ML20066A445
| ML20066A445 | |
| Person / Time | |
|---|---|
| Site: | 05000083 |
| Issue date: | 10/27/1982 |
| From: | Diaz N FLORIDA, UNIV. OF, GAINESVILLE, FL |
| To: | Thomas C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8211060057 | |
| Download: ML20066A445 (40) | |
Text
c NUCLEAR FACILITIES DIVISION COLLEGE UNIVERSITY OF FLORIDA j OF DEPARTMENT OF NUCLEAR ENGINEERING SCIENCES 102 NUCLEAR REACTOR BUILDING
',' ENGINEERING GAINESVILLE, FLORIDA 32611 AREA CODE 904 PHONE 392-1429
, October 27, 1982 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Cecil O. Thomas, Chief Standardization and Special Projects Branch Division of Licensing Re: University of Florida Training Reactor Facility License R Docket No. 50-83
Dear Sir:
The enclosed package contains proposed corrections for fourteen (14) pages of the new UFTR Technical Specifications as transmitted with your letter dated August 30, 1982 granting a 20 year license renewal. These corrections / changes are requested as being necessary to make these Tech Specs conform with those submitted to and verbally approved by the NRC Staff. In addition to a series of typographical and nomenclature errors, certain corrections are needed to make these Tech Specs conform to desired and approved UFTR operational requirements as well as to avoid unnecessary vagueness at several points.
The actual corrections requested (26 items) are summarized in Table 1 (Attachment I to this letter) as to where the error has occurred and the probable cause of the error. Most changes correct simple typographical .
errors, omissions or misinterpretations which render the Tech Specs vague, incorrect or incomplete at the point involved. These are marked with an asterisk (*) in Table 1. Several other changes are requested as necessary because of errors in the Tech Specs submitted with the UFTR Safety Analysis Report. These three items are marked with checks (/) in Table 1; however, as explained in Table 1, none of them represent any change in currently accepted UFTP operation.
The actual corrected pages of the UFTR Technical Specifications (Pages 2, 5, 8, 9, 10, 12, 15, 17, 20, 21, 23, 24, 28, and 35) are designated as Attachment II to this letter. All corrected sections are clearly marked by vertical lines in the right hand margin to provide easy reference. These changes will not have any significant negative effect upon the safe operation of the UFTR and do not alter the original intent of the Tech Specs. Indeed, thesecorrectionswillmaketheTechSpecsclearerandmoreeasilyapplied.P0 l t NILS J. DIAZ, DIRECTOR (904)392-1406 8211060057 821027 EMPLOYMENT OPPORTUNITY / AFFIRMATIVE ACTION EMPLOYER PDR ADOCK 05000083 P PDR
Office of Nuclear Reactor Regulation Page 2 October 27, 1982 The entire enclosure consists of
- 1. Three (3) signed originals and nineteen copies of this letter of transmittal.
- 2. Twenty-five copies of the proposed corrections to the new UFTR Technical Specifications.
We appreciate your connideration of these changes needed to complete our compliance with the licensing requirements and conditions. Please let us know if you need additional information.
Sincerely, lr j
N s h
.j Diaz Directoi of Nuclea Facilities NJDebg Enclosures m) /v n /[mhJ October 27, 1982
/ (notary public) (date) btasy M ,SteleOI E'A gy commission Esses 4 27,1985 m4 =. sw set === m
Table 1 Errors / Proposed Corrections to the New UFTR Technical Specifications Facility License R Docket No. 50-83 Page Number Discussion / Probable Cause of Error Importance 2 1. Word Interpretation (Measured versus Measuring Channel)
- has resulted in an erroneous definition which should be changed as shown.
- 2. Using the word " sufficient" in place of " insufficient"
- has led to a non-sequitur in the definition of a
" secured reactor"; the error was corrected in UFTR SAR Revision #1 to read that a secured reactor is one which has insufficient fissile fuel.
5 1. Again, UFTR SAR Rev #1 corrected the impossible require-
- ment to have water over the fuel boxes; rather, it should be over the fuel only.
8 1. Change corrects typographical error so that parentheses
- are closed (210%).
- 2. Typographical errors as well as NRC Interpretation of
- requirements for UFTR trips based on Secondary Cooling System flow rates have led to errors in the Secondary Cooling Specification which should read as now corrected.
