ML20215C167
| ML20215C167 | |
| Person / Time | |
|---|---|
| Site: | Rensselaer Polytechnic Institute |
| Issue date: | 06/30/1986 |
| From: | Harris D, Rodriguezvera, Wicks F RENSSELAER POLYTECHNIC INSTITUTE, TROY, NY |
| To: | |
| Shared Package | |
| ML20215C151 | List: |
| References | |
| NUDOCS 8610100133 | |
| Download: ML20215C167 (39) | |
Text
. ._.
\
4 RENSSELAER POLYTECHNIC INSTITUTE CRITICAL FACILITY SAFETY ANALYSIS REPORT Docket No. 50-225 i
License No. CX-22 Prepared by: Donald R. Harris, Ph.D., Director Frank Wicks, Ph.D., Supervisor Francisco Rodrigue:-Vera Matt A. Coleman June 1986 l
8610100133 860630 PDR ADOCK 05000225 P PDR i
L
A Table of Contents Topic Page
- 1. Introduction 1
- 2. Facility Description 2
- 3. Fuel Storage Vault 3
- 4. Reactor Components 4 4.1 Control Rod Drives 4 4.2 LEU Reactor Core 5 4.3 Fuel Pins 6 4.4 Control Rods 7
- 5. Accident Analysis 8 5.1 General Summary 8 5.2 Scenario for Reactivity Insertion 9 5.3 Analytical Method 12 5.4 Core Nuclear Characteristics 12
- 6. Technical Specifications 17
- 7. Startup Testing and Evaluation 18 7.1 Radiation Monitoring 18 7.2 Physical Inspection 19 7.3 Instrument Calibration 19 7.4 Interlock Verification 20 7.5 Control Rod Drop Time 20 7.6 Moderator Dump Time 20 7.7 Core Loading 20 7.8 Control Rod Calibration 21 7.9 Fuel Pin Worth 22 7.10 Temperature Coefficient of Reactivity 22 7.11 Void Coefficient of Reactivity 22 7.12 Relative Flux Mapping and Power Calibration 23 References Cited 24 (i)
e
- 1. INTRODUCTION.
Construction of the Reactor Cettical Facility (PCF) was completed in July of 1956 by ALCO Products, Inc. Originally the facility was constructed as a laboratory in which reacter experiments, necessary fcr the design and development of military and commercial power plants, could be performed in a safe and efficient manner. The experiments performed here were "zero-power" experiments, all of which took place at very low power levels. In 1964 Rensselaer Polytechnic Institute (RP1) hssumed operation of the facility for the instruction of students in the Institute's Department of Nuclear Engineering and Science, and for research and testing purposes. Up to the present the reactor has operated with a highly enriched uranium (HEU) fuel..
Recently the Nuclear Regulatory Commission (NRC) mandated that all NRC-licensed non power reactors using highly enriched uranium (HEU), including the RPI Critical Facility, convert to low enriched uranium (LEU) fuel unless compelling reasons can be given for continued use of HEU. The rule was set down to addresc an increasing concern with the possibility that HEU, widely used in non-power reactors around the world, might be diverted from its intended peaceful uses. Accordingly. RPI intends to refuel the present critical facility core with LEU as part of the reactor upgrade supported by the U.S. Department l
l (1) l l
c-e e
of Energy (DOE) and by RPI. This Safety Analysis Report is l
submitted to support the Institute's proposal for core 1 conversion to LEU fuel pursuant to 10 CFR 50.64. Physical modification of the reactor will begin subseqent to both satisfactory qualification of the subject LEU SPERT fuel pins and issuance of a conversion order by the Nuclear Regulatory Commission. Efforts to qualify the fuel are being pursued by Argonne National Laboratory under the direction of DOE.
This report addresses only pertinent changes to the RCF and its nuclear characteristics, imposed by the conversion to LEU.
The balance of information normally contained in a Safety Analysis Report (SAR) is still accurately recorded in the SAR submitted by RPI in January, 1983 and complements the material presented herein'.
- 2. FACILITY DESCRIPTION.
The RPI Reactor Critical Facility (RCF) is situated on the south bank of the Mohawk River, adjacent to the property leased by General Electric (GE) in the city of Schenectady, New York.
The geographic orientation of the RCF is best viewed in Figure 2.1. Exclusion areas depicted in Figure 2.1 are divided into two zones. The inner zone is enclosed by a chain linked fence
~
with two controlled access gates. It is roughly a 100 foot by 100 foot square, with a minimum distance of 40 feet to the (2)
e reector itself. The civil e::c l u s i o n one enconcasses the access road to the PCF'from Maxon Road and the facility pcrting lot; the minimum distance to the reactor is 50 feet. The civil exclusion one is bordered by the perimeter of the GE property shown in Figure 2.1. Both the civil exclusion zone and the GE area are open to the river on the northwest side.
- 3. FUEL STORAGE VAULT.
