ML20029B674
| ML20029B674 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Rolla |
| Issue date: | 03/05/1991 |
| From: | Miraglia F Office of Nuclear Reactor Regulation |
| To: | MISSOURI, UNIV. OF, ROLLA, MO |
| Shared Package | |
| ML20029B669 | List: |
| References | |
| NUDOCS 9103140068 | |
| Download: ML20029B674 (32) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
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Docket No. 50-123 University _of Missouri-Rolla
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Facility Operating License No. R-79
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(University-of Missouri-Rolla
)
Amendment No. 9 Research-Reactor)
ORDER MODIFYING LICENSE I.
University of Missouri-Rolla (the licensee) is the holder of Facility Operating License No. R-79 (the itcense) issued on November 11, 1961, and subseqt.ently renewed on April 16, 1985, by the U.S. Nuclear Regulatory Commission (the Commission).
The license authorizes operation of the University of Missouri-Rolla Research Reactor (the facility) at a power level of up to 200 kilowatts (kw) thermal (t).
The facility is a training reactor located in Rolla, Missouri, and is contained in the Nuclear Reactor Facility,
- which is flocated on the east edge of the campus of University of Missouri-Rolla.
The mailing address is Nuclear' Reactor Facility, University of Missouri-Rolla, Rolla, Missouri 65401-0249.
II.
On February 25,-1986, the Commission promulgated a final rule in Section 50.64 ofLTitle-10 of the Code of Federal Regulations (10 CFR) limiting the use of high enriched uranium (HEU) fuel in domestic research and test reactors (non-power reactors) (see 51 FR 6514).
The rule, which became effec-
.tive on March 27, 1986, requires that each licensee of a non-power reactor replace HEU fuel at its facility with low-enriched uranium (LEU) fuel acceptable to the 9103140068 910303 PDR ADOCK 05000123 P
2-7 Commission (1) unless the Commission has determined that the reactor has a l
- unique purpose and (2) contingent upon Federal Government funding for conversion-related costs. -The rule is intended to promote the common defense and security by reducing the risk of thef t and diversion of HEU fuel used in non power reactors and the adverse consequences to public health and safety and the environment from such theft or diversion.
l Sections 50.64(b)(2)(1) and (ii) require that a licensee of a non power reactor (1) not -initiate acquisition of additional HEU fuel, if LEU fuel that is acceptable to the O mmission for that reactor is available when the licensee proposes that acquisition, and (2) replace all-HEU fuel in its possession with available LEU fuel acceptable to the Commission for that reactor, in accordance with a schedule determined pursuant to 10 CFR 50.64(c)(2).
Section 50.64(c)(2)(1) of the rule, among other things, requires each i
licensee of a non power reactor, authorized to possess and to use HEU fuel, to develop and to submit to the Director of tho Office of Nuclear Reactor Regulation (Director) by March 27, 1987, and at 12-month intervals thereafter, a-written proposal (proposal) for meeting the rule's requirements.
Section 50.64(c)(2)(i) also requires'the licensee to include,the following in its proposal:
(1) a certification that Federal Government funding for conversion is available through the U.S. Department of Energy (D0E) or other-appropriate Federal agency and (2) a schedule for conversion, based upon availability of replacement fuel acceptable to the' Commission for that reactor and upon consideration of other factors such as the availability of shipping casks, implementation of arrangements for the available financial support, and reactor usage.
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, Section_50,64(c)(2)(iii) requires the licensee to include in its proposal, to the extent required to effect conversion, all necessary changes to the license, to the facility, and to the licensee's procedures (all three types of changes hereafter called modifications).
This paragraph.also requires the licensee to provide supporting safety analyses so as to meet the schedule established for conversion.
Section 50.64(c)(2)(iii) also requires the Director to review the licensee's_ proposal, to confirm the status of Federal Government funding, and to determine a final schedule, if the licensee has submitted a schedule for conversion.
Section 50.64(c)(3) requires the Director to review the licensee's supporting safety analyses and to issue an appropriate enforcement order directing both the conversion and, to the extent consistent with protection of the public health and safety, any necessary modifications.
In the statement of considerations of the final rule,-the Commission explained that in most cases,
-if not-all, the enforcement-order would be an order to modify the license under
-10 CFR 2-204 (see 51 FR 6514).
Section 2.204 provides, among other things, that the Commission may modify a license by issuing an amendment on notice to the licensee that it may demand a hearing with respect to any part or all of the amendment within 20 days from the date of the notice or such longer period as the notice may provide.
The amendment will become effective on the expiration of this 20-day-or-longer period.
If the. licensee requests a hearing during this period, the amendment will become effective on the date specified in an order made after the hearing.
. Section 2.714 states the requirements for a person whose interest may be affected by any proceeding to initiate a hearing or to participate as aLparty.
4 III.
On November 16, 1988, as supplemented on May 8, 1990, May 30, 1990,
. August 9, 1990 and October 25, 1990, the Director received the licensee's
. proposal, including its proposed modifications, supporting safety analyses, and senedule for conversion, The conversion consists of replacement of high-enriched with low-enriched uranium fuel elements. The fuel elements contain materials testing reactor (MTR)-type fuel plates, with the fuel meat consisting of uranium silicides dispersed in.an aluminum matrix.
These plates contain an enrichment of less than 20 percent with the uranium-235 isotope.
The Attachment to -this Order-includes the changes to the licensing conditions I
and technical specifications that are needed to amend the facility license.
