ML061720498
| ML061720498 | |
| Person / Time | |
|---|---|
| Site: | 05000083 |
| Issue date: | 06/19/2006 |
| From: | Vernetson W Univ of Florida |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML061720498 (11) | |
Text
UNIVERSITY OF FLORIDA Nuclear Facilities 202 Nuclear Sciences Center Department of Nuclear and Radiological Engineering P.O. Box 118300 Gainesville, Florida 32611-8300 Tel: (352) 392-1408 Fax: (352) 392-3380 Email: vemet@ufl.edu June 19, 2006 Attn: Document Control Room UFTR Emergency Plan U.S. Nuclear Regulatory Commission Revision 14 Washington, DC 20555 University of Florida Training Reactor (UFTR)
Facility License R-56, Docket No. 50-83 The enclosed package contains Revision 14 to the approved UFTR Emergency Plan. Revision 14 has been reviewed by UFTR management and the Reactor Safety Review Subcommittee (RSRS) to assure Revision 14 does not decrease the effectiveness of the UFTR Emergency Plan. The changes are considered relatively minor in nature; they are all associated with the conversion from using high enriched uranium (HEU) to low enriched uranium (LEU) fuel in the UFTR.
Revision 14 consists of a set of updates and revisions to the title page, page v, pages 1-1, 1-6, 1-12, 1-13, 1-14, 5-1 and 6-1. The new pages are marked with the usual vertical lines in the right margin for easy location of specific changes except for Table 1.1 which is entirely new. In this letter, the page number and line references are to those in the current copy of the Emergency Plan.
First, the title page is updated to reflect inclusion of Revision 14. Then on page v, in the List of Tables, the title of Table 1.1 is updated to reflect the table contents based on the HEU-to-LEU analysis.
Second, on page 1-1, in section 1.3.1, in the first paragraph, the reactor core description is updated to reflect the approximate 20 kg (uranium) of LEU fuel in the core, the fact it is 19.75% uranium silicide-aluminum fuel and that the fuel bundles are 14 plates each versus the previous 11 plates.
Third, on page 1-6, in section 1.5, the Maximum Hypothetical Accident (MHA) is considered as this section is reorganized and rewritten to reflect the new HEU to LEU conversion analysis with verb tenses in both the first and second paragraphs changed for writing style consistency and a typographical error in the word "various" corrected in line 6 of the first paragraph.
Fourth, on page 1-12, as section 1.5 continues, in the second paragraph, line 1, the verb tense is changed. Previously, for the Fuel Handling Accident (FHA), the reactor was assumed to operate at 100 kW steady state power for 30 days to build up equilibrium fission products. For the current FHA, the assumption is that operation is at 100 kW steady state power for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per day for 30 days. In the same paragraph, the last sentence is deleted, since Table 1.1 is being updated for the HEU to LEU analysis results, and several new sentences are inserted describing the new table. The third paragraph is then updated to summarize the whole body and thyroid dose calculations for public and worker doses presented in Table 1.1. A k) -- 2-0 An Equal Opportunity Institution -AV--q
P*e 2 Paace 2 Jn 19-.2006 June 920 NRC fncuiiment Cnntrol Des~k Fifth, also on page 1-12, in item (2), a sentence is added to reflect the practice, now to be part of the Tech Specs following the HEU to LEU analysis, that the last two rows of shield blocks over the core area will not be removed for at least 3 days after the last operation at power, and item (3) is updated based on the HEU to LEU analysis results. Consequently, with so many added lines of information on page 1-12, the last paragraph was moved and is now the first paragraph on page 1-14.
Sixth, on page 1-13, the new Table 1.1, "Summary of Occupational and Public Dose Results for the Fuel Handling Accident (FHA) for the LEU Fueled Core" reflects new dose calculations for the public and workers for the fuel handling accident (FHA) assumed for the LEU core.