- 3. The specification under shield tank should specify a
- low water level for rod-drop trip as in the text.
The added correction reading 6" below established normal level meets this requirement which was omitted accidentally.
9 1. Staff Interpreation of Log-N period and Power Levl
- Safety Channel scram functions has resulted in an indentation in the first entry in Table 3.2 which needs to be corrected as shown to indicate that two distinct channels are to be tested for scram function.
- 2. UFTR Staff Understanding of Safety System Operability /
Test on Loss of Primary Coolant Flow as performed in weekly versus quarterly checkouts requires that the Loss of Primary Coolant Flow scram function be checked quarterly. This scram function was erroneously reported to be performed with the daily checkout in the UFTR SAR.
Since a complete scram function test is only performed quarterly, as intended, this change is requested to be made as shown.
ATTACHMENT I s
Table 1 (Cont.)
Page Number Discussion / Probable Cause of Error Importance 9 3. To distinguish types of manual trips (See Table 3.1 of
- Tech Specs), the manual scram bar function must be
-designated here. Since the manual scram bar scram function and not the key off scram function is always tested with the daily checkout, this change to desig-nating the scram function as the " manual scram bar" is needed to meet the intent of the SAR as well as for consistency with Table 3.1, page 8 of the Tech Specs.
10 1. A typographical error in the UFTR SAR has been propagated.
For this first line to make sense, it must read " wide range drawer," as shown and not " wide range count."
- 2. The UPTR can be and has been run with no air-conditioning; /
the requirement was intended to be and should be only for adequate ventilation.
12 1. NRC Staff rewrite / interpretation omitted the word " MONITOR"
- in Air Particulate MONITOR as part of Reactor Vent System.
The full title is required for clarity in this section.
15 1. Correct usage requires that the word "Cladded" be changed to " clad" as shown.
17 1. Meaningful radioactivity surveillance must have a frequency on them. The frequency on surveillances for air-borne particulate contamination was inadvertently omitted in the UFTR SAR. Since this surveillance is performed as part of the weekly checkout of the UFTR, this frequency is included in the changed page.
The comma after monitors on item (3) makes the meaning
- 2.
clearer as corrected.
20 1. As inadvertently proposed in the UFTR SAR and written in
- the Tech Specs, " quarterly at intervals not to exceed 8 tonths" is incorrect. Since interlock testing is part of the weekly checkout, a change to this effect is proposed.
- 2. As indicated on Page 10, correction #2, the UFTR can be /
and has been run with no air-conditioning. Therefore, the item should read as proposed that the rasctor vent system is shut off "when the evacuation alarm is actuated."-
ATTACHMENT I
Table 1 (Cont.)
Page Number Discussion / Probable Cause of Error Importance 20 3. The word " staff" is a typographical error; " staff" should
- be " stack" as in " stack monitoring system."
21 1. Section 4.2.8 (2) , paragraph (c) , of the Tech Specs should
- properly be a third section designated . (3) as shown' in' the corrected page; this change clarifies this section since item (3) as proposed is a distinct item in the category of secondary water radioactivity checks and should be clearly distinguished from primary water radioactivity checks.
23 1. The specification on overall bundle size is incorrect stated;
- a careful check of drawings indicates the proposed change on line 1 of the table is correct. The values given were for the incorrect Figure 4-11 of the SAR. Updated drawings show 2.50 in.
- 2. The Tech Spec statement that all partial elements consist of 5 fueled plates contradicts the verbiage in Section 5.4 and is unnecessarily and unintentionally restrictive to the point that fuel loading following the table specification would be less safe than the proposed changed specification.
24 1. The word "alluminum" is corrected in the change to reflect proper spelling: " aluminum."
- 2. UFTR SAR Revision #1 corrected this error which is still reflected in the Tech Specs. The control blades are tipped with cadmium, not aluminum, as stated in the proposed changed page.
- 3. Control rod drive position indicators are both mechanically AND electronically geared to the rod drives; therefore, the i
Tech Specs should be changed to agree with this condition as approved.
! 28 1. The safe has cadmium " plates," not cadmium " plants" to separate layers of fuel bundles; therefore, the word change to plates is required for proper meaning 35 1. The word " alternatives" is a simple typographical error; for proper meaning the word " alternates" is needed as stated in the proposed changed page.