The NRC staff.has reviewed the licensee's submittals and the requirements of 10 CFR 50.64 and has determined that the public health and safety and 'the common defense and security require the licensee to-convert the facility from the use of-HEU-to LEU fuel pursuant to the modifications stated in the Attachment in accordance with the schedule-included herein following.
IV.
Accordingly, pursuant to Sections 51, 53, 57, 101, 104, 161b., 1611., and 1610. of the Atomic Energy Act of 1954, as amended, and to the Commission's regulations in 10 CFR 2.204 and Section 50.64, IT IS HEREBY ORDERED THAT:
On the later date of either receipt of LEU fuel elements by the licensee
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, or 30 days following the date of publication of this Order in the Federal Register Facility _ Operating License No. R-79 is modified by amending the license conditions and' Technical Specifications as stated in the Attachment'to this Order.
V.
Pursuant to the Atomic Energy Act of 1954, as amended, the licensee or any other-person adversely affected by this Order may request a hearing within 30 days of the date of this Order.
Any request-for a hearing shall be submitted to the Director,. 0f fice of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,. Washington, D.C. 20555, with a copy to the Assistant General Counsel for Hearings and Enforcement at the same address.
If a person other than the licensee requests-a hearing, that person shall set forth with particularity in accordance with 10 CFR 2.714 the manner in which the person's interest.is_ adversely affected by this Order.
If a hearing is-requested by the licensee or-a person whose interest is-adversely _affected, the Commission shall issue an order designating the time and place of any hearing.
If a hearing is held, the issue to be considered at such hearings is whether-this Order should be sustained.
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. This Order shall become effective on the later date of either the receipt of LEU fuel elements by the licensee or 30 days following the date of publica-tion of this Order in the Federal Register or, if a hearing is requested, on the date specified in an order following further proceedings on this Order.
FOR THE NUCLEAR REGULATORY COMMISSION f6+t '. [.
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Frank J. Miraglia, eputy Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland this 5th day of March 1991 Attachments:
As stated w
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ATTACHMENT TO ORDER MODIFYING FACILITY OPERATING LICENSE NO. R-79 A.
License Conditions Revised and Added by this Order 2.B.2 Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material,"'to receive, possess and use up to a maximum of 5.00 kilograms of contained uranium-235 at various enrichments, up to a maximum of 200 grams of plutonium-239 in the form of sealed plutonium-beryllium neutron sources in connection with operation of the reactor, and to possess, but not separate, i
such special nuclear material as may be produced by the operation Lof tne facility.. Without exceeding the foregoing maximum possession limits, the maximum limits on specific enrichments of t,-235 are as follows:
Maximum
'U-235 (kilograms)
% Enrichment Form 4.95
< 20%
MTR-type fuel 0.05 Any Fission chambers and flux--foils used in connection with operation of
-the reactor 2.B.4 Pursuant to the Act ano 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material," to possess, but not use, a= maximum of_4.90-kilograms of contained uranium-235 at greater than 20%
enrichment in the form of MTR-type _ reactor fuel until the existing inventory of MTR-type reactor fuel is removed from the facility.
2.C.2 Technical Specifications The Techr;ical Specificatfor:s contained in Appendix A, as revised L
through Amendment No. 9 are hereby-incorporated in this license.
The licensee shall operate the facility'in accordance with tk L-Technical Specifications.
.- B.
Technical Specifications Revised by this Order
.1.3 Definitions scram time - the elapsed time between reaching a limiting safety system i
set point and the time when a control rod is fully inserted.
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~2-shutdown margin - the minim,'m shutdown reactivity necessary to provide confidence that the reactor can be made subtritical by means of the control and safety systems starting from any permissibie operating condition with the maximum worth scrammable control rod and any non-scrammable control rod in their fully withdrawn positions and that the reactor will remain subtritical without further operator action.
3.1 Reactor Core Parameters Specifications (3) The minimum shutdown margin under any condition of operation with the highest worth control rod and any non-scrammable control rod i
fully withdrawn shall be no less than 1.0% delta k/k.
5.1 Site and Facility Description 5.1.2 The reactor is housed in a steel-framed, double-walled aluminum building designed to restrict leakage.
All air and other gases will be exhausted through vents in the reactor bay ceiling 30 feet (or 11 meters) above grade.
The Reactor Building free volume is approxi-mately 1700 cubic meters.
5.3 Reactor Core and Fuel 5.3.2 Fuel Elements (1)
Plate fuel elements of the MTR type are used.
The overall dimensions of each element are approximately 3 inches by 3 inches by 36 inches.
The active length of fuel is approximately 24 inches and the fuel is clad in aluminum alloy.
The fuel elements have 18 fuel plates joined to two side plates.
The whole assembly is joined at the bottom to a cylindrical nose piece that fits into the core grid plate.
The low enriched uranium (LEU) fuel meat is U,Si,ly 20% U-235.
dispersed in an aluminum matrix and is enriched to approximate The U.S. N.R.C. has approved the use of LEU fuel elements of this type in NUREG-1313.
(2) There are control rod fuel elements which are similar to the elements described in (1) with th'e exception that the center eight plates have been removed and have been replaced with guide plates such that the control rod cannot come in contact with fuel plates.
