-2 1.3 Reactor Location.......
1-3 1.4 Shared Facilities and Equipment and Special Location Features.
1-3 1.5 Comparison with Similar Facilities...........
1-3
- 1. 6 Nuclear Waste Policy Act of 1982............
1-3 2
SITE CHARACTERISTICS....................
2-1 2.1 Geography... 1 2.2 Demography................,
2-1 2.3 Nearby Industrial, Transportation, and Military Facilities..
2-1 2.4 Meteorology......................
2-1 2.4.1 Tornadoes 2-2 2.5 Geology.
2-2 2.6 Hydrology...
2-3 2.7 Seismology.
2-3 3
DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS........
3-1 3.1 Wind Damage.
3-1 3.2 Water Damage..
3-1 3.3 Seismic-Induced Reactor Damage..
3-1 3.4 Mechanical Systems and Components.
3-1 3.5 Conclusion..
3-2 4
REACTOR..
4-1 4.1 Summary Description.....
4-1 4.2 Reactor Core.
4-1 4.2.1 Fuel Elements.....
4-1 4.2.2 Control Rods.....
4-2 4-3 4.2.3 Conclusion...................
4.3 Reactor Tank.
4-3 4.4 Support Structures............
4-3 OVAMC SER v
TABLE OF CONTENTS (Continued)
P_iBLe 4.5 Shielding...
4-4 4.6 Reactor Instrumentation................
4-4 4.7 Dynamic Design Evaluation.
4-4 4.7.1 Excess Reactivity, Experiment Worth, and Shutdown Margin 4-5 4.7.2 Conclusion.
4-5 4.8 Functional Design of Reactivity Control System....
4-5 4.8.1 Control Rod Drive..
4-6 4.8.2 Scram-Logic Circuitry...
4-6 4.8.3 Conclusions...................
4-6 4.9 Operational Practices.
4-7 4.10 Conclusions..
4-7 5
REACTOR COOLING SYSTEM...
5-1 6
ENGINEERED SAFETY FEATURES......
6-1 7
CONTROL AND INSTRUMENTATION SYSTEMS.............
7-1 7.1 Systems Summary....................
7-1 7.2 Control Console...
7-1 7.3 Control System.
7-1 7.3.1 Nuclear Control System.
7-1 7.3.2 Process Control Systems.
7-2 1
7.4 Instrumentation System......
7-2 7.4.1 Nuclear Instrumentation 7-2 l
7.4.2 Process Instrumentation..
7-3 7.5 Conclusions.
7-4 8
ELECTRICAL POWER SYSTEM.
8-1 9
AUXILIARY SYSTEMS.
9-1 l
9.1 Fuel Handling and Storage.
9-1
)
9-1
}
9.2 Fire Protection System.
9.3 Conclusion..
9-1 10 EXPERIMENTAL PROGRAMS.
10-1 l
l 10.1 Experimental Facilities.
10-1 i
l OVAMC SER vi
TABLE OF CONTENTS (Continued)
Page i
10.1.1 Pool Irradiations...............
10-1 10.1.2 Pneumatic Transfer System.
10-1 10.1.3 Rotary Specimen Rack.....
10-1 1
10.1.4 Central Thimble.
10-2 10.2 Experimental Review.
10-2 10.3 Conclusions.
10-3 11 RADI0 ACTIVE WASTE MANAGEMENT................
11-1 11.1 Waste Generation and Handling Procedures.
11-1 11.1.1 Airborne Waste.
11-1 11.1.2 Liquid Waste.
11-1 11.1.3 Solid Waste.
11-2 11.2 Conclusion.
11-2 12 RADIATION PROTECTION PROGRAM.
12-1 12.1 ALARA Commitment.
12-1 12.2 Health Physics Program.
12-1 12.2.1 Health Physics Staffing.
12-1 12.2.2 Procedures.
12-1 12.2.3 Instrumentation.
12-1 12.2.4 Training.
12-2 12.3 Radiation Sources.
12-2 12.3.1 Reactor.
12-2 12.3.2 Extraneous Sources.
12-2 12.4 Routine Monitoring.
12-2 12.4.1 Fixed-Position Monitors.
12-2 12.4.2 Experimental Support.
12-3 12.5 Occupational Radiation Exposures..
12-3 12.5.1 Personnel Monitoring Program.
12-3 12.5.2 Personnel Exposures.
12-3 12.6 Effluent Monitoring.
12-3 12.6.1 Airborne Effluents.
12-3 12.6.2 Liquid Effluents.
12-3 12.7 Environmental Monitoring'.
12-3 OVAMC SER vii
r TABLE OF CONTENTS (Continued)
Page 12.8 Potential Dose Assessments..............
12-4 12.9 Conclusions.............
12-4 13 CONDUCT OF OPERATIONS....................
13-1 13.1 Overall Organization.
13-1 13-1 13.2 Training.
13.3 Emergency Planning...
13-1 13.4 Operational Review and Audits.............
13-1 13.5 Physical Security Plan.
13-1 13.6 Conclusion.....
13-2 14 ACCIDENT ANALYSIS.
14-1 14.1 Natural Phenomena..
14-1 14.2 Rapid Insertion of Reactivity....
14-2 14.3 Loss of Coolant / Shielding Water 14-2 14.4 Misplaced Experiments.................
14-4 14.5 Mechanical Rearrangement of the Fuel.........
14-4 14.6 Effects of Fuel Aging.
14-4 14.7 Handling Irradiated Fuel...............
14-6 14.7.1 No Environmental Release..
14-7 14.7.2 Environmental Release.
14-8 14.7.3 Conclusion.
14-8 i
14.8 Conclusions......................
14-8 15 TECHNICAL SPECIFICATIONS.........
15-1 16 FINANCIAL QUALIFICATIONS...
16-1 17 OTHER LICENSE CONSIDERATIONS.......
17-1 17.1 Prior Reactor Utilization..
17-1 17.2 Multiple or Sequential Failures of Safety Components.....................
17-2 18 CONCLUSIONS.........
18-1 19-1 19 REFERENCES.........
OVAMC SER viii l
LIST OF FIGURES Page 1.1 Reactor Laboratory Basement.
1-4 4.1 Reactor and Pit.
4-8 4.2 Core and Reflector Assembly..
4-9 4.3 Control-Rod Drive Mechanism.
4-10 5.1 Reactor Cooling Systems.....
5-2 7.1 Block Diagram of Instrumentation 7-5 13.1 Facility Organization....................
13-3 LIST OF TABLES 2.1 Number of Tornadoes, Tornado Days, and Tornado Deaths in Nebraska (1956-1973).
.2-4 7.1 Minimum Reactor Safety Channels.
7-6 12.1 Recent Exposure History for Reactor Facility Personnel...
12-5 14.1 Radiation Doses Within the Reactor Laboratory:
No Environmental Release.
14-9 14.2 Radiation Exposures in the Plume from the Design-Basis Accident.
14-9 OVAMC SER ix
1 INTRODUCTION The Omaha Veterans Administration Medical Center (OVAMC/ licensee) submitted an application for 10 year renewal of the Class 104 operating license (R-57) (NRC Dncket No. 50-131) for its TRIGA research reactor facility to the U.S. Nuclear Regulatory Commission (NRC) (staff) by letter (with, supporting documentation) dated May 10, 1979.
The application was signed and notarized by the Director of the Omaha Veterans Administration Medical Center.
The renewal application referenced the information regarding the design of the facility included in the application for the original construction permit and included supplements to the Safety Analysis Report, information for an Environ-mental Impact Appraisal, proposed modified Technical Specifications, an Emer-gency Plan, an Operator Requalification Program, a Fiscal Statement, and, under separate cover, a Physical Security Plan, which is protected from public dis-closure under Title 10 of the Code of Federal Regulations (10 CFR) 2.790.
The staff's technical safety review with respect to issuing a renewal operating license to OVAMC has been based on a visit to the facility and the information contained in the renewal application and supporting appendices plus responses to requests for additional information.
This material is available for review at the Commission's Public Document Room at 1717 H Street N.W., Washington, D.C.
This Safety Evaluation Report (SER) was prepared by R. E. Carter, Proj-ect Manager, Division of Licensing, Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission.
Major contributors to the technical review include the Project Manager and A. Chu (NRC) and J. E. Hyder, C. Linder, and C. C. Thomas, Jr. of the Los Alamos National Laboratory under contract to NRC.
The purpose of this SER is to summarize the results of the safety review of the OVAMC research reactor and to delineate the scope of the technical details con-sidered in evaluating the radiological safety aspects of continued operation.
This SER will serve as the basis for renewal of the license for operation of the OVAMC facility at power levels up to and including 18 kW.
The facility was reviewed against Federal regulations (10 CFR 20, 30, 50, 51, 55, 70, and 73),
applicable regulatory guides (Division 2, Research and Test Reactors), and appropriate accepted industry standards (American National Standards Institute /
American Nuclear Society (ANSI /ANS) 15 series).
Because there are no specific accident-related regulations for research reactors, the staff has at times com-pared calculated dose values with related standards in 10 CFR 20, " Standards for Protection Against Radiation," both for employees and the public.
1.1 Summary and Conclusions of Principal Safety Considerations The staff's evaluation considered the information submitted by the licensee, past operating history recorded in annual reports submitted to the Commission by the licensee, and reports by the Commission's Office of Inspection and Enforcement.
In addition, as part of its licensing review of several TRIGA reactors, the staff obtained laboratory studies and analyses of several acci-dents postulated for the TRIGA reactor.
The resolution of principal issues reviewed for the OVAMC reactor were OVAMC SER 1-1
(1) The design, testing, and performance of the reactor structure and systems and components important to safety during normal operation are inherently safe, and safe operation can reasonably be' expected to continue.
(2) The expected consequences of a broad spectrum of postulated credible acci-dents have been considered, emphasizing those that could lead to a loss of integrity of fuel-element cladding.
The staff performed conservative analyses of the most serious credible accidents and determined that the calculated potential radiation doses outside the reactor room wauld not exceed 10 CFR 20 doses for unrestricted areas.
(3) The licensee's management organization, conduct of training and research activities, and security measures are adequate to ensure safe operation of the facility and protection of special nuclear material.
(4) The systems provided for the control of radiological effluents can be operated to ensure that releases of radioactive wastes from the facility are within the limits of the Commission's regulations and are as low as reasonably achievable (ALARA).
(5) The licensee's Technical Specifications, which provide operating limits controlling operation of the facility, are sucn that there is a high degree of assurance that the facility will be operated safely and reliably.
(6) The financial data provided by the licensee are such that the staff has determined that the licensee has sufficient revenues to cover operating costs and eventually to decommission the reactor facility, (7) The licensee's program for providing for the physical protection of the facility and its special nuclear material complies with the requirements of 10 CFR 73.
(8) The licensee's procedures for training reactor operators and the plan for operator requalification are acceptable.
These procedures give reasonable assurance that the reactor facility will be operated with competence.
(9) The licensee has submitted an Emergency Plan in compliance with the exist-ing applicable regulations.
This item is discussed further in Section 13.3.
- 1. 2 Reactor Description The OVAMC TRIGA facility is an open tank-type heterogeneous, light-water-cooled reactor.
The core is moderated by zirconium hydride and water and reflected by water and graphite.
It is located near the bottom of a steel tank in a cylin-drical pit below ground level.
The concrete-lined tank rests on a concrete slab. The reactor core currently consists of 56 standard aluminum-clad uranium-zirconium-hydride (U-ZrH ) fuel elements.
The elements are spaced in grid plates x
so that about 33% of the core volume is occupied by water.
Shielding above the reactor core is provided by 16 ft of water, and the core is cooled by natural convection of the water in the tank.
The reactor is designed and licensed to operate at a steady-state power level of 18 kW with a maximum available excess reactivity of 1.00$.
It attained criticality with 54 fuel elements containing about 1.9 kg of 2ssU.
I in the fuel is enriched to less than 20% 23sU.
For purposes of testing and calibration, the reactor may be operated at power levels up to 19.8 kW.
The reactor was constructed in 1959; its initial operating license was issued on June 26, 1959, and its most recent renewal license (R-57) was issued on June 2, 1969, for a period of 10 years.
1.3 Reactor Location The TRIGA reactor facility is located in the Omaha Veterans Administration Medical Center, City of Omaha, Douglas County, Nebraska.
The reactor is housed in the basement of the southwest wing of the hospital building (see Figure 1.1).
1.4 Shared Facilities and Equipment and Special Location Features The reactor room is on the basement floor of the medical center hospital build-ing, where it is used primarily in research and radioisotope production related to the diagnosis and treatment of disease.
Utilities such as municipal water and nonradioactive sewage, natural gas, and electricity are provided for joint use in the entire building.
The reactor room has a separate ventilation system exhausting through filters to the outside environment and a chiller system dedi-cated to removing heat from the reactor water.
Research and preparation labora-tories share the reactor room, and the chemical hoods in these laboratory rooms are separately exhausted on the hospital building roof.
The medical center hospital building is located several hundred meters from the nearest dwelling in the unrestricted area.
- 1. 5 Comparison with Similar Facilities The reactor core geometry is similar to that of most of the 58 TRIGA reactors in operation throughout the world, 27 of which are in the United States.
How-ever, the cladding for the OVAMC reactor fuel is aluminum, which was used only in the first few TRIGA reactors.
The instruments and controls are similar to those on other research reactors licensed by the NRC.
1.6 Nuclear Waste Policy Act of 1982 Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 provides that the NRC may require, as a precondition to issuing or renewing an operating license for a research or test re3(tor, (that the applicant shall have entered into an agreement with the Dep-Je v 9t of Energy (DOE) for the disposal of high-level radioactive waste a-e Spa '. nuclear fuel.
DOE (R. L. Morgan) has informed the NRC (H. Denton) by 'C to-
!ated May 3, 1983, that it has determined that universities and
..t,mr
-roment agencies operating nonpower reactors have entered into contracts with DOE that provide that DOE retain title to the fuel and is obligated to take the spent fuel and/or high-level waste for storage or reprocessing.
