ML19208D384
| ML19208D384 | |
| Person / Time | |
|---|---|
| Site: | Idaho State University |
| Issue date: | 08/30/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19208D378 | List: |
| References | |
| NUDOCS 7909280346 | |
| Download: ML19208D384 (12) | |
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UNITEC STATES y))e(3 g NUCLEAR REGULATORY COMMISSION g ' Q j]. j WASHINGTON. D. C. 20555
%,' w 4 SAFETY EVALUATION AND ENVIRONMENTAL INPACT APPRAISAL BY THE OFFICE' 0F NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 2 TO FACILITY OPERATING LICENSE NO. R-110 IDAHO STATE UNIVERSIT.
DOCKET NO. 50-284 Introduction By letter dated August 10, 1978, is suppiementta November 21, 1978, Idaho State University (ISU or the licensee) requested that Facility Operating Licensa No. R-110 be amended to permit operation of the AGN-201 nuclear reactor at power levels up to 5 watts (thermal).
Operation at this power level would require modification of the reactor instrumentation and control system and installation of additional shielding.
Most of the existing reactor facility systems and components have the capability for operation at 5 watts (thermal); therefore, only minor modifications would be required. Upon completion of the modifications for 5-watt operation, the reactor would be designated Model AGN-20lM, Serial No. 103.
Discussion The AGN-201 reactor is a portable, self-contained reacter using a hemogeneous fuel mixture of uranium oxide and polyethylene enriched with U-235. The reactor is presently authorized to operate at a maximum power level of 100 milliwatts (thermal).
In preparation for 5-watt operation, the University proposes to consider any location within the reactor room floor as a high radiation area.
The instrumentation and control system and shielding require modification to operate at the neutron flux level resulting from 5-watt operation.
The following delineate 5 these modifications:
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J (a) Instrumentation tonsole and instrumentation modifications for continuous 5-watt operation are as follows:
The three neutron monitoring channels of the AGN-201 reactor will be modified to allow operation of the reactor at the higher power levels. The three detectors and associated circuitry are identified as Channels 1, 2 and 3 and are treated below. All channels have inter-locks on the "zero set" and " calibrate" twitches to pre-vent deactivation of any instrument during operation except as noted below. High level scram settings will be adjusted to prevent operation 200% above maximum power levels.
(1) Channel 1 The count rate meter channel will be unchanged for opera-tion between source and.1 watt. Both high and low level scrams will be active. A switch will be incorporated to to deactivate Channel 1 for operation above.1 watts to prevent saturation and burnup. Channel 1 may be bypassed in accordance with the Technical Specifications (TS) when-ever the reactor control or safety rods are not fully withdrawn.
(2) Channel 2 Present range capability of Channel 2 permits its use for operation of power levels up to 5 watts.
No change will be made except to extend the upper limit trip setting to scram at 10 watts (200% of 5 watts). All scram circuits will remain unchanged.
(3) Channel 3 No changes will be made in Channel 3 except to activate the existing range settings up to a level to permit 5 -
watt operation. All scram circuits will remain unchanged.
Channel 3, also may be bypassed in accordance with the TS under the same conditions as Channel 1.
The foregoing provides for at least two operating channel during reactor operations. This ene out of two logic for operations above 0.1 watts is acceptable for an AGN-20lM type reactor.
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,[ g a U10 M U Mb (b) Shielding Additional shielding has been provided by a concrete wall constructed of 4" x 8" x 16" concrete blocks and 4" x 8" x 12" barytes concrete blocks. The blocks were held to close
. dimensional tolerance in manufacture and stacked in such a manner that voids in the complete wall are at a minimum. Near the beam ports and glory hole, high density biccKs have been used between 40" and 112" above the base.
The use of these blocks further reduces radiation level in these areas. Over-head shielding is provided by 8" thick barytes blocks (minimum density 3.7 gicc).