(3) There are half fueled elements, which have nine LEU fueled plates (either the front ones er the rear ones as appropriately marked) and nine dummy (or unfueled) plates, i
I
. (4) There is an irradiation fuel element whien has six fuel plate positions left blank (that is plate positicos 11 through 16),
plates 10 and 17 are dummy (or fueled) and all the others (1 through 9 and 18) are LEU fueled.
5.3.4 Control Rod Drive Mechanisms (1) The shim / safety rod drives have a maximum vertical travel of 24 inches and a withdrawal rate of approximately 6-inches per minute.
The chim/
safety rods are magnetically coupled to the drive niechanisms and drop into the core, by gravity, upon a scram signal.
(2) The regulating rod drive has a maximum vertical travel of 24 inches and a withdrawal rate of approximately 24 inches per minute.
The regulating rod is mechanically coupled to its rod drive and does not respond to a '; ram signal.
5.4 Fissionable Material Storage 5.4.1 The fuel storage pit, which is located below the floor of the r,' actor pool and at the end opposite from the core, wl be capable of storing the complete fuel inventory of either n;ghly-enriched uranium (HEU) fuel or of low-enriched uranium, but not both.
The neutron multiplication factor of the fully loaded storage pit shall not exceed 0.9 under any conditions.
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ENCLOSURE T0__ LICENSE' AMENDMENT NO. 9
' FACILITY OPERATING LICENSE NO. R-79 DOCKET N0. 50-123 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.. The revised pages are identified by Amendment number and
. contain a vertical-line indicating the area of change.
Remove Paces Insert Pages 6
6 7-7 10 10 L
34 34 35 35 t
35A 36-36 4
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C 1% delta k/k in the referenco core condition and the reactivity worth of all experiments is accounted for.
reference core condition - when the core is at ambient tempera-ture and the reactivity worth of xenon is negligible (<0.21%
delta k/k).
regulating rod - a low reactivity-worth control rod used prA-narily to maintain an intended power level.
Its position may be varied either by manual control or by the automatic servo-controller.
reportable occurrence - any of the conditions described in sec-tion 6.5.2 of these specifications, safety channel - a measuring or protective channel in the reactor safety syrtem.
safety limits (SL) - limits on important process variables which are found to be necessary to reasonably protect the integrity of certain physical barriers which guard against the uncontrolled release of radioactivity.(6)
(The principal physicel barr_er is the fuel cladding.)
scramtime-theelapsedtimebetweenreachingalimitingsafetyl uystem set point and the time when a control rod is fully inserted.
secured experiment - any experiment, experimental facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means.
The restraining forces must be substantially greater than those to which the axperiment might be subjected.
senior operator - an individual who is licensed to direct the activities of reactor operators, such an individual is also a reactor operator.
l Mendment No. 9
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shall, should and may - the word "shall" in used to denote e requirement; the word "should" to denote a recommendation; and i
the word "may" to dencte permission, which is neither a require-ment nor a recommendation.
shim / safety rods - high reactivity-worth rods used primarily to provide coarse reactor control.
They are connected electro-magnetically to their drive mechanisms and have scram capabili-ties.
shutdown margin - the minimum shutdown reactivity neestsary to provide confidence that the reactor can be made suberitical by means of the control and safety systems starting from any per-missible operating conditio.7 with the maximum worth scrammable rod and any non-scramnable control rod in their fully withdrawn positions and that the reactor will remain suberitical without further operator action.
startup source - a spontaneous source of neutrons which is used to provide a channel check of the startup (fission chamber) channel.
surveillance time intervals -
t; "iar (interval not to exceed 30 months).
ai r e.11y (J nterval not to exceed 15 months).
sv 4 annually (interval not to exceed 7 1/2 wonths).
quarterly (interval not to exceed 4 months).
monthly (interval not to exceed 6 weeks).
weekly (interval not to exceed 10 days).
4411y (must be done during the working day).
true value - the actual value of a parameter.
unscheduled shutdown - any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to con-l ditions which could adversely affect safe operation, not includ-ing shutdowns which occur during testing or check-out operations.
l Amendment No. 9
10 3.
LIMITING CONDITIONS FOR OPERATION i
3.1 Reactor C?re Parametern Appljeability:
These specifications apply to the reactivity condition of the reactor and the reactivity wcrths of control rods and experiments.
Objectives:
To ensure that the reactor can be operated safily and to ensure that it can be shut down at all times.
Specificatio.na t The reactor shall not be operated unless the following conditions exist:
(1) The fuel loading shall be such that the excess reactivity above the reference core condition will be no more than 1.5%
uelta k/k, except that the excess reactivity may be increased up to a maximum of 3.5% delta k/k for purposes of control rod calibration only.
This increase in excess reactivity above 1.5% delta k/k will be permitted no more than twice a year and for no more than five consecutive working days each time.
The reactor shall be operated only by a licensed Senior Operator when the excess reactivity is greater than 1.5%.
(2) The recreor shall be operated only when all lattice positions internal to the active fuel boundary are occupied by either a fuel element or control rod fuel element or by an experimen-tal facility.
(3) Tne minimum shutdown margin under any condition of operation with the highest worth control rod and any non-scrammable control rod fully withdrawn shall be no less than 1.0% delta k/k.
(4) The regulating rod shall be worth no more than 0.7% delta k/k in reactivity.
/cendment No. 9
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l 5.
DESIGN FEATURES 1
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Only those design features of the facility describing materials of construction and geometric arrangements, which if altered or modified would significantly affect safety (and which are not included in sections 2, 3 or 4 of the Technical Specifications),
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are included in this section.