An a foot by 10 foot Fuel Storage Vault in one corner of the reactor rcom is provided with wall.s. ceiling, and floor made of 1 foot thick reinforced concrete. A storage rack constructed of unistrut is mounted against one wall of the vault opposite the vault access door. Figure 3.1 displays the storage arrangement. Eighty-one stainless steel tubes, 5 inches in diameter and wrapped in 0.015 inch thick cadmium sheaths, are bolted to the unistrut frame in parallel rows. Half of the tubes and their cadmium sheaths are 42 inches long, increased in length by about 14 inches over those previously used, to ensure the comparatively longer LEU fuel pins are completely encased when stored in the vault. The vault can safely accomodate, in physical weight and reactivity, the total inventory of both the HEU fuel plates and the anticipated maximum quantity of 600 LEU fuel pins for the new core. After destructive testing of five randomly selected LEU fuel pins, 595 pins have been identified for shipment to RPI.
This constitutes a total U"3' inventory of 20.94 kg. The (3)
e fuel storage vault was originally designed to safely store HEU fuel with 81 kg o f UC3' . Conservative calculations of the infinite multiplication factor for the vault, whenhousing both these inventories under completely flooded conditions, yield a value of much less than 0.90. Section 5.4 summari:es the methods employed to calculate fuel storage vault criticality data.
4 REACTOR COMPONENTS.
4.1 Control Rod Drives.
The overhead control rod d.ives, four in number, are mounted on the reactor tank as shown in Figure 4.1. Figure 4.2 offers a more detailed view of one such mechanism. The drives are supported by rigid cantilevers with three. degrees of freedom to allow positioning of the rods anywhere in the tank.
Structurally, the drives consist of a 1/20 horsepower motor, gear box. magnetic clutch, drive shaft, pinion gear, and control rod rack. Control rod position is determined by a pair
'of geared anti-backlash synchromotors. Electrically the control rods operate on demand from the control room with powor supplied to the magnetic clutches from the safety amplifiers.
A minimum holding current is adjusted for each drive individually to minimize magnet decay time and ther e fo r e rod drop time. This current is interrupted on receipt of any scram signal or on power failure.
(4)
The rod drive equipment utilized is essentially the sama gear used previously, except that modifications to the gear asnembly and re-calibration of the synchromoter' position indicators is necessary to provide for the increased lergth of rod travel. Maximum axial rod motion is limited to 36 inches for the LEU core design, the effective height of the core, as compared to 2E inches of travel in the HEU core.
4.2. LEU Reactor Core.
The core and support structure are designed for critical experiments using variable arrays of fuel pins. Fuel, in the form of 4.8w/o enriched UOn pellets in stainless steel clad, is arranged in roughly a cylindrical fashion with four control rods placed symmetrically about the core periphery. Fuel pins, with an effective length of 36 inches, are set on a square pitch of 1.4859 cm to yield an effective core radius of approximately 35 cm. Two fuel pin arrangements are discussed hora. The first, referred to as " Core A", is the solid array of pins shown in Figure 4.3. The second, referred to as " Core B", is the annular array of pins shown in Figure 4.4 with the central 5 y 5 square of fuel pins removed. Calculations show Core A to contain 421 pins during operation; Core B is calculated to hold 424 pins during operation. The cptional Core B permits use of the interior region for experimental purposes. The pins themselves are supported and positioned on a fuel pin support plate, drilled with 1/4 inch diameter holes (5)
to accept tips on the end of each pin. The support plate rests en a thick carrier plate which forms the base of a three-tiered overall core suppcrt structure. An upper fuel pin lattice l plate, depicted in Figure 4.5, rests on the thick top plate and is drilled through with 1/2 inch diameter holes on the prescribed pitch to secure the upper ends of the fuel cins.
The carrier plate, the lower fuel pin support plate, a middle plate, the top plate, and the upper fuel pin lattice plate are secured with tie rods and bolts. The entire core structure is anchored by four posts set in the floor of the reactor tank.
Finally, in the event that the fuel pins are bowed but still satisfactory for use in the core, a plastic spacer plate may be installed on the middle plate. Figure 4.6 depicts the total core assembly.
In general structure, the LEU core design compares favorably with that of the HEU core in that bo th are arranged in roughly a cylindrical fashion on a radius of about 14 inches. Both core designs also employ the three-tiered steel plate structure for mechanical support, with the additions of a
" support plate" and a " lattice plate" to provide for pin location in the LEU core. The most conspicuous physical differences are the increased axial length of the LEU core, 36 inches versus 22 inches in the HEU design, and the use of cylindrical pins rather than plates.
4.3 Fuel Pins.
(6)
The core will employ 4.9 w/o enriched SPERT(F-1) fuel rods.
contingent upon their satisfactory qualification by the COE and NRC. Each fuel rod is made up of sintered UCa pellets, encased in a stainless steel, tube, capped on both ends with a stainless steel cap and held in place with a chromium-nickel spring. An aluminum oxide (AlaGu) insulator between the fuel pellets and stainless steel caps on each end of the red is installed. Gas gaps to accommodate fuel expansion are also provided at both the upper end and around the fuel pellets.