Because the Veterans Administration is an agency of the Federal government, it is in conformance with the Waste Policy Act of 1982.
OVAMC SER 1-3
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2 SITE CHARACTERISTICS 2.1 Geography The OVAMC is built on a knoll, at an elevation of ~370 m above mean sea level (MSL) in a commercial area within the city limits of Omaha, Douglas County, Nebraska.
Omaha, which is on the eastern border between Nebraska and Iowa on the Missouri River, is at an elevation ~300 to 390 m above MSL.
Thus, the hospital is on some of the highest ground within the city.
To the north is a large county hospital, to the south a commercial district, to the west a resi-dential area, and to the east a golf course.
The OVAMC grounds are large, so the nearest offsite dwelling is several hundred meters away.
2.2 Demography The metropolitan area of Omaha includes suburbs in both Nebraska and Iowa with a 1980 population of 570,000, which has shown s5% increase since 1970.
Popu-lation within the city of Omaha itself has decreased from 347,000 to 312,000 during the same decade.
The change in population since the 1930s has shown a definite trend towards the northwest, west, and southwest.
2.3 Nearby Industrial, Transportation, and Military Facilities Omaha has no major heavy industry, but there is a large railroad yard and main-tenance shop $3.2 km to the se:.theast.
There is a main line of the Missouri Pacific railroad $300 m to the east, and the headquarters of the Strategic Air Command, Offutt Air Force Base is some 13 km to the southeast.
The Omaha air-port is more than 10 km from the medical center facility.
Because there are no industrial or military facilities in the near vicinity of the reactor site that could directly or indirectly cause accidental damage to the reactor itself, the staff concludes that such accidents need not be hypothe-sized and examined in detail.
2.4 Meteorology Omaha is situated on the west bank of the Missouri River; the river level at Omaha is normally $293 m (965 ft) above MSL.
The rolling hills in and around Omaha rise to $395 m (1,300 ft) above MSL.
The climate is typical continental, with relatively warm summers and cold, dry winters.
It is situated midway between two distinctive climatic zones--the humid east and the dry west.
Fluctuations between these two zones produce periods of weather conditions that are characteristic of either zone or combi-nations of both. Omaha is also affected by most storms that cross the country.
This causes frequent and rapid changes in weather, especially during the winter.
The prevailing winds at the Omaha Airport are SSE during most of the year, shifting to NNW during the winter quarter.
OVAMC SER 2-1
a Most of the precipitation falls during sudden showers or thunderstorms from April to September.
Of the total precipitation, about 75% falls during this 6-month period, predominantly as evening or night showers and thunderstorms.
Although winters are relatively cold, precipitation is light, with only 10% of the total annual precipitation falling during the winter.
Sunshine is fairly abundant, ranging from around 50% of the possible in the winter to 75% of the possible in the summer.
The mean date of the last killing freeze in spring is April 14, and the mean date of the first killing freeze in autumn is October 20.
The longest freeze-free period on record is 219 days in 1924, and the shortest period, 152 days in 1885.
The average length of the freeze-free period is 188 days.
2.4.1 Tornadoes For the OVAMC reactor site, the annual tornado frequency is 2.0 (H.C.S. Thom, 1963) in a 1 degree square centered near the site.
Therefore, the probability of a tornado hitting the site in any one year is 0.00158 with a return frequency about once every 633 years.
However, tornadoes have been recorded in the gen-eral area of the site.
A total of 20 tornadoes occurred in a 1 degree square centered near the site from 1953 to 1962.
On May 6, 1975, the tornado that hit Omaha was one of the most destructive, resulting in three deaths and up to $500 million in property damages.
Table 2.1 presents the tornado statistics for Nebraska from 1965 through 1973.
Because the OVAMC reactor facility is in the basement of the hospital building, surrounded by poured concrete walls with no windows and with 7 to 11 cm of concrete overhead, tornado damage to the reactor itself is very unlikely.
2.5 Geology The area lies within the Dissected Till Plains of the Central Lowland Physio-graphic Province of the United States.
The topography is gently rolling, and the ground surface at the OVAMC lies at an elevation ~370 m above MSL.
This elevation represents some of the highest ground within the city limits of Omaha, being s80 m above the level of the Missouri River.
The surface soils in the Omaha area are primarily loess and glacial drift deposits.
Two stages of glaciation, the Nebraskan and the Kansan, left thick deposits of till overlying bedrock.
It is believed that much of the glacial till has been eroded.in the vicinity of the OVAMC and that not more than 30 m remains.
The till consists mainly of lean and gravelly clays with a few lenses of sand gravel.
The exact depth to bedrock directly below the OVAMC site is not known but is estimated to vary between MSL elevation of 300 and 320 m, on the basis of the nearest top bedrock information.
The loess at the site is of Peorian and Loveland Formations of the late Pleisto-cene Epoch.
The soil classification of the Peorian indicates that the material consists predominantly of clayey silts and lean clay.
The soil of the Loveland formation varies from clayey silt to fat clay with minor amounts of sand and clayey sand in the basal part of the formation.
At the OVAMC site, the Peorian is from 10 to 15 m thick and the Loveland is over 20 m thick. This would mean that the total thickness of the overburden is $60 m.
OVAMC SER 2-2
Bedrock in this area is limestone and' shale of the Pennsylvania period.
The surface of the bedrock is very irregular because of an extensive period of erosion that followed the uplift of the area in early Pennsylvania time and continued to the Pleistocene Epoch.
This uplift brought the granite basement to within 180 m of the surface in certain areas, forming a ridge known as the Nemaha Ridge or Arch.
A major structure, the Humboldt Fault,.which has a throw of over 275 m, is associated with the Nemaha Arch.
The Humboldt structure zone is assumed to continue through a point in the southeastern city limits of Omaha.
The Humboldt Fault has not been considered to be a capable fault within the meaning of Appendix A,10 CFR 100, based on investigations for the Cooper, Fort Calhoun, and Wolf Creek Nuclear sites.
2.6 Hydrology Because no piezometers were installed or observation wells drilled at the site, there is no definite information as to the exact depth of the water table.
How-ever, on the basis of logs of borings drilled in 1946, the zone of saturation appears to be below 20 m, although there is some indication of perched water levels in the soil strata as high as 4.6 m.
Furthermore, because the OVAMC is sited on a knoll, there is reasonable assurance that neither surface nor gound-waters make the location unsuitable for the reactor facility.
2.7 Seismology The OVAMC site is located in Seismic Risk Zone 1 of the United States (Alger-missen, 1969), which is defined as " Minor damage, distant earthquakes may cause damage to structures with fundamental periods greater than 1.0 seconds, corresponding to intensities V and VI of the Modified Mercalli (MM) scale."
Intensity VI on the MM intensity scale is described as " Felt by all; many frightened and run outdoors.
Some heavy furniture moved; a few instances of fallen plaster or damaged chimneys.
Damage slight."
Historically there have been about 63 earthquakes reported within 320 km of the site.
The four largest oi these have had maximum MM intensities of VII.
The closest of these four events was about 100 km from the site and it is estimated that the intensity at the site from this event was about MM V.
There have been no reports or physical evidence of earthquakes at the site.
Searches of local historical newspaper accounts and other published materials on earthquake effects have led to the following conclusions:
i (1) no significant building damage (2) no loss or threat of loss of life (3) no livestock or crops affected This tectonically stable region is characterized by relatively low intensity as well a relatively low frequency of earthquakes.
The staff concludes that the history of no significant earthquake damage in the site region supports the conclusion that seismic-induced hazards to the OVAMC reactor are not significant.
OVAMC SER 2-3 I
l l
Table 2.1 Number of tornadoes, tornado i
days, and and tornado deaths in Nebraska (1956-1973)
Parameters Values Tornadoes Total number 590 Annual average 33 Greatest number (1958) 54 Least number (1966) 10 Tornado days Total 283 Annual average 16 Tornado deaths Total 14 Annual average 1
Source:
U.S. Department of Commerce,
" Climatological Data," National Summary NOAA Vol. 24, 1973.
9 l
l OVAMC SER 2-4
3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS Chapter 3 of the licensee's Safety Analysis Report (SAR) provides information on the design and functions of the reactor building, reactor systems, and auxiliary systems.
3.1 Wind Damage Meteorological data indicate a low frequency of tornadoes and effects of tropi-cal disturbances.
The OVAMC reactor facility is in the basement of the hospital building, surrounded by poured concrete walls with no windows and with 7 to 10 cm of concrete overhead, making tornado damage improbable.
3.2 Water Damage The OVAMC reactor building is s'ituated on a knoll s370 m above sea level, which is much higher than most of the ground within the Omaha city limits and s80 m above the level of the Missouri River.
Therefore, the staff concludes that there is reasonable assurance that damage to the reactor by flood or ground-water is small.
3.3 Seismic-Induced Reactor Damage Analyses of newspaper accounts since 1867 indicate that the site is in a tec-tonica11y stable region characterized by low level as well as low frequency of earthquakes.
The site is located in Seismic Risk Zone 1 of the United States.
There is a risk of slight damage, principally to poorly built or designed structures.
Because of the features of the OVAMC reactor building described in Section 3.1, the staff concluded that the risk of seismic damage to the reactor facility is small.
3.4 Mechanical Systems and Components The mechanical systems of importance to safety are the neutron-absorbing con-trol rods suspended from the reactor superstructure.
The motors, gear boxes, electromagnets, switches, and wiring are above the level of the water and readily accessible for testing and maintenance.
An extensive preventive main-tenance program has been in operation for many years at the OVAMC reactor to conform and comply with the performance requirements of the Technical Specifications.
The effectiveness of this preventive maintenance program is attested to by the small number and types of malfunctions of equipment over the years of operation.
These malfunctions have generally been one of a kind (that is, no repeats) and/
or of components that were fail safe or self annunciating (see Inspection Reports from the Office of Investigation and Enforcement and reports of Report-able Occurrences from the licensee, Docket No. 50-131).
Therefore, the staff concludes that there appears to be no significant deterioration of equipment with time or with operation.
Thus, there is reasonable assurance that continued OVAMC SER 3-1
operation for the requested period of renewal will not increase the risks to 4
the public.
3.5 Conclusion The OVAMC reactor facility was designed and built to withstand all credible and probable wind and water damage contingencies associated with the site.
A seis-mic event has a small likelihood of' occurring and the radiological consequences of such occurrence would not be significant; therefore, the staff has concluded that they need not be evaluated explicitly.
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t OVAMC SER 3-2 t
4 REACTOR 4.1 Summary Description The Omaha Veterans Administration Medical Center (OVAMC) TRIGA reactor is a research reactor designed and fabricated by General Atomic Company.
The reactor first achieved criticality on June 30, 1959.
It is a below grade, open-tank-type, light-water moderated, cooled, and shielded reactor that currently is authorized to operate in the steady-state mode at thermal power levels up to 18 kW with an excess reactivity limitation of 1.00$ (beta-effective for this reactor is assumed to be 0.79% Ak/k).
The reactor is used as a source of ionizing and neutron radiation for research in biology and medicine, including nuclear medicine, clinical chemistry, radio-biology, and biomedical applications of neutron analytical techniques.
The discussion in the following sections is based on information obtained from licensee reports and during visits to the facility and discussions with licensee personnel.
4.2 Reactor Core The reactor core consists of a relatively compact array of standard aluminum-clad TRIGA fuel elements, graphite dummy elements, 3 boron carbide control rods, control rod guides, a startup neutron source, and irradiation facilities (Figures 4.1 and 4.2).
The fuel elements are spaced so that about 33% of the core volume is occupied by water, yielding a fuel-to-water ratio resulting in a critical mass near the minimum value for 20% enriched uranium fuel.
The ele-ments are held in concentric rings by an upper and a lower grid plate.
The reactor currently requires 56 fuel elements to achieve criticality and to pro-vide the authorized excess of reactivity (1.00$) necessary to meet operating requirements.
The balance of the 85 fuel element positions is occupied by experimental facilities or graphite-reflector elements.
The latter are ele-ments in which the U-ZrH fuel is replaced by graphite.
x The core is surrounded by a cylindrical graphite reflector that is completely encased in a welded aluminium container.
The flooding of the reflector con-tainer because of a leak would decrease reactivity.
4.2.1 Fuel Elements The OVAMC reactor currently uses TRIGA aluminum-clad cylindrical fuel elements.
The active part of each fuel element is N3.6 cm in diameter by 0.36 m long and is a solid homogenous mixture of a U-ZrH alloy containing 8 weight percent uranium enriched to less than 20% in 2ssU.
The hydrogen-to-zirconium ratio is approximately 1.0.
A thin aluminum wafer at each end of the active fuel con-tains samarium oxide as a burnable poison.
Each element is jacketed with a 0.076-cm-thick aluminum can.
Ten-cm sections of graphite are inserted in the can above and below the fuel to serve as top and bottom neutron reflectors for the core.
Aluminum end fixtures are attached to both ends of the can.
The OVAMC SER 4-1
overall length of each fuel element is *0.72 m.
Visual examination of fuel elements on a quarterly basis has shown no indication of any deterioration or swelling.
An alternative TRIGA fuel element uses stainless-steel cladding and is the cur-rent standard element.
Like the aluminum-clad elements, the stainless-steel-clad fuel elements are a homogenous mixture of U-ZrH alloy containing approxi-x mately 8.5 weight percent uranium enriched to less than 20% in 2ssU.
The nomi-nal weight of 2ssU in each fuel element is 38 g.
The hydrogen-to-zirconium ratio is approximately 1.65 to 1.7.
The active part of each fuel element is
~3.6 cm in diameter by 0.38 m long.
Aluminum-samarium wafers are located at each end of the active fuel as a burnable poison.
Each element is jacketed with a 0.05-cm-thick stainless-steel can.
Graphite reflector plugs (49 cm long) are located above and below the fuel and serve as neutron reflectors.
Stain-less-steel end-fixtures are attached to both ends of the can.
