ML20059J908
| ML20059J908 | |
| Person / Time | |
|---|---|
| Site: | University of Iowa |
| Issue date: | 09/12/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059J902 | List: |
| References | |
| NUDOCS 9009210031 | |
| Download: ML20059J908 (9) | |
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r y SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i SUPPORTING CONVERSION ORDER TO CONVERT FROM HIGH-ENRICHED TO LOW-ENRICHED URANIUM FUEL FACILITY OPERATING LICENSE NO. R-59 1 IOWA STATE UNIVERSITY DOCKET NO. 50-116
- 1. O INTRODUCTION In accordance with Section 50.64, of Title 10 of the Code of Federal Regulations I (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 Iowa State University (ISU or licensee) has proposed to convert the fuel in
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- its Argonaut-Type Universal Training Reactor (the reactor, UTR) from high-enriched uranium (HEU) to LEU. On November 28, 1988, ISU submitted a safety analysis report (SAR) and revised technical specifications (TS) dealing'with-
. the changes needed to convert to LEU fuel.2 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 ISU's SAR and the proposed TS and on information provided by ISU on March 5 and March 7, 1990, in-response to the staff's questions'.2 This .
Ematerial is available for review at the Commission's Public Document Room at'
'2120 L. Street, N.W., Washington, D.C. This Safety Evaluation (SE) was prepared
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by T. S. Michaels, Project Manager, Division of Reactor Projects III, IV, V and Special Projects, Office of Nuclear Reactor Regulation, U.S.. Nuclear Regulatory Commission. Major contributors to the technical review include W. R. Carpenter ,
.and R. W. Carter of EG&G, Idaho National Engineering Laboratory (INEL).
l i 2.0 EVALUATION' f 2.1 General l The ISU UTR-10 is licensed for operation at thermal power levels not to. exceed 10 kW. The reactor uses plate-type fuel and is cooled by forced circulation of water'atfa nominal flow rate of 10 gpm. The licensee has proposed no changes to any reactor system or operating characteristics except for replacing the HEU i fuel elements with new LEU. fuel elements. The following evaluations and ' conclusions are based on that assumption. F e
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2.2 Fuel Construction and Geometry , The HEU fuel elements currently installed at the ISU UTR-10 contain 12 plates each, in which the fuel meat is a 92 percent enriched uranium aluminum alloy. l Each fuel plate contains approximately 22 g of U-2?5 for a total U-235 loading of about 264 g per fuel element, if no dummy fuel plates are used. The new LEU l fuel elements will have the same outer dimensions as the HEU fuel elements, but will contain 24 plates each, with the fuel meat in the form of uranium silicide (enriched to 19,75 percent U-235) dispersed in an aluminum matrix. The LEU ) fuel plates will each contain approximately 12.5 g of U-235 for a total loading 3 of about 300 g of U-235 per fuel element assembly that contains no dummy fuel plates. The geometries, materials, and fissile loadings of the current HEU l fuel elements and the replacement LEU fuel elements are shown in Table 1. The standardized LEU plates are thinner than the HEU, with a narrower water gap between the LEU plates than between the HEU plates. The resulting metal-to-water ratio for the LEU fuel element assemblies is 0.29 compared to 0.20 for the HEU assemblies. The Argonne National Laboratory (ANL) developed the fuel elements with plates and a uranium composition that are essentially identical with the proposed ISU UTR-10 plates especially for the conversion program for U.S. nonpower reactors (NPRs). The U.S. Nuclear Regulatory Commission (NRC) has reviewed and approved the performance of the ISU UTR-10 plates. These fuel elements have been tested extensively and irradiated to relatively high burnup in the Oak Ridge Research Reactor (ORR) with no failures having a safety significance3. Dummy plates are used as needed in the fuel element assemblies to achieve the required excess reactivity. The ISU UTR-10 may use up to 19 dummy plates in the LEU core to limit the licensea excess reactivity to 0.05 percent Ak/k. 2.3 Fuel Storage All of the HEU and LEU fuel at the ISU reactor facility can be stored in the 12 spaces in the core tanks, and the 16 spaces in the dry storage pit. Two
. criticality calculations performed for the dry storage pit yielded a value for k
aN$yofHEUfuel,andakthat was less than 0.41 if the water-flooded pit filled with an infinite value of 0.46 if two LEU fuel elements are placed in each of the 16 water-fl8Med pit spaces. Both of these calculations demon-strate the 7.5-inch. separation between fuel locations in the pit storage facility effectively isolates the individual pits from neutron transport interactions, thus providing a criticality-safe area to store both HEU and LEU fuel. The NRC staff (staff) thus concludes that ISU can safely store both the HEU and LEU feel in the core tanks and the dry pit storage' area. 2;4 Critical' Operating Masses of U-235 The ISU UTR-10 HEU core contains 12 fuel element assemblies with 12 fuel plates per assembly 'ncluding half-load, quarter-load, and dummy plates, as needed, to achieve the required and licensed excess reactivity. Its critical
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, - l TABLE 1. Comparison of parameters for the HEU and LEU cores at i
-lowa State University ti[11(Measured) LEM (Calculated)
GENERAL Critical mass (g U-235) 2947 3300 Excess Reactivity (% Ak/k). 0.5 0.5 A,ff 0.00645 0.00763 j Neutron lifetime, '/., (ps) 135 165 i/A,ff (s) 0.021 0.022 ' FUEL ELEMENTS i Number of assemblies 12 12 Number of' plates per assembly 12 24 , Fuel plate dimensions (in.) 26 x 3 x ').08 26 x 3 x 0.05 (cm) 66 x 7.6 x 0.20 66 x 7.6 x 0.13 Enrichment-(%) 92 19.75 ' Mass of U-235 per plate (g) 22 12.5-Water gap (in.) 0.4D 0.17 Fuel Thickness (in.) 0.04 0.02 (cm)- 0.10 0.05' Aluminum Cladding Thickness (in.) 0.02 0.015-(cm) 0.05 0.038 > Uranium Density (g/cc) 0.61 3.50 Fuel Matrix UA14 +Al U3Si2+Al ! R00 WORTH (% Ak/k) Regulating 0.12 0.26 Safety (nominal) 0.62 0.45 REACTIVITY COEFFICIENTS Temperature. Coefficient ~(% Ak/k/'C) -0.0068 -0.019 . Fuel ' Temperature' Coefficient (% Ak/k/*C) N/A -1.25 x 10-3 Void Coefficient (%Ak/k per % void) -0.172 -0.22 a
.Q 4 mass is approximately 3.20 kg uranium containing 2.95 kg U-235. The critical mass of the LEU reactor is predicted to be approximately 16.7 kg uranium containing 3.3 kg U 235 in 12 fuel element asremblies with 24 fuel plates per assembly. The core may contain up to 19 dummy plates to achieve the licensed
. excess reactivity of 0.5 percent ak/k. The increased LEU U-235 loading is ,
necessitated by the large increase in concentration of U-238, which absorbs ! low energy (thermal) and resonance energy (epithermal) neutrons, and causes a
" hardening" of th9 thermal neutron spet.trum. To increase the uranium mass, the concentration cf uranium increased.in the fuel matrix. The proposed concentration is similar to the concentration that was successfully achieved and tested by the U.S. Deportment of Energy (DOE) in the Oak Ridge Research Reactor (0RR).
l 1 The computed LEU reactor parameters reported in the ISU SAR amendment result j from computations on the 150 computer system using codes supplied by ANL. t ANL performed additional computations to furnish benchmark calculations so the - ISU computations could be verified. To perform the statistical calculations, ! ISU used LEOPARD 4 and EXTERMINATOR 5 computer codes. ISU completed transient 1 analyses using a modified version of the PARET6 7 computer code. A comparison l of the calculated prompt neutron lifetime and the effective delayed neutron l fraction for the HEU core and the LEU cores show that these parameters are very l similar. As shown in Table 1, both of these parameters have increased slightly ! for the LEU core. However, the ratios of these parameters, which determine transient response, are virtually unchanged. The ratio of neutron lifetime to beta effective is approximately 3 percent higher for the LEU' core, which 1 indicates that the rise in transient power for a given reactivity insertion.is i slightly slower for the LEU core. The staff concludes that the results contained in this documentation adequately demonstrate the basic neutronic similarity ! between the HEU and LEU cores at the Iowa State Reactor Facility. 1 ! 2.5 Hydraulic and Thermal Hydraulic Characteristics ! As noted in Section 2.2, the ISU LEU fuel plates are thinner and contain less l U-235 per plate than the HEU fuel. Because of the decreased fuel loading, each plate generates less heat. In addition, the thinner plates and cladding l- imp' rove the heat transfer to the coolant and compensate for the effect of the l increased number of LEU fuel plates in each element. Therefore, the two types j of fuel produce similar cross sections for coolant flow, with the cross section ! ! for the LEU fuel being slightly smaller (approximately 5 percent). The flow through the reactor pre remains nominally the same for the LEU as for the HEU i l l: core (10 gpn.). However, because the LEU plates and cladding are much thinner, i resista. ice to heat flow into the coolant will be less than half that of the l l current HEU fuel plates. At the licensed power of 10 kW and assuming one-half normal flow (5 gpm), licensee calculations show that cladding and fuel tempera- e tures are very close to the coolant temperature throughout the core, differing i by a maximum ~of less than 1 F. The staff finds that the hydraulic and thermal i hydraulic characteristics of the LEU core acceptable. l l
j , a 2.6 -Power Generation and Power Peaking The design of the ISU UTR-10 is such that the fueled regions contain no extraneous materials such n control rods or experiment volumes that would tend to perturb the clean profile of the core neutron flux. Therefore, the power distribution is primarily a function of core geometry. Because the HEU core and LEU core have the same geometry with respect to the size, number, and
. placement of fuel elements in the core, power peaking factors and power generation per element are also virtually the same. However, because of the changes in fuel loading and plate 'hickness, the power density in a typical LEU fuel plate will be e out-10 percen higher than in the power density of the corresponding HEU phte. The staff finds the power generation and peaking acceptable.
2.7 Control Rod Worths The ISU UTR-10 has four control rods to regulate core reactivity: two safety , rods, one shim-safety rod, and one regulating rod. All four control rods use "boral" as the neutron poison. The LEU core uses the same four control rods that are used for the HEU core along with all their associated hardware, drives, and mounts. Tne calculated worths for the HEU core are -0.12 percent Ak/k for the regulating rod and -0.62 percent Ak/k for each of the safety rods. For the LEU core, the calculated worths are -0.26 percent Ak/k and
-0.45 percent Ak/k, respectively. The LEU rod worths are fully acceptable for safe reactor operation and control.