ATTACHMENT I
. i 1
ATTACHMENT II Technical Specifications Proposed Corrections Pages 2, 5, 8, 9, 10, 12, 15, 17, 20, 21, 23, 24, 28 and 35 i
Inhibit: An inhibit is a device that prevents the withdrawal of control blades under a potentially unsafe condition.
Measured Value: The measured value of a parameter is the value as it appears at the output of a measuring channel.
Measuring Channel: The measuring channel is the combination of sensor, lines, amplifiers, and output devices that are connected for the purpose of measuring the value of a process variable.
Movable Experiment: A movable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating, or having incore components during operation.
Nonsecured Experiment: A nonsecured experiment, where it is intended that the experiment should not move while the reactor is operating, is held in place with less restraint than a secured experiment.
Operable: A system or component is operable when it is capable of perfoming Tts intended function in a nomal manner.
Operating: A system or component is operating when it is performing its intended function in a normal manner.
Reactor Operating: The reactor is considered to be operating whenever it is not secured or shutdown. .
! Reactor Safety System: The reactor safety system is that combination of measur-ing channels and associated circuitry that forms the automatic protective action to be initiated, or provides information which requires the initiation of manual protective action.
Reactor Secured: The reactor is secured when it contains insufficient fissile material or moderator present in the reactor, adjacent experiments or control rods, to attain criticality under optimum available conditions of moderation and re-flection, pr_
(1) the reactor is shut down, (2) electrical power to the control blade circuits is switched off and the switch key is in proper custody, (3) no work is in progress involving core fuel, core structure, installed control rods or control rod drives unless they are physically decoupled from the control rods, and (4) no experiments are being moved or serviced that have, on movement, a reactivity worth exceeding the maximum value allowed for a single experiment or one dollar, whichever is smaller.
Reactor Startup: A reactor startup is a series of operator manipulations of reactor controls (in accordance with approved procedures) intended to bring the reactor to a k efr fo 0.99 or greater. It does not include control blade manipu-lations made for purposes of testing equipment or component operability within a k
eff f 0.99 or less.
2
, ., , ; y473 g .- ~
g i g
s.
' N Specifications: The limiting safety system settings shall be i j -
, s (1) Power level at any flow rate shall not exceed 125 kW.- ~' .'xs a n (2) The primary coolant flow rate shall be greater than 30 gpm'at al) powerpy ,
levels greater than 1 watt. , j- -
- e. ,
(3) The average primary coolant outlet temperature shall not exr.eed 155*PWhen 'T measured at any fuel box outlet. ;., . ,
(4) The reactor period shall not be faster than 3 sec. -
(5) The high voltage applied to Safety Channels 1 and 2 neutron chambers shall '
be 90% or more of the established normal value. -
]
4 M
(6) The primary coolant pump shall be energized during reactor operations. c (7) The primary coolant flow rate shall be menitored at the return line.
(8) The primary coolant core level shall be at least 2 in. above the fuel. .
(9) The secondary coolant flow shall satisfy the following conditions whe'nN the reactor is being operated at power levels equal to or larger than 1 kW: .
(a) Power shall be provided to the well pump and the well water flow rate shall be larger than 60 gpm when using the well system for secondary ,
cooling. ,,
Q 9E. . , A (b) The water flow rate shall be larger than 8 gpm when using the city water system for secondary cooling. 4 "
\
(10) The reactor shall be shut down when the main alternating current (ac) power is not operating.
(11) The reactor vent system shall be operating during reactor operations. ,, [
(12) The water level in the shield tank shall not be reduced 6 in. belok th [-
established normal level. _
Bases: The University of Florida Training Reactor (UFTR) limiting safety system settings (LSSS) are established from operating experience and safety considera-tions. The LSSS 2.2.3 (1) through (10) are established for the protection of the fuel, the fuel cladding, and the reactor core integrity. The primary and secon-dary bulk coolant temperatures, as well as the outlet temperatures for the six fuel boxes, are monitored and recorded in the control room. LSSS2.2.3(11)are established for the protection of reactor personnel in relation to accumulation of argon-41 in the reactor cell and for the control of radioactive gaseous ef-fluents from the cell. LSSS 2.2.3 (12) are established to protect reactor per-sonnel from potential external radiation hazards caused by loss of biological shielding.