The Safety Analysis Report contains the details necessary for establishing criteria for the following Technical Specifications.
i 5.1 Sjte and racility_JMicriotion 4
5.1.1 The Nuclear Reactor. Building is located on the east side of the University of Missouri-Rolla campus in Rolla, l
Missouri, near 14th Street and Pine Street.
5.1.2 The reactor is housed in a steel-framed, double-walled aluminum building designed to restrict leakage.
All air and other gases will be exhausted through vents in the reactor bay ceiling 30 feet (or 11 meters) above grade.
The Reactor Building free volu.te is approximately 1700 cubic meters.
5.2 Reactor Coolant Svsten 5.2.1 The minimum temperature of the reactor pool should be no 0
0 less than 15.5 C (60 r) when the reactor is operated.
5.3 Reactor Core and ruel 5.3.1 Core Configurations Various core configurations may be used to accommodate exper-1ments.
4 5.3.2 ruel Elements i
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l Amendment No. 9
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l 35 1
(1) Plate fuel elements of the MTR type are used.
The overall i
dimensions of each element are approximately 3 inches by 3 inches by 36 inches.
The active length of fuel is approxi-mately 24 inches and the fuel is clad in aluminum alloy.
The fuel elements have 18 fuel plates joined to two side plates, i
The whole assembly is joined at the bottom to a cylindrical nose pieco that fits into the core grid plate.
h The low-enriched uranium (LEU) fuel meat is U Si 3
2 dispersed in an aluminum matrix and is enriched to approximately 20%
The U.S. NRC has approved the use of LEU fuel elements l
Of this type in NUREG 1313.
i I
I (2) There are control rod fuel elements which are similar to.the elements described in (1) with the exception that the center eight plates have been removed and have been replaced with l
4 4
guide plates such that the control rod cannot come in contact with fuel plates.
(3) There are half fueled elements, which have nine LEU fueled plates (either-the front ones or the rear ones as appro-priately marked) and nine dummy (or unfueled) plates.
(4) There is an irradiation fuel element which has six fuel plate positions left blank (that is plate positions 31 through 16), plates 10 and 17 are dummy (or unfueled) and all the others (1 through 9 and 18) are LEU fueled.
5.3.3 Control Rods (1) Poison sections of the three shim / safety rods are stainless steel and contain approximately 1.,5% natural boron.
The rods have nominal dimensions of 2-1/4 inches by 7/8 inches in cross section and are L1 inches long.
(2) The poison section of the regulating rod is a stainless steel oval-shaped tube, 25 inches long, having a wall thickness of 65 mils, and is mechanically coLpled to the rod drive.
l Amendment No. 9 l'
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35A 5.3.4 Control Rod Drive Mechanisms (1) The shim / safety rod drives have a maximum vertical travel of 24 inches and a withdrawal rate of approximately 6-inches per minute. The shim /saf ety rods are magnetica1]y coupled to the drive mechanisms and drop into the core, by gravity, upon a scram signal.
(2) The regulating rod drive has a maximum vertical travel of 24 inches and a withdrawal rate of approximately 24 inches per minute.
The regulating rod is mechanically coupled to its rod drive and does not respond to a scram signal.
(3) Lights are provided on the operator's console to indicate (Go to page 36.)
i
/cendment No. 9
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36
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upper limit, lower limit, shim range, and magnet contact for each shim / safety rod.
5.3.5 Start-up Source A neutron source is available of such a strength as to satisfy the requirements that the count rate is greater than 2 counts per second during a cold reactor start-up.
5.4 Fissionable Material Storage 5.4.1 The fuel storage pit, which is located below the floor of the reactor pool and at the end opposite from the core, will be capable of storing the complete fuel inventory of either highly-enriched uranium (HEU) fuel
-or of low-enriched uranium, but not both.
The neutron multiplication factor of the fully loaded storage pit shall not exceed 0.9 under any conditions.
Mendment No. 9
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULAT!0tj SUPPORTING CONVERSION ORDER TO CONVERT FROM
,HIGH-ENRICHED TO LOW-ENRICHED URANIUM FUEL FACILITY OPERATING LICENSE NO. R-79 UNIVERSITY OF MISSOURI - ROLLA DOCKET NO. 50-123
- 1. 0 INTRODUCTION In accordance with Section 50.64 of Title 10 of the Code of Federal Regulations (10 CFR) which requires that licensees of non power reactors convert these reactors to a low enriched uranium (LEU) fuel, except under certain conditions, the University of Missouri - Rolla (Rolla or licensee) has proposed to convert the fuel in its University of Missouri - Rolla Research Reactor (UMRR or the reactor) from high-enriched uranium (HEU) to LEU.
On November 16, 1988, Rolla submitted a safety analysis report (SAR) and revised Technicc1 Specifications (TS) dealing with the changes needed to convert to LEU fuel.1-5 The staff's safety review with respect to issuing an order to convert from HEU to LEU fuel has been based on an analysis of Ro11a's SAR and the proposed changes to the TS and on information provided by Rolla on May 8, 1990,6 May 30, 1990,7 August 9, 1990,8 and October 25, 1990,9 in response to the staff's questions.
This material is available for review at the Commission's Public Document Room at 2120 L Street, N.W., Washington, D.C. 20555.
This Safety Evaluation (SE) was prepared by A. Adams, Jr., Project Manager, Division of Advanced Reactors and Special Projects, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission (NRC).
