ML20216H781
| ML20216H781 | |
| Person / Time | |
|---|---|
| Site: | U.S. Geological Survey |
| Issue date: | 03/16/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216H772 | List: |
| References | |
| NUDOCS 9803230176 | |
| Download: ML20216H781 (7) | |
Text
u s W Ez p t UNITED STATES g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20665-0001
- s e,,s *f SAFETY EVALUATION BY THE OFFICE OF NULuEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 8 TO FACILITY LICENSE NO. R-113 UNITED STATES GEOLOGICAL SURVEY I
DOCKET NO. 50-274 )
I
1.0 INTRODUCTION
By letter dated May 5,1993, as supplemented by letters dated February 23 and August 17,1995, and July 16 and November 4,1997, the United States Geological Survey (USGS or the licensee) applied for an arnendment to the operating license for its TRIGA research reactor located in Denver, Colorado. The licensee requested that the technical specifications of the license be changed (1) to authorize the use of standard TRIGA fuel elements containing up to 12 weight percent (w%) uranium, in addition to continuing to authorize use of the currently approved elements containing 8.5 w% uranium; (2) to add a limiting condition for operation that would forbid reactor operation if the calculated steady-state thermal power produced in any fuel element in the reactor would exceed 22 kilowatts; and (3) to specify conditions and intervals for recalculation of the thermal power per fuel element.
2.0 EVALUATION l
2.1 Introduction l The USGS facility is a MARK I TRIGA reactor licensed to operate at steady-state thermal l ,
power levels up to 1 megawatt, and in the pulse mode with reactivity insertions up to
- 2.1% Ak/k. The reactor is currently authorized to use standard stainless steel clad
( uranium-zirconium hydride fuel containing 8.5 w% low-enriched uranium. The USGS
- reactor was initially licensed in 1967 and operates on the grounds of the Denver Federal SE Center, which includes other Federal research and administrative functions not related to operation or utilization of the reactor. Since initiallicensing of the USGS facility, the :
o reactor vendor, General Atomics, has developed and obtained approval of the U.S. Nuclear O Regulatory Commission (NRC) to test and use TRIGA fuel with uranium ioadings up to g 30 w%. The new higher loaded fuels contain bumable poisons so that fuellifetimes are much longer, but the neutronics characteristics for steacy-state operation are not gg n< significantly different from the original fuel. However, the higher loaded fuels may exhibit N different power density and pulsing characteristics. From among the spectrum of low-
@$ - enriched TRIGA-type fuels available, the USGS has requested authorization to use the N current 8.5 w% up to 12 w% fuel. The submittal from the USGS provides up-to-date analyses and discussions of the thermal properties and safety considerations related to its proposed reactor operation with mixtures of fuelloadings. The 12 w% TRIGA-type fuel
l l
! 2 l
i j was used extensively in the annular core pulse reactor at Sandia National Laboratories, and the NRC has previously evaluated and licensed a TRIGA reactor at Pennsylvania State University similar to the USGS TRIGA reactor to use mixtures of these two fuel types.
2.2 Steady-State Operation if all of the 8.5 w% fuel elements (rods) currently in the reactor were replaced at one time by fresh 12 w% elements, the new core would require fewer elements, and hence a smaller core. If so, power densities in the fuel elements would be correspondingly increased, but there would be no other significant changes in the thermal power distributions or the temperature distribution. Therefore, previous thermal-hydraulic analyses of a TRIGA reactor with one type fuel would be applicable. However, the USGS )
application would allow adding one new fuel element at a time, and if a single fresh 12 w% )
element were to be inserted into a fuel position in the existing 8.5 w% core, thermal power j and temperature in that element could peak 30 percent to 45 percent above its near neighbors. Because the proposed license amen'iment would allow any possible mixtures of new and old fuel elements, limiting possible configurations were analyzed by the licensee.
l These analyses have been reviewed, evaluated, and found acceptable by the staff.
General Atomics (GA 9064) has calculated that departure from nucleate boiling would not ,
occur if the thermal power developed in a fuel element did not exceed 43 kW in a typical l 100-element cylindrical core with inlet coolant at 33*C under natural thermal convection.