Figure 4.7 depicts a single fuel pin and its pertinent dimensions.
4.4 Control Rods.
Four control rods are provided, spaced 90 degrees apart at the core periphery. Each rod consists of a E.75 inch square stainless steel tube, effectively 36 inches in length, which passes through the core and rests on a hydraulic buffer on the bottom carrier plate of he support structure. Housed in each of these " baskets" are two enriched baron absorber sections, one positioned above the o ther as depicted in Figure 4.9. The 9 t '> poison contained in each absorber section is held in a stainless steel cermet that is also clad with stainless steel.
Each of the four rods has approximately the same reactivity effect, and pertinent nuclear characteristics are detailed later in this report.
(7)
Control rods used in the HEU core were 14 inches shorter and contained fuel followers in the f o rin of uranium plates.
Changes to the control rods were necessary to accommodate the larger axial dimension of the LEU core.
- 5. ACCIDENT ANALYSIS. .
5.1 General Summart.
Several potentially serious accident scenarios were evaluated and. even in the worst event sequence considerad, no release of a significant quantity of radioactivite fission products to the reactor cell would occur. Effects due to natural phenomena, mechanical rearrangement of the fuel, and reactivity insertion were all analyced.
While the new core configuration is clearly different from previous design, the characteristics of the site and the RCF physical structure remain essentially the same. Adequate protection against the potential effects of natural phenomena including fires. windstorms, floods, and earthquakes is thereby provided. Radiological hazards to the public from these events t
are not significant. ;
Mechanical rearrangement of the fuel to obtain a supercritical configuration, inadvertently or with intent. is not a credible occurrence. Core desigr' objectives included minimizing the critical mass and, therefnre. other configurations of the fuel would be less reactive and not (8)
__ .J
.- n 1
likely to support a large reactivity excursion. Physical damage to fuel pins is precluded both by each element's cladding and the stainless steel support plates /stanchiens that compose the basic core structure. In the unlikely event that sufficient force to break one or 'mcre of the fuel pins wac developed, the very low fission product inventory accumulated in the fuel elements would not cause a significant off-site hazard.
The potentially most severe accident analyzed was due to reactivity insertion and, hence, this became the limiting case for design purposes. Hypothesizing that an unsecured experiment causes $0.60 reactivity tc be instantaneously inserted while the reactor was operating at maximum pcwer, the resultant excursion induces a negligible rise in fuel temperature. This scenario and the details of the analysis are discussed below.
5.2 Scenario for Reactivitv Insertion.
The most extreme scenario hypothesired consists of the worst reactivity excursion coincident with a single failure in the reactor protection system.
The worst reactivity excursion results from an unsecured experiment with a reactivity worth equal to the maximum excess reactivity allowed by the Techt:ical Specifications of 30.60.
Specifically this could result from an experiment in which a strip of poison, such as baron, was placed in the core, the (9)
control rods pulled all the way out to obtain just critical conditions, thereupon the baron strip falls out of the cera resulting in a step reactivity insertion of the specified amount. A pre-accident power level of 200 watts was assumed, based upon the Technical Specification limit of 100 watts and incorporating a factor of two te account for the cumulative uncertainties associated with instrument calibration. For analytical purposes, the reactivity feedback effects of temperature and void formation were neglected so that conte:1 rod insertion was necessarily the terminating event.
The open circuit failure of the ion chamber serving log power and period channel 2(PP2), coincident with the beginning of the accident, was also assumed. Because this one ion chamber supplies the input to the circuit that provides both the log power (135 watts) and the short period (5 seconds) scram, these scram relays are assumed to be disabled. The failure chosen, then. is the " worst case" single instrument malfunction. Remaining scram protection is provided only by the two linear power channels (LP1, LPS), each of which initiates a scram if its respective meter indicatior exceeds 9 0. o f the selected scale. Commonly the operator upscales these meters by factors of three as power increases during a directed power increase to preclude an inadvertent shutdown.
For purposes of the accident scenario LP1 and LP2 are assumed to indicate a value of 10% on the highest selectable scale at (10)
the onset of the accident, roughly correlating with 200 watts in-core power (LOO watts indicated with factor of two uncertainty). Thus the power must increase by a factor of nine from this pre-accident level to prompt the linear power channel scram activation. Notably, because of the nature of the accident. its severity is not sensitive to variation in initial power. The single insertion of a fixed amount of positive reactivity quickly puts the reactor on a constant positive period, so that both the value af reactor power and its rate of increase when scram is initiated are unrelated to power levels immediately beforehand. Hence selection of a very low power, visible yet well below the point of adding heat, would not have aggravated the results of the analysis.