Major contributors to the technical review include W. R. Carpenter and R. W. Carter of EG&G, Idaho National Engii.eering Laboratory (INEL).
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l i 2.0 EVALUATION 2.1 Introduction The UMRR is licensed for operation at thermal power levels not to exceed 200 kW(t), using thin plate-type fuel, and is cooled by natural thermal convection of the pool water, with eventual heat-dump to the reactor room I
atn.osphere primarily by evaporation.
The licensee has proposed no changes to any reactor system or operating characteristics except for the replacement of the HEU fuel elements by new LEU fuel elements.
The following evaluations and conclusions are based on that assumption.
2.2 Fuel Construction and Geometry The standard HEU fuel elements in use at the UMRR are of a typical materials testing reactor (MTR)-type design, consisting of 10 fuel-bearing curved plates attached to supporting aluminum side platte.
Each fuel plate is a sandwich consisting of a 0.051-cm (0.020-in) thick layer of-a dispersion of aluminum-uranium oxide completely c1cd in a 0.051-cm (0.020-in) thick cover of aluminum.
The uranium in the fuel meat is tnriched to 90 percent uranium-235, and each
. plate contains 17 g of this isotope.
The standard LEU fuel elements will be of a similar design with the same outer dimensions, but will contain 18 aluminum-clad fueled plates.
Each of these plates will consist of uranium silicide (U Si ) dispersed in aluminum and 3 p completely clad in aluminum alloy.
In these plates, the fuel meat will be 0.051-cm (0.020-in) thick, and the cladding will be 0.038-cm (0.015-in) thick.
The uranium in the fuel meat is enriched to less than 20 percent uranium-235, and each plate contains 12.5 g of this isotope. 'Both the HEU and LEU cores also contain control-rod fuel. elements with several middle plates missing to provide. space for control rods.
Table I compares the geometries, materials, and fissile loadings of the current HEU fuel ano the planned LEU fuel.
Figure 1 presents a schematic cross section of the two cores.
Note that one fuel element has been moved to a different location in the LEU core to help achieve a more
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. uniform power density dt +ribution across the core and to increase the worth of the shim / safety rod, in c'idition to the change in the chemical form of the uranium, the cladding allo) on the LEU fuel will be 6061 Al instead of 1100 A1.
The Argonne National Laboratory (ANL) developed fuel elements with plates identical to the proposed UMRR LEU fuel especially for the U.S. non power reactor fuel conversion programs.
These fuel elements were te.ted extensively under relatively hostile environmental conditions in the Oak Ridge Research Reactor (ORR) with no failures having a safety significance.
The NRC reviewed and approved their performance.10 In addition to the full fuel elements and the four control-rod fuel elements, l
two partial fuel elements are planned for the UMRR to allow for more precise reactivity control.
Both the HEU and LEU partial elements are constructed the same way, with one-half of the fueled plates replaced by dummy aluminum plates.
These partial elements are to be used in locations at the edge of the core.
Because of the minimal flux or reactivity perturbation in this region, such use has been considered by the staff and judged acceptable.
The fuel elements are constructed with a cylindrical fitting that positions and supports the fael element on the grid plate.
HEU and LEU fuel elements are identical in this respect.
2.3 Fuel Storage At one end, the reactor pool is deeper for fuel storage.
The reactor pool contains two racks that can hold up to 15 elements each in a suberitical geometry.
This storage pit contains no drainage pipes, and even if all water were removed from the main reactor pool, stored fuel would still be submersed
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below at least 16 ft of water.
Measurements have shown that the neutron multipli:ation of each of these storage racks is less than 0.6 when completely filled with the standard HEU fuel elements, and there is no significant neutron coupling between them.
The LEU fuel should not be higher in reactivity than the HEU fuel.
The two racks can store th current inventory of irradiated HEU fuel 1
in a criticality-:;afe geometry with adequate heat dissipation.
In the event l
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. that both cores are at the site simultaneously, the LEU core would be expected to be in either the shipping containers or in the core.
The staff considers these fuel storage facilities to be acceptable for either core.
2.4 Critical Operating Masses of U-;'35 The UMRR HEU core contains 19 fuel elen.ents, including 1 partial element and 4 control-rod elements, as shown in Figure 1.
The current U-235 operating mass of approximately 2.87 kg is contained in 169 fueled plates and has less than the limit of 1.5 percent ok/k excess reactivity specified in the TS.
The i
operating mass of uranium-235 using the LEU fuel is predicted to be 3.66 kg, contained in 293 fueled plates.
The additional uranium-235 is required to compensate for the absorption of both epithermal and thermal neutrons in the uranium-238 of the LEU, and is achieved partially by the increase of uranium concentration indicated in Table I.
The calculated change in fuel loading is as expected, and is consistent with other conversions from HEU to LEU fuel.
Such parameters as the prompt neutron lifetime and the effective delayed neutron fraction are changed only slightly by changing from HEU to LEU, and their values are as expected on the basis of the larger uranium loading and the increased metal-to-water ratio in the LEU core.
Therefore, the staff concludes that these and other calculated results confirm the basic neutronic similarity between che HEU and the proposed LEU cores of the research reactor at Rolla.
2.5 Hydraulics and Thermal-Hydraulics The UMRR core is cooled by the natural convection of the pool water.
The UMRR has no external secondary cooling system, so the major mechanisms for heat dissipation are the evaporation of the pool water into the reactor bay and the air exhaust by a ventilation f an.