2.8 Shutdown Margin The NRC requires there be reasonable assurance that a nonpower reactor can be shut down from any operating condition, even if the control or safety rod of maximum worth is in its most reactive position (fully withdrawn). On the basis of. the computed control rod worths and the authorized excess reactivity, the 150 reactor would be subcritical by approximately 0.66 percent Ak/k with the rod of maximum worth fully withdrawn. This margin is acceptable because it is substantially larger than the technical specification margin of at least 0.35 - percent Ak/k. To be conservative in its calculation of shutdown reactivity, 150 has not included the backup shutdown reactivity provided by draining the moderator from the core tanks, even though the operability of the fail-safe dump valve is a limiting condition for operation in the technical specifications. 2.9 Excess Reactivity Additional reactivity above the cold, clean critical state is required to allow a reactor to perform programmatic and academic. functions. The excess reactivity permitted by the ISU UTR-10 technical specifications is 0.5 percent Ak/k for both the HEU and LEU cores. The LEU core may need up to 19 dummy fuel plates to limit the actual excess reactivity to 0.5 percent Ak/k. Because the authorized maximum excess is only 0.5 percent Ak/k, inadvertent insertion of all of this excess will not allow the reactor to promptly become critical. Therefore, any credible transient power increase would be quickly terminated by a power-level automatic reactor shutdown or an operator intervention and would only slightly increase fuel temperatures (Section 2.12-1), which is acceptable for both the HEU and the proposed LEU cores.
j d l .. 6-2.10 Reactivity Feedback Coef ficients The licensee computed the temperature coefficient of reactivity and the void
' coefficient of reactivity for both the HEU and LEU cores. Both coefficients are more negative than required by the technical specifications. The void coefficient of the LEU core is significantly more negative than for the HEU core because the LEU core would be even more under moderated. The licensee's calculated temperature coefficient of the LEU core is also more negative than
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in the HEU core, partly because of the Doppler effect in broadening the neutron capture resonances of the relatively much more abundant U-23B that is present in the LEU fuel. Because the Doppler feedback occurs within the fuel, and is prompt, this feedback is more effective in countering a reactor transient in the LEU core than is the moderator temperature coefficient in the HEU core. Thus, these predic+; reactivity coefficients for the LEU core are acceptable because they are le- r and more effective in leading to reactor stability than the reactivity coefficients for the HEU core. 2.11 Fission Product Inventory and Containment The total inventory of fission products will not be significantly different between the HEU and LEU cores. Furthermore, because each LEU fuel element contains 24 plates instead of the 12 plates in each HEU fuel element, the inventory per plate will be less in the LEU core. However, the aluminum cladding is thinner on the LEU plates, which may reduce the integrity of the fission product barrier. However, cladding of this thickness has been used on HEU fuel for many years in many of the licensed research reactors. Because no failures or significant releases of fission products have been attributable to the integrity resulting from this cladding thickness, thtre is reasonable assurance that the new LEU fuel will perform satisfactorily in containing fission products in the relatively benign environment of the 10 kW ISU UTR-10 reactor.
~2.12 Potential Accident Scenarios Among the various possible accidents considered by the licensee and the NRC staff when the original license was issued to ISU, only two could be affected by the conversion from HEU to LEU fuel. These two scenarios are addressed in the following paragraphs.
2.12.1 Int ertent Insertion of Excess Reactivity In the ISU UTR-10 safety analysis report (SAR) for HEU fuel, the hypothetical
-accident begins with the sudden insertion of all the licensed excess reactivity into the critical reactor operating at full power. The reactor operator takes no corrective action before the automatic safety system trip that is activated at 150 perceret of full power. The same accident scenario is presented in the revised SAR for the LEU fuel. The power-transient analyses indicate a maximum power level of 36 kW for the HEU fuel and 26 kW for the LEU fuel. In both cases, the maximum temperature increase is less than 1 C. Therefore, postulated transients do not create conditions that would endanger personnel or render equipment inoperative. The staff has' reviewed these analyses and agrees with ISU's conclusion that the consequent parameters for a 0.5 percent Ak/k step reactivity insertion in the ISU LEU core are acceptable because they are less )
severe than in the currently licensed ISU HEU core. I
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1 1 2.12.2 Fuel Handling Accident In the ISU UTR-10 SAR8 that was issued in August 1981 and submitted to the NRC in November 1981, the licensee considered the maximum hypothetical accident (MHA) to be an accident that results in damage to the structure of the fuel plate in the reactor room air and that is caused by the equivalent of one fuel plate being stripped of its cladding. The licensee also considered this same accident to be the MHA in the revisions to the SAR for conversion to LEU fuel. The staff has concurred that thit is a conservative and acceptable MHA. The only significant difference between the inventory of radioactivity in the LEU and the HEU fuel is that the Pu-239 formed by neutron capture in the U-238 is much more abundant in the LEU fuel, nowever, because cf the licensed power level and consequent burnup of fuel at ISU, the additional buildup of the inventory of plutonium in the LEU fuel is still radiologically insignificant. The release of the fission products, including this plutonium, from one damaged LEU fuel plate could result in maximum radiation exposures in the unrestricted areas only slightly higher than those exposures reported in the 1981 SAR and still only about 5 percent of the 10 CFR Part 20 guidelines. Therefore, damage
' LEU fuel in place of HEU fuel would cause no significant change in the risk to $ % health and safety of the public, which was already acceptably low for the t.. ' eat 4EU core.
3.0 CR . ION 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 ISU UTR-10 reactor. The staf f 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
4.0 REFERENCES
- 1. Iowa State University, Amendment to the SAR for the UTR-Y Reactor Reactor Facility at Iowa State University, Docket No. Ld 116, submitted to the U.S. Nuclear Regulatory Commission in accordance with requirements of-10 CFR.50.64, November 1988.
2. Letter from Richard A. Hendrickson, to T. Michaels, U.S. Nuclear Regulatory Comn.ission, " Questions Regarding HEU/ LEU Conversion at Iowa State University," March 5, 1990. l 3. R. R. Hobbins, et al, Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Nonpower Reactors, NUREG-1313, February 1988. l-l l L l l l l'
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4., R. F. - Berry, LEOPARD - A Spectrum Non-spatial Depletion Code for the IBM-7094, WDD-3741, Westinghouse Electric Corporation, Atomic Power Division, 1963. 5. T. B. - Fowler, et al. EXTERMINATOR - A Multioroup Code for Solvino Neutron and' Diffusion Equations in One and Two Dimensions, ORNL-TM-842, Oak 3 Ridge National Laboratory, 1966. l i l
- 6. C
. F. Obenchain, PARET - A Program for the Analysis of Reactor Transients, I00-17282, Idaho National Engineering Laboratory,1969. I 7.
W. L. Woodruff, "A Kinetics and Thermal-Hydraulics Capability for the Analysis of Research Reactors," Nuclear Technology 64, pp. 196-206, 1 (February 1984); and W. L. Woodruff, "The PARET Code and the Analysis of SPERT I Transients," Proceedings of the International Meetino on Research and Test Reactor Core Conversions f rom HEU to LEU Fuel, Argonne National-Laboratory, 1983, Argonne, IL, November 8-10, 1982, ANL/RERTR/TM-4, CONF-821155, pp. 560-578. 8. R. A. Hendrickson, et al. Safety Analysis Report for the Trainina Reactor UTR-10, Iowa State University, August 1981. r r I 1 i r l
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,j. 8 ATTACHMENT TO LICENSE AMENDMENT NO. 8 TO FACILITY OPERATING LICENSE NO. R-59 DOCKET NO. 50-116-The Technical Specifications (TS) have been reformatted in their entirety, Changes made to the TS are identified by vertical lines indicating the areas of change.
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TECHNICAL SPECIFICATI0ilS i for the UTR-10 REACTOR FACILITY. , at IOWA STATE UNIVERSITY - 4
' Docket No. 50-116
, License No. R-59 ,
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Original: August 1983 - Amendment 1: November 1909 1 t 1' h
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, 7 1.0 DEFINIT 10NS' The term; Safety Limit, limiting Safety System Setting, and Limiting Condition for Operation are as defined in paragraph 50.36 of 10 CFP Part 50.
CHANNEL TEST - The introduction of a signal into the channel for verification
-that it is operable. .
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CHANNEL CALIBRATION - The adjustment of the channel such that its output corresponds with acceptable accuracy to known values of the parameter'which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall- be deemed to include a Channel Test. r CHANNEL CHECK.- A qualitative verification of acceptable performance by
- obiervation of channel behavior. This verification, where poss'ible, shall include'the comparison of the channel with other independent channels or systems measuring the same variable.
CONFINEMENT B0UNDARY - The surface surrounding the reactor facility defined by the interior partition walls of' offices and laboratories on the north, east
, - and south sides of the building and by the west interior wall which isolates the basement, first floor, and the west corridor of the second floor from the centr'al bay.
CONTROL R0D - A plate fabricated with Boral as the neutron absorbing material which-is used to establish neutron flux changes and to compensate.for routine
- reactivity losses. This in.cludes safety-type and regulating rods.
CORE - The portion of.the reactor volume which includes the graphite reflector, core tanks, and control rods. The thermal column and shield tank duct are not included. 1-1 4 e v , -, - r ,
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DELAY TIME - The elapsed time between reaching a limiting safety system setpoint and the initial movement of a safety-type rod. ' DELAYED NEUTRON FRACTION - When converting between absolute- and dollar-value reactivity units, a beta of 0.00763 is used. DROP TIME - The elapsed time between reaching a limiting safety system sctpoint and the full insertion of a safety-type rod. EXCESS, REACTIVITY - That amount of reactivity that would exist if all control rods (control, regulating, etc.) were moved to the maximum reactive condition from the point where the reactor.is exactly critical. EXPERIMENT - Any operation, hardware, or target (excluding devices such as , detectors, foils, etc.) which is des'igned to investigate non-routine reactor characteristics or which is intended for irradiation within the core region, on or in a beam port or irradiation facility and which is not rigidly secured to a core or shield structure so as to be a part of their design. - MEASURED VALUE - The value of a parameter as it appears on the output of a
- channel.
MEASURING CHANNEL - The combination of sensor, line, amplifier and output devig,es which are connected for the purpose of measuring the value of a parameter. MOVABLE EXPERIMENT - An experiment 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. ODERABLE - A component or system 1.s capable of performing its intended function.
' OPERATING:- A component or system is performing its intended function.
1-2 . i
- i. : p REACTIVITY LIMITS - Those limits imposed on reactor core excess reactivity.