With the reactor operating initially at 200 watts, the insertion of $0.60 positive reactivity causes power to promptly jump to 600 watts and then increase on a period of 3.0 seconds to 1800 watts, at which point LPl and/or LP2 generate a scram signal. Allowing 1.5 seconds thereafter for the rods to be bottomed (Technical Specification is 900 msec), analysis conservatively assumed the instantaneous insertion of $1.000 negative reactivity (less than the core shutdown margin) at 5 seconds after the excursion began.
Maximum power reached during the transient is slightly below 3050 watts, depositing about 10 kJ of energy in the core and inducing a fuel temperature rise of less than 0.1 'C above (11)
an initial value of 20 SC. This energy deposition is roughly a factor of 10;' less than the core safety limit identified in the Technical Specifications. Figure 5.1 portrays changes in power for the stated reactivity insertion transient, annotated with pertinent events. Clearly the integrity of the fuel is not in question. Additionally, while feedback effects were intentionally disregarded in the analysis, the very small i'
temperature change encountered would have made their cumulative effect negligible. This conclusion is valid for both the Core A and Core B pin arrangements.
5.3 Analvtical Method.
The transient analyses conducted in support of this report employed the "FRKGB" computer code model', developed at RPI specifically for low power pool reactors. The model includes recent developments in transient thermal hydraulics models and correlations, utilizing Runge-Kutta time stepping methods to derive numerical solutions. The program was initially benchmarked against a set of Gaussian, Nordheim-Fuchs, and SPERT type bursts.
5.4 Core Nuclear Characteristics.
Tables 5.1 through 5.3 list pertinent nuclear and physical characteristics o t' the core which are relevant to safe operations.
(12)
The core physics design and the fuel vault criticality calculations were carried out using the LEOPARD code (wi th ENDF/9-4 based data) to compute few group diffusion constaats.
the PL AT AB to compute equivalent few group diffusion constants for strong absorbers (this code used detailed flux spectra from LEOPARD), and the DIFXY' code to apply few group diffusion code theory in X-Y geometry.
Figures 5.2 and 5.3 display graphs of the temperature coefficient of reactivity for the solid (Core A) and annular (Core B) core fuel pin arrangements, respectively. The curves portray data derived from the computer codes referenced above.
(13)
O o
Table 5.1 Nuclear and Physical Characteristics of the RPI LEU Core Effective Delayed Neutron Fraction G = 0.00765 Effectiva Neutron Lifetime 1* = 1 2 . 2 x 1 0 - s e c Delayed Neutron Data Group No. 13L1" Decav C o n t- t a n t < t >
1 0.041 3.01 2 0.115 1.14 3 0.396 0.301 4 0.196 0.111 5 0.219 0.0305 6 0.033 0.0124 Reactor Power P = 100 watts A.ial Power Shape Chopped Sine Coolant Temperature T = 20*C (1) G.R. Keepin, " Physics of Nuclear Kinetics". Addison Wesley (1965).
(14)
c -
5.2 Kinetics Parameters of RPI LEU Core and Technical Soecificationn Kinetics Parameter LEU Core Value Technical Specification Excess Reactivity 0.00468 <O.00469 at 680F Reactivity with One <-0.005 <-0.005 Stuck Rod
, Shutdown Margin >O.02 >0.02 Core Average Isothermal <O for T)91oF- <O for T>100*F Temperature Coefficient of Reactivity Core Average Void -9.99x10-3/cm3 <-3.3x10-^/cm' Coefficient of at 570F Reactivity
- Integrated Reactivity 1.073 x 10-' ~<1.148 x 10 -':2 Due to Temperature Change, 500F-T(a r Or ) '*
Reactivity Worth of <0.039 <O.039 Standard Fuel Assembly +
Value cited is for the Core B arrangement. Values for Core A are less restrictive.
4 (15)
Table 5.3 Calculated Feedback Coe ff ic i ent s fo- RPI LEU Core 0.7647 pcm/cm 3 Core Average Void Coefficient :+
of Reactivity Radial
- Values of the Average Void Coefficient of Reactivity:
Distance from Average Void Coefficient Core Center (cm) (ocm/cma) 0.00 -1.2795 2.97
-1.26078 5.94 -1.14822 8.92 -0.97842 11.89 -0.77206 14.86 -0.56474 17.83 -0.27250
" Values cited along a radial from the core center outward toward a control rod with symmetry assumed.
~~~---------------------------
Isothermal Temperature Coefficient for LEU Core A:
ar(*C) = 1.825x10-"T" - 4.2x10-^T + 6.932x10-5 and a r < O fer T < 16*C(61*F)
Isethermal Temperature Coefficient for LEU Core B:
a r(*C) = 2.113x10-"T" - 5.Ox10-^T 4 1.423x10-
and a r < 0 for T < 32*C(91*F)
(16) m _ . _ _ . - -~
. 6. ]ECHNICAL ScECIFICATIONS.
The attached. Technical Specifications are essentially unchangeo since November 1983, with the following enceptions:
(1) Section 1.0 definitions reflect physical changes in core and control rod make-up, in particular the use o f fuel pins in lieu of plates and the deletion of control rod fuel followers.