The reactor pool volume is sufficient to allow up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> continuous operation at 200 kW(t) before the water tempera-ture reaches the TS limit of 57 C (135 F).
Measurements of coolant temperatures with the HEU core showed a 12 C (22 F) temperature rise through the core, and indicate about 0.1 m/s (.3 ft/s) coolant flow.
Because the LEU fuel elements contain more plates than do the HEU fuel elements, they will allow for only about two-thirds of the area for coolant flow that was allowed by HEU.
- However,
4 5-the heat flux from a fuel plate to the coolant will decrease to about 0.56 of that with HEU.
The combination of these factors leads to coolant temperatures and fuel plate temperatures not significantly different between the two cores.
In both cases, the natural convection coolant flow is low enough so that no adverse hydraulic forces are applied to the thin fuel plates.
The staff concludes that the small changet in the thermal-hydraulics parameters between the HEU and LEU cores at the 200 kW(t) power level are well within safe limits and, therefore, acceptable.
?.6 Power Density and Power Peaking BotP the UMRR staff and the
.nputed power densities and power peaking, based on the assumed nuci a; arameters for the HEU and LEU cores.
The power dis tribution among the fuel elements is very nearly the same for the two c' res, with a small shift because of the relocation of one partial fuel element.
For both cores, the peak power density is found on an inner plate of one of the control-rod fuel elements and is caused by the thermal-neutron flux peaking in the wider water gap.
The ratio of peak-to-average power densities was about 2.2 for HEU, and increases to about 2.3 for LEU.
In a theoretic 61 sense, the slightly higher peaking dues reduce safety margins and would be a consideration if the UMRR were operating at power levels suf ficient to cause the coolant temperature to be near to departure from nucleate boiling (DNB) in the hot channel.
However, because the UMRR operates at 200 kW(t), the thermal hydraulic safety margins are very high, and the staff concludts that the slight increase in the peaking factor of the LEU core does not represent a significant safety risk.
Furthermore, the higher peaking in the LEU core was included in the calculation of the inadvertent transient accident discussed in Section 2.12.1 herein.
2.7 Control Rod Worths The UMRR has four control rods to regulate core reactivity:
three shim / safety rods that use boron-steel as the neutron-absorbing poison and a regulating rod i
that uses stainless steel as the neutron poison.
The LEU core uses the same four control rods as used for the HEU core along with all their associated hardware, drives, mounts, and other peripherals.
The calculated reactivity
6-worths for the HEU core are 2.6 percent to 3.4 percent ak/k for each of the shim rods, with a total worth of approximately 9-percent Ak/k.
The calculated worth of the regulating rod is 0.31 percent Ak/k.
The measured reactivity worths for the HEU cnre are 2.63-percent to 3.36 percent Ak/k for each of the shim rods, and 0.34-percent AL/k for the regulating rod.
For the LEU core, the calculated worths of the shim / safety rods are reduced by about 20 parcent, and the worth of the regulating rod is increased by about 60 percent.
The worths are somewhat different for the LEU core, parti, because of the different core arrangement that was made by relocation of one partial fuel element to enhance toe regulating rod worth.
In general, the measured rod worths with the HEU core for both the shim and regulating rods are higher than the calculated worths.
Therefore, the measured worths of the control rods in the LEU core should be only a little different from the HEU core and will remain acceptable to the staff.
However, the licensee will carefully take measurements of the rod worths during the startup tests with the LEU core.
- 2. 8 Shutdown Marain The NRC requires that there be reasonable assurance that a non power reactor can be shut down from any operating condition, even if the control / safety rod of maximum worth and any non scrammable rod are in the most reactive positions (fully withdrawn).
On the basis of the computed control rod worths and the authorized excess reactivity, the Rolla reactor would be subcritical by approxi-mately 3.5 percent Ak/k with these two referenced rods fully withdrawn.
This margin is significantly larger than the T5 shutdown margin of at least 1 percent Ak/k, and
- Judged acceptable by the staf f.
2.9 Excess Reactivity Additional reactivity above cold, clean critica is required to allow a reactor to perform its intended programmatic functions.
The licensee discussed the amount required to compensate for various operational losses of reactivity.
The licensee's submittal also discussed calculated changes in reactivity caused by varying the uranium-235 loading of the partial fuel elements, which is the most expeditious way for the licensee to adjust core excess reactivity.
The calculations indicate there is reasonable assurance that the UMRR can achieve I
7 the excess reactivity permitted by the TS, which is the same for the LEU and HEU cores.
The authorized maximum excess is 1.5 percent ok/L except under special circumstances that are explicitly included in the previously approved 1
T5.
Because all of the 1.5 percent ok/k excess reactivity could be inserted quickly, the licensee computed the response of the reactor to such an event.
This calculation is discussed in Section 2.12.1 herein, where the staff concludes that maintaining the same authorized excess reactivity in the LEU core as has been the case with the HEU core is even safer and, therefore, acceptable.
2.10 Reactivity Feedback Coefficients The licensee computed the moderator and fuel temperature coefficients of reactivity and the void coefficient of reactivity for both cores.
The licensee compared these values both with measured
'.as for the HEU core and with values calculated at ANL for the LEU core.
All of the coefficients are at 1 cast as negative as required by the TS.
The calculated moderator temperature coefficient for the LEU core was considerably less negative than for the HEU core, which was almost twice as negative as the measured value for HEU.