Seventh, on page 1-14, as section 1.5 concludes, the existing two paragraphs are unchanged but they are moved down the page as a result of the last paragraph on page 1-12 being moved to page 1-14 due to the additional material added on page 1-12.
Finally, on page 5-1, in section 5.0, in line 2, the guideline limits for whole body dose are updated from 1 Rem to 0.5 rem and for the thyroid dose, from 5 rem to 3 rein per 10 CFR 20. Similarly, on page 6-1, in section 6.0, the same changes are made on lines 2 and 3 for the whole body dose and the thyroid dose, respectively.
As indicated, all these changes have been reviewed by UFTR management and by the Reactor Safety Review Subcommittee to assure they do not decrease the effectiveness of the UFTR Emergency Plan.
In general, these changes update the Plan to reflect the conversion from HEU to LEU fuel and make the Plan better suited to assure a proper response to emergencies at the University of Florida Training Reactor.
If there are any questions, please let us know. Thank you for your consideration.
Sincerly, .
William G. Vernetson Director of Nuclear Facilities WGV/dms Enclosures cc (letter only): A. Adams, Sr. Project Manager, NRC Reactor Safety Review Subcommittee Sworn and subscribed this k _day of June 2006 Da Commission Dampier l1 1-'..:DanaL.
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,O'ttjZExpires July 22009 Notary Public
EMERGENCY PLAN for the UNIVERSITY OF FLORIDA TRAINING REACTOR Updated through Revision 14, 6/06 FACILITY LICENSE: R-56 DOCKET NUMBER: 50-83 W. G. Vemetson University of Florida October 1983
UFTR EMERGENCY PLAN LIST OF FIGURES Figure 1.1 Isometric Sketch of the UFTR with Shielding Removed .................................... 1-2 Figure 1.2 First Floor Plan for the University of Florida Training Reactor Building .................................................................................................. 1-4 Figure 1.3 Second Floor Plan for the University of Florida Training R eactor Building .................................................................................................. 1-5 Figure 1.4 Relative Geographic Location of Alachua County and Gainesville in the State of Florida ....................................................................... 1-7 Figure 1.5 Map of Greater Gainesville Area showing Placement of the University of Florida Campus and Major Landmarks ......................................... 1-8 Figure 1.6 UFTR Building Placement on University of Florida Campus with Respect to Major Campus Arteries and Buildings ....................................... 1-9 Figure 1.7 University of Florida Campus Map with Building Locations, Primary Landmarks and Boundaries .................................................................. 1-10 Figure 1.8 UFIR Building Access Roads .......................................................................... 1-11 Figure 3.1 UFTR Organization Chart,................................................................................... 3-2 Figure 3.2 UFTR Emergency Organization Including Extensions ........................................ 3-6 LIST OF TABLES Page Table 1.1 Summary of Occupational and Public Dose Results for the Fuel Handling Accident (FHA) for the LEU Fueled Core ................................. 1-13 Table 5.1 UFTR Emergency Classification Guide ............................................................... 5-2 Table 8.1 Equipment Available from Radiation Control Office for Emergency Dose and Radiation Level Assessment ....................................... 8-2 Table 8.2 Equipment Typically Available In the UFTR Facility for Dose and Radiation Level Assessment.......... ...................................................... 8-3 Table 10.1 Maintenance and Calibration Schedule for Radiation Detection and Measuring Equipment ................................................................................. 10-3 Table 10.2 UFTR Safety System Operability Tests ............................................................. 10-4 Table 10.3 Decontamination Room Emergency Equipment Inventory ............................... 10-6 REV 7, 12/91 REV 8, 12/92 REV 12,8/01 v REV 14, 6/06
1.0 INTRODUCTION
1.1 Scope of the UFTR Emergency Plan The University of Florida Training Reactor (UFTR) Emergency Plan is designed to cope with emergencies which arise as a result of, or in connection with, reactor operations. Where possible, it adopts the standard campus procedures that are widely practiced and understood by campus emergency response teams. However, the plan deals primarily with emergency responses that are required by the unique nature of the research reactor facility and the credible accidents that might arise within the facility.