(2) Section 2.1, which identifies reactor core safety limits, has been changed to reflect a criterion that limits the amount of energy which may be accumulated in the core to preclude material damage. The limit is consistent with experimental determination and considerate of cladding effects.*
(3) Section 3.1, which specifies limiting conditions for operation, has been ammended to incorporate limits en reactor power as well. The values cited are lower than those previously identified in the 1983 SAR. Section 3.1 also limits the maximum
- reactivity worth of a fuel " pin" vice an " assembly"(a term which described the HEU Core).
(4) Section 3.2, which specifies core reactivity coefficients, requires the temperature coefficient to be negative above ;OO'F vice the value of 90aF previously listed. This recognizes the computed behavior of the temperature coefficient for the LEU Core B design as d4 scribed in Figure 14.4. For similar reactns the integrated temperature coefficient, between minimum operating temperature and the temperature at which it becomes negative, is required to be less than 30.15 vice the previous value af $0.11.
(5) Section 5.4, which pertains to the physical characteristics if the reactcr core, reflects the installation of fuel pins in plat e of the plated fuel assemblies formerly utilized.
(6) Section 6.1, which addresses administrative controls and organizational structure, has been altered to reflect the current hierarchy as it has evolved since 1983.
None of the above changes is substantial nor aggravates the predicted consequences of accident scenarios analy:ed. Moreover, core physics and kinetics parameters fall well within the envelopes of the previous Technical Specifications'.
Experimental verification of all specified parameter va'ues will be undertaken as described in Section 7, in conjunctio ,,i th LEU Core loading.
(17)
- 7. STARTUP TESTING AND EVALUATION.
To safely affect the LEU conversion and measure nuclear characteristics of the new core, a sequence of test and evaluation actions are planned. These actions are summarily described below and presented in the sequence they are to be accomplished during the prescribed test and evaluation period in the work schedule. All testing shall be carried out under the direction of the RCF Supervisor and be the general responsibility of the RCF Director. Results of all testing will be clearly documented and signed by the individual (s) completing the test, the RCF Supervisor, and the RCF Director.
These documents will be maintained on file by the RCF Direc tor for the life of the facility and available for review.
Methods employed in testing closely resemble those used previously and outlined in the " Manual of Experiments" associated with the curriculum". In all cases, results obtained through experimentation and testing are compared with limits preucribed by the Technical Soecifications to ensure compliance or prompt development of new specifications as appropriate. No distinction is made between tests performed for the LEU Core A or Core B configuration, but it is intended that all specified testing shall be carried out on each of these configurations separately.
7.1 Radiation Monitorino.
(18)
The Office of Radiation and Nuclear Safaty shall, upon receipt of the LEU fuel at RPI, and periodically thereafter during conversion, monitor i- 1 record rad i a t io n e.<po sur e levels in the vicini ty of the fuel storage vault and throughout the RCF. Particular attention shall be given to the development of a dose assessment under routine operating conditions--fuel transfer evolutions, core loading, and RCF critical operation.
Contamination surveys shall also be incorporated in the monitoring program. Information gained through such monitoring shall be used to minimi:e personnel exposure both during and after the conversion task.
7.2 Physical Inspection.
In addition to the procedural inspections during the core conversion, a final physical inspection of all reactor components for proper installation and integrity shall be made by a qualified Senior Reactor Operator. The inspection will be made with reference to a checklist, identifying pertinent dimensions, orientations, instrument readings, fastener torques, electrical connections, and such other i n fo rma t i o n as the RCF Director deems appropriate. No subsequent testing or core fuel loading, as described in this section of the SAR, is authorized until the physical inspection is satisfactorily completed.
7.3 Instrument Calibration.
(19)
Area gamma monitors, particulate air monitors, and the RCP criticality detectors shall be verified to.have been calibrated within six months prior to the scheduled initial LEU core loading. (Such calibration is a semi-annual requirement of the Technical Specifications.)
7.4 Interlock Verification.
Immediately prior to LEU core loading and operatico thereafter each of the interlocks described under " Limiting Conditions for Operation" in the Technical Specificaticns shall be verified to be operable.
7.5 Control Rod Drop Time.
The insertion time of each control rod from its fully withdrawn position, with the reactor vessel filled to a normal water level, shall be measured and verified to be less than the maximum limit allcwed in the Technical Specifications.
7.6 Moderatne Dump Time.
Measured from an initial normal-reflector moderator water level. the time required for the auxiliary reactor scram (moderator-reflector water dump) to insert negative reactivity shall be verified less than the maximum time allowed by the Technical Specifications.
7.7 Core Loading.
An inverse multiplication method will be utilized for core loading and initial approach to criticality. The initial loading step will consist o f o ne-hundred fuel pins, and (20)
E subsequent fuel additions will be determined by the extrapolation of inverse multiplication plots until the minimum critical mass is reached. The follow restrictions are imposed en the loading process:
(1) Fuel additions will be limited to one-half the difference between the loaded mass and th extrapolated critical mass or to fifty fuel pins, whichever is less.