These analyses also indicated that the axial power peaking was primarily dependent on fuel rod lengths and, therefore, should be the same for all standard-length TRIGA fuel elements.
On the basis of this information and the related fuel temperatures, the USGS proposed that )
operation of its reactor should be limited so that the integral power in any element would L not exceed 22 kW. In the justification analyses, the licensee assumed inlet coolant at 25'C, and included the positive probable error of tile calculations within the 22 kW.
Operation within these limits would ensure that the departure from nucleate boiling ratio at the hottest po;nt in the reactor fuel would not fall below two, an acceptably safe condition.
l Thermal power developed in a single fuel rod is not easily measurable in a research reactor, !
! so USGS proposed that 22 kW calculated thermal power per rod be added to the technical specifications as a limiting condition for operation. Because temperature has been accepted as the limiting safety parameter for TRIGA fuel, the licensee also showed that the 22 kW power limit would ensure that the hottest point in the fuel would not reach the accepted safety limit (1150*C) for allowed operating conditions, including any arbitrary mixture of 8.5 w% and 12 w% fuel elements.
To perform these calculations, the licensee proposed using a Monte Carlo computer program, Monte Carlo N-Particle (MCNP), that was developed and tested at a Department of Energy nationallaboratory and has recently been benchmarked against measurements at i TRIGA facilities. Calculations of power per element in the USGS reactor were performed and presented by the licensea to help justify the use of the MCNP program. The licensee's calculations showed that if the core sizo was much below 100 fuel elements, the maximum power in at least one element could exceed 22 kW at 1 MW reactor power. Therefore, the licensee also proposed adding a technical specification that would require power calculations if reactor operations with power above 100 kW were planned with a core size
~
3 of less than 100 elements. The staff has determined that these considerations give reasonable confidence that the power levels can be calculated with acceptable validity and that steady-state reactor operation within the revised technical specifications will not lead to loss of fuel integrity and is therefore acceptable.
2.3 Pulsing Operations The licensee plans to continue operating the reactor in the pulse mode but proposes no changes in the operating conditions or applicable technical specifications directly associated with pulsing. The 12 w% fuel elements, with their increased absorption of thermal neutrons, will lead to a decrease of both the neutron lifetime and the temperature coefficient of reactivity. Changes in these parameters can change pulse characteristics such as fuel temperature and pulse size for a fixed reactivity insertion in a 12 w% core compared to an 8.5 w% core. However, most pulse parameters are a function of averages or integrals over the entire core, so the two parameters noted above would vary slowly as the fuelloading was changed from mostly 8.5 w% fuel to mostly 12 w% fuel. However, because of the power and temperature peaking if a small number of 12 w% elements were loaded into an otherwise 8.5 w% core, the licensee showed that the predicted maximum temperature during any authorized pulse, though somewhat higher than with all 8.5 w% l fuel, still would not reach within several hundred degrees of the accepted safety limit of 1150"C for pulsing TRIGA fuel. The USGS Technical Specification D.3. requires that the i fuel temperature be measured and limited not to exceed 800"C during reactor pulsing.
]
Because power densities and temperatures in 12 w% fuel elements will generally exceed j those in 8 w% fuel, the licensee proposed changing Technical Specification D.3. to include a requirement that the instrumented element contain 12 w% uranium if any other element in the core contains that concentration. This requirement would provide reasonable assurance that the temperature of the measured fuel element in the B or C ring would not be exceeded by others in the core. On the basis of these considerations, the staff ]
concludes that the USGS technical specifications related to pulsing continue to provide {
reasonable assurance that fuel damage due to allowed pulsing will remain acceptably unlikely with the proposed use of 12 w% fuel.
2.4 Postulated Accidents The licensee has analyzed the loss of cladding integrity in the reactor room air of the most ;
irradiated fuel element. This is the standard maximum hypothetical accident (MHA) j l
adopted by the staff for TRIGA reactors, which places a conservative upper bound on the consequences of a fission product release accident. Because the MHA is hypothetical, a scenario that would lead to the MHA has not been identified by the staff. To show the i effect of the proposed use of 12 w% fuel, the failure of a 8.5 w% fuel element in air is l compared with the failure of a 12 w% fuel element in air.