The overall length of the fuel element is the same as that of the aluminum-clad fuel ele-ment (*0.72 m).
Although the OVAMC reactor currently uses aluminum-clad fuel elements, any replacement requirements will necessarily be stainless-steel cladding because aluminum-clad elements are not readily available from the fabricator (General Atomic Company).
The licensee's revised Technical Specifications provide for use of aluminum-and/or stainless-steel-clad fuel elements.
The staff has re-viewed the use of cores containing fuel elements with both types of cladding and with mixtures of cladding types.
The staff finds that substitution of stainless-steel-clad elements will not result in the degradation of the reactor performance and concludes that there is reasonable assurance that the OVAMC reactor is capable of safe operation, as limited by its revised Technical Speci-fications, with a core containing either or both types of fuel element cladding.
The staff notes that there is extensive operating experience with both types of fuel elements under conditions (power-level and pulsing) that are more severe than those experienced under the operating conditions prevalent in the OVAMC reactor.
4.2.2 Control Rods The power levels in the OVAMC TRIGA reactor are regulated by three boron-carbide (neutron-absorbing material) control rods.
The control rods operate in perfo-rated aluminium guide tubes.
The guide tubes are attached to the bottom grid plate, and the upper grid plate provides lateral support.
Each control rod has an extension tube that connects to a drive mechanism thorough an anvil and electromagnet system.
One control rod, designated as the safety rod, is com-pletely withdrawn from the core during normal operation.
A second control rod, the shim-safety rod, is used for coarse reactor power level control.
The safety and shim-safety rods are each worth approximately 1.90$.
The third control rod is used as a regulating rod for fine control of reactor power and has a worth q
of approximately 0.45$.
The regulating rod is used in conjunction with a servo-amplifier system to provide automatic control of the reactor power and to bring the reactor up to power on preset periods of 30 or 60 sec.
Quarterly visual examination of control rods before December 1973 showed pitting, which required replacement of the rods.
General Atomic Company concluded on the basis of analysis of a rod replaced in 1966 that the pits were probably a result of iron particles embedded in the surface of the rod during manufacture.
The manufac-turing and inspection process was modified to decrease the likelihood of such OVAMC SER 4-2
inclusions.
Periodic inspection since the last replacement of a contr61 rod (November 1973) has revealed no evidence of pitting.
4.2.3 Conclusion The staff has reviewed the information regarding the reactor fuel, core arrange-ment, and reactivity control elements and found that design and performance capability of the components is adequate to ensure the safe operation of the reactor during the proposed license renewal period.
4.3 Reactor Tank The reactor is near the bottom of a below grade cylindrical pit (Figure 4.1) located in the basement of the 11-story OVAMC hospital building.
The pit con-tains a 2.1-m-inside-diarneter steel tank with a 0.64-cm-thick wall.
The tank rests on a 0.28-m-thick concrete slab.
An s0.25-m-thick poured concrete wall surrounds the outside of the tank.
The concrete slab and wall provide a pro-tective barrier between the tank and the surrounding soil.
The inside of the tank is covered on the sides by a layer of "gunite" s5 cm thick and on the bottom by a layer of poured concrete ~10 cm thick.
The entire inner surface is coated with two applications of a waterproof epoxy resin coating.
Visual observation of the tank with binoculars shows no evidence of deterioration of the tank during the past 24 years, and there has been no indication of unex-plained loss of water.
The reactor tank contains s4,000 gal of water with a normal shielding depth of 4.9 m above the top grid plate.
A low-level alarm is actuated if the pool water level falls to 3.6 m above the top of the core.
The natural thermal convection of this water adequately disperses the heat generated in the core by the normal operation of the reactor.
The pool water is pumped through a chiller unit (refrigerator) that ultimately disposes of the heat to the atmosphere outside of the Medical Center building.
The pool water outlet pipe in the cooling system is no more than 1 m below the top of the pool, thereby eliminating the possibility of lowering the water level by more than 1 m in the event of a coolant piping rupture and the resultant siphoning of water out of the reactor tank.
If the external cooling system fails with the reactor operating at 18 kW, the rate of rise of the temperature of the water in the reactor tank will be less than IC per hour providing adequate time for corrective action to be taken.
In the event of the loss of all coolant, the natural convection of air through the core will maintain its temperature below the cladding failure level, and all the fission products that would have been generated during extended opera-tion will be retained within the individual elements (GA-6596, 1965).
Three emergency storage pits are located in the reactor room floor adjacent to the reactor tank.
The pits are vertical 10-in.-diameter steel pipes 3 m long and are lined with an organic coating.
The pits may be filled with water and used for temporary storage of irradiated specimens or irradiated fuel elements before their ult; mate disposal.
4.4 Support Structures A fixed bridge spans the reactor tank at floor level in the reactor room (Fig-ure 4.1).
The control rod drives, specimen-removal drive mechanism, rotary OVAMC SER 4-3
specimen-rack drive mechanism, central irradiation thimble, pneumatic transfer tube system, and control sensors are located on and/or suspended from the bridge.
The fuel elements are supported by top and bottom grid plates.
The bottom grid plate supports the weight of the fuel element, the pneumatic transfer tube, the central thimble, and the control-rod guide tubes (Figure 4.2).
The top grid plate provides lateral position control only.
The bottom grid plate is attached to the underside of the reflector container.
The reflector is mounted on structural supports that rest on and are attached to the reactor tank bottom.
A recess is provided in the reflector can for the rotary specimen rack.
4.5 Shielding Because the reactor tank is entirely imbedded in earth and concrete below floor level, the only area of personnel access is from the top, in the reactor room.
The usual 4.9 m of water above the core provides more than adequate attenuation of the neutrons and gamma rays between the core and the working area at the top of the pool.
4.6 Reactor Instrumentation The operation of the reactor core is monitored by four separate detector chan-nels.
A fission chamber, two boron-lined compensated ion chambers, and a boron-lined uncompensated ion chamber constitute the reactor core monitoring system.
These detectors monitor the neutron-flux density of the core and provide trip signals to the safety circuits.
The nuclear control instrumentation and process control instrumentation are discussed in Section 7.
4.7 Dynamic Design Evaluation The safe operation of a TRIGA reactor during normal operations is accomplished by the control rods and is monitored accurately by the core power-level detec-tors.
A backup safety feature is the reactor core's inherent large, prompt, negative temperature coefficient of reactivity resulting from an instrinsic molecular characteristic of the U-ZrH all y at elevated temperatures.
Because x
of the large, prompt, negative temperature coefficient, step insertion of excess reactivity resulting in an increasing fuel temperature will be compensated for by the fuel matrix rapidly and automatically.
This will te
'nate the resulting excursion without any dependence on (1) the electronic or nw nanical reactor safety systems or (2) the actions of the reactor operator. Because of the large, prompt, negative temperature coefficient of reactivity, changes of reactivity resulting in a change in fuel temperature during steady-state operation also will be rapidly compensated for by this special fuel mixture, thus limiting the reactor steady-state power level (GA-E-117-833, 1980; Simnad et al., 1976).
Similarly, this inherent characteristic of the U-ZrH fuel has been the basis for designing TRIGA reactors with a pulsing capability as one normal mode of operation, although the OVAMC reactor is not authorized for or provided with the transient rod and instrument systems to implement pulse-mode operation.
Abnormal operations (accidents) are discussed in Section 14.
OVAMC SER 4-4
4.7.1 Excess Reactivity, Experiment Worth, and Shutdown Margin The maximum power excursion transient that could occur wauld be one resulting from the inadvertent rapid insertion of the total available excess reactivity.
The OVAMC TRIGA fuel loading is limited by the licensee's revised Technical Specifications and the SAR to 1.00$ excess reactivity under clean-cold critical conditions. A reactivity transient accident based on this limitation is ana-lyzed in Section 14.
The revised Technical Specifications limit the reactivity worth of individual experiments to 1.00$.
The sum of the absolute worths of all experiments in the reactor also is limited to 1.00$.
Irradiations (which are defined as a sub-class of experiments meeting specific requirements, such as dose rates upon removal from the reactor of no more than 10 mR per hour at 1 ft, encapsulation, and irradiation periods of less than 15 days) are limited to a maximum reactivity worth of 0.25$.
The minimum shutdown margin, as required by the licensee's revised Technical Specifications, is 0.51$ with (1) the highest worth experiment in its most reactive state (2) the reactor in the cold-critical xenon-free condition (3) the highest worth control rod fully withdrawn 4.7.2 Conclusion The staff concludes that the inherent large, prompt, negative temperature coefficient of the U-ZrH fuel m derator provides a basis for safe operation of x
the OVAMC reactor in the steady-state mode.
Furthermore, on the basis of the above information, the staff concludes that (1) the limitation on total excess reactivity of 1.00$, (2) a limitation on total absolute experiment reactivity worth of 1.00$, (3) a limitation of 1.00$ on individual experiments, (4) a limi-tation of 0.25$ on individual irradiations as defined in the Technical Specifi-cations, and (5) operation in compliance with Technical Specification minimum shutdown margin requirements provides assurance that operation of the OVAMC reactor to support the experimental program will pose ne threat to the health and safety of the public.
In addition, the staff concludes that the minimum shutdown margin of 0.51$ under cold-critical xenon-free conditions with the control rod of highest worth fully withdrawn and the experiment of highest worth in its most reactive position is sufficient to ensure that the reactor can be adequately shut down under all likely operational conditions.
4.8 Functional Design of Reactivity Control System The power level in the OVAMC reactor is regulated by the use of three standard control rods, which contain boron as the neutron-absorbing material.
The con-trol rods are driven vertically in or out of the core by electro-mechanical drive mechanisms.
Each control rod drive system is energized f rom the control console through independent electrical cables and circuits, which tends to limit the probability of multiple malfunctions of the drives.
Any or all of the three controls rods will be released through the safety circuitry to fall by gravity into the core on the receipt of a scram signal.
OVAMC SER 4-5
4.8.1 Control Rod Drive The contrel-rod drive mechanisms (Figure 4.3) are located on the bridge at the top of the reactor pool structure and consist of a motor and reduction gear that drive a rack-and pinion system.
Potentiometers provide rod position infor-mation for the shim-safety and regulating rods at the control console.
The control-rod extension tube and dash pot are connected to the rack through an electromagnet and armature.
In the event of electrical power failure or a scram signal, all of the electromagnets are deenergized and the control rods fall into the core by gravity.
The drive motors are nonsynchronous, single phase, and reversible.
Electrical dynamic and static braking on the drive motors are used for fast stops.
Limit switches on the drive assembly indicate the up and down positions of the magnet, the down position of the rod, and armature-magnet contact.
The control-rod drive mechanisms have a stroke of approximately 0.38 m. The maximum rod withdrawal rate is 30 cm per minute, yielding a maximum reactivity insertion rate of about 0.04$ per second.
4.8.2 Scram-Logic Circuitry 4
The scram circuitry ensures that essential reactor core and operational condi-tions are satisifed for reactor operation to occur or continue.
The conditions that must be satisfied are specified in the OVAMC Technical Specifications.
The scram-logic circuitry is a fail-safe system such that any scram signal to or component failure in the scram-logic system will result in the loss of con-trol-rod magnet power, releasing the control rods and causing a reactor shut-down.
Input signals to the scram circuitry are supplied from several process variables and power level sensors that operate independently of each other to ensure redundancy.
(The details of individual sensors are presented in Sec-tion 7.)
The time between activation of the scram-logic system and the full insertion of each control rod is limited to less than 2 sec by the Technical Specifications to ensure adequate safety for the reactor and fuel elements for the range of anticipated operations at OVAMC.
4.8.3 Conclusions The OVAMC reactor is equipped with safety and control systems typical of most nonpower reactors.
Therefore, the staff concludes that there is sufficient redundancy of control rods so that the reactor can be brought to a safe shut-down even if the most reactive control rod fails to insert upon receiving a scram signal.
More than one nuclear instrumentation channel monitors the neutron density (power level), providing redundancy to mitigate consequences of single malfunctions.
In addition to the active electro-mechanical safety control for normal and abnormal operation, the large, prompt, negative temperature coefficient of reactivity inherent in the U-ZrH fuel m derator discussed in Section 4.6 pro-x vides a backup safety feature.
The reactor shutdown mechanism of this fuel limits the steady-state power level and terminates inadvertent transients that
' produce large increases in temperature.
Because this inherent shutdown mecha-nism acts to limit the magnitude of a possible transient accident, it will miti-gate the consequences of such accidents and can be considered to be equivalent to a fail-safe engineered safety feature.
OVAMC SER 4-6
In accordance with the above discussion, the staff concludes that the reactivity control systems of the OVAMC reactor are designed and will function adequately to ensure safe operation and safe shutdown of the reactor under all probable normal and off-normal operating conditions.
4.9 Operational Practices The OVAMC has implemented a thorough preventive maintenance program that is supplemented by a detailed preoperational checklist to ensure that the reactor is not operated at power without the appropriate safsty-related components operable.
The reactor is operated by NRC-licensed personnel in accordance with explicit operating procedures, which include specific responses to any reactor control signal.
All proposed experiments involving use of the reactor are reviewed by the Reactor Safeguards Committee for potential effects on the reactivity of the core or damage to it as well as for possible effects on the health and safety of employees and the general public before installation in the reactor or its experimental facilities.
Similarly, proposed irradiations are reviewed for compliance with the fechnical Specification limitations on irradiations by a health physics qualified licensed senior reactor operator or by a licensed senior reactor operator and a person qualified in health physics before the irradiation can be performed.
4.10 Conclusions On the basis of the information presented abcve, the staff concludes that the OVAMC TRIGA reactor is designed and built according to good industrial practices.
It consists of components representing hundreds of reactor years of operation and includes redundant safety-related systems.
The (caff review of the OVAMC reactor facility has included studying its spe-cif'.c design and installation, its controls and safety instrumentation, its specific preoperational and operating procedures, and its operational limita-tions as identified in the original and revised Technical Specifications and all other pertinent documents associated with the license renewal.