5
1 s
. T
\ ,
Table 3.1 Specifications for reactor shfety system trips,
\s r- , Type of safety Specification
~
9 A -
system trip 4
Automatic Trips _ .
[
Period less than,3 sec \ s - Full Q
Power at 125% of full power,' ,*
, Full Full Loss of' chamber high voltage (110%)
' 7' l.oss of electrical" power to cy:ntrol Full
,s . - % console
,. ,. ce P,imary r cooling system _ ,, . Rod-drop t, ,, N <
Loss of pump power .
.g ..
. Low-water level in core (< 4?,9")
" No outlet flow .
5 Low inlet water flow (< 30' gpm)'
, +
/' Secondar.y cooli,ng system Rod-drop i
i (at power levels above 1 kW)
/ .
~
,' .i .
Loss of" flow (well water < 60 gpm, city water < 8 gpm) s Loss of pump; power
~' ~-
High, primary coolant average outlet temperature Rod-drop
, Es155*F)
Shield Tank Rod-drop Low Water level (6" below established normal level,)
l Ventilation system Rod-drop Loss of power to dilution fan Loss of power to core vent system Manual Trips Manual scram bar Rod-drop l Console key-switch 0FF (two blades off bottom) Full l
8 i
Table 3.2 Safety system operability tests.
Component or scram function Frequency Log-N period channel Before each reactor startup follow-Power level safety channels ing a shutdown in excess of 6 hr, and after repair or deenergization caused by a power outage 10% reduction of safety channels high voltage 4/ year (4-monthmaximuminterval)
Loss of electrical power to console 4/ year (4-monthmaximuminterval)
Loss of primary coolant pump power 4/ year (4-monthmaximuminterval)
Loss of primary coolant level 4/ year (4-monthmaximuminterval)
Loss of primary coolant flow 4/ year (4-month maximum interval)
High average primary coolant outlet tem- With daily checkout ture Loss of secondary coolant flow (at power With daily checkout levels above 1..kW)
Loss of secondary coolant well pump power 4/ year (4-month maximum interval)
Loss of shield tank water level 4/ year (4-month maximum interval)
Loss of power to vent system and dilution 4/ year (4-monthmaximuminterval) fan Manual scram bar With daily checkout 3.2.4 Bases The reactor control system provides the operator with reactivity control devices to control the reactor within the specified range of reactivity insertion rate and power level. The operator has available digital blade position indicators for the three safety blades and the regulating blade. The three safety blades can only be manipulated by the UP-DOWN blade switches (manual); the regulating blade can be manually controlled or placed under automatic control, which uses the linear channel as the measuring channel, and a percent of power setting control. The two independent reactor safety channels provide redundant protection and information on reactor power in the range 1%-150% of full power. The linear power channel is the most accurate neutron instrumentation channel, and provides a signal for reactor control in automatic mode. The percent of power infonnation is displayed by the linear channel two-pen recorder. It does not provide a protective function. The log wide range drawer provides a series of information, inhibit, and protection function from extended source range to full power. The safety channel 1 signal and the period protection signal are derived from the wide range drawer. The wide 9
range drawer. The wide range drawer provides protection during startup through the source count rate interlock (2 cps),10-sec period inhibit and the 3-sec period trip. The primary and secondary coolant flow rate, temperature and level sensing instrumentation provides information and protection over the entire range of reactor operations and is proven to be conservative from a safety view-point. The key switch prevents unauthorized operation of the reactor and is an additional full trip (manual scram) control available to the operator. The core level trip provides redundant protection to the primary flow trip. The core level trip acts as an inhibit during startup until the minimum core water level is reached.
3.3 Reactor Vent System These specifications apply to the equipment required for controlled release of gaseous radioactive effluent to the environment via the stack or its confine-ment within the reactor cell.
3.3.1 Specification (1) The reactor vent system shall be capable of maintaining an air flow rate between 1 and 400 cfm from the reactor cavity whenever the reactor is operating and as specified in these Technical Specifications. The reactor air cavity flow shall be periodically calibrated to minimize argon-41 releases to the environment while maintaining a negative pressure within the reactor cavity to minimize radioactive hazards to reactor personnel.