Quantities are referenced to a Reference Core Condition, f
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REACTIVITY WORTH OF AN EXPERIMENT - The maximum absolute value of the ' reactivity change that would occur as a result of intended or anticipated changes or credible malfunctions that alter experiment positian or a
. configuration. '
REACTOR OPERATING - The reactor is operating whenever it is not secured or shutdown. REACTOR OPERATOR (RO) - An~ individual who is licensed to manipulate the controls of a reactor. L REACTOR SECURED - A reactor is secured when: i (1) It contains insufficient fissile material or moderator present in the reactor to attain criticality under optimum availible conditions of moderation and reflection, or (2) A combination of the following:
.a. The minimum number of neutron absorbing control rods' are fully inserted-or other safety devices are in shutdown position, as required by technical specifications, and r s b. The magnet power keyswitch is in the off position and the key is removed from the lock, and
- c. 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 p d. No experiments in or near the reactor are being moved or i serviced that have, on movement, a reactivity worth exceeding l the maximum value allowed for a single experiment or 0.763% l Ak/k whichever is smaller. I 1-3 )
's REACTOR SHUTDOWN ~ The reactor is shutdown if it is subcritical by at least.-
0.763% Ak/k in the Reference Core Condition and the reactivity worth of all experiments'.is accounted for. - 1
- REACTOR SAFETY' SYSTEMS.- Those systems, including their associated input ,
. channels, which are designed to initiate automatic reactor. protection or to provide information for initiation of manual protective action.
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READILY AVAILABLE ON CALL - Applies to an individual who: (1) Has been specifically designated and the designation known to the operator on duty, and i (2) Keeps the operator on duty informed of where he or she may be rapidly contacted (e.g., by phone, etc.), and i (3) It capable of getting to the reactor facility within a reasonable time under. normal conditions (e.g., 30 minutes). REFERENCE CORE CONDITION -'The condition of the core when it is at ambient-temperature (cold) and the reactivity worth of xenon is negligible, less than 0.23%'Ak/k. REGULATING ROD - A low-worth control rod used primarily to maintain an intended power-level that does not have scram capability. Its position may be varied manually or-by the servo-controller. SAFETY CHANNEL - A measuring or protective channel in the reactor safety system.; r SAFETY-TYPE-ROD - A rod that can be rapidly inserted by cutting off the holding current in its electromagnetic. clutch. This applies to safety #1, safety:#2, and' shim-safety. 1-4
i SECURED EXPERIMENT - Any experiment, experiment 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 gr' ester than those to which the experiment might be subjected by hydraulic, pneumatic, buoyant, or other forces which are normal to the operating environment of the ; experiment, or by forces which can arise as a result of credible malfunctions. SENIOR REACTOR OPERATOR (SRO) - An individual who is licensed to direct the activities of a Reactor Operator (RO) and to manipulate the controls or a reactor. SHALL, SHOULD, AND MAY - The word "shail" is used to denote a requirement, the word "should" to denote a recommendation, and the word "may" to denote , permission, neither a requirement nor a recommendation. l - SHUTDOWN MARGIN - The minimum shutdown reactivity necessary to provide , [ confidence that the reactor can be made suberitical by means of the' control dnd safety systems starting from any permissible operating Condition although the most reactive rod is in its most reactive position, and that the reactor will remain subcritical without further operator action. TRUE VALUE - The actual value of a parameter or variable. UNSC' H EDULED SHUTDOWN - Any unplanned shui.down of the reactor caused by
- actuation of the reactor safety system, operating error, equipment malfunction, or a manual shutdown in response to conditions which could ,
adversely affect safe operation, not including shutdowns which occur during testing or check-out operations. s 4 e 1-5
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>2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ,
2.1 Safety limits- ! 2.1.1 Appl _icability These specifications apply to the variables that affect thermal and thy,draulic p'erformance of the core. . 2.1.2 Objective To assure fuel cladding integrity. 2.1.3 Specifications A. The' true value 'of the steady power level under various. flow . , conditions 'shall not" exceed 15 kilowatts. B. The true value of the' primary coolant flow rate shall not be , less than 3.5 gpm for periods. greater than 5 minutes at all power levels greater than one kilowatt. C. The true value of the primary coolant outlet temperature shall not exceed 180 0 s F N l* ' l~ 2.1.4 Bases i Specifications A and B provide limits which protect the fuel cladding i from damage'due to excessive heat flux and surface temperature if the primary. coolant pump fails. There is sufficient time for the operator to take i corrective action before' saturated pool boiling begins since the rate of temperature rise.is approximately 9.8 0F per hour per kilowatt (SAR: 6.2); the time to increase-from the maximum allowable core inlet temperature of.160 0F to the boiling temperature when operating at 10 kilowatts would be t - 2-1
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- approximately 32 minutes. Even if boiling did occur, the maximum critical heat flux ratio (critical heat flux divided by the maximum heat fl0x i'n the core) is so large (on the order of 1000) that damage to the cladding would be -
very unlikely. Specification C provides a limit for core outlet coolant temperature under forced convection cooling. If the primary coolant flow rate was as low as 3.5 gpm,and the core inlet temperature was 160 0F at 10 kilowatts, the l temperature rise across the core would be nearly 20 0, F As coolant temperatures reach 180 0 F (which is also the dump tank limit) and above, the corrosion rate increases, thus accelerating the loss of fuel plate cladding. O 4 b n 2-2 4 w- - m
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS (continued) 2.2 Limitina Safety System Settinas 2.2.1-Applicability This specification applies to the setpoints of safety channels. 2.2.2 Objective To assure that automatic trip action is initiated and that the
' operator is warned to take protective action to prevent a safety limit from being exceeded.
2.2.3 Specifications ' i L v The limiting safety system settings are the following:
. a A. Maximum power level trip setpoint shall not exceed 12.5 I kilowatts.
i B. Minimum primary coolant flow rate trip setpoint shall not be 'I less than 5 gpm. s C. Maximum primary coolant outlet-temperature trip'setpoint shall not exceed 170 0F, 1. 2;2.4 Bases The trip setpoints provide adequate margins for the limits specified
,in 2.1.3. Trip-setpoint-A initiates. automatic scram. Trip setpoints B and C lc initiate alarms signaled-by'a horn and lighted annunciator. Operator intervention in the non-scram trips provides timely response due to the slow variation of temperature even in the most adverse case discussed in 2.1.4.
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-3.0 LIMITIh'c CONDITIONS FOR OPERATION
~ 3.1 Reactor fatt Parameters-i P
3.1.1 Applicability Th'ese specifications apply to the parameters which describe the i reactivity condition of the core. !
'3.1.2 Objective i
To ensure that the. reactor can not achieve prompt criticality and'that o it can be safely shut down under any condition. ' 3.1.3 Specifications .
. The' reactor snall not be made critical unless 'the following conditions exist: ^
A. The total' core excess reactivity with or without experiments [ shall' not exceed 0.50% Ak/k. j L B. The. minimum shutdown margin provided by control rods in the :- ' reference core condition shall' not belless than 0.35% Ak/k. 4 E 'C. The fuel loading pattern and experiment apparatus inserted in the [ core shall be approved by the Reactor Use Committee. 3.l.4 Bases
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Specifications A and B are based on values used in the power transient.
- analysis (SAR: 6.3) where it is assumed that all of the excess reactivity, ,
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4
.i 0.50% Ak/k, is suddenly inserted as a positive step function. The; safety I system response 10 assumed to result in the minimum shutdown margin, 0.35% Ak/k, being st.pplied by rapidly inserted safety type control rods, assuming the rod with the-greatest worth is not available, 6 Specification f. limits the changes in core configura*. ion to those
- approved by the committee. charged with review and approval of experiments.
t
?
i i s e 'I 3-2
- , ;,. .. 0 y 1 3.0LIMITINGCONDITIONSFOROPERATION(Continued) !
s 3.2 Reactor Control and Safety System .
--3.2.1 Applicability
)
1 These. specifications apply to the reactor safety system and safety- j related instrumentation. J
' j 3.2.2 Objective 1
To specify the lowest acceptable level of performance or the minimum-number of acceptable components for the reactor safety system and safety-related instrumentation. 3.2.3LSpecifications l The reactor sball not be made critical unless the following conditions exist:' A. Tne reactor safety channels and safety-related measuring channels ,
- f. shall be operable in accordance with Table 3-1, including the h minimum number of channels and the indicated maximum or' minimum.
g '. setpoints. ' l [ B. All three safety-typeicontrol rods shall be operable and have L the following response time capabilities:
- (1) Delay time shall not exceed 100 milliseconds. ,
(2) Drop time shall not exceed 600 milliseconds. l l C. The reactivity insertion rate for a single control rod shall not. exceed 0.019% Ak/k/sec. 3-3
i_j,.. h. ' .i s c, , b . D. The dump valve shall be operable and shall be capable "o'f reaching its normally-opened position in not more than 600 ' milliseconds after the scram signal is initiated. ! E. The following bypasses may be applied to the channels indicated
, provided the appropriate compensation is employed:
(1) During measurements of control rod. worth, the startup j sequencc for removal of safety-type rods with no position indication may be altered if elapsed Othdrawal times are . observed as the rod that establishes criticality is maneuvered. 1 (2) During measure'ments of reactor thermal power or control rod , worth, the signal from the multirange linear power channel
, neutron detector may be used exclusively for measurement data recording if another detector of equivalent-
- c characteristics is used as a substitute.
t 3.2.4 Bases . - T Specification A provides assurance that the reactor safety. instrumentation channels which may be needed to shutdown.the reactor-are operable. In addition, other channels which are important to safe operation-because of interl'ock or alarm action are included.- Each channel, along with :! the setpoint, minimum number required,: and function,. is listed in Table'3-1.
--- The control rod withdrawal inhibit assures that the ' operator has an- ,
operable channel and appropriate neutron flux levels during startup. .
-- The' integrity of the startup neutron source is protected, and excessive radiation levels are avoided by the coincident power and source / closure scrams.
---The period scram limits the rate of power-level increase to values-which are manually controllable without ' reaching excessive power levels or fuel f
4 3-4
. = i- -
wmperatures.
-- The linear percent power scrams provide automatic protective a6 tion to prevent exceeding the safety limit (2.1.3 A) on reactor power, j
-- The multirange linear power channel provides information to guide the !
-operator in establishing a set power level with greater precision than that available from other power level monitoring channels.
-- The scr.am derived for the loss of high voltage to the neutron detectors '
provides a conservative response to an instrumentation system failure. The , recommended operating voltage serves as the guide to detect a significant loss in pow 6r supply potential.
-- The alarm response to a fault in the scram circuit provides notice to the operator that the sc*am bus may not be operable if a subsequent fault develops.
Operators are directed by procedure to shut down the reactor when this alarm is noted.
-- The moderator level channel inhibits control rod withdrawal until the '
moderator reaches an appropriate level above the fuel plates during startup-operatio,n. This minimum level restricts variations in moderator level at startup which could produce *significant changes in reactivity balance and neutron detector response. (See also 3.3.4) -
-- The moderator high level scram provides automatic shutdown and the
. subsequent draining of the moderator from the core tanks if the level exceeds the setpoint. Accidental floqding of the graphite reflector and uncontrolled loss'of coolant are avoided.
-- The shim-safety position indicator channel must be operable to permit the operator to determine the excess reactivity from the critical rod position and' rod calibration information.
-- The earthquake scram is provided to put the reactor in a shutdown condition before the protection system components are subjected to forces
-which might make them inoperable.