The design features of the reactor are similar to those typical of the research reactors of the TRIGA type operating in many countries of the world, more than 20 of which are licensed by NRC. On the basis of this review of the OVAMC reactor and experience with these other facilities, the staff concludes that there is reasonable assurance that the OVAMC reactor is capable of safe operation, as limited by the revised Technical Specifications, for the period of the license renewal.
OVAMC SER 4-7
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5 REACTOR COOLING SYSTEM The reactor cooling system is shown schematically in Figure 5.1.
The reactor core is maintained in a tank of demineralized water and cooled by natural convection.
The reactor heat is removed from the cooling water by a 5-ton refrigeration unit and discharged to the atmosphere by an air-cooled condenser outside the building.
Cooling water flows from the reactor tank to a water monitor chamber where tem-perature, conductivity, and gamma radioactivity are measured.
The circulating pump takes water from the monitor and discharges it through a refrigerated heat exchanger, the filter, the demineralizer, the flowmeter, and back into the reactor tank.
There is a line bypassing the filter and demineralizer so that coolant can circulate when one or both of these units are being serviced.
There is a skimmer at the pool-surface. When the skimmer is operating, part of the coolant loop flow is through the skimmer, cleaning the pool surface of debris or contamination.
Reactor coolant lost by evaporation from the pool surface is replaced by manu-ally pouring demineralized water directly into the reactor pool.
The cooling water loop takes suction from the reactor pool at a point about 1 m below the pool surface.
Thus, a piping rupture could draw the pool down only 1 m and still leave the reactor core adequately cooled and shielded.
The instrumentation and controls associated with the reactor cooling system are described in detail in Section 7.
The staff concludes that the reactor cooling system is adequate to remove heat from the fuel and prevent melting under all normal and probable off-normal operating conditions.
It is concluded further that, inasmuch as the fuel tem-perature increases less than 1C per hour if the coolant loop is shut down, complex safety systems or interlocks are unnecessary.
There is reasonable assurance that the system can continue to function adequately for the duration of the proposed license renewal.
OVAMC SER 5-1 i
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6 ENGINEERED SAFETY FEATURES The only engineered feature designed to mitigate the consequences of a nuclear accident at the OVAMC reactor facility is the reactor room ventilation system.
The reactor room ventilation system provides heated or cooled 100% outside air to the reactor laboratory at the rate of 4,266 ft3 per minute through six ceil-l ing outlet ducts.
The exhaust effluent of 4,960 ft3 per minute exits the reactor room into the outside air by means of an exhaust fan installed in the outside wall of the building.
In addition, two laboratory fume hoods exhaust a total of approximately 500 fta per minute by means of fans installed on the roof of the hospital building.
Thus, the reactor area is kept at a slight negative pressure with respect to the rest of the hospital.
The reactor area exhaust fan is operated continuously and has a starter switch mounted on the reactor console so that it can be started or stopped manually. The fan is equipped with a gravity-operated damper on the exhaust side, so that the exhaust damper will close when the fan is switched off or its power is interrupted.
In addi-tion, when the fan is stopped, a duct pressure control closes an absolute damper in the air supply duct and simultaneously causes an alarm to be initiated on the hospital Honeywell Delta-2000 control system, which is monitored continually.
Thus, a single switch on the reactor console can stop air from entering or leav-l ing the reactor laboratory and if the exhaust fan stops, the hospital ventila-tion engineers are immediately notified by the Honeywell computer system.
The staff concludes that the reactor ventilation system equipment and procedures are adequate to control the release of airborne radioactive effluents and to minimize releases of airborne radioactivity in the event of abnormal or acci-dent conditions and that the public will be adequately protected from airborne radioactive hazards related to reactor operations.
OVAMC SER 6-1
7 CONTROL AND INSTRUMENTATION SYSTEMS 7.1 Systems Summary The control and instrumentation systems for the OVAMC TRIGA reactor are similar to those used in other research reactors in the United States.
The nuclear fission process is controlled by using three control rods.
The control and instrument systems are interlocked to provide automatic and manual scram capa-bility in case of reactor malfunction and to provide the means for operating the various components in a manne~r consistent with design objectives.
A sche-matic of the OVAMC instrumentation and scram system is shown in Figure 7.1 and the minimum required reactor safety channels, functions, and set points are shown in Table 7.1.
7.2 Control Console The reactor control console contains the control, indicating, and recording instrumentation required for operation of the reactor.
All of the reactor's essential functions are controlled from the console.
On the control panel are (1) rod control switches for raising and lowering the control rods; (2) rod-position indicators to show the position of the shim-safety and regulating rods to within 0.2%; (3) annunciator lights to indicate the up or down position of each rod and rod-magnet contact; (4) linear and log-N power recorders; (5) pe-riod, power level, pool temperature, and log-count rate meters; (6) monitor alarm lights; and (7) additional pilot lights to indicate power on, cooling system on, and startup source strength.
Other annunciator lights on the con-sole indicate the source of a scram signal.
Automatic scram is initiated by (1) an excessive reactor power level as indi-cated by either power level channel, (2) a reactor period less than a preset value, (3) an ion-chamber power supply failure, or (4) an electrical power failure.
Manual scram can be initiated by the operator by means of the console scram button or the magnet current key switch.
The magnet current key switch breaks the rod-magnet circuit so that the console may be operated without rod withdrawal if the switch is off.
Following a scram or the dropping of a rod and after the rod reaches the full-in position, the drive mechanism automat-ically follows the rod down to reestablish contact.
- 7. 3 Control System The control system is composed of both nuclear and process control equipment and is designed for redundant operation in case of failure or malfunction of components essential to the safe operation of the reactor.
7.3.1 Nuclear Control System The nuclear control system consists of the safety, shim-safety, and regulating rods and their associated drive mechanisms.
A detailed discussion of the con-trol mechanism is presented in Section 4.
The logic of the control instrumenta tion includes the following:
OVAMC SER 7-1
(1) The drive mechanisms consist of motors and reduction gears driving racks and pinions.
(2) The control rods are magnetically coupled to the drive shafts and can be scrammed.
(3) The control rods can only be withdrawn singly.
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(4) A rod interlock prohibits withdrawal of the shim-safety or regulating rods unless the safety rod is fully withdrawn.
Similarly, the safety rod can not be withdrawn unless the other two control rods are fully inserted.
(5) The speed of rod withdrawal is limited to about 0.3 m per minute to ensure a conservatively safe rate of reactivity insertion.
The regulating rod may be used for automatic servo-control of reactor power at a constant preset level or for bringing the reactor to power on a preset period.
7.3.2 Process Control Systems The process control systems consist of the circuitry and devices required to energize and deenergize the coolant pump and ventilation system.
7.4 Instrumentation System The instrumentation system is composed of both nuclear and process instrumenta-tior. circuits.
The annunications and/or indications provided by the electronics system on the control console are discussed in Section 7.2.
7.4.1 Nuclear Instrumentation This instrumentation provides the operator with the necessary information for proper manipulation of the nuclear controls (Figure 7.1).
(1) The startup channel comprises a fission-counter chamber, power supply, preamplifier, linear amplifier, and log count-rate meter.
This channel indicates up to 10,000 counts per second and covers the power range from about 10 4W (source level = 10 8W) to about 1 W.
A minimum source-count interlock from this circuit prevents rod withdrawal unless the measured source level exceeds a predetermined value.
(2) The log-N and period channel comprises a compensated ion chamber, power supply, log input circuit, micromicroammeter, recorder, period amplifier and meter.
Log N power is indicated on the recorder over a range of 6-decades up to full licensed power level.
Over this power range, period is indicated on a meter in the range from -40 sec through infinity to
+7 sec.
The period scram is obtained from this circuit, as is derivative i
control for the servoamplifier in the automatic regulating circuit.
(3) Power level and scram channel no. 1 comprises a compensated ion chamber, power supply, micromicroammeter with range switch, and linear recorder.
This channel is a combination information channel and scram channel on the reactor power level.
By means of a range switch, it indicates the power level from source level to full licensed power and will produce a 0VAMC SER 7-2
scram if the power level becomes significantly greater than full scale on any range. The output of this channel also furnishes power-level informa-tion to the automatic regulating channel.
J (4) Power level and scram channel no. 2 comprises an uncompensated ion chamber, power supply, and power-range adjustment control and meter to indicate power level from 0 to 110% of full licensed power.
Scram level on this channel may be adjusted from 20 to 110% of full power.
(5) The automatic regulating channel consists of a servoamplifier that controls the regulating rod and thus keeps the reactor power level constant.
The servoamplifier is activated by an error signal that is governed by the setting of the power-demand control in relation to the actual reactor power level.
Because period information also is employed, the servo amplifier may be used to automatically bring the reactor up to power level, within the limits of the worth of the regulating rod, on a preset period of either 30 or 60 sec.
Automatic changes in power level on these periods are possi-ble. The servo amplifier will allow quick recovery to bring the power level back to within 1% of the original value, even when a step change in reactivity of up to severai tenths of 1% of Ak/k is made.
All neutron-sensing chambers are hermetically sealed in aluminum cans and mounted on the outside of the reflector so that their positions are vertically adjustable in order to change sensitivity.
7.4.2 Process Instrumentation This instrumentation is used for (1) sensing and monitoring parameters asso-ciated with the pool water and (2) radiation monitoring.
(1) The water-radioactivity monitor comprises a gamma-radiation detector and a count-rate-meter circuit that gives both audible and visible alarms if the gamma activity in the pool water reaches a preset value.
(2) The water-conductivity monitor consists of a conductivity probe and Wheat-stone bridge circuit.
Daily measurements of the conductivity are made to ensure that neutron activation of pool water impurities will be small and that chemical corrosion of fuel clsdding is limited.
Experience has shown that the buildup of radioactivity i< neglible if the conductivity does not exceed 5 micromhos per centimeter asaraged over a month.
l (3) The water-temperature monitor consists of a resistance-bulb thermometer that senses the bulk pool temperature.
Temperature indication is provided on the control console.
This system is required to be operational whenever the reactor is in operation.
The reactor is shut down if the temperature exceeds 35 C.
Previous experience and calculations for TRIGA reactors operating at power levels up to 250 kW indicate that with the bulk pool temperature limitation of 35'C, the maximum temperature of OVAMC reactor fuel under steady-state operating conditions will be well below the phase change temperature of 550 C for the ZrH 1 alloy.
(4) The water-level monitor consists of a float-switch and associated circuitry.
This provides both an audible and visual alarm if the water level is less than 3.6 m above the top of the core.
OVAMC SER 7-3
l (5) The area radiation monitor is a calibrated, nonjamming gamma-ray monitor with an audible alarm located in the reactor laboratory.
The monitor is positioned a short distance from the reactor isotope removal tube.
The alarm set point is 2 mR per hour.
(6) The continuous airborne radiation monitor is a calibrated, recording, continuous air monitor located in the reactor laboratory near the top of the reactor.
The monitor can detect both gaseous and particulate radio-activities.
The monitor location is con'sistent with the expected area of maximum airborne activity under both normal and abnormal conditions.
The alarm set point is 0.1% of the 41Ar equivalent concentration listed in Appendix B, Table I, of 10 CFR 20 (2 x 10 8 pCi/ml).
The monitor also contains a charcoal filter to provide the capability of monitoring for airborne radioiodines.
The OVAMC revised Technical Specifications require the radiation monitoring channels to be in operation whenever the reactor is operating.
However, for periods of time required for maintenance, the Technical Specifications permit the replacement of the monitoring device with portable gamma-sensitive instru-ments having their own alarms or that are kept under observation.
Other radia-tion monitoring devices that are not part of the required area and airborne monitoring systems are available.
7.5 Conclusions The control and instrumentation system at the OVAMC reactor facility is well designed and maintained.
The quality of workmanship evident in the installa-tion is acceptable.
All power and instrumentation wiring is protected from physical damage by conduit and/or cable trays.
The specifications of the indi-vidual components are in excess of minimal requirements for the overall system.
Redundancy in the important area of reactor power measurements is ensured by overlapping ranges of the log-N and linear power channels.
The control system is designed so that the reactor is automatically shut down if electrical power is lost.
On the basis of an analysis of the control and instrumentation systems, the staff concludes that both systems comply with the requirements and the perfor-mance objectives of the Technical Specifications and that they are acceptable to ensure safe operation of the facility.
OVAMC SER 7-4
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Table 7.1 Minimum reactor safety channels Safety channel Function Set point Percent power Scram 100% of licensed power Linear power level Scram 100% of full-scale recorder Log-N and period Scram Minimum period of 7 sec Startup Prevents with-Minimum of 2 counts per sec drawal of any control rod Console scram button Scram Manual Ion chamber power Scram Failure of power supply Magnet current key Scram Manual switch Simultaneous manual Prevents with-withdrawal of two drawal rods
- Withdrawal of shim-Prevents with-safety or regulating drawal rod with safety rod not all the way out or seated
- Withdrawal of safety Prevents with-rod with shim-safety drawal or regulating rod not seated
- Alarm when water level Pool level is less than 3.6 m above top of core Bulk pool temperature Meter indi-Reactor shutdown if cation temperature -35 C (manual)
- May be defeated for control rod calibration.
OVAMC SER 7-6
8 ELECTRICAL POWER SYSTEM The electrical power requirements of the reactor facility are supplied by three circuits from the hospital electrical distribution system.
The reactor facility nas no emergency electrical power system except for two battery powered lanterns that activate when the building power fails.
In the j
event of loss of electrical power, the control rods are released to fall into 1
the core by gravity, causing safe shutdown of the reactor.
The staff concludes that the electrical power system is acceptable and that an emergency power system is not necessary to ensure safe shutdown of the reactor.
l l
I l
OVAMC SER 8-1 l
9 AUXILIARY SYSTEMS 9.1 Fuel Handling and Storage Handling of fuel elements is done in the pool water by using long-handled tools.
If a fuel element is to be removed from the pool, it would first be placed in a shielded cask.
If a fuel element were to become damaged, it would be removed from the pool and placed in one of three emergency pits in the reactor room floor.