As detailed in the Commission's Safety Evaluation dated February 11, 1957, supporting 5-watt operation for Aerojet-General Nuclecnics. Docket 50-32, an 18" additional concrete shield wall was sublicient to maintain acceptable radiation levels external to the shield when operating at 5 watts.
Sub-sequent analysis by Aerojet-General Nucleonics indicated that 16" of ordinary concrete shielding was sufficient.
Twelve inches of barytes concrete is more effective than 18" of ordinary concrete. The essential features of the shielding are shown in Figures 2, 3 and 4 of the licensee's application.
The radiation levels associated with 5-watt operation (peak thermal flux of 2.5 x 108 n/cm2.sec) have been calculated by Aerojet-General Nucleonics.
The table below gives Aerojet-General results for gamma dose for 18" corirete shields:
Energy (Mev) u ux B
e "*
Dose (= rem /hr) 3.0 0.081 3.64 1.9 0.027 1.5 2.2 0.105 4.73 2.2 0.0038 1.0 Total = 2.3 mres/hr (13" concrete)
Changing the e~"* for a 16" concrete wall and using the same buildup factors yields the following table for the proposed shield:
17eriy ux 3
e,,
Cose (cre /hrl Inery ("evi 3.0 0.081 3.30 1.9 0.037 2.
2.2 0.103 4.2S 2.2 0.014 1.2 Total = 3.3 ares /hr (le" concrete)
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, Although calculations and operating surveys of similar facilities show that at 5 watts power there will be no area on the reactor floor outside the concrete shield where the total radiation exceeds acceptable levels; nevertheless, the reactor floor will be a control area with restricted access.
Procedural controls have been established to ensure no un-authorized entry into the restricted area.
The licensee has shcwn in the safety analys s in their application for 5-watt i
operations and based on previous gamma and neutron dose calculations and surveys, the total radiation level in the reactor rocm is computed to be less than 10 mrem /hr. This is considerably lower than the limits set forth in 10 CFR Part 20 for restricted areas. Other AGN-201 facilities operating at 5 watts have reported survey data showing that the actual radiation levels have not exceeded 1.0 mrem /hr.
The initial approach to 5-watt power will be controlled with complete surveys conducted of the shielding and restricted area to determine actual dose rates and shield integrity. The licensee will advise the Commission of survey results indicating significant inadequacies in shielding if they should occur.
If the licensee encounters any radiation levels exceeding 10 CFR Part 20, the reactor will be shutdown and a safety analysis conducted. The licensec has committed to submit to the Commission for review ar,y required modifications prior to reactor operation greater than 0.1 watt (thermal).
I.
SAFETY EVALUATION The present facility has not significantly changed frem that described in license Anendment No. I to Facility License No. R-110, dated April 28,1976, which authorized the operation of the AGN-201 reactor in the Lillibridge Engineering Laboratory, Pocatello, Idaho.
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. The low power (5 watts) will produce a negligible fission product inventory.
This and the strong negative temperature coefficient of reactivity, substantiates that AGN-201 reactors do not present significant hazards to the public.
Their safety and reliability have been demonstrated in several facilities operating at 5 watts for many years.
The proposed TS have been reviewed and revised for 5-watt operation.
The TS generally incorporate the design features, characteristics and operating conditions described in the original Hazards Summary Report for the AGN-201 reactor (1) submitted in support of Dockets F-15 and F-32 and referenced in the licensee's application and features of TS from other AGN-201 reector facilities operating at 5 watts.'
Inclusion of comprehensive Turveillance requirements and administrative controls will assure ac eptable performance of safety related equipment and require safety related reviews, audits and operating procedures. Record keeping and reporting require-ments will provide sufficient information to permit an assessment by the Commission of safety related activities and changes.
There are, however, several differences between the acconpanying TS and the original AGN documentation. These are discussed below.
The AGN-201 Preliminary Des'1gn Report (2), submitted on the F-15 docket, mentioned thermal fuses in the control and safety rods and e boron-loaded polyethylene sheet surrounding the graphite reflector.