(2) The diluting fan shall be operated whenever the reactor is in operation and as otherwise specified in these Technical Specifications, at an
{
exhaust flow rate larger than 10,000 cfm.
(3) The air conditioning / ventilation system and reactor vent systems are auto-matically shut off whenever the reactor building evacuation alarm is auto-matically or manually actuated.
(4) All doors to the reactor cell shall normally be closed while the reactor is operating. Transit is not prohibited through air lock and control room doors.
3.3.2 Bases Under normal conditions, to effect controlled release of gaseous activity through the reactor vent system, a negative cell pressure is required so that any build-ing leakage will be inward. Under emergency conditions, the reactor vent system will be shut down and the damper closed, thus minimizing leakage of radioactivity from the reactor cell.
3.4 Radiation Monitoring Systems and Radioactive Effluents 3.4.1 Area Radiation Monitors The reactor cell shall be monitored by at least three area radiation monitors, two of which shall be capable of audibly warning personnel of high radiation level s. The output of at least two of the monitors shall be indicated and 10
3.4.3 Reactor Vent System The reactor vent system shall be operated at all times during reactor operation.
In addition, the vent system shall be operated until the stack monitor indicates less than 10 counts per second (cps). Whenever the reactor vent system is operat-ing, air drawn through the reactor vent system shall be continuously monitored for gross concentration of radioactive gases. The output of the monitor shall be indicated and recorded in the control room. The reactor vent system shall be inmediately secured upon detection of: a failure in the monitoring system, a failure of the absolute filter, or an unanticipated high stack count rate.
3.4.4 Air Particulate Monitor The reactor cell environment shall be monitored by at least one air particulate monitor, capable of audibly warning personnel of radioactive particulate airborne contamination in the cell atmosphere.
3.4.5 Liquid Effluents Discharge (1) The liquid waste from the radioactive liquid waste holding tanks shall be sampled and the activity measured before release to the sanitary sewage system.
(2) Releases of radioactive liquid waste from the holding tanks / campus sanitary sewage system shall be in compliance with the limits specified in 10 CFR 20, Appendix B, Table 1, Column 2, as specified in 10 CFR 20.303.
3.4.6 Solid Radioactive Waste Disposal Solid radioactive waste disposal shall be accomplished in compliance with appli-cable regulations and under the control of the Radiation Control Office of the University of Florida.
3.4 Limitations on Experiments Applicability: These specifications apply to all experiments or experimental devices installed in the reactor core or its experimental facilities.
Objectives: The objectives are to maintain operational safety and prevent damage.
to the reactor facility, reactor fuel, reactor core, and associated equipment; to prevent exceeding the reactor safety limits; and to minimize potential hazards from experimental devices.
Specifications:
(1) General The reactor manager and the radiation control officer (or their duly appointed representative) shall review and appreva in writing all proposed experiments prior to their performance. The reactor manager shall refer to the Reactor Safety Review Subcomnittee (RSRS) the evaluation of the safety aspects of new experiments and all changes to the facility that may be necessitated by the requirements of the experiments and that may have safety significance. When experiments contain substances that ir-radiation in the reactor can convert into a material with significant 12
(1) The evacuation alarm is actuated automatically when two area radiation monitors alarm high (3 25 mrems/hr) in coincidence.
(2) The evacuation alarm is actuated manually when an air particulate monitor
, is in a valid alarm condition.
(3) The evacuation alarm is actuated manually when a reactor operator detects '
a potentially hazardous radiological condition and preventive actions are required to protect the health and, safety of operating personnel and the general public.
Bases: . To provide early and orderly evacuation of the reactor cell and the reactor building and to minimize radioactive hazards to the operating personnel and reactor building occupants.
3.7 Fuel and Fuel Handling Applicability: These specifications apply to the arrangement of fuel elements in core and in storage, as well as the handling of fuel elements.
Objectives: The objectives are to establish the-maximum core loading for reactivity control purposes, to establish the fuel storage conditions, and to establish fuel performance and fuel-handling specifications with regard to radiological safety considerations.
Specifications:
(1) The maximum fuel loading shall consist of 24 full fuel elements consisting.
of 11 plates each containing enriched uranium and clad with high purity aluminun.
(2) Fuel element loading and distribution in the core shall comply with the fuel-handling procedures.