1.2 Basis for the UFTR Emergency Plan The emergency planning requirements for research reactors are specified in 10 CFR, Part 50, Appendix E.I Applicable guidance in emergency planning is set forth in Revision 1 to Regulatory Guide 2.6, "Emergency Planning for Research and Test Reactors" (March 1983)(21 and in ANSIIANS-15.16-1982, "Emergency Planning for Research Reactors."[31 These docu-ments were used as the basis for development of the UFTR Emergency Plan.
The UFTR Emergency Plan and associated Standard Operating Procedures meets or exceeds the requirement and guidelines delineated in these documents.
1.3 Characteristics of the UFTR Facility 1.3.1 Reactor Characteristics The UFTR is of an Argonaut-UTR type, with some modifications to adapt it to the university training program. The reactor core is heterogeneous in design currently using about 20 kg (uranium) of 19.75% enriched uranium silicide-aluminum (U3Si 2-Al) fuel contained in aluminum cladding. Water is used as the coolant and also as the moderator. The remainder of the moderator consists of graphite blocks which surround the boxes containing the fuel plates and the water moderator. The graphite also serves as a reflector. The fuel is contained in MTR-type plates assembled in bundles and contained in 6 water-filled boxes surrounded by reactor grade graphite. Each bundle is composed of 14 fuel plates, each of which is a sandwich of aluminum cladding around a uranium silicide-aluminum "meat."
There are four swinging-arm-type control blades (three safety and one regulating) consisting of four cadmium vanes protected by magnesium shrouds that operate by moving in a vertical arc within the spaces between the fuel boxes. These blades are moved in or out by mechanical drives or they may be disconnected by means of electromagnetic clutches and allowed to fall into the reactor. The drives, located outside the reactor shield for accessibility, are connected to the blades by means of long shafts. An isometric sketch of the UFTR facility with the shielding removed is presented in Figure 1.1.
The biological shield is made of cast-in-place concrete 3 to 6 ft. thick. Access to the ends and top of the reactor is provided by removal of concrete blocks cast to fit openings and to prevent radiation streaming.
1-1 REV 14, 6/06
1.4 UFTR Facility Location The UFTR building is located on the campus of the University of Florida at Gainesville, in Alachua County. Figure 1.4 shows the geographic location of Alachua County with Gainesville at its center in the north central portion of the Florida peninsula. Figure 1.5 shows the location of the University of Florida campus within the City of Gainesville. As shown in Figure 1.5, the University of Florida campus is in the southwestern quadrant of the greater Gainesville area approximately one mile from the center of the city (University Avenue and Main Street).
The Nuclear Sciences Center (Building 634) is annexed to the reactor building which is labeled
[UFTR (Building 557)" in Figure 1.6. Distances to key campus structures are shown via concentric circles drawn with the UFTR as the center, the first circle having a 250 ft. radius and the rest being at 500 ft. increments from the reactor building located at the center point. A detailed UF campus map showing all major arteries along with building locations, landmarks and boundaries is shown in Figure 1.7. Emergency vehicular approach to the reactor building is via one of three service drives delineated in Figure 1.8: the reactor service drive leading from Gale Lemerand Drive to an area west of the reactor building, the Journalism lot service drive leading from Stadium Road to an area east of the reactor building (limited during construction to enlarge the Journalism building), and the Nuclear Sciences Center service drive also leading from Gale Lemerand Drive to an area south of the reactor building.
1.5 Credible Accidents and Consequences Credible accidents for the UFTR were discussed in the University of Florida Safety Evaluation Report (SER).t 51 That discussion was based largely upon a generic study of various postulated types of accidents leading to cladding failure. Generic accidents analyzed by Battelle Pacific Northwest Laboratory include:
(1) Insertion of Excess Reactivity (2) Explosive Chemical Reaction (3) Graphite Fire (4) Fuel-Handling Accident (5) Core-Crushing Accident For case 1, the conversion SAR concludes that no credible nuclear excursion could lead to fission product release since there would be no fuel or clad melting. The second scenario was considered impossible because rapid metal-water reactions will not occur in the UFTR.