The function of the thermal fuses in the control and safety rods was to cause the rods to fall from the core in the event of excessive temperatures produced in a nuclear excursion. They would, therefore, serve as a backup to the core thermal fuse which already serves as a backup to the normal scram system. The function of the baron-loaded sheet was to absorb thennal neutrons thereby reducing gama ray production from neutron capture in the shield water and the resulting radiation level outside the shield.
These design features were not mentigngd in subsequent submittals including the Hazards Surcary eporttl J, the AGN-201 Reactor Manual (3) and the Shield Cesign Reportl4 They were not referred to in the original AEC Hazards Analysis ( } or subsequent safety evaluatien.
They were not incorporated into the assembled AGN reactors and are not included in the proposed TS.
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. !4ny years of experience operating AGN-201 research reactors without themal fuses in the control and safety rods and without a baron-loaded polyethylene sheet surrounding the graphite reflector has established that these reactors can operate safely, as arsembled, at the proposed licensed power within acceptable radiation levels to both operating personnel and the general public. Based on our review and the above considerations, we have concluded there is reasonable assurance that operation without thermal fuses in control and safety rods and the baron-loaded polyethylene sheet referred to in the Preliminary Design Report will not endanger the health and safety of the public.
0-3) limited the total availa The original AGN-201 documentation excess reactivity to 0.2% ak/k. We have previously evaluated (6) ble-and authorized an increase in the excess available reactivity including contributions from positive wortn experiments, to 0.65% ak/k and this value has been incorporated into the IS. Because of the self-limiting action of the large negative temerature coefficient, an instantaneous reactivity insertion as high as 2.0% ak/k would not result in core damage or radioactivity release. Limiting the total available excess reactivity to 0.65% ak/k assures that the reactor will not become prompt critical and that the reactor periods will be sufficiently long such that the reactor protection system and/or operator action can effectively scram the reactor well before any safety limits are exceeded.
The licensee has submitted a revised safety analysis report in the application that incorporates the foregoing changes. No changes made to the facility have resulted in a decrease in margins of safety.
Furthermore, reactors virtually identical to this one with similar TS have been licensed for operation for periods of up to 40 years.
Hence, the bases and conplusions with respect to the safety of operation that were termined in our Safety Evaluation supporting the original license, as amended, in support of the current operating license remain unchanged. The revised TS are more definitive than the original and will provide the necessary controls and surveillance requirements to ensure safe operation.
The fuel consists of polyethylene material with uranium dioxide (enriched to 19.9% in U-235)unifornly dispersed throughout the polyethylene. Polyethylene is an organic material that can sustain radiation damage when exposed to fission product bombard-ment. Test data was provided by Aerojet-General Nucleonics of samples of core material exposed in the Argonne National Laboratory 1052 288
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CP-5 reactor. The CP-5 reactor is a 5 megawatt (flux-10 n/cm sec) reactor. Tests included exposures at full power for periods up to one week continuous operation. Analyses of these tests re-vealed that radiation damage was evident in a reduced density and there was some loss of hydrogen from the polyethylene. An extrapolation of these results, assuming that the integrated flux-time (nyt) is responsible for the damage, for continuous operation at 100 watts equates to a core life of six years prior to any damage occurring. At 5 watts continuous operation the core life would be approxfmately 120 years. As the normal operating cycle is less than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week, or less than 24%, the projected life approaches 600 years. From this analysis it is reasonable to con-clude that the AGN-201 core operating 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week at 5.0 watts (flux - 2.5 x 108 n/cm2 - sec) would sustain no radiation damage over the remaining 17 years of reactor operation authorized by the licensee's license.
We have postulated a design basis accident (DBA) for an AGN-201 operating at 5 watts to be the failure of the polyethylene moderator /
cladding of a single fuel disc in air.