(3) Fuel elements exhibiting release of fission products because of cladding rupture shall, upon positive identification, be removed from the core.
Fission product contamination of the primary water shall be treated as evidence of fuel element failure.
(4) The reactor shall not be operated if there is evidence of fuel element failure.
(5) All fuel shall be moved and handled in accordance with approved procedures.
(6) Fuel elements or fueled dev ces shall be stored and handled out of core in i
a geometry such that the keff is less than 0.8 under optimum conditions of moderation and reflection (7) Irradiated fuel elements or fueled devices shall be stored so that tempera-tures do not exceed design values.
15
1 The Radiological Environmental Monitoring Program shall be conducted as specified below and under the supervision of the radiation control officer.
3.9.2 Radiological Environmental Monitoring (1) Monthly environmental radioactivity surveillance outside the restricted area shall be conducted by measuring the gama doses at selected fixed locations surrounding the UFTR complex with acceptable personnel monitoring devices.
A minimum of six independent locations shall be used. A review of potential causes shall be conducted whenever a measured dose of over 40 mrems/ month at two or more locations is determined and a report shall be submitted to the RSRS for review.
(2) Radioactivity surveillance of the restricted area (reactor cell) shall be conducted as follows:
(a) Surface contamination in the restricted area shall be measured by taking random swipes in the reactor cell during the weakly checkout.
Measured surface contamination greater than 100 dpm/cM beta-gama or greater than 50 dpm/100 cm2 alpha are limiting conditions for operation requiring review and possible radiological safety control actions.
(b) Airborne particulate contamination shall be measured using a high volume air sampler during the weekly checkout. Measured radioactive airborne contamination 25% above mean normal levels are limiting conditions for operation requiring review and possible radiological safety control actions.
(3) The following radioactivity surveys, using portable radiation monitors, are limiting conditions for operation:
(a) Surveys measuring the radiation doses in the restricted area shall be conducted quarterly, at intervals not to exceed 4 months, and at any time a change in the normal radiation levels is noticed or expected. Radiation exposures shall be within 10 CFR 20 limits for radiation workers.
(b) Surveys measuring the radiation levels in the unrestricted areas surrounding the UFTR complex shall be conducted quarterly, at intervals not to exceed 4 months, and at any time a change in the normal radiation levels is noticed or expected. Doses shall be within 10 CFR 20 limits for the general public.
3.9.3 Bases The bases for establishing the Radiological Environmental Surveillance Program are the established limits for internal and external radiation exposure and requirements that radiation doses be maintained ALARA.
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(d) primary coolant flow measuring system (c) primary coolant temperature measuring system (9) Following maintenance or modification to the reactor safety system, a channel test and calibration of the affected channel shall be performed before the reactor safety system is considered operable.
4.2.3 Reactor Vent System Surveillance (1) The reactor vent system flow rates shall be measured annually at intervals not to exceed 14 months, as follows:
(a) reactor cavity exhaust duct flow (1 cfm < flow rate < 400 cfm)
(b) stack flow rate > 10,000 cfm (2) The following interlocks shall be tested as part of the weekly checkout:
(a) core vent system damper closed if diluting fan is not operating (b) reactor vent system shut off when the evacuation alarm is actuated.
4.2.4 Radiation Monitoring Systems and Radioactive Effluents Surveillance (1) The area radiation monitor channels, the stack monitor, and the air parti-culate monitor shall be verified to be operable before each reactor startup, as required by the daily checkout. Calibration of radiation monitoring channels shall be performed quarterly at intervals not to exceed 4 months.
(2) The Ar-41 concentration in the stack effluents shall be measured semiannually at intervals not to exceed 8 months.
(3) Releases of radioactive liquid waste from the holdup tanks shall be monitored before discharging to the sanitary sewage system to ensure compliance with 10 CFR 20 regulations.
(4) The reactor shall be placed in a reactor shutdown condition whenever Specification 4.2.4(1) is not met.
(5) The reactor vent system shall be imediately secured upon detection of fail-ure of the stack monitoring system.
4.2.5 Surveillance of Experimental Limits (1) Surveillance to ensure that experiments meet the requirements of Section 3.5 shall be conducted before inserting each experiment into the reactor.