These pits may be filled with water.
The emergency pits can also be used for storage of undamaged irradiated fuel.
9.2 Fire Protection System Wall mounted fire extinguishers and a fire hose make up the facility fire pro-tection system.
Any fires that cannot be controlled by operating personnel using this equipment will be dealt with by the Omaha, Nebraska, municipal fire department providing general support to the OVAMC.
Personnel of the fire depart-ment are briefed on special hazards, including radiological hazards, that might be encountered in fighting fires at the OVAMC.
9.3 Conclusion The staff concludes that the fuel-handling and storage facilities are appro-priate for the reactor size and use.
The staff also concludes that the fire protection system is adequate for the facility.
OVAMC SER 9-1
10 EXPERIMENTAL PROGRAMS The OVAMC reactor serves as a source of neutrons for special radionuclide pro-duction.
In addition to in pool irradiation capabilities, the experimental facilities include a pneumatic transfer system, a rotary specimen rack, and a central thimble.
The effect of any experiment or sample on excess reactivity is limited by Sections 3.6 and 3.7 of the Technical Specifications.
10.1 Experimental Facilities 10.1.1 Pool Irradiations The open pool of the reactor permits bulk irradiations in the water outside the cylindrical graphite reflector.
The decision to perform experiments in the reactor pool (as opposed to using the pneumatic transfer system or the specimen rack) is dictated by specimen size and the type and radiation source strengths required.
The actual placement of experiments or samples in the core region also may be controlled by their effect on excess reactivity.
10.1.2 Pneumatic Transfer System A 1.25 in. outside diameter pneumatic transfer tube is provided for the rapid transport of samples to and from the region of the reactor core.
The sample holders can be inserted or removed while the reactor is in operation through a constant exhaust system that is vented through a filter to the exhaust duct.
The system has automatic timing controls.
This facility is used principally for the production of isotopes with short half-lives.
The specimens are in-serted into and removed from the pneumatic system in the reactor laboratory, with shielding as necessary.
10.1.3 Rotary Specimen Rack The rotary specimen rack consists basically of an aluminum ring that can be rotated around the core.
Forty evenly spaced aluminum cups are hung from the ring and serve as irradiation specimen holders.
The ring can be rotated manu-ally from the top of the reactor pool so that any one of these cups can be aligned with the single isotope-removal tube that runs up to the top of the reactor.
This tube is used for removing and replacing irradiation specimens.
An indexing and keying device is provided to ensure positive positioning of the cups.
The rotary specimen rack is enclosed completely in a welded aluminum container.
The aluminum ring is located at approximately the level of the top grid plate, with the specimen cups extending from the ring down to about 0.1 m below the top of the active lattice.
In the radial direction, the centers of the cups are about 0.1 m from the inner edge of the graphite reflector assembly.
The container enclosing the rotary specimen rack has been designed to ensure that it will remain watertight.
Furthermore, the system is designed so that flooding this container will decrease the reactivity of the reactor.
CVAMC SER 10-1
10.1.4 Central Thimble A central thimble is provided to permit irradiations or experiments in the region of maximum neutron flux density and maximum statistical weight.
It con-sists of a vertical 3.4-cm-inside-diameter aluminum tube leading from the top of the reactor pool through the center of the reactor core and terminating below the bottom of the core.
The bottom of the tube is capped, but holes drilled in the wall of the tube directly above the upper grid plate ensure that the portion within the fueled region will be filled with water during reactor operation.
The shield water can be removed from the portion of the central thimble above the upper grid plate using air pressure to force the water out of the tube through the holes in the tube wall.
This provides a highly collimated beam of neutron and gamma radiation for experiments.
The radiation dose on the next floor directly above the reactor is 1.5 mR per hour with the thimble, which has 5 cm of lead shielding above it, operating as a beam tube.
Lead bricks (2 in. thick) are stacked around the central thimble before the shield water is removed, and the radiation dose in the reactor room when the thimble is used as a beam tube is <2 mR per hour.
Both of these radiation levels are well within the limits of 10 CFR 20 for restricted access.
The central thimble has been used once for a beam to determine radiation dose levels during such use and has demonstrated that with minimal additional shield (2 to 4 in.
of lead) above the thimble it can be used with no significant hazard to the hospital staff or to the public.
10.2 Experimental Review A Reactor Safeguards Committee (RCS) that reports to the Director of the Omaha Veterans Administration Medical Center provides an independent review of (1) changes in reactor operating procedures, (2) proposed changes in Technical Specifications or other license conditio's, and (3) all experimentation affect-n ing reactor operation.
This committee is composed of individuals collectively having a broad spectrum of expertise in radiation and/or reactor-related tech-nology (see Section 6.5 of the revised Technical Specifications).
The limitations on experiments are specified in Section 3.6 of the Technical Specifications.
Each experimental request involving reactor operation must be approved through submittal of an experiment plan to the RCS.
After RSC's re-view, it must be approved by the Reactor Supervisor.
Irradiations are a sub-class of experiments that fall within Section 3.7 of the Technical Specifica-tions.
Irradiations may be approved by a licensed senior reactor operator with training in health physics or a licensed senior reactor operator and a person with training in health physics.
In addition to ensuring safe and licensed reactor use, the review and approval processes provide for personnel specifically trained in radiological safety and reactor operations to consider and recommend alternative operatior.al conditions (such as different core positions, power levels, or irradiation times) that might decrease personnel exposure and/or the potential release of radioactive materials to the environment.
OVAMC SER 10-2
10.3 Conclusions The staff concludes that the design of the experimental facilities, combined with the detailed review and administrative procedures applied to all research activities, is adequate to ensure that experiments (1) are unlikely to fail, (2) are unlikely to release significant radioactivity to the environment, and (3) are unlikely to cause damage to the reactor systems or its fuel.
There-fore, the staff considers that reasonable provisions have been made su that the experimental programs and use of the experimental facilities do not pose a sig-nificant risk of damage to the reactor or of uncontrolled release of radio-activity to the environment.
l s
OVAMC SER 10-3
11 RADI0 ACTIVE WASTE MANAGEMENT Radioactive waste resulting from OVAMC reactor facility operations is discharged to the environment in dilute gaseous form, released as liquid to the OVAMC sani-tary sewer system or packaged as solids and transferred to an approved disposal site in accordance with applicable regulations.
11.1 Waste Gene.'ation and Handling Procedures 11.1.1 Airoorne Waste The potential radioactive airborne waste includes neutron activated gaseous 41Ar and 18N and dust particulates in the dry experimental facilities.
No fission products escape from the fuel cladding during normal operations.
No measurable amount of 16N is produced at the licensed power level of the OVAMC reactor.
The radioactive airborne waste is produced principally by the neutron irradiation of the pool water and air dissolved in it and of the air and air-borne particulates in the pneumatic transfer system.
The staff has reviewed the licensee's assumptions and computations on the pro-duction rate of 42Ar and agrees that the concentration in the room air will remain well below the value quoted in 10 CFR 20, Appendix B, for unrestricted areas.
Because the average temperature of the pool water does not increase much during operation, the rate of exchange of dissolved gases between the water and the room is not large.
Hence, most of the dissolved 41Ar will beta decay within the water.
Furthermore, because most of the 4 tar produced in the air in the rotary specimen rack will not exchange rapidly with room air, it will decay in situ.
As a result of these factors, it is estimated that the annual airborne release of 41Ar to the unrestricted environment is less than 0.1 Ci.
Exposure of personnel to the small amount of 41Ar generated by the reactor is further limited by a high constant airflow rate through the reactor room and experimental area.
A separate ventilation system is provided for the two radio-chemistry hoods.
During normal operations, the exhaust air from the reactor room is discharged to the outside environment at ground level.
Under emergency conditions, the exhaust fan can be turned off and dampers close off the air inlet to and the air exhaust from the reactor room.
11.1.2 Liquid Waste Many activities associated with the normal research operations that are con-ducted within the reactor facility are capable of generating liquid radioactive waste.
There is essentially no liquid radioactive waste resulting from opera-tion of the reactor.
Most liquid radioactive waste associated with the reactor results from the experiments performed with the reactor.
General OVAMC use of other radionuclides (centrolled under licenses other than the reactor operating OVAMC SER 11-1
license) leads to a major fraction of any liquid waste released to the unre-stricted environment.
All potentially radioactive liquid waste is initially stored to allow radio-active decay and then released to the OVAMC sanitary sewer system after it has been determined that it complies with the limits governing the release of by-products from reactor operations.
If necessary, dilution of the stored waste with the sewer liquid with which it combines is made to ensure that only enough is released to be in compliance with applicable state and federal regulations for release of radioactive materials.
11.1.3 Solid Waste Low-level solid waste generated as a result of reactor operations consists primarily of ion exchange resins, filters, potentially contaminated paper and gloves, and occasional small, activated components.
These items are monitored; if active, they are packaged in accordance with applicable NRC and Department of Transportation (DOT) regulations and are transferred from the OVAMC in accordance with applicable regulations.
11.2 Conclusion The staff concludes that the waste management activities of the OVAMC reactor facility have been conducted and are expected to continue to be conducted in a manner consistent both with 10 CFR 20 and with as-low-as-is-reasonably-achievable (ALARA) principles.
Among other guidance, the staff review has followed the methods of ANSI /ANS 15.11, 1977, " Radiological Control at Research Reactor Facilities."
Because 41Ar is the principal potentially significant radionuclide released by the reactor to the environment during normal operations, the staff has reviewed the history, current practice, and future expectations of operations.
The staff concludes that the potential doses in unrestricted areas as a result of actual releases of 41Ar have never exceeded or even approached the limits specified in 10 CFR 20 when averaged over a year.
Furthermore, the staff's conservative com-putations of the hypothetical doses beyond the limits of the reactor room give reasonable assurance that potential doses to the public as a result of 41Ar would not be significant, even if there were a major change in the operating schedule of the reactor at the current licensed maximum power level.
OVAMC SER 11-2
12 RADIATION PROTECTION PROGRAM The OVAMC reactor facility has a structured radiation safety program with a health physics staff equipped with radiation detection equipment to determine, control, and document occupational radiation exposures at the reactor facility.
The facility monitors airborne effluents to ensure that the releases are in compliance with applicable regulations.
The facility has an environmental monitoring program (1) to verify that radiation expotures in the unrestricted areas around the facility are within regulations and guidelines and (2) to confirm the results of calculations and estimates of environmental effects resulting from the research programs.
12.1 ALARA Commitment The OVAMC administration has formally established the policy that all reactor-related operations are to be conducted in a manner to keep all radiation expo-sures as low as reasonably achievable (ALARA).
This strong policy is imple-mented by a set of specific guidelines and procedures.
All proposed experiments and procedures at the reactor are reviewed for ways to minimize the potential exposures of personnel.
Any unanticipated or unusual reactor-related exposures are investigated by the RSO, the reactor operations staff, and the Reactor Safeguards Committee to develop methods to prevent recurrences.
12.2 Health Physics Program 12.2.1 Health Physics Staffing The normal, full-time health physics staff at the OVAMC reactor consists of one professional and one technician with additional support as needed.
The onsite staff has sufficient training and experience to direct the radiation protection program for this research reactor.
The RSO has been given the responsibility, the authority, and adequate lines of communication to arovide an effective radiation safety program.
The staff has determined that the OVAMC health physics staff is acceptable for the proper support of the research efforts within this facility.
12.2.2 Procedures Detailed written procedures have been prepared that address the health physics activities required for the OVAMC reactor facility.
Copies of these procedures are readily available to all personnel.
12.2.3 Instrumentation The OVAMC reactor facility has a variety of detecting and measuring instruments available for monitoring potentially hazardous ionizing radiation.
The instru-ment calibration procedures and techniques ensure that any credible type of radiation and any significant radiation intensities will be detected promptly and measured correctly, OVAMC SER 12-1
12.2.4 Training All reactor facility peisonnel are given an indoctrination in radiation safety before they assume their work responsibilities.
Additional radiation safety instructions are provided to those who will be working directly with radiation or radioactive materials.
The training program is designed to identify the particular hazards of each specific type of work to be undertaken and methods to mitigate their consequences.
Retraining in radiation safety is provided as well.
As an example, all reactor operators currently are given an examination on health physics practices and procedures at least every year.
The level of any retraining given is determined by the examination results.
12.3 Radiation Sources 12.3.1 Reactor Sources of radiation directly related to reactor operations include radiation from the reactor core, ion exchange columns and filters in the demineralizer system, and radioactive gases (primarily 41Ar).
The reactor U-ZrH alloy fuel is contained in aluminum (or stainless steel) x cladding.
Radiation exposures from the reactor core are reduced to acceptable levels by water and concrete shielding.
The ion exchange resins and the filters are changed routinely while only low levels of radioactive materials have accu-mulated, thereby minimizing personnel exposure.
Personnel exposure to the radiation from chemically inert 41Ar is limited (1) by dilution of this gas in the reactor room, (2) by prompt removal of this gas from the reactor room and experimental areas, and (3) by its discharge to the atmosphere, where it is diluted and diffused further before reaching occupied areas off site. The maximum equilibrium 42Ar concentration in the reactor room would be less than 3 x 10 8 pCi/cc, which is 7% of the guideline concentration given in 10 CFR 20, Appendix B, Table II, Column 1 (for unrestricted areas).
12.3.2 Extraneous Sources Sources of radiation that may be considered as incidental to the normal reactor operation, but are associated with reactor use, include radioactive isotopes produced for research, activated components of experiments, and activated samples or specimens.
Personnel exposure to radiation from intentionally produced radioactive mate-rial as well as from the required manipulation of activated experimental compo-nents is controlled by rigidly developed and reviewed operating procedures that use the standard protective measures of time, distance, and shielding.
12.4 Routine Monitoring 12.4.1 Fixed-Position Monitors The OVAMC reactor facility uses several fixed position radiation monitors.
Area radiation monitors are placed at strategic locations throughout the reactor room.
One of these provides audible and visual alarms immediately adjacent to the reactor.