-- The manual scram and the magnet power keyswitch provide two methods for the reactor operator to manually shut down the reactor if an unsafe or abnormal condition should occur.
Specification B is based on values used in the power transient analysis (SAR: 6.3) where it is assumed that two safety-type control rods are 3-5 , O _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ . . . - -,-__.t. - - - . - - - - -- ----%
I i inserted as a ramp function. The safety system is assumed to initiate rod . motion within 100 milliseconds after reaching the limiting safety' system setting and to have the rods full inserted within a total time (delay plus
-insertion) of 600 milliseconds. ,
Specification C is based on a conservative value used for many years as a limit for reactors of the same type as the UTR 10. The 'imit assures a-safe rate of power change during startup and during power ascensions. Specification 0 assures that the moderator can be drained from the core tanks following a scram and provide backup shutdown action. When the y dump valve opens in 600 milliseconds or less, the water is drained from the core tanks in approximately 4 seconds. Specification E provides for bypasses of. a startup interlock and a normal instrumentation signal connection.
-- The startup sequence requires a fixed order of safety-rod removal: Safety
#1 full out, safety #2 full out, theri partial removal, depending on the excess c l
reactivity, of = either the shim-safety or regulating rods. To measure the I maximum worth of the shim-safety, the startup sequence interlock may be bypassed to allow removal of a safety rod and then the shim-safety. The remaining safety _ rod (neither safety rod is equipped with. intermediate position indication) can be safely maneuvered to the critical position by keeping cumulative withdrawal time. l
-- The normal connection of the multirange linear power channel can- be bypassed with no redu'ction in the performance capability of the channel by usindanother detector of equivalent characteristics, located in another but comparable position with relation to the fuel region, as the signal source for ,
- power level information. The changeover is completed at low power,' and any change in calibration factor noted for later use at higher power levels. This-bypass is used to obtain detector current information at high power levels for '
therma 1 power measurements and calibrations, and for control rod worth measurements. 3-6
-- ~ --
='
s, , t
~
' i l:- ;
' Table'3-1. Required Safety Channels and Safety-Related Channels. <
. Channel' Setpoint Min. Operable . Function u
NUCLEAR ' I Lbg.% power . Inhibits control l
. Min. Count, rate. 20.1 mW 2 rod withdrawal.
i
~ Power level 31_W 2 Scram iff aU closures are not' 1 L
, seated o'-scurce is not stored. ]
Period' 25 seconds 2 Scram
^
o Linear % power s12.5 kW 2 Scram Q I' Multirange linear power -- 1 Power information. /
-High. voltage loss. ,
to neutron detectors 290% V(a) ~2 Scram y .; l- Scram circuit failure, . Fault to gnd 1 Alarm
^
PROCESS i Moderatorlevel(b). Inhibi.ts rod
; Normal op level 242 inches 1- ' magnet current.
High level 555 inches 1 Scram Shim-safety ~ -- 1 Excess reactivity'
' position :information. 1
-Earth' quake- -s4 Richter 1 Scram i MANUAL Manual scram switch -.. 1 Scram. -
Magnet, power keyswitch -- 1 Scram
. (a)Recommendsdoperatingvoltage.
(b) Measured from the core tank base' plate.
. a 4 h
. h 3-7 :
s >. 7
.__a
l
-3.0 LIMITING CONDITIONS FOR OPERATION (Continued)
. i
- 3.3 Coolant Systems 3 d.1 Applicability These specifications apply 'o the minimum operating equipment and .
limits of operation for the cooling system. 1 3.3.2 Objective ' To ensure that the reactor fuel can be adequately cooled with water of high quality. ;
'3.3.3 Specifications t
The reactor shall not be made critical unless the following conditions 4 exist: ' A. The coolant system instrumentation channels shall be operable in . accordaned with Table 3-2, including the minimem number of
- i channels and the indicated setpoints. ,
p s B.-The primary coolant inlet temperature shall.be. maintained in the range from a lower value determined by.sptcification 3.1.3 A to 150 0F, arid the primary coolant outlet temperature shall not exceed 160'0F ~ C. The primary coolant flow rate. shall be maintained in the range from 5 to 15 gpm, except that the flow rate may'be less th'n a 5 gpm if the power level is less than one kilowatt and an approved experiment requires the reduced flow condition. D. The primary coolant temperature at the deionizer inlet shall not 3-8 4 4 - , , . . - , . _ ~ , _ -
-- . .-- -- - . . - ~ .- .- - . -. . .
. g 1. ? , c L x .
j[ i g exceed 140 0F, and the deionizer flow shall be cut off until the m temperature is below the limit. f t
-E. The primary coolant conductivity shall not exceed 2 micromhos per '
centimeter, except for periods not to exceed 7 days when the value shall not exceed 10 micrombos per centimeter. F. The radiation exposure rate observed by the deionizer column detector shall not exceed five times the nominal value measured ] during normal full power operation, or the exposure shall not ! exceed 10 milliroentgens in one hour, whichever is the smaller value, i s G. The net detection rate of confirmed fission product activity in the primary coolant shall not. exceed the " decision" limit ~for the detection system used in the analysis. aa 3.5.4' Bases' . Specification A provides assurance that the cooling system _ h
' instrumentation channels are operable. Each channel, along with the setpoint, minimum number required, and function is' listed in Table 3-2.-
-- The moderator (primary coolant) level channel inhibits control rod-withdrawal until- the moderator reaches an appropriate level above the fuel ,
.platesdhringthestartupoperation. This minimum level interlock ensures -
ample coolant level to provide heat transfer for,the fuel plates, _ and-it also .- restricts the variations-in moderator level at startup which could produce i significant changes in reactivity balance and neutron detection rates (SAR: L4.5.3). ,
-- The primary cool'nt a inlet-temperature channel permits compliance with specifications 3.1.3 A and B by initiating a low-level alarm and providing the- .
Joperator with information to establish a minimum coolant temperature which-avoids an: excessive reactivity inventory.
. 'i The primary coolant' outlet temperature channel initiates an alarm signal Y
3-9 , f 1 i 0
t,. ,
- \
at the high temperature setpoint of 160 0F. This provides an adequate margin to avoid the the safety limit specified in 2.1.3 C. ;
- The primary coelant flow rate channel initiates a low flow alam to we.rn the operator to reduce power in compliance with safety limit 2.1.3 B, if the power level is at or above one kilowatt (SAR: 6.2).
- The primary coolant conductivity channel initiates an alarm when the specific conductance exceeds 2 micrombos per centimeter. Operation may cor)tinue at a high'er level for a limited time as. indicated in 3.3.3 E.
- The radiation equipment detector located near the deionizer initiates an .
alarm when the exposure rate exceeds five times the nominal value observed i during normal full power operation (See 3.7.4). . Specification B is based on values of primary coolant temperature : which must be maintained to avoid violating the limit on excess reactivity (3.1.3 A) at the lower end of the range, and to ' avoid the high temperature safety limit 2.1.3 C (180 0F) which also is the limit on the d>/mp tank (SAR: 4.3.2). Specification C provi6s a range on the primary coolant flow rate - which will adequately cool the fuel plates and avoid safety limit 2.1.3 B, and also provide fl u ibility for low power experiments which may require an essentially stagnant coolant. It incorporates, through its lower limit of 5 , gpm, an implied coolant leak detection provision since a significant loss of prletry coolant (which is held in the process pit until analyzed) reduces the suction head on tiie pump to the point where the minimum flow rate cannot be maintained. Specification D provides a limit to prevent damage to the deionizer [ resins and possible transport of fractured resin beads past the filter and into the primary coolant stream. The flow through the deionizer will have to be restored at a temperatura below the limit if the conductivity limit is approached. Specification E is based on experience at many facilities with similar coolant systems; this value is known to be a satisfactory upper limit for ) normal operations. Trace mineral activation products do not exceed acceptable limits and corrosion rates are negligibly low when the upper limit is not l exceeded (SAR: 4.2.2, 4.5.3 and 6.1.2). Provision for conductivity transients
. t 3 - 10
. . - , , - - - . . - . - - , . _ . - . - . - ~ , - -
i l l due to crud releases adds flexibility to the limit. Specification F is based on the assumption that the increase in exposure level is due to either fission product activity or radioactive trace - minerals normally present in the primary coolant being concentrated in the : deionizer column. The trip setpoint is based on local conditions and must be determined so that it detects significant activity with respect to normal detection rates wit. rut causing too-frequent false alarms. Since the deionizer . column is located near the boundary of the restricted area, the 10 mR/h upper limit provides a conservative margin to avoid exceeding the requirements of . paragrtph 20.105 of 10 CFR Part 20 on radiation uoses in unrestricted areas. Specification G provides a limit based on statistical hypothesis testing and it depends on the detection system being used to evaluate the coolant sample. The term used in NCRP Report No. 58, pp. 275-279, is the
" decision limit", and it can be used to determine if the not detection rate of the sample is st tistically different from background at a confidence level of 95%, when equation (7.8) is used. The background in this case is taken to be the detection rate of samples without fission product activity. When the sample detection limit does exceed this limit, the leaking fuel assembly must be identified and removed from the reactor (see 3.7.3 C).
l % l l i l i f l. I l 3 - 11 9
+ --
l Table 3-2 Required Coolant System Instrumentation Channels. Channel Setpoint Min. Operable Function Moderatorlevel(a) Normal op level 142 inches 1 Inhibits rod m9gnet current; -
<,tablishes
'rinum coolant !
level. Primary coolant inlet as required to 1 Thermal power temperature satisfy 3.1.3 information i Primary coolant outlet temperature sl60 0F 1 Alarm Primary coolant flow rate h5 gpm 1 Alarm Primary coolant. conductivity s2 micromhos/cm 1 Alarm Radiation level As required to 1 Alarm Dstonizer unit satisfy 3.3.3 F (a) Measured'fromthecoretankbaseplate. N 4 I l i 1 3 - 12
l
~
3.0 LIMITING CONDITIONS FOR OPERATION (continued)
; )
3.4 Confinement l i 3.4.; Applicability i Thi.s specification applies to the operations that require confinement and to the equipment needed to achieve confinement. l 3.4.2 Objective l To ensure that the confinement boundary can be secured when needed. : i 3.4.3 Specifications 1 A. The reactor confinement boundary shall be operable whenever the reactor i.s' operating. B. The reactor confinement boundary shall be secured during fuel transfer operations. l 3.4.4 Bases + l i s Specification A is based on the assumption that the doors and windows located in the building walls that define the confinement boundary may need to be secured due to the accidental release of radioactive material generated during reactor operation. i Specification B is based on the hypothetical acetdent (SAR: 6.4) that , occurs during movement of a fuel assembly and the importance of having the confinement.boundhry secured prior to the fuel transfer operation. e S 3 - 13 , t
,_. - , - - , - - - - ~ " ' ' ~
- l i
I 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) 3.5 Ventilation systems , 1 There is no forced air circulation system in the reactor room or the building housing it. i 9
?
l - \ . s 1 4 l , l' O 3 14
3.0 LIMITING CONDITIONS'FOR OPERATION ~(Continued) . 3.6 Emeroency EME ! 3.6.1 Applicability I ( These specifications apply to the emergency power supply for the radiation monitoring system. 3.6.2 Objective To specify the source of emergency electrical power and the minimum ' operating time. 3.6.3 Specifications i t
. j l
The reactor shall not be made critical unless the following conditions ' exist:' . l A. The battery-powered standby AC power supply for the radiation 4 u monitoring system shall be operable and shall have the following ' operating time capabilities: s (1) Operating time without the radiation-evacuation horn being ' I activated shall be not less than eight hours. (2) Operating time with the radiation-evacuation horn being activated shall'be not less than two hours. , 3.6.4 Bases Specification A requires that the standby AC power system, which ! consists of at least two lead acid storage batteries', a charger-transfer. unit, L 3 - 15 9
f and an inverter, be capable of providing a tripless switchover for supplying AC power to the radiation monitoring systems, and that the power source be able to sustain operation for the specified intervals. There are no systems, other than radiation monitoring, that need emergency power since the reactor is automatically shutdown when AC power failure occurs. The radiation ' evacuation horn imposes a large incremental load on the power source and , severely re. duces the operating time; however, the evacuation signal, if I needed, would be of sufficient duration to accomplish its intended purpose. : 4 I e 0 1-t 4 3 - 16 1 .