(2) Each fuel element will be placed in a lattice position that will preserve, as nearly as possible, a symmetric Core.
(3) The number of fuel pins to be added at any given step can never exceed that of the previous step.
(4) Each fuel pin will be added in a lattice position such that the minimum surface to volume ratio is achieved.
(5) After the minimum critical mass is attained and until the core is loaded to achieve the desire response, fuel additions will be limited to four fuel pins.
7.8 Control Pod Calibratien.
Control rod calibration curves shall be derived for individual rods and for the entire bank utilizing the positive period method and sub-critical multiplication technique.
Initial calibration will examine reactivity changes above critical using small incremental adjustments in rod height to generate a range of periods to which reactivity can be related. Extrapolations of this data will then be utilized to assess control rod reactivity worth belcw critical using sub-critical multiplication calculations over a full range of (21)
red heights. Specific parameters to be determined are excess reactivity, differential / integral rod worths, shut-down margin, and control rod reactivity insertion rates.
A second check of the above measurements will employ a rod / bank drop method, relating the decay of the neutron flur after the event to reactivity worth.
7.9 Fuel Pin Worth.
The reactivity worth of individual selected fuel pins on two separate radii and along the core periphery will be determined by evaluating the required change in critical rod height with the designated pins removed.
7.10 Overall Temperature Coefficient'of Reactivitv.
After attaining critical conditions at a uniform temperature at or above the ninimum allowed for operation, the reactor tank's immersion heaters are utilized to induce a temperature rise which thereby results in a departure from criticality. Movements of a calibrated control rod are then affected ta compensate for reactivity changes due to temperature rise, permitting computation of the overall isothermal temperature coefficient. The reactivity change will be measured from ambient to approximately 140*F.
7.11 Void Coefficient of Reactivity.
Voids are introduced selectively by positioning polystyrene foam in desired locations in one quadrant of the core. A plastic stringer of known volume and material density is (22)
-~ -.- - - _ _ _ -
inserted in the fuel region, and the resultant difference in critical rod heights (under otherwise similar conditions) is compared to infer reactivity worth. In the center of the core, voids are introduced at various heights to also examine any axial variation and permit an average void coefficient determination for the entire core.
7.12 Relative Flux Mappino and Power Calibration.
Initially the absolute thermal flux at the center of the reactor core is determined by irradiating / activating a stable isotope of an element with thermal neutrons and then measuring the activity of that element. (Au"'7 is typically utilized as the target element.) A computation of power is made be relating this activity to the neutron flux incident upon fuel nuclei, utilizing mathematical equivalences and known cross-section values.
A relative flux map is created by measuring the activity of selected fuel pins on contact at prescribed axial locations, accounting for different decay times with a relative decay factor to normalize all data. This information, coupled with the previous calculation o f the absolute thermal flux, perin i t s estimation of overall core power and verification o f instrument readings. Safety system maximum power trip settings are based on this data.
(23) i f
n - , .- ,, -
. - ,-.-..e ,- , , - - - - - -,,-r- , - - , . . . _ , . , , ,
REFERENCES CITED
- 1. D.R. Harris and F. Wicks, "Rensselaer Polytechnic Institute Critical Facility Safety Analysis Report." Docket No.
50-225. License No. CX-22. January 1983
- 2. D.R. Harris, O.C. Jones, F.E. Wicks, A.B. Harris, F.
Rodriguez-Vera, and C.F. Chuang, " Design Basis Transient Analysis for Low Power Research Reactors", Proc. of Int.
Symposium on Use and Development of Low and Medium Flur Research Reactors, Cambridge, Mass., Oct. 16-19, 1983.
Atomkemenergie, Kerntechnik, 44, 450 (1983).
- 3. L.E. Strabridge and R.F. Barry, Nucl. Sci. and Eng., al, 58 (1965).
- 4. D.R. Harris, "PLATAB, a Code for Computation of Equivalent Diffusion Theory Parameters for Strong Absorbers," Tech.
Apl. Associates TAA-1, 1996.
- 5. D.R. Harris, "DIFXY, a Multigroup Diffusion Code for X-Y Geometry," Tech. Apl. Associates, TAA-1, 1995.
- 6. P.E. MacDonald, R.K. McCardell, Z.R. Martinson, R.R.
Hobbins, S.L. Seiffert, and B.A. Cook, " Light Water Reactor Fuel Response Druing Reactivity Initiated Accident Experiments", Proc. ANS Topical Meeting, Portland, Oregon (1979).
- 7. P.R. Nelson and D.R. Harris, " Reconfiguration of the RPI Critical Facility," Nucl. Tech., 60. 320 (1983).
- 8. Richard M. Kacich, "A Manual of Experiments for the Rensselaer Reacter Facility," RPI Department of Nuclear Engineering, May 1989.
(24)
l l
~
- - ~
I i S
D
- N A
\ l t
u o
t .
r -
a S
R A
M NY A W
,, D NR A
' O S -
R N S E
E N C V C N I A
R
\
\ .