CVAMC SER 12-2
When the reactor is operating, additional radiation monitors are required by the Technical Specifications. These include a continuous air monitor, located in the reactor room, and a primary coolant water monitor.
e 12.4.2 Experimental Support The health physics staff participates in experiment planning by reviewing all proposed procedures for ways to minimize personnel exposures and limit the generation of radioactive waste.
Approved procedures specify the type and degree of health physics involvement in each activity.
As examples, operating procedures require that changes in experimental setups include a survey by health physics personnel using portable instrumentation, and all items removed from the reactor room must be surveyed.
12.5 Occupational Radiation Exposures 12.5.1 Personnel Monitoring Program The OVAMC reactor facility personnel exposures are peasured by the use of film badges assigned to individuals who might be exposed to radiation.
In addition, non-self-reading pocket ion chambers are used.
Instrument dose rate and time measurements are used to ensure that administrative occupational exposure limits are not exceeded.
These limits are in conformance with the limits specified in 10 CFR 20.
12.5.2 Personnel Exposures The OVAMC reactor facility personnel annual exposure history for the last few years is given in Table 12.1.
l 12.6 Effluent Monitoring 12.6.1 Airborne Effluents As discussed in Section 11, radioactive airborne effluents from the reactor facility consist principally of activated gases.
The airborne radioactivity is monitored to provide prompt indication of any abnormal concentrations being discharged to the environment.
This is accomplished by withdrawing a represen-tative stream from a point near the top of the reactor through a continuous air monitor.
This monitor i so is provided with a charcoal filter for monitoring the presence of iodine radioactivity.
The output of the monitor is indicated on a meter having adjustable alarm set points, and a continuous record also is provided.
12.6.2 Liquid Effluents The reactor generates no radioactive liquid effluents.
Radioactive liquid waste generated in the research program is discussed in Section 11.1.2.
12.7 Environmental Monitoring Radioactive gas is the only potentially radioactive material released to the environment as a result of the routine operation of the OVAMC reactor.
The CVAMC SER 12-3
routine gaseous effluent measurements consist of those recorded by the contin-uous air monitor, and the monthly exposure data obtained from film badges located within the reactor room, at the exhaust port output, and at the water treatment pit output.
The latter represents the airborne exhaust to the envi-ronment because the reactor room air is discharged through the water treatment pit.
The net integrated exposure at the water treatment pit output for 1982 was 40 mrad.
A continuous air sampler was operated by the State of Nebraska Health Depart-ment's Division of Radiological Health on the roof of the Omaha-Douglas County Hospital (300 m from the OVAMC) for a number of years, primarily as a weapons testing fallout monitor.
At no time was any activity detected that could be attributed to the OVAMC reactor operation.
This monitoring program has been discontinued because of funding problems.
12.8 Potential Dose Assessments Natural background radiation levels in the Omaha area result in an exposure of about 80 mrems per year to each individual residing there.
At least an addi-tional 10% (approximately 8 mrems per year) will be received by those living in a brick or masonry structure.
Any medical diagnosis x-ray examination will add to this natural background raniation, increasing the total accumulative annual exposure.
Conservative calculations by the staff based on the maximum possible amount of radioactivity released by the reactor operations predict an exposure to an indi-vidual in the unrestricted areas of much less than 1 mrem per year based on annual 41Ar release rates of 0.1 Ci per year.
The staff considers this release rate to be a reasonable estimate on the basis of its knowledge of the facility and its operating schedule.
12.9 Conclusions The staff has determined that radiation protection receives appropriate support from the OVAMC administration.
The staff concludes that (1) the program is adequately staffed and equipped, (2) the OVAMC reactor health physics staff has adequate authority and lines of communication, (3) the procedures are integrated correctly into the research plans, and (4) surveys verify that operations and procedures achieve ALARA principles.
The staff also coricludes that the effluent monitoring programs and environmental monitoring procedure are adequate to promptly identify significant releases of radioactivity and to predict maximum exposures to individuals in the unrestricted These predicted maximum levels are a small fraction of applicable regula-area.
tions and guidelines specified in 10 CFR 20.
Additionally, the staff concludes that the OVAMC reactor radiation protection program is acceptable because the staff has found no instances of reactor-related exposures of personnel above applicable regulations and no unidentified
'significant releases of radioactivity to the environment.
Furthermore, the staff considers that there is reasonable assurance that the personnel and procedures will continue to protect the health and safety of the public from routine operations.
OVAMC SER 12-4
Table 12.1 Recent exposure history for reactor facility personnel Number of individuals in each range Whole-body exposure (rem / year) 1978 1979 1980 1981 1982 No measurable exposure 2
2 1
4 2
Measurable exposure
<0.1 2
1 3
0 1
0.1 to 0.25 0
1 0
0 0
>0.25 0
0 0
0 0
Total number of individuals monitored 4
4 4
4 3
4 B
OVAMC SER 12-5
13 CONDUCT OF OPERATIONS 13.1 Overall Organization Responsibility for the safe operation of the reactor facility is vested within the chain of command shown in Figure 13.1.
The Reactor Supervisor is delegated responsibility for overall facility operation.
He delegates the succession to this responsibility during his absence.
13.2 Training Most of the training of reactor operators is done by inhouse personnel.
The licensee's Operator Requalification Program has been reviewed, and the staff concludes that it meets the applicable regulations (10 CFR 50.34(b)).
13.3 Emergency Planning 10 CFR 50.54(q) and (r) require that a licensee authorized to possess and/or operate a research reactor shall follow and maintain in effect an emergency plan that meets the requirements of Appendix [ to 10 CFR 50.
At the staff's request, as part of the application for license renewal, the applicant sub-mitted a plan following guidance contained in RG 2.6 (1978 For Comment Issue) and in ANS 15.16 (1978 Draft).
In 1980, new regulations were promulgated, and licensees were advised that revised guidance would be forthcoming.
On May 6, 1982, an amendment to 10 CFR 50.54 was published in the Federal Register (47 FR 19512, May 6, 1982) recommending these guides and establishing new submittal dates for Emergency Plans from all research reactor licensees.
The deadline for submittal from a licensee in the OVAMC class <2 MW was November 3, 1982.
The applicant / licensee transmitted an updated Emergency Plan by letter dated October 21, 1982, thereby complying with existing appli-cable regulations.
13.4 Operational Review and Audits The Reactor Safeguards Committee (RSC) provides independent review and audit of facility activities.
Alternate members may be 40 pointed by the RSC Chairman to serve on a temporary basis.
The RSC or a subcommittee meets at least once per calendar quarter.
The committee must review and approve plans for modifica-tions to the reactor, new experiments, and proposed changes to the license or to procedures.
The committee also is responsible for analyzing and for review-ing audits of reactor facility operations and management and for reporting the results thereof to the OVAMC Hospital Chief of Staff.
13.5 Physical Security Plan The OVAMC reactor facility has established and maintains a program to protect the reactor and its fuel and to ensure its security.
The NRC staff has reviewed the plan and concludes that the plan, as amended, meets the requirements of OVAMC SER 13-1
10 CFR 73.57 for special nuclear material of low strategic significance.
OVAMC's licensed authorization for reactor fuel falls within that category.
Both the Physical Security Plan and the staff's evaluation are withheld from public disclosure under 10 CFR 2.790(d)(1) and 10 CFR 9.5(a)(4).
Amendment 7 to the facility OL R-57, dated February 6,1981, incorporated the Physical Security Plan as a condition of the license.
13.6 Conclusion On the basis of the above discussions, the staff cencludes that the licensee has sufficient experience, management structure, and procedures to provide reasonable assurance that the reactor will be managed in a way that will cause no significant risk to the health and safety of the public.
k I
OVAMC SER 13-2
3 M
P; DIRECTOR OMAllA V.A.
MEDICAL CENTER I
CHIEF 0F STAFF j
ASSOCIATE CHIEF 0F STAFF RESEARCil REACTOR SAFEGUARDS COMMITTEE REVIEW AUDIT FUNCTION FUNCTION RADIOLOGICAL REACTOR
~~~-~~1~~-~j SAFETY OFFICER SUPERVISOR *
-~--
REACTOR OPERATIONS
- Responsible for Facility Operation Figure 13.1 Facility organization
14 ACCIDENT ANALYSIS In establishing the safety of the operation of the OVAMC reactor, the licensee analyzed potential accidents to ensure that these events would not result in potential hazards to the reactor staff or the public.
The licensee's analysis has included the potential effects of natural hazards as well as potential acci-dents involving the operation of the reactor.
In addition, the NRC staff has obtained an independent analyses of accidents with TRIGA-fueled reactors (NUREG/CR-2387), and has analyzed a fuel-handling accident.
Among the potential accidents considered to be credible, the one with the greatest potential effect on the environment and the unrestricted area outside of the OVAMC reactor facility is the loss of the cladding integrity of an irra-diated fuel rod in air in the reactor laboratory.
This has been designated as the design-basis accident and, for purposes of classification, is referred to as the " fuel-handling accident." The licensee and the staff have evaluated other possible accident sequences that originate in the intact reactor core; none pose a significant risk of cladding failure.
However, it is possible that an operator, when removing a fuel element from the core or relocating one pre-viously removed following irradiation, could experience an accident that would break the integrity of the fuel cladding.
If this cladding were ruptured, then noble gases and halogen fission products could escape into the pool.
This will be designated as the design-basis accident (DBA) for a TRIGA reactor.
A DBA is defined as a postulated accident with potential consequences greater than those from any event that can be mechanistically postulated, Thus, the staff assumes that the accident occurs but does not attempt to describe or evaluate deter-ministically the mechanical details of the accident or the probability of its occurrence.
Only the consequences are considered.
The following potential accidents or effects were considered for evaluation and analysis.
(1) natural phenomena (2) rapid insertion of reactivity (nuclear excursion)
(3) less of coolant (4) mechanical rearrangement of fuel (5) effects of fuel aging (6) handling of irradiated fuel 14.1 Natural Phenomena The licensee has considered the potential effects of natural phenomena such as earthquakes and tornadoes on the OVAMC reactor and concluded that the hazards to the reactor are negligible.
The area is seismically stable, characterized by earthquakes of low intensity as well as low frequency.
Examination of seis-mic events since 1867 indicates no significant damage.
Tornadoes are more fre-quent than earthquakes; however, the fact that the reactor is in the basement of the OVAMC hospital building and is surrounded by poured concrete walls with no windows and witn 7 to 10 cm. of concrete overhead makes tornado damage OVANC SER 14-1
l improbable. The staff agrees with the licensee's conclusion that the hazards from natural phenomena are negligible.
14.2 Rapid Insertion of Reactivity (Nuclear Excursion)
The maximum power excursion (transient) that could occur would be one resulting from the inadvertent rapid insertion of the total available excess reactivity.
The OVAMC TRIGA fuel loading is limited by the current license to 1.00$ excess reactivity above clean-cold critical.
The staff believes that this is a reason-able limitation based on operational experience and that it should be included in the licensee's revised Technical Specifications specifically as well as by inference.
Because the Technical Specifications limitation on the reactivity worth of a single experiment is also 1.00$, it is conceivable that a step reac-tivity insertion of 1.00$ can be obtained.
Although neither the licensee nor the staf/ has been able to postulate a credible mechanism that would result in a step insertion that is rapid enough to cause a transient based on prompt neu-trons, it has been assumed for purposes of the analysis that such an event does occur.
The staff notes that the OVAMC reactor is neither authorized to nor equipped for pulse-mode operation.
The General Atomic Company demonstrated by direct experimentation with the proto-type Torrey Pines TRIGA reactor (GA-0531, 1958; GA-0722, 1959) that the insertion of 2.00$ excess reactivity caused no damage to the reactor nor any significant radiation exposure to individuals near the reactor or in the surrounding area.
This reactivity insertion yielded a reactor period of 10 msec and a peak power of approximately 250 MW.
The prompt, negative temperature coefficient limited the total energy release from the transient to a level that caused no fuel dam-age. Within 30 sec after initiation of the transient, the reactor power level had returned to a quasi-equilibrium level of 200 kW.
The maximum fuel tempera-ture was about 360 C, well below the temperature (550 C) at which the fuel with a hydrogen-to-zirconium ratio of 1.0 undergoes a phase transition (GA-4314, 1980).
Extensive further experience with TRIGA reactors has confirmed the inherent safety of the reactors under transient-mode operatior.. Thus, if all of the excess reac-tivity (1.00$) authorized for the OVAMC reactor is inserted rapidly, the result-ant transient would not approach those that have been demonstrated as safe for routine transient mode operation of other similar TRIGA-type reactors.
On the basis of the above considerations, the staff concludes that there is no credible nuclear excursion possible with the OVAMC reactor that could lead to fuel melting or cladding failure resulting from high temperature or high inter-nal gas pressure.
Therefore, there is reasonable assurance that fission product radioactivity will not be released from the fuel to the environment as a result of a reactor transient.
14.3 Loss of Coolant / Shielding Water Because there are a number of floors in the OVAMC hospital building immediately above the reactor that are normally occupied, the loss of the cooling / shielding water could result in potential radiation exposure to occupants of these areas.
In addition, such an accident would result in increases in temperatures of the fuel and cladding.
The licensee's analysis indicates that the loss of water accident can occur by only two mechanisms:
(1) the tank may be pumped dry or (2) a tank failure may allow the water tc drain into the soil.
OVAMC SER 14-2
The tank outlet water line extends only ~1 m below the normal water level.
Therefore, even if the water system is operated carelessly--if, for, example, it is operated when the pump discharge line has been disconnected for repairs--
the tank cannot be accidentally pumped dry.
This can only be done by delib-erate action.
In the unlikely event that it is necessary to drain the tank for repairs, the fuel will first be removed in shielded casks.
Because the recir-culating pump does not have sufficient suction head to drain the tank, another more powerful pump must be installed with its suction line inlet below the core.
Tank failure could possibly be caused by a severe earthquake or major settling of the building foundation.
As noted in Section 14.1, the hazard from earth-quakes is not significant.