Similarly, a serious graphite fire resulting in damaged fuel is dismissed because the set of required conditions is essentially not possible. In addition, the core-crushing accident was analyzed as the UFTR Maximum Hypothetical Accident (MHA) for the UFTR. For the so-called MHA, the UFTR was run at full power for 30 days to build up equilibrium fission products followed by assumed instantaneous release of 100% of the noble gases and volatiles produced within recoil range of the clad surface (or 2.7% of total volatile activity) in one fuel bundle. Such a scenario is not credible, however, and was intended only to demonstrate how dose rates at various distances are affected by various building leakage rates up to and including total failure. Although no fuel was melted, mechanical damage was assumed to cause effective cladding removal and resultant gaseous activity release as discussed above. Since the event is the unlikely dropping of a shield block, REV 12, 8/01; REV 13, 9/05; 1-6 REV 14,6/06
this event was considered extremely unlikely, again, it was used only as the maximum hypothetical accident, not a credible accident.
Therefore, in agreement with the Battelle study, it was concluded that the most credible accident was the loss of cladding on one fuel plate due to a fuel handling accident. The cladding loss accident lacks a detailed causal explanation, but intuition suggests that the outer plates of a fuel element are the most likely to suffer mechanical damage. The Battelle postulated cladding loss is equivalent to two sides of a single fuel plate. In the LEU core, for the Fuel Handling Accident (FHA), the reactor is assumed to operate at 100 kW steady state power for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per day for 30 days. Then the fuel element with highest power was selected for evaluation with the accident applied to the highest power fuel bundle with a 3 day delay since at least 3 days are required to pass after the last reactor operation at power before not only fuel handling but also before moving the last two layers of protective concrete blocks to access the fuel to limit possible potential consequences of fuel handling accidents and to preclude damaging a fuel bundle with a dropped shield block before 3 days have elapsed. For the FHA, the assumption continues that the cladding would be stripped from the selected LEU fuel bundle for the fuel handling accident.
As indicated in Table 1.1, the radiological exposure from the FHA calculated for a member of the public at closest approach would be much less than 1.0 mRem whole body dose from the noble gases and less than 0.522 Rem to the thyroid from the iodine gases.
Correspondingly, occupational radiological exposure would be less than 1.0 mRem whole body dose and less than 200 mRem to the thyroid. For these accidents, radiation doses to the public in unrestricted areas as well as workers would be far below the limits stipulated in 10 CFR 20.
Even so, the assumptions used in these calculations are believed to be very conservative for three reasons:
(1) First, it is highly unlikely that dropping a fuel element would be severe enough to cause fuel damage equivalent to stripping the cladding from an entire fuel plate.
(2) Second, fuel transfer operations cannot begin immediately after shutdown. The shielding blocks first must be removed from the structure to reveal the fuel elements in the core.
In addition, the UFTR does not shut down and immediately begin to manipulate fuel.
Typically, the UFTR will shut down from power operations for more than 7 days prior to commencing fuel-handling operations. In all cases, the reactor would be shutdown from power operations at least 3 days to allow substantial decay of fission product inventory.
In addition, the last two layers of shield blocks over the core area will not be removed for at least 3 days after the last operation at power.
(3) The UFTR would not usually operate 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s/day for a 30-day period. The reactor has a license limit of 23.5 MW-hours per month but the UFTR averaged less than 25.0 MW-hours per year for a typical ten-year period (9/81-8/91).