N -
- Y T
D I N
A I O C
$} .
R I V
'oc- N D N
s O A X E A T
'00 .
- M I S
2 1 2
e Y
'0 0
r u
g T 1 i
R F E .-
P -
O -
- R P 'o -
C E I L R A T
E C C N E S O L
, Z E Y -
T N L -
I O A L I R I -
S E 5 C U N 2
4 A E L E 6 1
F C G N X .
G O N
L Z E G A D I C R
~
L D I E I L -
L T N - V B -
I I I U N C R I B C 1 . l I
m G ,
M Es UNE I[jy i *l -
1
'5 e
c ,
}h j q .
2: .
,, 1
- 1. g ..
)
! j @ i ti e ! 9 @
1 f fj! [
om i;g =e Y oh f
pI I@If z.
j lik@
- t
/$
a n mm .
y..,p[ig$
[p kiM" $
y w 1;[4.d7' k l i
! pg .
9 i
4p a'"& ~
%@ _r.J w (ey J* &
t .! s
- c. :
j~. l fdh I
- g g ! j~ l @{gj g((
y gyd&
'.t;Q tt
[ ,Y.,
A M e a- }sunu%i{ h +
2 o$
3 1;
ti@{
~
.f;PRtuw.p .l %)k_4%$t.,4a3 e b s m -
l i
i . mi i$ ead '
i .d i
$ $w $ M d d aral E di s i -n d J A. 5
?& N6% 935 . .nJ!gRNIgN yAhy4..
,-ww------
hk -w,---- ,, .-n,--,- -- .- - kkb
- - - F
, e 9
E t ( s
~ ,,
Ew}- r 1 4i$ig T
~
F GE!!( g.)i;. icSH .g g.g sygt.3[nh; g
;-t -
iri d , h,.# ??g.:p i.ggy.g : .
.y qyj wa .s.
3,. .. ;
.. w @v.u:
r . I -- ppy f., Lw I i' , g. j q, u u p 7ha.,g}gg j yw I4/5 1, f- I k
..': .h>
$'Q 4; h9
'~
rgsf. .. -
"i b S l i' ,. ' t(4 d 1 ,,s
,1 -
H b' ' d' j}}fgg'$$5 4.W % l l$ k y [$lh$b '!lf)hEh) I 9 Figure 4.1 h Reactor Tank Structure
$hh .
--,,,,m --,c -,
.---e- y,n-,----,,----------,--.--------en n, - - - - - - - - - - - -
i
.:.:jdh,.,.+L i g hy;t.p-. yt ;'acs u8ccyjgthwg,. e .Q~a.p % ,
; .; .g.
. tywd
.%,d w .'.t f y 1 O...
y %r.clp p 4, ,N,nynxupps,9f.hDy.y:! r? r .. M...y@,9.,iM@W.p..~,3..,..,c.d.]%pigj u Q . m -1' 4 ,g 1@8 @Bg A @J Js,INOgiel y i ~. ql IV nddStwi' ,k - D~'M.fl}j1 ghigly;,gdi{
' .Npc 3
37 s 3 , .iiW.
. i r..
M;L .N$.a @n u -v Nihli'
.R,gl:i.$E. .
h Ipt - F wsy gudg ,;,Ln.'FdMV.yJ/',y rdpfM,.:.ji /MygipT![?$ y -
.Iif$
nng! y:-@v.$$fh f h- g, i t [' y .,, dE i ....syv$y?g$hu.$.bkff rh - !!!O.f!.$kih5,hhk;ksl (AQWu f, ??.L.,g ' n9 n. p. %{a =. 2 ;p7w !fjDQ6TR ~.g . . . lh. 4,. :, e - -# g:;* fM*kd Ajp%y!;WQ i!Gi#;/e,fME i&zd a tcpl,F .,
.... nyd, >
- i. N'i a;g.k,,;.. g ru
.r
. :qqgysp:.n
- i. i rt tp % d m W gaC @. M.. . . . W: ~ - p*)
( d-v3.;gJSy:t;jgClig?f. 3.h(LM"h,g(
%Iifd. %.,g5'A.Jh
,1)(p kI$d h[y NIk., I@fFamMgmmmngEggm., & N. hk
.W..qQ .