. i
i {
. i 3.0LIMITINGCONDITIONSFOROPERATION(Continued) 'i
- i 3.7 Radiation Monitorina Systems d Effluents l
3.7.1 Applicability i I These specifications 6pply to the radiation monitoring systems and to j the limits on effluent releases. . 3.7.2 Objective , ; To specify the minimum number of acceptable components or the lowest -
- acceptable level of performance for the radiation monitoring systems and the j limits for releases of effluents. '
,3.7.3 Spectfications '
The reactor shall n6t be made critical unless the following conditions exist: . l A. The radiation monitoring channels and components shall be . operable in accor. dance with Table 3-3, including the minimum ;
. _ number of channels or components, and their setpoints.
s < B. The cumulative energy production of the reactor shall not exceed 4760 kilowatt-hours in any twelve-month interval nor exceed 100
- kilowatt-hours in any 7 day interval to limit the generation and release of argon-41.
C. If evidence exists that the limit in 3.3.3 G will be exceeded, the reactor shall be shutdown and the leakage source found and eliminated; however, the reactor may be operated intermittently to assistin determining the source of leakage. 1 i 1 3 - 17 . l 1
.- . . .- ~ . . . -
l . . 1 l 3.7.4 Bases I , Specification A provides assurance that the required radiation monitors are operable.
-- The air-particulate monitor is placed in service and operated continuously when designated experiments are being performed, viz., those which could produce air. borne radioactivity. The alarm setpoint is influenced by the normal background reading while the reactor operates at the required power level and is based on the same reasoning as given for the deionizer monitor ;
setpoint.
-- The radiation detector located near the deionizer initiates an alarm when the exposure rate exceeds five times the nominal .value observed during normal [
) full power operation. The trip value is sufficient for significant radiation events, yet not too sensitive to produce frequent false alarms. (See also ; 3.3.3 F.) This monitor would be the first to sense a release of fission products into the coolant.
-- The radiation area monitors are placed around the perimeter of the r'eactor -
room. All four units 'are able to initiate an alarm signal at or above 5 mR/h whenever the reactor console is energized. The south and west units initiate a radiation evacuation alarm at or .bove 50 mR/h when the reactor is in operation; when the console is not energized, the radiation evacuation setpoint is 5 mR/h. The 5 mR/h limit is based on the minimum value permitted for criticality monitoring of SNM in storage and applies when the area is unattended, while the 50 mR/h limit is based on the radiation level associated with the emergency action level for the alert classification.
-- The doorway radiation monitor serves a's a frisker to detect abnormal levels of radiation when a person passes the detector. The increasing aural signal alerts the reactor operator and the affected individual that further assessment must be initiated.
-- The radiation film badge (or its equivalent) provides iadiation dose information at the perimeter wall of the reactor room and serves as a control for the film badges used by personnel in the restricted area.
Specification B provides a conservative limit on the generation and release of argon-41 and is based on measurements at this facility (SAR: k 3 - 18
- , . . - - - - - .._,e, .-,--- ep,--w , -- w ------+v. p
l 4.5.4). Argon 41 is the only significant radioactive affluent produced during j normal operation of the reactor, and the limits provided meet the requirements )
. of paragraphs 20.103 and 20.106 of 10 CFR Part 20. The first part of
. i specification B is based on the assumption that the reactor operates l continuously at 10 kW for 476 hours and that the dilution factor from diffusion of the air in the enclosure is only 10; for these conditions, the argon-41 concentration averaged over one year is about 50% of the value listed ,
for unrestricted areas in Table II, Appendix B of 10 CFR Part 20. The second part of specification B uses the assumptions that the reactor operates continuously at 10 kW for 10 hours for one 40 hour week; these conditions yield an average concentration in the enclosure of 50% of the value listed for restricted areas in Table 1, Appendix B of 10 CFR Part 20.
~
Spech'ication C allows a search for a leaking fuel element to be conduc'ted by using the reactor to the extent needed to detect the source of fission nroducts. t I e i [ l e 4 6 4 3 - 19
.: . l t
Table 3 3. Required Radiation Monitoring Channels or Components. ,, ! Channel Setpoint Min. Operable Functic.i i Air-Particulate (a) ; unit As required 1 Alarm i Delonizer(b.) unit As required to 1 Alarm ,_ satisfy 3.3.3 F . t Area units (c)(d) 5(50) mR/h 4 Alarm Doorway monitor - 1 Warn of abnormal ; radiation level. 1 Environmental 1 Integrated dose in res-Film badge or equivalent -- restricted area l i t (3)This unit is activated whenever designated experiments *are being performed. (b)This unit serves as the fission product monitor a's specified in 3.3.'3 F. _ (C)When either the north or east area monitoring units are inoperable, portable instruments may be substituted for periods up to 48 hours.
'(d)The normal setpoint 'i'sshown. The parenthetical value is the naaximum
.setpoint to be used depending on local conditions. Use of higher than normal setpoints requires approval of the Reactor Manager. The south'and i west units monitor the fuel storage area and are reset to the normal value
.after reactor shutdown.
i r e e 9 9 I 9 e 9 e 3 - 20 F e i
l
. l 3.0 LIMITING CONDITIONS FOR OPERATION (Continued) !
i 3.8 Exoeriments . i 3.8.1 Applicability - These specifications apply to the experiments installed in the reactor and its experiment facilities. 3.8.2 Objective [ To prevent damage to the reactor and excessive release of radioactive material in the event of experiment failure, and to avoid exceeding any safety limit. 3.8.3 Specifications Experiments installed in the reactor shall meet the following , conditions: A. Prior to initiation, each type of experiment utilizing the , reactor shall be approved by the Reactor Use Committee. B. Operational limits peculiar to an experiment shall be included l in instructions to the reactor operator. C. The reactivity worth of any single experiment, or group of experiments, installed in the core shall be limited to
-0.48% Ak/k to +0.14% Ak/k.
I l. D. Significant amounts of special materials used in experiments, including fissionable material, explosives or metastable materials capable of significant energy release, or materials I that are corrosive to reactor components.or highly reactive with 3 - 21 i
l I the coolant, shall conform to established special requirements, i
- +
E. Credible failure of any experiment shall not result in releases or ' exposures in excess of established limits nor in excess of the ' annual limits established in 10 CFR Part 20. ! i F. Experiments shall be designed so that they will not contribute to , the failure of other experiments, core components, or principle physical barriers to uncontrolled release of radioactivity. Also, no credible reactor transient shall cause an experiment to fail in such a way as to contribute to an accident. 3.8.4 Bases i Specification A, B, D, E, and F are based on requirements stated in the standard for The Development of Technical Specifications for Research i Reactors, ANSI /ANS 15.1-1982. Specification C is based on the effect of the failure on an experiment, or group of experiments, on the reactivity of the reactor. 'In the ; case of an experiment failing with 40.14% Ak/k inserted as a step function, the resulting period would be 30 seconds, which can be easily managed with control rod movement. If the -0.48% Ak/k step insertion occurs because of experiment failure, the reactor excess reactivity would drop to 0.02% Ak/k, an j amount sufficient to maintain criticality and to continue operation if j neceNsary, i l i 3 - 22
l
. 4.0 SURVEILLANCE REQUIREMENTS
. \
Surveillance tests, except those specifically required for safety when - the reactor is shut down, may be deferred during reactor shutdown; however, they must be completed prior to reactor startup. 4.1 Reactor [sra Parameters 4.1.1 Applicability These specifications apply to the surveillance activities required for ' reactor core parameters. 4.1.2 Objective To specify the frequency and type of testing to assure that' reactor core parameters conform to the specifications of section 3 of these Specifications. ' 4.1.3 Specifications A. The excess reactivity shall be measured at least annually and following significant core or control rod changes, s B. The shutdown margin shall be measured at least annually and following.significant core or control rod changes. , 4.1.4 Bases
- The measurements required in specifications A and B are sufficient to provide assurance that the reactor core parameters are maintained within the specifications 3.1.3 A and B since the fuel burnup rate is extremely low and-important changes in the core parameters can be detected on a timely basis.
4-1 S
, ,,..,,n e ~ - . , ,
7 3 e , t
.t Core or control rod changes are not considered to be complete until the excess
{ reactivity and shutdown margin measurements are finished. ; j r I i i i c
\
i r !
- e y
e 1 1 P-l t l
- 9 ,
b P 9
+
4-2 -
.i I
O i l
. l 4.0SURVEILLANCEREQUIREMENTS(Continued) ,
4.2 Reactor (gatrgl ud Safety System 4.2.1 Applicability ' These specifications apply to the surveillance activities required for the reactor' control and safety system. 4.2.2 Objective To specify the frequency and type of testing or calibration to assure that the reactor control and safety system conforms to the specifications of { section 3 of these Specifications. 4.2.3 Specifications I
. 1 A. The reactivity worth' of each control rod shall be measured at least annually and whenever the core configuration is changed by fuel l
assembly replacement or rearrangement. l l i l B..The drop time and delay time of each control rod, except for the regulating rod, and the withdrawal time of each control rod shall s be measured at least annually and whenever any maintenance on a l control rod which may affect its motion is completed. ' C. The operability of the control rods and the dump valve shall be L tested daily when the reactor is operating. ' i D. An operability test, including trip action, of each safety channel ; L listed in Table 31 that provides a scram function shall be ; completed prior to each reactor startup following a period when the , reactor is secured in excess of 24 hours or at least weekly during i continuous operating periods. O 4-3 1
- j. i l.' ;
. _ . . . _ _ _ . . _ . . . ~ . . _ _ _ _ , _ _ . . . _ _ , _ _ _ . , . _ _ _ . _ . . _ , , .. , . _ , - .