At the time of construction of the reactor facility, there was no evidence of foundation failure during the previous 8 years of the building's existence.
Subsequent examination of the reactor tank has shown no evidence of deterioration.
As described in Section 4.3, the reactor tank has five barriers that prevent leakage from the tank.
Two of these barriers are waterproof--the epoxy resin coating and the welded-steel tank. The other three barriers (gunite, reinforced concrete, and the surrounding soil) would present a very high resistance to water leakage.
The core drilling made at the reactor I
location before construction shows the soil to be clayey silt and glacial clay, both of which are essentially impervious to water flow (Abbot, 1956).
Even though the probability of a loss of shielding water is believed to be exceedingly small, the licensee performed a calculation to evaluate the radio-logical hazard associated with this accident.
If the reactor had been operating for a long period of time at 10 kW before instantaneously losing all of the shielding water, the integrated dose that an individual would receive in the first-floor area immediately above the reactor would be about 20 R in the first hour.
Isolation of the area would reduce this potential exposure.
The radia-tion from the unshielded core would be highly collimated, so that if an individ-ual did not expose himself directly to the core, he could work in the immediate vicinity of the tank for several hours.
He could fill the tank with water from a fire hose and view the interior of the tank with a mirror while making the necessary emergency repairs.
To ensure that personnel would be alerted to this accident, a float switch is installed in the reactor tank to actuate an audible alarm located at the con-trol console if the water level falls to within 3.6 m of the top of the core.
There is reasonable assurance that the operator would be able to institute corrective action before the core would be completely uncovered.
Because the water is required for adequate neutron moderation its removal would terminate any significant neutron chain reaction.
However, the residual radio-activity would continue to deposit heat energy within the fuel.
In the loss-of-coolant accident scenario, it is assumed that sufficient water is lost to l
uncover the core and that subsequent heat removal from the fuel is provided only by air convection.
From the data of Table IV, TID-14844 (AEC, 1962), for example, it is estimated that the initial rate of fission product heating in an average OVAMC fuel element would be less than 5 W, if loss of all water and reactor shutdown were simultaneous and instantaneous.
On the basis of this power level, it is further estimated that air circulation would prevent the fuel temperature fron rising more than 100C.
This temperature, superimposed on the operating temperature at 18 kW steady state would be well below any safety linits for the OVAMC fuel.
OVAMC SER 14-3
l On the basis of the above considerations, the licensee concluded that the pos-sibility of loss of coolant / shielding water is extremely unlikely and that consequences would be unlikely to cause damage to the reactor or to result in serious radiation exposure to staff or occupants of the hospital.
The staff concurs with these conclusions and concludes that a rapid loss of coolant from the reactor tank following extended operation at 18 kW would not result in melting of the fuel or cladding or other loss of cladding integrity from other causes.
14.4 Misplaced Experiments This typa of potential accident is one in which an experimental sample or device is inadvertently located in an experimental facility where the irradia-tion conditions could exceed the design specifications.
In that case, the sample might become overheated or, develop pressures that could cause failure of the experiment container.
As discussed in Section 10, all new experiments at OVAMC are reviewed before insertion, and all experiments in the region of the core are isolated from the fuel cladding by at least one barrier--such as the pneumatic transfer tube, the central thimble, or the reflector can.
The staff concludes that the experimental facilities and the procedures for experiment review at the OVAMC reactor facility are adequate to provide reason-able assurance that failure of experiments is not likely, and, even if failure occurred, breaching of the reactor fuel cladding will not occur.
Furthermore, if an experiment should fail and release radioactivity within an experimental facility, there is reasonable assurance that the amount of radioactivity re-leased to the environment would not be more than that from the accident dis-cussed in Section 14.7.
14.5 Mechanical Rearrangement of the Fuel This type of potential accident would involve the failure of some reactor system or could involve an externally originated event which disperses the fuel and, in so doing, breaches the cladding of one or more fuel elements.
The staff has not developed scenarios for accidents such as these.
Thus there is no logical basis for deciding if any arbitrary scenario is credible. Instead, a later sec-tion of this chapter discusses a scenario assuming the failure of the cladding of an element after extended reautor operation and evaluates potential doses resulting from various hypothetical scenarios for release of the inventory of radioactivity.
This approach should address the spectrum of fuel cladding failures.
The staff concludes that no mechanical rearrangement that is credible would lead to an accident with more severe consequences than those accidents con-sidered in Section 14.2 or 14.7.
14.6 Effects of Fuel Aging The staff has included this process in this section so all credible mechanisms related to loss of integrity of fuei cladding are addressed.
However, as dis-cussed in more detail in Section 17, fuel aging should be considered normal with use of the reactor and is expected to occur gradually.
The reactions ex-ternal to the cladding that might occur are addressed in Section 17.
In this section the possibility of internal reactions is discussed.
There is scme OVAMC SER 14-4
evidence that the U-ZrH fuel tends to fragment with use, probably as a result x
of the stresses caused by high temperature gradients and high rate of heating during pulsing (GA-4314, 1980).
Some of the possible consequences of fragmenta-tion are (1) a decrease in thermal conductivity across cracks, leading to higher central fuel temperatures during steady-state operation (temperature distributions during pulsing would not be affected significantly by changes in conductivity because a pulse is completed before significant heat redistribu-tion by conduction occurs), and (2) more fission products would be released into the cracks in the fuel.
With regard to the first item above, hot cell examination of thermally stressed hydride fuel bodies have shown relatively widely spaced radial cracks that would cause minimal interference with radial heat flow (GA-4314, 1980).
- However, after pulsing, TRIGA-type reactors have exhibited an increase in both steady-state fuel temperatures and power reactivity coefficients.
At power levels of 500 kW, temperatures have increased by approximately 20C and power reactivity coefficients by approximately 20% (GA-5400, 1965).
General Atomic has attrib-uted these changes to an increased gap between the fuel material and cladding caused by rapid fuel expansion during pulse heating, which reduces the heat transfer coefficient.
Experience has shown that the observed changes occur mostly during the first several pulses and have essentially saturated after 100 pulses.
Because these effects are small and have been observed in many TRIGA-type reactors operated at pulses up to 5.00$ and power levels as high as 1.5 MW and because the OVAMC reactor is not operated in the pulse mode, they are not considered to pose any hazard during continued operation of the OVAMC reactor.
Two mechanisms for fission product release from TRIGA fuel meat have been pro-posed (GA-4314, 1980; GA-8597, 1968).
The first mechanism is fission fragment recoil into gaps within the fuel cladding.
This effect predominates up to about 400 C and is independent of fual temperature.
OVAMC operating fuel temperatures have never exceeded 400 C; thus, this will be the main effect.
General Atomic has postulated that in a closed system such as exists in a TRIGA fuel element, fragmentation of the fuel material within the cladding will not cause an in-crease in the fission product release fraction (GA-8597, 1968).
The reason for this is that the total free volume available for fission products remains con-stant within the confines of the cladding.
Under these conditions, the forma-tion of a new gap or widening of an existing gap must cause a corresponding narrowing of an existing gap at some other location.
Such a narrowing allows more fission fragments to traverse the gap and become embedded in the fuel or cladding material on the other side.
In a closed system in which the density of the fuel meat is constant, the average gap size and therefore the fission product release rate remains constant, independent of the degree to which fuel material is broken up.
Above approximately 400 C, the controlling mechanism for fission product release is diffusion, and the amount accumulated in the gap is dependent on fuel temper-ature and fuel surface-to-volume ratios.
In the OVAMC fuel this mechanism is not signifir ant bec.luse of the low fuel temperature and low utilization factor.
OVAMC SER 14-5
The staff concludes that the likely processes of aging of the U-ZrH I"*I
- d' x
erator under low power, steady-state, nonpulsing operation would not cause sig-nificant changes in the operating temperature of the fuel or affect the accumu-lation of gaseous fission products within the cladding.
Therefore, the staff also concludes that there is reasonable assurance that fuel aging will not significantly increase the likelihood of fuel-cladding failure, or the quantity of gaseous fission products available for release in the event of loss of cladding integrity.
14.7 Handling Irradiated Fuel This potential accident (the design-basis accident) includes various incidents to one or more fuel elements, with the reactor shut down, in which the fuel cladding might be breached or ruptured.
The staff's worst case scenario assumes that an accident occurs after a long run at full licensed power so that the inventories of all radionuclides of significance in the scenario are at their maximum (saturation) values.
Also to remain general, the staff did not try to develop a detailed scenario, but simply assumed that the cladding of one fuel element certainly fails and that all of the fission products accumulated in the gap are released abruptly.
Several series of experiments at General Atomic Company have obtained data on the species and fractions of fission products released from U-ZrH under various x
conditions (Simnad et al., 1976; Foushee and Peters, 1971; Baldwin, Foushee, and Greenwood, 1980).
The noble gases were the principal species found to be released, and, when the fuel specimen was irradiated at temperatures below about 350 C, the fraction released could be summarized as a constant equal to 1.5 x 10 s. The species released did not appear to depend on the temperature of irradiation, but tne fraction released increased significantly at much higher temperatures.
General Atomic has proposed a theory describing the release mechanisms in the two temperature regimes that appears plausible, although not all data agree in detail.
It seems reasonable to accept the interpretation of the low tempera-ture results, which implies that the fraction released for a typical TRIGA fuel element will be a constant, independent of operating history or details of operating temperatures, and will apply to fuel whose temperature is not raised above approximately 400 C for any appreciable time.
This means that the 1.5 x 10 5 could be reasonably applied to TRIGA-type reactors operating up to at least 800 kW steady-state and is currently applicable to the OVAMC 18 kW steady-state reactor.
Because the noble gases do not condense or combine chemically, it is assumed that any released from the cladding will diffuse in the air until their radioactive decay. On the other hand, the iodines are chemically active, and are not volatile below about 180 C.
Therefore, some of the radioiodines will be trapped by materials with which they come in contact, such as water, and structures.
In fact, evidence indicates that most of these iodines will eithee not become or not remain airborne under many accident scenarios applicable to nonpower reactors (RG 3.35).
However, to be certain that the fuel-cladding-failure scenarios discussed below led to upper limit dose estimates for all events, the staff assumed that 100% of the iodines in OVAMC SER 14-6
the gap become airborne. This assumption will lead to computed thyroid doses which may be at least a factor of 100 too high in son.e cases.
The staff has reviewed the various acceptable methods for computing the dose within and beyond the confines of the OVAMC reactor facility in case of a fis-sion product release. Although the methods outlined in RGs 1.3, 1.4, and 3.35 give results that are very conservative for nonpower reactors the dose calcu-lation and ground-level atmospheric diffusion models given in these regulatory guides were used in the calculations.
The radionuclides and their associated average beta and gamma energies per disintegration were taken from Table 1 of RG 3.35.
Radionuclide inventories were calculated using the method of TID-14844 (U.S. AEC Report TID-14844, 1962) and the fission product yields were taken from General Electric Company's report (NE00-121541, 1974).
Two scenarios were considered by the staff and are discussed in the following sections.
14.7.1 No Environmental Release A single-fuel-element cladding failure in air occurs immediately after an extended reactor operation at 18 kW.
The analysis is based on the following assumptions:
(1) All radionuclides in the gap, including the iodines, are released into the air (release fraction = 1.5 x 10 5).
(2) The OVAMC reactor laboratory and radiochemical hood exhaust fans are shut down, and the damper in the air supply system is closed, so that the air-borne radioactivity is confined to the reactor room.
(3) The radionuclides released are uniformly instantaneously distributed throughout the reactor laboratory volume of 7.3 x 108 cm,s (4) Ten minutes are required for the reactor staff to exit the reactor laboratory.
(5) The iodine dose conversion factors and breathing rate factors are based on the recommendations in the International Committee on Radiological Protection (ICRP), Committee II report (1959).
(6) A core array of 55 fuel elements with the failed element leading the average core power per element by a factor of 2 (0.67 kW).
I (7) No decrease in source strength resulting from radioactive decay occurs during the exposure period.
The resulting concentrations of airborne radionuclides are of the same order of magnitude as the corresponding MPCs for restricted areas given in 10 CFR 20, Appendix B.
The results of the calculations are given in Table 14.1.
OVAPC SER 14-7
l 14.7.2 Environmental Release In this accident it is assumed that the same initial loss-of-cladding-integrity event occurs but that the reactor laboratory air exhaust fan does not shut down and the air supply system damper does not close.
The radiochemical hood fans are assumed to close.
Dilution of the airborne radioactivity resulting from intake air is ignored; thus the time to totally exhaust the reactor laboratory 6
3 air is taken as 5 min based on an exhaust rate of 2.34 x 10 cm per second.
A ground-level release with the most exposed individual 10 m from the release point is assumed.
Horizontal and vertical diffusion coefficients were estimated from the curves in Figures 3.2 and 3.3 of Turner's publication (1970), assuming the atmospheric conditions recommended in RG 3.35.
These extrapolations are not very reliable, but this is offset by the conservative nature of the remain-der of the assumptions used in the calculation.
The model and atmospheric con-ditions recommendec in RG 3.35 for an exposure time ranging from 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> were used in the calculations.
The building wake effect was also neglected.
However, including this ef fect would further reduce the computed dose.
Other assumptions for this scenario are similar to those given in the previous case.
The results of this computation are presented in Table 14.2.
14.7.3 Conclusion In accordance with the discussions and analysis above the staff concludes that if one fuel element from the OVAMC reactor were to release to the room air all the noble gaseous and iodine fission products accumulated in the fuel-cladding gap, radiation doses to both occupational personnel and to the public in unre-stricted areas would be far below the guidelines and limits of 10 CFR 20.
This conclusion is valid even for the very unlikely accident scenario selected, namely, that the clad failure occurs immediately after an extended full power operation, that all of the gap radioactivity, including all iodines, escapes to the environment in 5 min, and the exposed individual in the unrestricted area is 10 m from the release point.
The staff further notes that the assumptions and methods used in the calculation, having been developed for power reactors, are very conservative for research reactors.