If all elements ruptured and all fission products were released, the activity would be approxi-mately 150 millicuries, principally Iodine-131. This equates to about 1.5 x 10-4uCi/ml if unifomly dispe.rsed throughout the reactor room (106 liters). The release from the postulated DBA when unifomly dispersed throughout the reactor room is 2.5 x 10-7 uCi/ml which is less than the Part 20 limiting dose for a restricted area. Any release to the outside atmosphere would be only a small fraction of this dose and therefore present no hazard to the public.
Although the DBA is only remotely possible; should it occur, the additional precautions of evacuating the reactor room will be accom-plished and emergNcy procedures implemented.
Moreover, due to the fact that:
(1) no unusual problems have avisen during over 11 years of authorized operation at 0.1 watt, (2) the revised TS require surveillance and periodic testing of safety related equipment to assure conticued safe operation of the reactor at 5 watts and to assure that any significant component degradation will be detected in a timely manner, and (3) other AGN-201 reactors of this type also have considerable operating experience at 5 watts without evidence of any unusual problems, we have concluded that the Idaho State University AGN-201 reactor can be operated in a safe.ranner at 5.0 watts.
Furthermore, based on the foregoing considerations, we have concluded that the estimated life of the facility will extend far beyond the end of the current license period. Therefore, from a reactor safety standpoint, the amendment is acceptable.
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, Conclusion on Safety We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation at 5 watts in the proposed manner, and (2) such activities will be conducted in com-pliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
II.
Environmental Imoact Acoraisal The environmental impact associated with operation of research reactors has been generically evaluated in the attached memorandum (Reference 7). This memorandum concludes that there will be no significant environmental impact associated with the licensing of research reactors to opera'.e at power levels.up to 2fMt and that no environmental impact stataments are required to be written for the issuance of construction pernits or operating licenses for such facili ties. We have determined that this generic evaluation is applicable to operating of the Idaho State University AGN-201 reactor at an increased power letel of 5 watts, and that there are no special or different features which would preclude reliance on the generic evaluation. Consequently, we have determined that the conclusion reached in the generic evaluation is equally applicable to this license action and that an environmental impact statement need not be prepared. Furthermore, ba.ed on our review of specific facility items which are considered for potential environmental impact, we have concluded that this license action is insignificant frem the standpoint of environmental impact.
Facility There are no pipelines or transmission lines entering or leaving the site above. grade. All utility services (water, steam, electricity, telephone and sewage) are below grade and are comparable to those required for typical campus laboratories.
Heat dissipation is accomplished by radiation in a large water tank which serves as the heat sink and is a sealed unit. The reactor is desicned as a sealed system, and in normal operation does not'have any gaseous or liquid radioactive effluent.
Solid, low-level radioactive waste cenerated in the research effort will be packaced in accordance with USNRC and Department of Transportation (00T) regulations and shipced off-site for storage at NRC approved sites. The transportation of such waste will be done in accordance with existing NRC-DOT regulations in approved shipping containers. Chemical and sanitary waste systems are similar to those existing at other university laboratories and buildings.
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Environmental Effects of Facility Operation Release of thermal effluents from a reactor of 5.0 watts will not hava a significant effect on the environment. This small amount of waste heat is rejected to the surrounding water tank and eventually to the atmosphere by means of conduction and radiation.
There will be an insignificant amount of release of gaseous or liquid effluents.
Yearly doses to unrestricted areas from external radiation will be at or below established limits of 10 CFR Part 20. Solid radio-active wastes generated in the research program will be shipped to an authorized disposal site in approved containers. These wastes should not amount to more than a few shipping containers a year.
No release of potentially hannful chemical substances will occur during normal operation. Small amounts of chemicals and/or high-solid content water may be released from the facility through the sanitary sewer from laboratory experiments.
Other potential effects of the facility, such as esthetics, noise and societal or impact on local flora and fauna are expected to be too small to measure.
Environmental Effects of Accidents Accidents ranging from the failure of experiments up to the largest core darage and fission product release considered possible result in doses of only a small fraction of 10 CFR Part 100 quidelines and are considered negligible with respect to the environment.