. e. , .
3, i : 4 i E. A calibration of the channels listed in Table 31 that can be calibrated shall be performed at least annually and whenever any ; maintenance on a channel which may affect its performance is I compl eted, l l F. The thermal power output of the reactor shall be measured at least l 4
'nnually.
a i 4.2.4 Bases Specification A requires rod worth measurements for all rods at annual l intervals; no significant changes in the worths of the control rods are likely to occur during that time. Since changes in the fuel loading in the vicinity l of a control rod may cause a significant change in its worth, a measurement l after the fuel change is appropriate. The rod drop and delay time measurement intervals required in sp,ecification B verify the limits in specification 3.2.3 B and are appropriate l to detect abnormal performance as can be shown by experience at this facility. . 'i l Withdrawal tiine measurements provide data to determine if specification 3.2.3 ) C is being violated. l Specification C verifies the operability requirements in specification 3.2.3 8 and D during each day of operation. s In specification D each channel capable of generating a scram signal l's tested during the precritical procedure, prior to startup, so that the l conditions of specification 3.2.3 A are satisfied. . Specification E requires calibration of safety and safety-related l channels at an interval which is appropriate and justified by experience at ) this facility.
- The annual verification of reactor thermal power output, as required
, by specification F, is appropriate and justified by experience at this I facility.
9 44
. _ , . ,. ._ _ , ...- ..,y..,,,
. I 4.0 SURVE!LLANCE REQUIREMENTS (Continued) {
4.3 Coolant Systems i 4.3.1 Applicability i These specifications apply to the surveillance activities required for .! the reactor coolant system. 4.).2 Objective ! l To specify the frequency and type of testing or calibration to' assure that the reactor coolant system conforms to the specifications of section 3 of 3
- these Specifications. - I 4.3.3 Specifications A. The coolant system instrumentation channels listed in Table 3 2
)
shall be calibrated at 1sast annually and whenever any maintenance
- on a channel which may affect its performance is completed.
B. The primary coolant temperature, flow rate, conductivity, and y l l radiation level at the deionizer shall be measured and recorded at i l s startup and at least every four hours when the reactor is operating. '
. i
- i C. A primary coolant sample shall be analyzed for radioactivity 'at t t-least quarterly and whenever the exposure rate at the deionizer >
exceeds the limits of specification 3.3.3 F. : 4.3.4 Bases ' i Specification A requires calibration of the coolant system instrumentation channels at an interval which is appropriate and justified by e l- 4-5 ,
' ~ . . j experience at this facility. :
Specification B requires verification of the operating limits of f spe'cifications 3.3.3 8 F at an interval which is appropriate and justified by experience at this facility. ' Specification C relates to the monitoring for fission products and other activated materials in primary coolant samples. Experience at this t facility shows that the sampling interval is appropriate. i r 4 k i e
- 9 T
I f
+
3 e 4 e 4 4-6
l
. l 4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.4 Confinement k
4.4.1 Applicability t
. l This specification applies to the surveillance activities required for ;
the reactor' confinement.- 4.4.2 Objective . i To specify the frequency and type of testing to assure that the reactor confinement conforms to the specifications of section 3 of these Specifications. 4.4.3 Specification A. The doors and windows in the confinement boundary shall undergo testing for normal closure at least once every quarter.* i l . 4.4.4 Bases s This specification requires that the doors and windows in the confinement boundary be tested to verify that they can be closed when needed. The testing interval is adequate to verify operability based on experience'at this facility.
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o .. ; 4.0 SURVEILLANCE REQUIREMENTS (continued) - l 4.5 Ventilation Systems i This specification does not apply to this facility. G s k e 1 t 9 i
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4.0 SURVEILLANCE REQUIREMENTS (Continued)
,- 4.6 fR t9taCX E ntt 4.6.1 Applicability These specifications apply to the surveillance activities required for the emergency power system.
4.,6.2 Objective To specify the frequency and type of testing to assure that the emergency power system conforms to the specifications of section 3 of these Specifications. 4.6.3 Specifications . These surveillance activities are required for safety when the reactor is not being operated. , A. The battery powered AC standby power supply shall be tested for switch-over action, and for voltage and specific gravity characteristics at least quarterly, s B. The batteries shall be tested for full discharge at least every three years. 4.6.4 Bases
, Specification A requires verification of operability of the standby power supply to complete the switch-over from normal AC power to the batteries at an interval which is appropriate based on experience at this facility. The measured values of voltage and' specific gravity give adequate warning of reduced battery performance within the testing interval.
4-9
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i A full discharge test of the batteries every three years, as required l in specification B, is appropriate for the type of battery used in, the power ! supply; the interval is well within the normal 4 5 year warrantet ';ife for : conditions much more severe than those encountered in this application, j l
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i l 4.0 SURVEILLANCE REQUIREMENTS (Continued) 4.7 Radiation Monitorina System And Effluents i 4.7.1 Applicability J These specifications apply to the surveillance activities required for the radiation monitoring system and effluents released from the facility. 4.7.2 Objective To specify the frequency and type of testing to assure that the radiation monitoring system and effluent releases conform to the
-specifications of section 3.of these Specifications.
4.7.3 Specifications These surveillance activities (except E) are required for safety when , the reactor is not being operated. I l A. A calibration of the channels listed in Table 3-3 that can be ! calibrated shall be performed at least annually and whenever any maintenance on a channel which may-affect its performance is i completed. ' B. An operability test, including source checks, of the radiation - monitoring channels listed in Table 3-3 shall be performed at least monthly, t C. The radiation levels at the area and deionizer units shall be measured and recorded'at startup and at least every four hours when . .; the reactor is operating. D.-The environmental film badge cited in Table 3 3 and smetisurveys 1
, 4 - 11
.~. . . . - -- . - . - - . _ . . . _ . . . . . _. .
l l l in and around the reactor enclosure shall be analyzed at least ! quarterly. ; ! l E. The cumulative entray conversion shall be computed and recorded at I least quarterly, and it shall be computed on a weekly basis to 3 monitor short-term argon-41 releases. ! 4.7.4 Bases l l
. Based on experience at this facility and the average usage pattern of the reactor, specifications A D are adequate to verify that the operations ,
conform to the specifications of 3.7.3. Specification E requires verification that the cumulative energy , conversion limit of specification 3.7.3 8 is not exceeded; this is an indirect
- eethod of monitoring the generation and release of argon-41.
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i 4.0 $URVEILLANCE REQUIREMENTS (Continued) l 4.8 Exneriments - 4.8.1 Applicability These specifications apply to the surveillance sctivities required 'or experimenti installed in the reactor. - i i 4.8.2 Objective
- To specify the frequency and type of testing to assure that the experiments conform to the specifications of section 3 of those !
- Specifications.
4.8.3 Specifications ' { r A. The identification and location of all installed experiments shall be recorded prior to each reactor startup.
- B. Other specific surveillance activities shall be established during 3
the review and approval process specified in section 6.
-(.8.4 Bases apecification A requires that the reactor operator verify that the installed experiments are approved on a most conservative frequency basis.
- Specification B recognizes that detailed surveillance requirements !
will vary among experiments, and that the experiment review committee specifies the appropriate type and frequency of surveillance. 4 - 13 1 9 e -- , .
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t 5 5.0 DESIGN FEATURES , 5.1 1111 Ap.d facility Descriotion . The reactor is housed in the Nuclear Engineeri q Laboratory, which is located on the west edge of the main campus of Iowa State University, in Ames,
^
Iowa. The Nuclear Engineering Laboratory is a two-story, three-level building of brick construction, built in 1934. The reactor, a model UTR 10, was ; installed and first operated in 1959. It is fueled witn uranium enriched to approx,imately 19.75% in the U-235 isotope, moderated and cooled with light l water, reflected with graphite, and operates at a maximum thermal power of 10 kilowatts. The reactor is located on the ground floor level, central bay area of the Laboratory structure. The central bay is approximately 34 feet high and has a floor area of 37 feet by 56 feet of which a space approximately 37 [ feet by 38 feet is allocated to the reactor. A wall surrounding this area is . constructed of standard concrete block and reaches a height of 10 feet 4 inches on the north, east and south sides; the west boundary is a wall that reaches from the floor to the ceiling of the central b'ay region. The purpose of these walls is to limit access of unauthorized personnel to the immediate vicinity of the reactor and to deftr.e the outer perimeter of the restricted I area. The enclosure surrounding the reactor includes the central section of
- the building as defined by the interior partition walls of offices and labor,atories on the north, east and south sides of the building and by the
!- west interior wall which isolates the basement, first floor, and the west corridor of the second floor from the central bay. The purpose of the
- enclosure is to act as as a confinement volume and to help limit the release of radioactive materials to the environment. The enclosure volume is slightly less thart 2500 cubic meters, and the average infiltration rate for the ,
building is estimated to resrlt in two changes per hour. There is no central forced-air circulation systte in the building. The enclosure has_twc outside doors, one in the east wall and a large overhead door opening to the south. All interior doors leading into the enclosure are of a standard type used in interior construction. Other 5-1 ,
t t i significant penetrations into the enclosure consist of roof-level windows on ' the north and south sides which can be manually opened or closed, 3s a group l per side, in less that one minute per group. i
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y 5.0 DESIGN FEATUias (Continued)- L 5.2 Reactor Coolant System In normal operation, the prima'ry coolant is pumped (18 psig) from the dump tank (capacity 220 gallons) through the heat exchanger (10 gpm, 80 0F ) to
.the bottom of the core tanks, upward past the fuel plates, to the overflow pipe manifold and returned to the dump tank (0 psig, 87 0F at 10 kW and 10 I gpm); approximately 92 gallons are contained in the piping and core tanks
.during operation. A quick-opening dump valve in'the feed line to the core l: tanks is provided to allow draining of moderator (coolant) following a scram.