The staff has compared the results presented here with estimates based on finite sized plumes of airborne radioactivity, in order to confirm the conservative nature of these results (Lahti, 1981; NUREG-0851).
14.8 Conclusions On the basis of the above evaluation, the staff concludes that the credible possible accidents involving the OVAMC reactor do not pose significant hazards to the OVAMC staff or patients, the public, or the environment.
l OVAMC SER 14-8
Table 14.1 Radiation doses within the reactor laboratory:
no environmental release Thyroid
- Beta dose Gamma dose dose commit-Element (mrem)
(mrem) ment (rem)
Kr 0.2 0.3 Xe 0.3 0.2 I
0.2 0.8 0.2 Total 0.7 1.3 0.2
" Thyroid dose from krypton and xenon are negligible.
Table 14.2 Radiation exposures in the plume from the design-basis accident Thyroid
- Beta dose Gamma dose dose commit-Element (mrad)
(mrad) ment (rem)
Kr 0.40 7.2 Xe 0.71 0.58 I
0.46
- 1. 8 0.45 Total 1.6 9.6 0.5
l l
l DVAMC SER 14-9
15 TECHNICAL SPECIFICATIONS The licensee's Technical Specifications evaluated in this licensing action define certain features, characteristics, and conditions governing the con-tinued operation of this facility.
These Technical Specifications are explic-itly included in the renewal license as Appendix A.
Formats and contents acceptable to the NRC have been used in the development of these Technical Specifications, and the staff has reviewed them using the Standard ANSI /ANS 15.1-1982 as a guide.
On the basis of its review, the staff concludes that normal reactor operation within the limits of the Technical Specifications will not result in offsite radiation exposures in excess of 10 CFR 20 limits.
Furthermore, the limiting conditions for operation, surveillance requirements, and engineered safety fea-tures will limit the likelihood of malfunctions and mitigate the consequences to the public of offnormal or accident events.
OVAMC SER 15-1
16 FINANCIAL QUALIFICATIONS The Omaha Veterans Administration Medical Center is part of the Veterans Administration which is under the Executive Branch of the Federal Government.
It operates on an annual allocation of funds governed by the Bureau of the Budget.
The OVAMC TRIGA-type reactor is operated by the OVAMC.
Therefore, the staff concludes the funds will be made available as necessary to support continued operations, and eventually to shut down the facility and maintain it in a con-dition that would constitute no risk to the public.
The licensee's financial status was reviewed and found to be acceptable in accordance with the require-ments of 10 CFR 50.33(f).
OVAMC SER 16-1
17 OTHER LICENSE CONSIDERATIONS 17.1 Prior Reactor Utilization Previous sections of this SER concluded that normal operation of the reactor causes insignificant risk of radiation exposure to the public and that only an offnormal or accident event could cause some exposure.
Even a design-basis accident (defined as one worse than can be mechanistically justified) would not lead to a dose to the most exposed individual greater than applicable guidelines or regulations (10 CFR 20).
In this section, the staff reviews the impact of prior operation of the facility on the risk of radiation exposure to the public.
The two parameters involved are the likelihood of an accident and the consequences if an accident occurred.
Because the staff has concluded that the reactor was initially designed and constructed to be inherently safe, with additional engineered safety features, the staff must also consider whether operation will cause significant degrada-tion in these features.
Furthermore, because loss of integrity of fuel cladding is the design-basis accident, the staff must consider mechanisms which could increase the likelihood of tailure.
Possible mechanisms are:
(1) radiation degradation of cladding strength, (2) high internal pressure caused by high temperature leading to exceeding the elastic limits of the cladding, (3) cor-rosion or erosion of the cladding leading to thinning or other weakening, (4) mechanical damage as a result of handling or experimental use, and (5) degrada-tion of safety components or systems.
The staff's conclusions regarding these parameters, in the order in which they were identified above, are (1) The aluminum-clad low-hydride TRIGA fuel in the core has been in use since 1959 and has been subjected to less than 1% burnup of 2ssU.
TRIGA fuel at more extensively used reactors has been in use for many times as much burnup, with no observable degradation of cladding as a result of radiation.
(2) Because the OVAMC reactor operates at a maximum power level of 18 kW, the temperature of the fuel does not exceed 100*C during normal operation.
At this temperature, the pressure of the air and/or hydrogen within the cladding does not increase significantly.
(3) Water flow through the core is obtained by natural thermal convection, so the staff concludes that erosion effects as a result of high flow velocity will be negligible.
High primary water purity is maintained by continuous passage through the filter and demineralizer system. With conductivity below about 5 pmho-cm 1, corrosion of the aluminum cladding is expected to be negligible, even over a total 40 year period.
(4) The fuel is handled as infrequently as possible, consistent with periodic surveillance.
Any indications of possible damage or degradation are OVAMC SER 17-1
i l
investigated immediately.
The only experiments which are placed near the core are isolated from the fuel cladding by a water gap and at least one metal barrier, such as the pneumatic tubes or the central thimble.
There-fore, the staff concludes that loss of integrity of cladding through damage does not constitute a significant risk to the public.
(5) OVAMC performs regular preventive and corrective maintenance and replaces components as necessary.
Nevertheless, there have been some malfunctions of equipment.
However, the staff review indicates that most of these mal-functions have been random one-of-a-kind incidents, typical of even good quality electromechanical instrumentation.
There is no indication of sig-nificant degradation of the instrumentation, and the staff further con-cludes that the preventive maintenance program would lead to adequate identification and replacement before significant degradation occurred.
Therefore, the staff concludes that there has been no apparent significant degradation of safety equipment and, because there is strong evidence that any future degradation will lead to prompt remedial action by OVAMC per-sonnel, there is reasonable assurance that there will be no significant increase in the likelihood'of occurrence of a reactor accident as a result of component malfunction.
The second aspect of risk to the public involves the consequences of an acci-dent.
Because the OVAMC reactor has not and is not expected to operate on the maximum available schedule, the inventory of radioactive fission products will be far below that postulated in the evaluation of the design-basis accident both by the applicant and the NRC staff (see Section 14).
Therefore, the staff concludes (1) that the risk of radiation exposure to the public has been accept-able and well within all applicable regulations and guidelines during the his-tory of the reactor, and (2) that there is reasonable assurance that there will be no increase in that risk in any discernible way during this renewal period.
17.2 Multiple or Sequential Failures of Safety Components Of the many accide t scenarios hypothesized for the OVAMC reactor, none produce consequences more st ere than the design-basis accidents reviewed and evaluated in Section 14.
The only multiple-mode failure of more severe consequences would be failure of the cladding of more than one fuel element.
No credible scenario constructed by the staff has included a mechanism by which the failure of integ-rity of one fuel element can cause or lead to the failure of additional elements.
Therefore, if more than one cladding should fail, the failures would either be random, or a result of the same primary event.
Additionally, the reactor con-tains redundant safety-related measuring channels and control rods. Failure of all but one control rod and all but one safety channel would not prevent reactor shutdown to a safe condition.
The staff review has revealed no mechanism by which failure or malfunction of one of these safety-related components could lead to a nonsafe failure of a second component.
OVAMC SER 17-2
18 CONCLUSIONS Based on its evaluation of the application as set forth above, the staff has determined that (1) The application for renewal of Operating License R-57 for its research reactor filed by the Omaha Veterans Administration Medical Center dated May 10, 1979, as amended, complies with the requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regula-tions set forth in 10 CFR, Chapter 1.
(2) The facility will operate in conformity with the application as amended; the provisions of the Act, and the rules and regulations of the Commission.
(3) There is reasonable assurance (a) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public; and (b) that such activities will be conducted in compliance with the regulations of the Commission set forth in 10 CFR, Chapter 1.
(4) The licensee is technically and financially qualified to engage in the activities authorized by the license in accordance with the regulations of the Commission set forth in 10 CFR Chapter 1.
(5) The renewal of this license will not be inimical to the common defense and security or to the health and safety of the public.
OVAMC SER 18-1
19 REFERENCES Abbot, R. U., American Civil Engineering Practice, John Wiley and Sons, Inc.,
New York, 1956.
Algermissen, S.
T., " Seismic Risk Studies in the United States," in Proceedings of the Fourth World Conference of Earthquake Engineering, Vol.1, pp 14-27, Santiago, Chile,1969.
Baldwin, N.
L., F. C. Foushee, and J. S. Greenwood, " Fission Product Release from TRIGA-LEU Reactor Fuels," Seventh Biennial U.S. TRIGA Users Conference, San Diego, California, 1980.
Code of Federal Regulations, Title 10, " Energy," U.S. Government Printing Office, Washington, D.C.
Foushee, F. C., and R. H. Peters, " Summary of TRIGA Fuel Fission Product Release Experiments," Gulf-EES-A10801, San Diego, California, Sept. 1971.
General Atcmic Company GA-531, R. S. Stone and H. P. Sleeper, Jr., "TRIGA Transient Experiments, Interim Report," Sept. 19, 1958.
--, GA-722, " Hazards Report for Torrey Pines TRIGA Reactor," May 27, 1959.
--, GA-4314, M. T. Simnad, "The U-ZrH Alloy:
Its Properties and Use in x
TRIGA Fuel," E-117-833 (1974), Feb. 1980.
--, GA-5400, "Thermionic Research TRIGA Reactor Description and Analysis,"
Rev. C, Nov. 1, 1965, transmitted by letter dated Feb. 28, 1966 (Docket No. 50-227).
--, GA-6596, J. R. Shoptaugh, Jr., " Simulated Loss-of-Coolant Accident for TRIGA Reactors (transmitted by letter dated September 22, 1970, Docket No. 50-227-), Aug. 1965.
--, GA-8597, F. C. Foushee, " Release of Rare Gas Fission Products from U-ZrHx l
Fuel Material," Mar. 1968.
I General Electric Company, NED0-121541, M. E. Mak and B. F. Rider, " Compilation of Fission Product Yields Vallecitos Nuclear Center 1974," Jan. 26,1974.
f International Committee on Radiological Protection, " Permissible Dose for Internal Radiation," ICRP Publication 2, Report of Committee II, Pergamon Press, 1959.
Lahti, G. P., et al., " Assessment of Gamma-Ray Exposures Due to Finite Plumes,"
Health Physics, 41:319, 1981.
OVAMC SER 19-1
Simnad, M. T., F. C. Foushee, and G. B. West, " Fuel Elements for Pulsed TRIGA Research Reactors," Nuclear Technology, 28:31-56, 1976.
Turner, D. B., " Workbook of Atmospheric Dispersion Estimates," Public Health Service Publication No. 999-AP-26, Rev., 1970.
Thom, H. C. S., " Tornado Probabilities," Monthly Weather Review, 91:730-763, 1963.
U.S. Atomic Energy Commission report TID-14844, J. J. DiNunno, F. D. Anderson, R. E. Baker, and R. L. Waterfield, " Calculation of Distance Factors for Power and Test Reactor Sites," Mar. 23, 1962.
U.S. Nuclear Regulatory Commission, NUREG-0851, " Nomograms for Evaluation of Doses from Finite Noble Gas Clouds," Jan. 1983.
--, NUREG/CR-2387, S. C. Hawley et al., " Generic Credible Accident Analysis for TRIGA Fueled Reactors," Battelle Pacific Northwest Laboratories, Apr.
1982.
--, RG 2.6, " Emergency Planning for Research Reactors," For Comment Issue, 1978, Rev. 1, Mar. 1982.
--, RG 3.35, " Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear Criticality in a Plutonium Processing and Fuel Fabrication Plant," Rev., July 1979.
INDUSTRY CODES AND STANDARDS American National Standards Institute /American Nuclear Society (ANSI /ANS) 15 series.
American Nuclear Society ANSI /ANS 15.1, " Standard for the Development of Technical Specification for Research Reactors," 1982.
--,15.11, " Radiological Control at Research Reactor Facilities," 1977.
American Nuclear Society (ANS) 15.16, " Standard for Emergency Planning for Research Reactors, For Comment Issue 1978, Draft 2, Nov. 29, 1981.
OVAMC SER 19-2
'0" til au U.S. NUCLEAR [EGULATORY COMMISSION BIBLIOGRAPHIC DATA SHEET NUREG-0988 hasty"Tv"aiuaYiMEf5'E"RNa#@d'Io the Renewa1 of the
- 2. Item ei.n=>
Operating License for the Research Reactor at the Omaha Veterans Administration Medical Center
- 3. RECIPIENT'S ACCESSION NO.
- 7. AUTHOR (S)
- 5. DAIE REPORT COMPLETED M ON TH l YEAR July 1983
- 9. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS //nclude lip Codel DATE REPORT ISSUED Division of Licensing uOyrs l YEAR Office of Nuclear Reactor Regulation July 1983 U. S. Nuclear Regulatory Commission s (t,m 3,,,, ;
Washington, D.C.
20555
- 8. (Leave blank}
- 12. SPONSORING ORGANIZATION N AME AND MAILING ADDRESS (include 2,p Codel p
Same as 9. above
- 11. FIN No.
- 13. TYPE OF REPORT PE RIOD COVE RE D //nclusive dates)
Safety Evaluation Report
- 15. SUPPLEMEN TARY NOTES 14 (teave n/ank)
- 16. ABSTR ACT 000 words or less)
This Safety Evaluation Report for the application filed by the Omaha Veterans Administration Medical Center (0VAMC) for a renewal of operating license number R-57 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission. The facility is owned and operated by the Veterans Administration and is located in its Medical Center in the city of Omaha, Nebraska. The staff concludes that the TRIGA reactor facility can continue to be operated by OVAMC without endangering the health and safety of the public.
- 17. KEY WORDS AND DOCUMENT AN ALYSIS 17a DE SC HIP T O RS Non Power Reactor Omaha Veterans Administration Medical Center TRIGA License Renewal 17b. IDENTIFIE RS oPEN-EN DE D TE RYS 18, AVAILADlLITY STATEMENT 19 SE CURITY CL ASS ITh:s report) 21 NO OF PAGES
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