(2) The reactivity worth of an experiment shall be determined at approximately 1 W power level or as appropriate within limiting conditions for operation, before continuing reacter cperation with said experiment.
20
4.2.6 Reactor Building Evacuation Alarm surveillance (1) The coincidence automatic actuation of the two area monitors and the manual actuation of the evacuation alarm shall be tested as part of the weekly checkout.
(2) The automatic shutoff of the air conditioning system and the reactor vent system shall be tested as part of the weekly checkout.
(3) Evacuation drills for facility personnel shall be conducted quarterly, at intervals not to exceed 4 months, to ensure that facility personnel are familiar with the emergency plan.
4.2.7 Surveillance Pertaining to Fuel (1) The incore reactor fuel elements shall be inspected biennially at intervals not to exceed 30 months, in a randomly chosen pattern, as deemed necessary. At least two elements will be inspected.
(2) Fuel-handling tools and procedures shall be reviewed for adequacy before fuel loading operations. The assignment of responsibilities and training of the fuel-handling crew shall be performed according to written proce-dures.
4.2.8 Primary and Secondary Water Quality Surveillance (1) The primary water resistivity shall be determined as follows:
(a) Primary water resistivity shall be measured during the weekly checkout by a portable Solu Bridge using approved procedures. The measured value shall be larger than 0.4 megohm-cm.
(b) Primary water resistivity shall be measured during the' daily checkout at both the inlet and outlet of the demineralizers (DM).
The measured value, determined by an online Solu Bridge alarming in the control room, shall be larger than 0.5 megohm-cm at the outlet of the DM.
(2) The primary water radioactivity shall be measured during the weekly checkout for gross S-y and gross a activity.
(a) The measured a activity shall not exceed 50 dpm above background level .
(b) The measured S-y activity shall not exceed 25% above mean normal activity level.
(3) The secondary water system shall be tested for radioactive contamination during the weekly checkout according to written procedures.
21
prevent entrance during reactor operation. The freight door and panel shall not be used for general access to or egress from the reactor cell. This is not meant to preclude use of these doors in connection with authorized activities when the reactor is not in operation.
5.3 Reactor Fuel Fuel elements shall be of the general MTR type, with thin fuel plates clad with aluminum and containing uranium fuel enriched to no more than about 93% U-235.
The fuel matrix may be fabricated by alloying high purity aluminum-uranium alloy or by the powder metallurgy method where the starting ingredients (uranium-aluminum) are in fine powder form. -The fuel matrix also may be fabricated from uranium oxide-aluminum (U 0 -Al) using the powder metallurgy process. There shall be nominally 14.5 g of U 33$ per fuel plate.
The UFTR facility license authorizes the receiving, possession, and use of (1) up to 4.82 kg of contained uranium-235 (2) a 1-Ci sealed plutonium-beryllium neutron source (3) an up-to-25-Ci antimony-beryllium neutron source Other neutron and ganma sources may be used if their use does not constitute an unreviewed safety question pursuant to 10 CFR 50.59 and if the sources ' meet the criteria established by the Technical Specifications.
5.4 Reactor Core The core shall contain up to 24 fuel assemblies of 11 plates each. Up to six of
! these assemblies may be replaced with pairs of partial assemblies. Each partial assembly shall be composed of either all dummy or all fueled plates. A full assembly shall be replaced with no fewer than ten plates in a pair of partial assembl ies.
Fuel elements shall conform to these nominal specifications:
Item Specification overall size (bundle) 2.845 in. x 2.50 in. x 25.625 in.
clad thickness 0.015 in.
plate thickness 0.070 in.
water channel width 0.137 in, number of plates standard fuel element - 11 fueled plates partial element - no fewer than 10 plates in a pair of partial assemblies.
plate attachment bolted with spacers fuel content per plate 14.5 g U-235 nominal 23
The reactor core shall be loaded so that all fuel assembly positions are occupied.
The fuel assemblies are contained in six aluminum boxes arranged in two parallel rows of three boxes each, separated by about 30 cm of graphite. The fuel boxes are surrounded by a 5 ft x 5 ft x 5 ft reactor grade graphite assembly.
The top of the fuel boxes are covered during operations at power above 1 kW, by the use of the shield plugs and/or gasketed aluminum covers secured to the top of the fuel boxes. The devices function to prevent physical damage of the fuel, to minimize evaporation / leakage of water from the top of the fuel boxes, and to minimize entrapment of argon in the coolant water for radiological pro-tection purposes. .