The scenario used by the licensee includes the following conservative assumptions:
- a. The reactor operated at fulllicensed power for a long time (at least 40 days);
- b. A 12 w% fuel rod, having operated continuously at 22 kW, or 8.5 w% fuel rod, having operated continuously at 14.5 kW (the most exposed element), is instantly removed from the core to the reactor room;
l 4 )
I
- c. Tha cladding fails instantly, releasing all airborne fission products (halogens and noble gases) in the fuel-clad gap;
- d. The airborne fission products instantly disperse uniformly into the reactor room air; and
- e. The reactor room air exhaust system switches, as designed, to emergency mode of operation, releasing contaminated air to the environment in a controlled manner.
The licensee calculated the resultant doses to occupants of the reactor room, to a person assumed to be just outside the building, a person at the fence line of the Denver Federal Center and at the nearest residence.
The occupational doses in the reactor room due to the airborne radioactivity included whole body gamma-ray doses due to immersion in the noble gas fission products in the finite-sized room, and thyroid doses due to inhalation of iodines. If the occupants of the reactor room were evacuated within 1 minute, a reasonable assumption based on the results of emergency drills conducted by the licensee, the whole-body dose would not exceed 800 mrem for 12 w% fuel (550 mrem for 8.5 w% fuel). The committed dose equivalent to the thyroid from iodines would not exceed 26 rem for 12 w% fuel (17 rem for 8.5 w% fuel).
The calculations of the doses in the controlled area just outside the reactor building j assumed a semi-infinite hemispherical cloud for the noble gas whole-body immersion doses, and minimal dispersion of the cloud for the thyroid doses resulting from inhalation of iodines. If persons around the reactor building were evacuated within 5 minutes, the whole body dose would not exceed 30 mrem for 12 w% fuel (20 mrem for 8.5 w% fuel).
The committed dose equivalent to the thyroid from iodines would not exceed 900 mrem for 12 w% fuel (600 mrem for 8.5 w% fuel). Rapid evacuation and access control are both feasible at the USGS reactor facility because it is staffed by U.S. Government employees and is within a fenced Government compound. The licensee states that from numerous evacuation drills that have been conducted at the site, security persornel would arrive within 5 minutes to evacuate as much of the surrounding area as necenary:
The licensee calculated doses at the fence and at the nearest permanent residence using a method and code developed by a Department of Energy nationallaboratory for transient airborne releases of radioactive material. The staff has performed sample calculations with this computing program and obtained doses in reasonable agreement with those provided by the licensee. The staff also compared sample calculated doses with those obtained by a ,
method developed by NRC several years ago for finite clouds of noble gases. For l comparable conditions, the method used by the licensee gave doses from 5 to 15 times those obtained by the latter method. On the basis of these results, the staff considers the l licensee's computed doses at distances typical of the perimeter fence and the nearest permanent residence to be applicable conservative values. !
The nearest fence of the Federal Center is about 305 m (1000 ft) from the reactor building. l The licensee's calculated whole-body dose to a person at the fence for the entire release from the reactor room would not exceed 0.3 mrem for 12 w% fuel (0.2 mrem for 8.5 w%
fuel). The committed dose equivalent to the thyroid from iodines at the fence would not exceed 14 mrem, for 12 w% fuel (9 mrem for 8.5 w% fuel). The nearest residence is located about 640 m (2100 ft) from the reactor. The calculated whole-body dose to a a
5 person at this location for the entire release from the reactor building would not exceed 0.1 mrem and the committed dose equivalent to the thyroid from iodines would not exceed 3.5 mrem for either of the two fuels evaluated.
Separate standards do not exist for evaluating the effects of postulated accidents in a research reactor. The results of accident analysis have generally been compared with the 10 CFR Part 20 criteria in effect at the time the research reactor was licensed. For the USGS TRIGA, these criteria are found in 10 CFR 20.1 through 20.602 and related appendices. The limiting annual acceptable doses the staff has applied to the USGS TRIGA reactor is 5-rem whole-body and 30-rem thyroid exposure for occupationally exposed persons and 0.5-rem whole-body and 3-rem thyroid exposure for members of the public.