REV 7, 12/91 REV 9, 1/95 REV 11, 1/99 1-12 REV 14,6106
Table 1.1 Summary of Occupational and Public Dose Results for the Fuel Handling Accident (FHA) for the LEU Fueled Core Occupational Radiological Exposure Rate from LEU Core Thyroid Dose Rate Whole Body Dose 5-Minute 5-Minute Rate Exposure Rate Exposure Distance (rem/hr) (rem) (remLhr) (rem)
Inside Reactor Building 2.33 0.194 0.0045 0.00038 Limit: Thyroid = 30 rem, Whole Body 5 rem Radiological Exposure for the Public from LEU Core Time of Thyroid Dose (rem) Whole Body Dose (rem)
Distance Exposure Leak Rate N%Vol/hr) Leak Rate % Vol/hr (M) (hr) 10% 20% 100% 10% 20% 100%
4 16.5 2 0.109 0.199 0.522 7.8xl0"4 1.4xlO" 3.7x10-4 190.0 24 0.0193 0.0182 0.0205 1.3xlO"5 l.4xlO"5 1.5xlO"5 Limit: Thyroid = 3 rem, Whole Body = 0.5 rem 1-13 REV 14,6/06
Because of the conservative basis of the inventory and release calculations, the UFTR staff feels that it is extremely unlikely that members ofthe general public will receive radiation exposures greater than those permitted by 10 CFR 2061 when the reactor building is secured following such an accident. This position is in agreement with the UCLA staff position as described in their proposed Emergency Plan. Nevertheless, in keeping with UFTR Tech Specs and the ALARA criterion, the appropriate accident control strategy is to evacuate and secure the entire reactor building, including the reactor cell. There will be no pressure increases from a dropped element accident so maintaining the integrity of the reactor cell can greatly mitigate the radiation doses to the public. Of more direct concern is protecting personnel within the UFTR facility. Securing the facility limits releases and allows time to analyze a situation and to take advantage of decay of activity released to the cell atmosphere.
The UFTR Technical Specifications require an interlock to shut down the reactor cell air conditioning and the ventilation system when the evacuation siren is tripped whether initiation is automatic or manual.
Because the cell will be secured, any releases can be controlled and, if necessary, evacuation from nearby buildings can be effected before exceeding the limits of 10 CFR 20. For the so-called credible accident, described above, and other less serious accidents, the need for evacuation of large areas is totally unnecessary.
1-14 REV 14,6/06
5.0 EMERGENCY ACTION LEVELS (EAL-s)
There are no credible accident scenarios that lead to exposures exceeding the guideline limits of 0.5 rem to the whole body or 3 rem to the thyroid for any individual beyond the operations I boundary,[51 as shown by the values quoted in Table 1.1. Protective action guides for the general public and onsite personnel beyond the operations boundary are inappropriate. Somewhat similar concepts are employed internally for assessment of emergency status, as shown in Table 5.1.
Emergency Action Levels specified in Table 5.1 and described in subsections 7.2.1, 7.3.1 and 7.4.1 of Section 7.0 are considered to be EALs for activating the emergency organization and the initiation of protective actions at the level appropriate for addressing the emergency event in question.
REV 9, 1/95 REV 12, 8/01 5-1 REV 14,6/06
6.0 EMERGENCY PLANNING ZONE (EPZ)
Emergency planning zones for the UFTR are unnecessary as there are no credible accidents which lead to exposures exceeding the guideline limits of 0.5 rem to the whole body or 3 rem to the 31 thyroid in any regions beyond the operations boundary.t However, simply for planning purposes, 1
the operations boundary is established as an EPZ to conform with Table 2 of ANSI/ANS-15.16 [3 which represents an alternate method for determining the size of the EPZ. As indicated in the standard, this EPZ is selected based upon the postulated releases from credible accidents. In addition it should be noted that the UFTR authorized power level of 100 kW is well below the 2 MW limit in Table 2 for which the acceptable EPZ size is the operation boundary. The size of the area within the operations boundary is large enough to provide a response base that would support actions outside this area should this ever be needed. The predetermined protective actions for the EPZ are described in Sections 7.2.4 and 7.3.4.
6-1 REV 14,6/06