i A low-pressure (5 psig) steam heater and controller system for the dump tank-and a deionizer/ filter system for the purification loop, which operates-(1 gpm, <140 0F ) in parallel with the main loop are provided (Ref: Drawing RI-D-130). The operating temperature may range from about 80 0F to no more th:n.160 0F, with the lower end preferred to reduce the corrosion of aluminum. ; Moderator level, inlet and outlet temperature, flow rate and conductivity. sensors are installed at' appropriate locations and connected to the process instrumentationsystem(Ref: Drawing Rl-D-116). The primary coolant system is . essentially al'l-aluminum in construction; the pump casing and' impeller, some v al've parts, the dump tank heater element, and process instrumentation sensor - elements in contact with the water are stainless steel or similar corrosion- ' resistant materials. - I s .The energy transferred through the heat exchanger is dissipated to the building sewer system by once-through cooling water obtained from the campus water main. Secondary cooling flow is induced by water main pressure, and the ' l -flow rate is set-by a' motor-operated valve to control the amount of cooling in L the heat exchanger resulting in core inlet temperature control. To prevent . secondary water-from entering the primary system if a tube-leak should occur, - a pressure differential is maintained in the heat exchanger to allow pri. mary water to enter the secondary system. The process pit accommodates the equipment and instrumentation sensors
.for the process system (Ref: Drawing Rl-E-151). A sump, with a capacity of 9.5 gallons ~and a manually energized sump pump, can discharge liquids from 5-3 1
;g
- the process pit te another sump located in the basement floor. The basement
. sump also receives second'ary coolant outlet water and-acts like a dilution tank; it has a capacity of 123 gallons. 0utflow from the basement sumo passes through an overflow pipe connected to the building sewer system.-
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5.0 DESIGN FEATURES (Continued). 5.3 Reactor Core, fm l A.n_d Safety System > Core A graphite reflector surrounds the core tanks, except' a water qaf reflector of no less than 13 cm thickness is maintained above the fuel
. assemblies during reactor operation. The composition of the rngion between s I
the core tanks (the coupling region) can be changed by remo<a1 of graphite blocks a N insertion of other materials, and small-volume experiments can be placed in the water gap between plates in the fuel assembly, or in.the water ' ref_ lector above or beneath the fuel,-subject to specifications 3.8.3 (Experiments). A rabbit tube, oc larger than 10 cm outside diameter, penetrates the-graphite reflector' at the west face of the north core tank. A neutron source, providing a minimum of 1.0 E+6 neutrons /second is inserted !n into.the coupling region by means of the source positioner during the startup . I Loperation (Ref: Drawings Rl-E-154 and RI-E-161). L L ! Fuel Reactor fuel is contained in aluminum-clad flat plates, similar to p argonaut type _ fuel. Fuel meat is U3Si2, enriched to 19,75%-in the U-235 - b isotope, dispersed in aluminum.to achieve a uranium density of 3.47 g/cc. The.
- fuelme a t ,1 0.51 mm thick, is clad with 0.38 mm aluminum. Each fuel plate contains'12.5 grams of U-235. The core contains 12 assemblies.each with '
approximately 24 fuel plates dependin; upon.the measured critical configuration.- Sol.id aluminum plates and assemblies with missing fuel plates, ( y ' L Lforexperimentalpurposes,areusedtoadjustthecorefuel_loadingforthe l, licensed excess' reactivity of 0.50% Ak/k. (Ref: Drawing Rl-A-121-1) i k 5-5
u l Safety System ; j Four'Boral contro'1 rods, two safety, one shim safety, and one regulating, are positioned in the graphite external reflector adjacent to the
, outside face and near each outside corner of the core tanks assembly (Ref: 1 Drawings R1-R-212, R1-R-213, Rl-R-214). Each control rod is connected by a ,
stainless steel flat spring to a motor-driven drum. Each safety rod is coupled to its drive mechanism by an electrically energized magnetic clutch. Two safety rod drives have limit switches with console indicators showing full I withdrawal and full insertion. Shim-safety and regulating rod positions are. ; displayed on the control console. The moderator level measuring channel providec a signal (interlock) j which permits control rod drive magnets to be energized only after a minimum-moderator level setpoint is exceeded, and it provides a signal (scram) when , the moderator' level exceeds the high level setpoint. ' A , neutron-sensitive' power level measuring channel with a functional-
- range of 1. E-7 to 1.5 E+2 pe'rcent power, based on 10 kilowatts thermal, ,
provides a signal (interlock) which prevents withdrawal operation of the control rod drive motors if the minimum power level (minimum count rate) setpoint is not exceeded, a signal (scram) when the one-watt level is exceeded, and the. neutron _startup source is not in its storage position or all
. closures (two operating closures above the fuel, and one at-the end of the thermal column) are not properly seated; this channel provides a signal to L .the period channel to generate a signal (scram) when the period is less than th'e short-period setpoint. These signals are derived from the log percent l power channel.' .
A neutron-sensitive power level measuring channel, with a functional l range of 10 to 150 p,ercent of -10 kilowatts, provides a signal (scram) when a-high power level setpoint is exceeded. This signal is derived from,the linear l percent power channel. 5-6 .. 9 ____--_:__.___.------.---.-------__.___--_-_ _. - - -- v ~
, ^ 'L.-
5.0 DESIGN FEATURES (Continued) 5.4 Fissionable Material Storaae Fueled experiments and fuel devices not in the reactor are stored in a dry fuel storage pit monitored by radiation-and intrusion detectors (Ref: Drawing R1-E-194 and Physical. Security Plan). .The fuel storage array, under i all conditions of moderation and reflection with light water, has an effective ' multiplication factor less than 0.9. e t l- ' e t e
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7
- .. I 6.0 ADMINISTRATIVE CONTROLS 6.1 Oraanization 6.1.1 Structure :'
L The organization for the management of the reactor facility shall be structured as indicated.in Figure 6-1. Job titles are shown for illustration and may. vary. Levels of authority indicated divide responsibility as follows: Level 1: Responsible for the facility license and sits administration Level 2: Responsible for the reactor facility operation and management.
~
Level 3: -Responsible for daily operations. The Reactor Use Committee is appointed by, and shall report to the University Radiation Safety Committee Radiation safety personnel shall l report to. Level 2:or hig'her through an independent organizational channel. 6.1.2 Responsibility , The Executive Officer, Department of Nuclear Engineering, shall be
. responsible'for the.-facility license and site administration. I L
I Individuals:at the various management-levels shown in Figure 6-1, in- "
. addition to-having the responsibility for the policies and' operation of the-facility, shall be responsible for safeguarding the lic and facility y L f personnel from undue radiation exposures and for adhdPfng to all requirement's "
l of the Operating: License and.the! Technical Specifications. *
.In 'all instances, responsibilities of one level may'be assumed-by designated alternates, or by higher levels, conditional upon appropriate-qualifications.
6.l'.3 Staffing ,30 (1) The minimum staffing when the reactor is not secured shall be:. 3 m 6-1 e !
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- a. A licensed reactor operator in the control room.
b; A. licensed senior reactor opera, tor- readily available of call.
- c. A health. physics qualified individual readily available on call.
.(2) Events requiring the direction of a senior. reactor operator:
- a. Recovery from unplanned or unscheduled shutdown (in this instance, documented verbal concurrence from a SRO is required).
- b. Fuel transfer operations.
- c. Any maintenance activity involving the reactor safety system that could cause a significant increase in the reactivity of the reactor,
- d. Relocation of any in core experiment with a reactivity worth greater than 0.763% Ak/k.
3 (3) Events requiring the presence of a health physics-qualified individual: ;
- a. Fuel transfer operations.
- b. Installation, changing locations, or removal of an experiment.that 7 involves removal of a shield plug or closure. -
- c. Any maintenance activity involving the reactor safety system that L
could cause an cbnormal release of radioactive materials. . s L 6.1.4 Selection and Training of Personnel L i l- The selection, training and requalificat' ion of operations personnel shall meet or exceed the requirement of American National Standard for-Selection' and Training of Personnel for Research Reactors, ANSI /ANS-15.4-1977,- or its successor,. and be in accordance with the Requalification- Plan approved-l by the Nuclear Regulatory Commission. l' 6-2
4 6.0 ADMINISTRATIVE CONTROLS (Continued) a 6.2 Review And Audit , The Reactor Use comittee (RUC) shall perform the independent review and - audit the safety aspects of reactor facility operations, 6.2.1 Composition and Qualifications
. The Reactor Use Comittee shall be composed of the Reactor Manager and-a radiation health physicist, both ex officio (voting), and at least three other members having expertise in reactor technology. Comittee members shall l be appointed by the University Radiation Safety Comittee. (The Radiation
~
Safety Comittee is composed of a representative from each of five ' colleges in the university-in which_research in the physical and life sciences and in engineering is conducted,.plus three members with specific expertise i,n l radiation ~ protection. At least one of these members shall also represent university management. The college representatives are chosen from the . Colleges of Agriculture, Engineering, Sciences and Humanities, Home Economics, L and Veterinary Medicine. One of the three other members shali be the-University Radiation Safety Officer (RS0). The chair of the comittee shall , be appointed by the Vice President for Academic Affairs. The terms on the comittee for the RSO and chair are indefinite. All others are for three - yearA with reappointments being determined by the Vice President for Academic Affairs.) 6.2.2 Charter and Rules (1) The Reactor Use Comittee shall meet at least semiannually and more
' frequently as circumstances warrant, consistent with effective .
monitoring of facility activities. Written records of its meetings shall be kept and copies forwarded, in a timely manner, to the University Radiation Safety Comittee. L 6-3 1 1 - ---.---__--____----_-___-___--__---__.--_.____l--._--
1 it
.(2) A quorum shall be three tambers. Members of the operation staff t shall not be a. voting majority. #
(3) Any action recommended by the Reactor Use Comittee that may adversely affect the operations ar.d/or safety of the University ; comunity shall be reported by the RUC chairman to the University !
. Radiation Safety Comittee which shall have veto power over such a recomendation. I
' (4) The Reactor Use Comittee may appoint one or more qualified .
individuals.to perform the audit function, o , e 6.2.3 Review Function The following items shall- be reviewed: . (1) Determinations that proposed changes in equipment, systems, te~sts, t
' experiments, or procedures do not involve an unreviewed safety 1
. question.
-(2) All new procedures and major revisions thereto having safety .
significance and proposed changes in reactor facility equipment, or s systems having safety significance. (3) All new experiments or classes of experiments that could affect reactivity or result in the release of radioactivity. I
'(4) Proposed changes'in the Technical Specifications or the Operating License. ,
.(5) Violations of the Technical Specifications of the Operating d License. Violations of internal procedures or instructions having safety. significance. ,
6-4
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. i (6)-' Operating abnormalities having safety significance. :'
(7) Reportable occurrences listed in.6.6.2< (8) Audit reports. 6.2.4'A'udit Function
. The audit function shall include selective (but comprehensive).
examination.of operating records, logs, and other documents. Discussions with l ' cognizant personnel and observation of operations should also be used as 1 appropriate. In no case shall the individual immediately responsible for the L area, audit in that area. Deficiencies uncovered that affect reactor safety shall be reported immediately to the University Radiation Safety Committee. A L t - written report of the findings of the audit shall be. submitted to the Reactor Use Committee within 30 days after completion of the audit.- The following- o-items shall he. audited: 1
- 1 (1) Facility operations for conformance to the Technical' Fpecifications I
l and' applicable Operating License conditions, at least one per calendar year (interval between audits not to exceeq 15 months).
- 1. .
.(2) The retraining and requalification program-for the operating staff, at' least' once_ every other calendar year (interval' between audits-1 not to exceed 30 months).