The fuel temperature coefficient is due almost entirely to the doppler effect in uranium-238 neutron capture resonances, so its magnitude is negligible for HEU, but for LEU the doppler offect adds about 20 percent to the total temperature coefficient of reactivity.
The calculated void coefficient for the LEU core is about 30 percent more negative than for the HEU core, and the measured value for HEU is about 60 percent more negative than its calculated value.
Even though there is not good agreement between the various calculations, and between measurements and calculations for the existing HEU core, there is reasonable assurance that all of the coefficients are negative over the temperature range of operation of the UMRR and, therefore, will tend to decrease reactivity as the temperature of the reactor core increases.
Therefore, the staff finds the overall results of the calculations and measurements acceptable.
During the LEU reactor startup j
testing, measurements will be required to ensure compliance with the TS.
I
8-2.11 Fission product Inventory and Containment The total inventory of fission products from VMRR operation at 200 kW(t) will not be significantly different between the HEU and LEU cores.
Furthermore, because the number of fuel elements and the power distribution are nearly identical between the HEU and LEU cores, the fission product distribution will be very similar among the fuel elements of the two cores.
A principal offference is that the fission product inventory in each plate will be less in the LEU core because of the decrease in the fissile loading of uranium-235 in each plate.
Another slight difference, between the HEU and LEU fuels is in the fission product barrier, the cladding.
As indicated in Table I, the cladding thickness is not the same, and the HEU is fabricated with 1100 Al cladding and the LEU is fabricated with 6061 Al cladding.
The staff judges this to pose no significant additional risk because the U Si -Al fuel developed by the U.S.
3 2 Department of Energy (00E) and ANL and extensively tested in the ORR was clad with 6061 Al and experienced no failures attributable to cladding of this thickness and material.10 Therefore, the staff concludes there ib reasonable assurance that the new LEU fuel will perform satisfactorily in the UMRR in containing fission products under both normal operating conditions and the most i
severe postulated transient discussed in section 2.12.1, 2.12 possible Accident Scenarios Among the various possible accidents considered by the licensee or the NRC staff at the time of the 1984 license renewal for the UMRR,11 only one could be affected by the conversion from HEU to LEU fuel.
This scenario is addressed below.
l l
2.12.1 Inadvertent Insertion of Excess Reactivity The licensee's documentation presents the results of computations using the modified pARET code 10, 12 for stepwise intertion of reactivity into the proposed LEU core.
For the first computations, the licensee assumed that only inherent l
reactivity feedback mechanisms limit the transient, and the assumed values for the coefficients are given.
Since the TS limit excess reactivity to 1.5 percent ak/k for both the HEU and LEU cores, this value was used in the stepwise
e 1
9 reactivity insertion, even though no credible means has been identified to add all of the excess reactivity in such a fashion.
The computations for the LEU core reven'ed that the nuclear excursion was self-limiting before the fuel and cladding temperatures exceed 500*C (932*F) which is at least 50'C (90'F) below the temperature at which some cladding deterioration has been observed.10 The principal reactivity feedback mechanism was the heating, expansion, and partial voiding of the moderator.
This mechanism also limits an excursion in a HEU core, as demonstrated at the SPERT project, which the licensee cites.
Moreover, the licensee's calculations included the feedback caused by the doppler effect, which is significant only in an LEU core, and that f actor increased the feedback coef ficient by about 20 percent.
The licensee compared these results with relevant SPERT experiments and demonstrated that the UMRR LEU core produces lower fuel and cladding temperatures and less energy for each transient than does the UMRR HEU core.
The licensee also calculated an insertion of 1.5 percent ok/k with the scram system functional and obtained a termination of the power excursion that was appropriately in advance of the termination for HEU fuel.
The staff concludes that the licensee's analyses are appropriate, and that the postulated reactivity insertion in the present HEU core would have produced a higher fuel and cladding temperature than would occur with the proposed LEU core.
Therefore, the risk of core damage is decreased by the conversion from HEU to LEU.
The staff also concurs that the consequences of a 1.5 percent Ak/K reactivity insertion in either core would not lead to fuci damage or a release of fission products.
3.0 Cnanges to Technical Specifications The staf f is changing a number of TS in conjunction with the conversion to LEU fuel.
The definition of " scram" is changed to " scram time" to better describe the defiaition.
15e staff is clarifying the definition of " shutdown margin" and the statement in T' 3.1 (3) concerning the minimum shutdown margin by adding the condition that non-scrammable control rods should be considered withdrawn, which is their most reactive position, when calculating compliance with the shutdown margin limit.
=
e i The staff is correcting a typographical error in TS 5.1.2 concerning the free volume of the reactor building bv changing the value from 17000 to 1700 cubic meters.
The staff is making changes to TS Section 5.3.2 concerning fuel elements.
The staff is amending TS 5.3.2 (1) and 5.3.2 (2) and is adding TS 5.3.2 (3) and 5.3.2 (4) to physically describe the new LEU fuel elements.
The staf f is clarifying TS 5.3.2 (1) and 5.3.4 (2) concerning control rod drive mechanisms to include information on the control rod's ability to scram.
Finally, the staff is clarifying TS 5.4.1 concerning fissionable material storage to state that the fuel storage pit can hold the entire HEU or LEU core, but not both at the same time.
The staff has reviewed and agrees with these changes to the TS in connection with the conversion of the UMRR to LEU fuel.