Unavoidable Effects of Facility Ooeration The unavoidable effects of operation involve the fissionable material used 'in the reactor.
No adverse impact on the environ-ment is expected from these unavoidable effects.
Alternatives to Oceration of the Facilig To accomplish the objectives associated with research reactors, there are no suitable alternatives.
Scme of these objectives are training of students in the operation of reactors, production of radioisotopes, and use of neutron and ganra ray beams to conduct experiments.
Lone-Term Effects of Facility Construction and Ooeration The long-term effects of research facilities are considered to be beneficial as a result of the contribution to scientific knowledge and training.
Because of the relatively low amount of capital resources involved and the small ircact on the environment very little irreversible and irretrievable commi+ ment is associated with such facilities.
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, Costs and Benefits of Facility and Alternatives The monetary costs involved in operation of the facility are less than $15,000 per year. There will be very limited environmental impacts. The benefits include, but are not limited to, some com-bination of the following: conduct of activation analyses, conduct of neutron radiography, training of operating personnel and education of students. Some of these activities could be conducted using particle accelerators or radioacti~. sources which would be more costly and less efficient. There is no reasonable alternatives to a nuclear research reactor for conducting this spectrum of activities.
Conclusion and Basis for Necative Declaration Based on the foregoing analysis, we have concluded that there wil' be no significant environmental impact attributed to this proposed license power increase. Having made this conclusion, we have further concluded that no environmental impact statement for the proposed action need be prepared and that a negative declaration to this effect is appropriate.
III. Safety Evaluation and Environmental Considerations Succorting Additional Items Included in Amendment No. 2 Also included in the amendment are changes to the license, setting a limit on the authorized enrichment of fuel used ir, the reactor, adding the physical security plan as alicense condition, and deleting reporting and record keeping requirements in the existing license.
The inclusion of enrichment levels in the license is necessary as 10 CFR 573.47 and the safeguards upgrade rules criteria are established by not only the amount but also the enrichment of the special nuclear material. Addition of the physical security plan as a license con-dition is being done pursuant to 10 CFR 550.54. The previous license conditions pertaining to reporting and record keeping requirements are now included in the revised TS.
These changes are administrative. They do not involve significant new safety information of a type not considered by a previous Commission safety review of the facility.
They do not involve a significant increase in the probability or consequences of an accident, do not involve a significant decrease in a safety margin, and therefore do not involve a significant hazards consideration. We have also concluded that there is reasonable assurance that the health and safety of the public will not be endangered by these changes.
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. 6 We F.r.,'e also determined that these changes do not authorize a
_ change in effluent types or total amounts nor an increase in power level, and will not result in any significant environmental impact.
H1ving made this determination, we have further concluded that tnese changes involve actions which are insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)
(4) that an environmental impact statement, negative declaration or environmental impact appraisal need not be prepared in connection with these changes.
Dated: August 30, 1979 1052 293
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References 1.
" Hazards Sumary Report for the AGij-201 Reactor," Aerojet General Nucleonics, August 1956 (see Docket F-15).
2.
"AGN Model 201 Reactor, Preliminary Dee'gn Study," Aerojet General Nucleonics, May 1956 (see Jocket F-15).
3.
"AGN-201 Reactor Manual," Aerojet General Nucleanics, July 1957 (see Docket F-15).
4.
' Shield Design for the AGN-201 Reactor," Appendix F to Reference 1, September 1956 (see Docket F-15).
5.
AEC Memorandum accompanying License R-10, March 20,1957 (see Docket F-32).
6.
NRC Safety Evaluation on analysis of experiments conducted by Georgia Institute of Technology, Docket 50-276, dated August 25, 1977.
7.
D. Muller to D. Skovholt memorandum " Environmental Considera-tions Regarding the Licensing of Research Reactors and Critical Facilities," dated January 28, 1974.
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