Considering the conservative assumptions of both the MHA scenario and the dose calculations noted above, the staff concludes that there is reasonable assurance that this MHA would not pose significant radiological risk to the operating staff, the public, or the environment for any fuel mix authorized by the amended license.
2.5 Technical Spemfications The licensee proposed changes in the USGS technical specifications to ensure operation of the reactor within the limits that were shown not to lead to fuel damage with the use of up to 12 w% TRIGA fuel. The proposed technical specifications are as follows:
Specification D.1., Reactor Core. The underlined words are added to the existing specification and the words with strikeout are removed:
- 1. The core shall be an assembly of TRIGA &sesinen stainless steel clad fuel-moderator elements, nominally 8.5 to 12 w% uranium, arranged in a close-packed array except for (1) replacement of single individual elements with incore irradiation facilities or control rods; (2) two separated experiment positions in the D through E rings, each occupying a maximum of three fuel element positions. The reflector (excluding experiments and experimental facilities) shall be water or a combination of graphite and water. The reactor shall not be operated in any manner that would cause any fuel element to produce a calculated steady state power level in excess of 22 kW.
Specification D.3., Reactor Core. The underlined words are added in the existing specification:
- 3. Fuel temperatures near the core midplane in either the B or C ring of elements shall be continuously recorded during the pulse mode of operation using a standard thermocouple fuel element. The thermocouple element shall be of 12 w% uranium loading if any 12 w% loaded elements exist in the core. The reactor shall not be operated in a manner which would cause the measured fuel temperature to exceed 800*C, Specification D.7., Reactor Core. The following specification is added in its entirety:
- 7. The power produced by each fuel element while operating at the rated full power shall be calculated if the reactor is to be operated at greater than 100 kW with less than 100 fuel elements in the core. Recalculations shall be performed:
l
)
)
6 a) at 61 1 month intervals; or b) whenever a core loading change occurs.
Power per element calculations are not required at any time that the core contains at least 100 fuel elements or if reactor power is limited to 100 kW. If the calculations show that any fuel element would produce more than 22 kW, the reactor shall not be operated with that core configuration.
In a telephone conversation with the licensee on February 19,1998, the licensee stated
' that the removal of the words " Mark lil" from Technical Specification D.1 is an editorial
. change because the words are meaningless when referring to TRIGA fuel. The licensee's application did not contain a justification for this change.
The staff has reviewed and evaluated the application from the USGS for amendment of the operating license for its nuclear research reactor. The licensee requested approval to use standard uranium-zirconium hydride fuel containing 8.5 to 12 w% uranium in the reactor.
The licensee provided analyses of the operation, power production, and resultant fuel temperatures for both steady-state and pulsed operation of the reactor. Changes in the l licensee's existing technical specifications were proposed to accommodate and limit the operation of the reactor with mixed fuel cores. The staff evaluated the justification provided by the licensee for the proposed changes and concludes that the changes in the technical specifications'are both necessary and sufficient to provide reasonable assurance that continued operation of the reactor, as proposed, analyzed, and limited, will not pose significant or unacceptable radiological risk to the health and safety of the public, the reactor staff, or the environment. Therefore, the staff concludes that changes to the technical specifications to amend the operating license as proposed by the licensee are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves changes in the :nstallation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes in inspection and surveillance requirements. The staff has determined that this amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Purcuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
l
)
The staff has concluded, on the basis of the considerations discussed above, that (1) because the amendment does not involve a significant increase in the probability or l consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from any accident previously evaluated, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the
3 i l l !
l l i 7
l public will not be endangered by the proposed activities; and (3) such activities will be
! conducted in compliance with the Commission's regulations, and the issuance of this amendment will not be inirnical to the common defense and security or the health and i safety of the public.
Principal Contributors: R. E. Carter, INEEL A. Adams, Jr.
1 March 16, 1998 Date:
l l
l l
i i
l l
L L
t-1 1
l 1
1 I
i