4.0 CONCLUSION
The staff has reviewed and evaluated all of the operational and safety factors affected by the use of LEU fuel in the place of HEU fuel in the UMRR.
The staff concludes that the conversion, as proposed, would not reduce any safety margins, would not introduce any new safety issues, and would not lead to increased radiological risk to the health and safety of the public.
Therefore, the conversion to LEU V Si -Al fuel, as described, is acceptable.
3 2 Dated:
March 5, 1991 l
TABLE I.
COMPARISON OF THE HEU AND LEU FUEL ELEMENTS AT THE UMRR HEU LEU Number of plates / element 10 18 Number of standard fuel elements 14 14 Number of partial fuel elements 1
1 Number of control-rod elements 4
4 Fissile Loading / plate, g U-235 17 12.5 Fissile Loading / element, g U-235 170 225 Uranium density, g/cc 0.94 3.5 Enrichment, %
90 19.8 Fuel Meat composition U 0 -Al U Si -Al 3g 3 2 Cladding Material 1100 A1 6061 Al Fuel Meat dimensions Thickness, mm 0.51 0.51
-Width, mm
~ 63
~ 63 Length, mm
~ 600
- 600 Cladding thickness, mm 0.51 0.38
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4.0 REFERENCES
1.
" Safety Analysis Report for the University of Missouri-Rolla Reactor "
i Rev. 1, October 1988.
2.
L. Covington, "A Neutronic and Thermal-hydraulic Study of the Conversion of the University of Missouri-Rolla Reactor to Low Enriched Uranium Fuel," MS Thesis, University of Missouri-Rolla, December 1988.
3.
M. Straka and L. Covington, " Study of Neutron Physics:
Conversion of the University of Missouri-Rolla Reactor to Low Enriched Fuel," Trans.
Am. Nucl. Soc., Vol. 55, November 1987.
4.
M. Straka, " Reactivity Worth of the Flooded Isotope Production Element L
in a. Core of the UMR Reactor," UMRR/88-1, Nuclear Reactor Facility, University of Mi souri-Rolla,1988.
S.
M. Straka," Reactivity Accident Analysis of the UMRR-LEU Core,"
UMRR/88 2, Nuclear Reactor facility, University of Missouri-Rolla, 1988.
4 6.
Letter from Albert E. Bolon to Public Document Room, U.S. Nuclear Regulatory Commission, " Response to Request for Additional Information,"
May 8, 1990.
7.
Letter from David W. Freeman to Public Document Room, U.S. Nuclear Regulatory Commission, " Letters Regarding U.S. Department of Energy Funding Commitments for the HEU to LEU Conversion Project," May 30, 1990.
8.
Letter from David W. Freeman to Document Control Desk, U.S. Nuclear Regulatory Commission, August 9, 1990.
9.
Letter from David W. Freeman to Document Control Desk, U.S. Nuclear Regulatory Commission, October 25, 1990.
10.
NUREG-1313. "SER Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non-Power Reactors,"
July 1988.
11.
NUREG 1086, "SER Related to the Renewal of the Operating License for the Research Reactor at the University of Missouri-Rolla."
- 12.. W. L. Woodruff, "A Kinetics and Thermal-Hydraulics Capability for the -
Analysis of Research Reactors," Nuclear Technology, 64, 196, 1984.
(
c i.
I i
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~
n ENCLOSURE 4 OUTLINE OF REACTOR START-UP REPORT AND COMPARISONS WITH CALCULATIONS 1.
Critical Mass Measurement with HEU Measurement with LEU Comparisons with calculations for both LEU and HEU.
2.
Excess (operatiosial) reactivity Measurement with HEU Measurement with LEU Comparisons with calculations for both LEU and HEU.
3.
Control and regulating rod calibrations Measurement of differential and total rod worths, and comparisons with calculations for both HEU and LEU.
4.
Reactor power calibration Methods and measurements that assure operation within the license limit.
Comparison between HEU and LEU nuclear instrumentation setpoints, detector positions, and detector output.
5.
Shutdown margin Measurement with HEU Measurement with LEU Comparisons between these, and with computations for both.
6.
Partial fuel element worths for LEU Measurements of the worth of the partial loaded fuel elements.
7.
Thermal neutron flux distributions Measurements of the core and measured experimental facilities with HEU and LEU, and comparisons with each other calculations.
8.
Results of determination of LEU effective delayed neutrons fraction, temperature coefficient, and void coefficient.
Comparison with calculations and HEU core measurement.
9.
Comparison of the various results, and discussion of the comparison, including an explanation of any significant differences that could af fect both normal operation and possible accidents with the reactor.
10.
Measurements made during initial loading of the LEU fuel, presenting suberitical multiplication measurements, predictions of multiplication for next fuel additions, and prediction and verifi:ation of final criticality conditions, l
March 5, i
Dr. Albert E. Bolon Enclosed with the Order, which is being sent to the Federal Register for publication, is a copy of the safety Evaluation RepoII snd replacement pages for the Technical Specifications.
1 Sincerely, Original signed by Alexander Adams, Jr.
Project Manager Non-Power Reactors, Decommissioning and Environmental Project Directorate Division of Advanced Reactors Office of Nuclear Reactor Regulation
Enclosures:
1.
Order / Amendment No. 9 2.
Replacement pages for Technical Specifications 3.
Safety Evaluation Report 4.
Outline of start-up report cc w/ enclosures:
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