5.5 Reactor Control and Safety Systems Design features of the components of the reactor control and safety systems that are important to safety, as specified under Section 3.2 of these Technical Specifications, are given below.
5.5.1 Reactor Control System Reactivity control of the UFTR is provided by four control blades, three safety blades and one regulating blade. The control bladcs are of the swing-arm type consisting of four aluminum vanes tipped with cadmium, protected by magnesium shrouds. They operate in a vertical arc within the spaces between the fuel boxes. Blade motion is limited to a removal time of at least 100 sec and the insertion time under trip conditions is stipulated to be less than 1 sec. The reactor blade withdrawal interlock system prevents blade motion which will bxceed the reactivity addition rate of 0.06% Ak/k per sec, as specified in these Technical Specifications. The control blade drive system consists of a two-phase fractional horsepower motor that operates thruugh a reduction gear train, and an electrically energized magnetic clutch that transmits a motor torque through the control blade shaft, allowing motion of the control blades.
The blades are sustained in a raised position by means of this motor, acting through the electromagnetic clutch. Interruption of the magnetic current results in a decoupling of the motor drive from the blade drive shaft, causing the blades to fall back into the core. Position indicators, mechanically and electronically geared to the rod drives, transmit rod position information to the operator control console. Reactor shu*.down also can be accomplished by 4 voiding the moderator / coolant from the cora. Two independent means of voiding the moderator / coolant from the core are provided:
(1) water dump via the primary coolant system dump valve opening under full trip conditions (2) water dump via the rupture disk breaking under pressure conditions above design value The integral worths of the individual safety blades vary from about 1.3 to 2.3%
Ak/k depending on position in the core and individual characteristics. The regulating blade worth is about 1% Ak/k. The rod worths, drive speeds, and drop-time values are sufficiently conservative to ensure compliance with the 24
5.8 Fuel Storage 5.8.1 New Fuel Unirradiated new fuel elements are stored in a vault-type room security area equipped with intrusion alarms in accordance with the Security Plan. Elements are stored in a steel, fireproof safe in which a cadmium plate separates each layer of bundles to ensure subcriticality under optimum conditions of moderation and reflection.
5.8.2 Irradiated Fuel Irradiated fuel is stored upright in dry storage pits within the reactor build-ing in criticality-safe holes.
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(8) implementation of the Emergency Plan (9) procedures that delineate the operator action required in the event of specific malfunctions and emergencies (10) procedures for floading conditions in the reactor facility, including guidance as to when the procedure is to be initiated and guidance on reactivity control Substantive changes to the above procedures shall be made effective only after documented review by the RSRS and approval by the facility director (Level 2) or designated alternates. Minor modifications to the original procedures which do not change their original intent may be made by the reactor manager (Level 3) or higher, but modifications must be approved by Level 2 or designated alternates within 14 rays. Temporary deviations from the procedures may be made by the l
senior opeiating individual present, in order to deal with special or unusual circumstanc es or conditions. Such deviations shall be documented and reported to Leve! 2 or designated alternates.
6.4 Experiments Review and Approval (1) Experiments review and approval shall be conducted as specified under Section 3.5, " Limitations on Experiments," of these Technical Specifica-tions.
(2) The experiments review and approval shall ensure compliance with the requirements of the license, Technical Specifications, and applicable regulations and shall be documented.
(3) Substantive changes to previously approved experiments with safety signi-ficance shall be made a.ily after review by the RSRS, approval in writing i by Level 2 or designated alternates. Minor changes that do not signifi-
, cantly alter the experiment may be approved by Level 3 or higher.
t (4) Approved experiments shall be carried out li accordance with established approved procedures.
6.5 Required Actions 6.5.1 Action to be Taken in Case of Safety Limit Violation (1) The reactor shall be shut down, and reactor operations shall not be resumed until authorized by the Nuclear Regulatory Commission.
(2) The safety limit violation shall be promptly reported to Level 2 or desig-nated alternates.
l l (3) The safety limit violation shall be reported to the Nuclear Regulatory Commission.
(4) A safety limit violation report shall be prepared. The report shall describe the following:
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