L
..1
- -(3) The results of action taken to~ correct those deficiencies that may l occur in the. reactor facility equipment, systems, structures, or l methods of operations that' affect reactor safety,. at least once -
y per calendar year-(interval'between audits not to exceed 15 1 g. months). " l l L (4) The reactor facility Emergency and Physical Security Plans and t - l
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. implementing procedures at least once every other calendar year (interval not to exceed 30 months). :
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- 4-6.0 ADMINISTRATIVE CONTROLS (Continued) 6.3 Procedures Written procedures- shall be prepared, reviewed and approved prior to initiating any of the activities listed in this section. The procedures shall
- be reviewed by the Reactor Use Committee (see 6.2.3) and approved by the Retetor Man'ager or a designated alternate. These reviews and approvals shall.
be documented in a timely manner. Substantive changes to the procedures shall be made effective only after documented review by the Reactor Use Committee. c and approval by the Reactor Manager or a designated alternate. Minor modifications to the original' procedures which do not change their original
'i ntent may be made, but the modifications must be approved by the Reactor Manager or a designated alternate within 14 days. Temporary deviations from the procedures may be made by the on-duty SRO in order-to deal with special or unusual circumstances or conditioas. Such deviations shall be documented and i reported to the Reactor. Manager or a designated alternate.- Several-of the following activities may be inc1'uded in a single manual or set of procedures !
l or divided among various manuals or procedures: j' . i (1) Startup, operation and shutdown of the reactor.-
.(2) Fuel loading, unloading, and movement within the reactor, s
~(3) Routine maintenance of major components of systems that could have ,
an effect.on reactor safety. (4)~ Surveillance tests and calibrations required by-the Technical < Specifications or those that may have an effect on reactor safety. l L (5) Personnel radiation protection consistent with applicable regulations. l (6) Administrative controls for operations and maintenance and for the 9 6-7 L
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u -. ' iconduct;of irradiations and experiments that could affect reactor-safety or__ core reactivity. :
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E (7)- Implementation of the Emergency and Physical Security Plans, i l-e I 1 l-
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L' , - - - L-6.0 ADMINISTRATIVE. CONTROLS (Continued) l
-6.4 ExDeriment Review ADA ADDr0 val l
L Approved experiments shall be carried out in accordance with established
- and approved procedures.
L ! L (1) All new experiments or classes of experiments shall be reviewed by the Reactor Use Committee and approved in writing by the Reactor . . j Manager or a designated alternate prior to initiation. ( 2.) . Substantive changes to previously approved experiments shall'be made only after they are reviewed by the Reactor Use Committee and approved in writing by the Reactor Manager or_a designated alternate. Minor changes that do not significantly alter the experiment may be approved by the Reactor Manager or a designated alternate. , i l i l r e O S 6-9 1 s I
,, = ,
s 6.0 ADMINISTRATIVE (Continued)
'6.5 Reauired' Actions :
, 6.5.1 Action to be Taken in case of a Safety Limit Violation l
L (1) 'The rea'c tor shall be shut down and reactor operations shall not be resumed until authorized by the Nuclear Regulatory Commission
.. (NRC).
(2) The safety. limit violation shall be promptly reported to the Reactor. Manager or a designated alternate. (3) The. safety limit. violation shall be' reported to NRC. L (4) A safety' limit violation report shall be prepared. The report, and any follow up report, shall.be reviewed by the Reactor Use '
. Comittee and shall be. submitted to the NRC when authorization is sought to resume operation of the. reactor' The report shall-l describe the following:
L . a-
. Applicable circumstances leading to the violation, including, s when known, the cause and contributing factors..
i b. Effect of the violation upon reactor facility components, , systems, or structures and on the health-and safety of-personnel and the public. l c. Corrective action to be taken to prevent recurrence. i a . 6.5.2 Action'to be Taken in the Event of an Ogcurrence of.the Type l Identified in 6.6.2(1)b. and 6.6.2(1)c. 4 (1) Reactor conditions shall be returned to normal or the reactor shall
' be shut down. If. it is necessary to shut down the reactor to -
6 - 10 4 , 2 It
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correct the occurrence, operations shall not be resumed unless l authorized by the Reactor Manager or a designated alternate. ' (: (2)' Occurrence shall be reported to the Reactor Manager or a designated-L alternate and to the NRC. 1 1, L (3) Occurrence shall be reviewed by the Reactor Use Committee at its next scheduled meeting. 1 i ( 1 1 y l
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3 ;.. . . ;. L ; L 6.0ADMINISTRATIVECONTROLS(Continued). > [
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L' 6.6 Reports I 6.6.1 Operating Reports i l ( A routine operating report providing the following information shall be submitted to the Nuclear Regulatory Comission in accordance'with the - L ~ provisions of 10 CFR 50.59 at the end of each 12-month period: (1) A narrative sumary of reacccr operating experience including the i energy produced by the reactor. (2) The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence. (3) -Tabulation of major preventive and corrective maintenance '
. operations'having safety significance.
(4) Tabulation of major changes in the reactor facility and' procedures,
- and tabulation of new tests or experiments, or both, that are significantly different from those performed previously and are not described in the Safety Analysis Report, including. conclusions'tha't
.s no unreviewed safety questions.were involved.
..(5) A sumary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control.
of the owner-operator as determined at or before' the point of ^such release or discharge. .The sumary chall include to the extent. practicable'an estimate of ind.ividual radionuclides present in' the -l effluent. If the estimated average release after dilution or diffusion is less than 25 percent-of the-concentration allowed or: ' i. recomended, a statement to this effect is sufficient. . 6 - 12 m I - e v v- 6
,7_
i (6) A sumarized result of any environmental surveys performed outside ~ the facility. . (7) A sumary of exposures received by facility personnel and visitors where such exposures are greater than 25 percent of that allowed or
.I recomended.
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6.~6.2 SpecialfReports x
-(1) There will be a be.a report not later than the following working day by telephone to the appropriate NRC Regional Office:and confirmed in writing by telegraph or similar conveyance to the -
appropriate NRC Regional Office with a copy to the Director of ' Inspection and Enforcement to be followed by a written report that > describes the-circumstances of the event within 14 days of any of the following: '
- a. Violation of safety limits (see 6.5.1).
c .
- b. Release of radioactivity from the site above allowed limits (see 6.5.2).
i
- c. Any of the following (see 6.5.2): ;
l (i) Operation with actual safety system settings for required systems less conservative than the limiting safety system L settings specified in the Technical Specifications ' (ii) Operation in violation of limiting conditions for operation established in the Technical. Specifications 4 u unless prompt' remedial action is taken. (iii)- A reactor. safety system component malfunction'which i renders or could render th'e system incapable of performing
) ,
its intended safety function'unless the malfunction or ) condition is discovered during maintenance tests;or 1 periods'of reactor shutdown.
- , (iv)- An unanticipated or uncontrolled change in reactivity .
L greater tha.. the licensed excess reactivity, or _ 0.763% Ak/k, whichever is smaller.
-(v) Abnormal and significant degradation in reactor fuel, or. "
' cladding, or both, or coolant boundary which could result in exceeding prescribed radiation exposure limits of-g- personnel or environment, or both.
(- (vi) An observed inadequacy in the implementation of i L l, 6 - 14 i \~ t
4 administrative or procedural controls such that the ; inadequacy causes-or could have caused the existence or development of an unsafe condition with regard to reactor j
,; operations, i
'(2) A written report within 30 days to the appropriate NRC Regional Office with a copy to the Director of Inspection and Enforcement
'concerning the following: .
.. a. Permanent- changes in the organization involving. Nuclear -
Engineering Department Executive Officer, Reactor Manager, or p Radiation Safety Officer. If i
- b. Significant changes in the transient or accident analysis as described in the Safety Analysis Report. j l
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. 6.0 ADMINISTRATIVE CONTROLS M
6.7 Records 6.7.1 Records to be Retained for a Period of at least Five Years or the i Life of the ';omponent if Less than Five Years
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f- (1) Normal reactor facility operation (but' not including supporting j documents such as checklists, log sheets, etc., which shall be maintained for a period of a least one year). L [ I: (2) Principal maintenance operations. ' l. (3) Reportable occurrences. I (4) Surveillance activities required by the Technical Specifications. l; .. Reactor facility radiation and contamination surveys where required (5) 4 by applicable regulations. - + (6) Experiments performed with the reactor.. ' L , .
, (7) Fuel inventories, receipts, and shipments. >
i '(8) Approved changes in operating procedures. j" (9)- - Records of meetings and audit reports of the Reactor Use . Committee. , I W 6.7.2 Records to be Retained for at least One Training Cycle Retraining and requalification of licensed operators: Records of the ! most recent complete cycle shall be maintained at all. times the individual .is , employed. . 6 - 16 i T l
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6.7.3 Records to be Retained for the Lifetime of the Reactor facility - Applicable annual reports, if they contain all of the required information, may be used as records in this section . (1) Gaseous and liquid radioactive effluents released to the environs.- (2) Off-site environmental monitoring surveys ' required by the Technical
. Specifications.
(3) Radiation exposure for all personnel monitored. (4) . Drawings of the reactor facility. t I !
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}g' _fhN 1; e:- , - LEVEL 1 . >
't Univorelty ,
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Preeldent - 1
, -L i f a,
9 f T. Vice Preeldent .
. Ylee Preeldent . ,
Academio Affelte Buelness D Finance 1 r 1 r
- 1 r Director '
Dean of Radletion Safety " " ' "
- Engineering Committee He h D Ss -
I I ! r i r 3 l
. Nuc E Dept '
Reactor Use. l . Exec Officer - Committee , l - l '
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- LEVEL 2 l ~
L Redletion
. Reactor -
""" - d l Safety Officer
' Meneper l 1 I
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1 r 1 r' l
. LEVEL 3 i Health Physico b---
Reactor Staff .r Operatione Stoff
' - - -- Committee Memberehlp 3
l Figure 6-1 Organization structure a r e --e ,*v' wy rre- w w- ce*v.w-m- -
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p Enciesure 4 OUTLINE OF REACTOR START-UP REPORT AND COMPARISONS WITH CALCULATIONS
- 1. Critical Mass Measurement with HEU Measure'nent with LEU Comparisons with calculations for both LEU and HEU.
- 2. Excess (operational) reactivity Measurement with HEU Measurement with LEU Comparison with calculations for both LEU and HEU.
- 3. Control and regulating rod calibrations Measurements 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 limitc '
Comparison between HEU and LEU nuclaar instrumentation setpoints.
' detector positions, av.' 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 255, 37.55, 505 and 62.55. loaded fuel elements is the positions they are allowed to occupy.
- 7. Thermal neutron flux; distributions.
Measurements with HEU and LEU, and comparisons with each other and calculations.-
- 8. Discussion of how compliance with void and temperature coefficient values in Technical Specifications is to be assured. Comparisons with any calculations for both HEU.and LEU fuel.
- 9. Comparison of the various'results, and discussion of the com>arison, including an explanation of any significant-differences whici have an-impact on both norma 1' operation and potential 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 verification of final !
criticality conditions. 1
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