.h
,}!. 'b 1
' ,I -
//
e Ew k
- x. ..vyPhy p '
ht;smalsy{3 i.h
,. hbtIi'X:
'd1,?[$0 (b &- '
)%;,[hihihNiG,f&ll
- $lMh,hY//,$a[4 f of Q 1 ,
Mutidjff f
~h r
%. p e Ll ; g [ g i g g. a W Q
~
mfan i 1
) hh e f-
.d$ftkNrdF h$$jj kT 9 E# hI L y hb
.,, 5.hN;(k[ h5.h h L.hk % b ikilijrAd6 f i6H"iFnt![\
uen\ f, $ ynQ,q'R]?,qv1t.-@Nf{h@il$m 2 l h %. n c. ~?5:& w.lH Y 1
.w:.w , .t>
~
k,1.h t yy '
%w'fftq e
5
- n. i .f-a m(,-vi p~ , a - c wa '
MiW7 MM
~
C $%MsE$nM
$$iW3 ': k$$$$$$$$ WG
'2 '
'!' d l I lf[ ! .
t, h ... g i' p,41'it f ,h NM $dgrH.(yi ti # f . dbD .'. [ E55d J.- 3 i$d
- pE3 d'uBstys }'.II. g'd 6gpt,4h$pp1hprljfi.the$mEkk.
x$ $y..h:q$ in X ng h 'sp{lgj~Q!{%ro NI ),t ,huhp... ;gg%..;h 11i Control Rod Drive Figure 4.2
-l
CONTROL R0D CORE t HO 2 CONTROL R0D s oo o 0 O O
\ O o FAST NEUTRON FLUX o
O n ~ O e . O . Ti!ERMAL NEUTRON FLUX . o . j
.. . . . . . *
- o .
O . I o O O O A Figure 4.3 CORE A CONFIGURATION AND FLUX MAP
4 182 0
- i t -
C0tiTROL
- ROD .
CORE II2 0 l C0tiTROL ROD "o o O
- o O
va o
*O O
o FAST NEUIRoll FLUX O
- o
- TilERilAI, HEU1RUti
- FLUX * * *
. . o .
, e o .
. . o a
O . O o g Figure 4.4 CORE B CollFIGURATIO!! MID FLUX !!AP
5 %
. 11.. !,
a gg N ti 's s4 t a I ;y
=
, 3
. .._i _.
_ N !_
.E s i
- !! lI ai l g 00 o000000000000 I . 1 oc i
g. gg oo- ---- o_.__ _-._ee o ._..( o O- 0- - - cm - O, d o og
$- l$
a 00 te o' g ao ~ e o< y
< s, __ _ _ ._ _..g e 1
( .
.o. _ _ _ _ __ __ _ _
.o w
' pc _ __.._ _ _ _ _._._ ..
.O
.o.
c q _ _ _ _.._ _ __ __
- o. _ __ __. . _..__ _.
_g l l e o000000000000000000000000 o 3
-O so
. 1' -E _{ -- --- -
O - OL_ J
-o - -
o _ __ _.g-
- m e
- o- - e _--- o- o c - o(
.S
=
d'o- _ _ _ _ _ _ _ _ p ___ _ __
.oo g oo. - _ _
a _-oo e 09- p -
--oo a 4
O( o J O oo , 00
- 00< 000 4 o< ( 0 0 0 _ '
o ----E. .-p1'000o 00 .. o
.-. . . _- . . ._- u O
E -t
- O O
- .2 =
~
iis e
=
*INI! =
n = i i ! , i d E E i i E ;; ! ! t ! i ai g ;; a g ; , y
- 7' t
Figure 4.5 UPPER FUEL PLATE LATTICE i 4 I i. l y,--.-.. .a..,_.p_.,_.._., ..__,.,c ..~_.-...,.y__._,,.,_g ,m_, i,,..,a,,__,s.- , , , . . . - . , . ._ _ r, -._-_-.._y.__-4, r_ ._ ,,,
CI'tirPCt. ACD paskCr CCNTAINS AS$QAECH SCCTION ' *s
$ bi' k.. V 1s I h k
@ v a_ A ' 2 s
,/ o n
TCP FL ATE h.. f + ya vy. \ ~, h { 3 \ f g
/_._.-, ......
e ii" 11 p >-,. 'J 4 3.T \ 5. ~ ( 6 g 1 /l j Jl 1 Mij '
! ) } ;
,, .- ? ,
1
- h. O s~ I, M,,. \
, .J K-
\ r j , . * **
%> i 3r
#C, "rern
% t a _. _
~ v d : \st a nr .
k y 'c l f : : Q'1L .:,%gF 43 ...." gf - ,e
%, R, QO '
'/h f_ ~~:q;"'
. > s.
o [d,1)d s.
'L
,x cc"h'&cTo o
.. f a v.s s ' x.ess <i ,e
., /t- ))
4 j N/g'JqsM (.'4 "*""y pl.1f 'st arxTCa / 1 : n,,. I, / y ;
. u c g. ...:3 . .v.. A .
Figure 4.6 CORE SUFFORT STRUCIURE An alternative control rod arrangement is shown here. The actual control rod positions for Core A and Core B are rotated through
., a 45' angle about the core center as illustrated in Figures 4.3 ,4.5. . . . . . . . . . _ . . _ . . . _ . . _ . , . . . . . _ _ . . . . . . . . ._.--m .- -
s p 's a llole for Fuel llanxlling 'Ibol u Spring for lloiding U02 Pellets
+ Void / Gap Healon witich allows for tlic Depansion of Fission rasses
<) / %, J.
r/- O, . ~
- '(
- L Yh..
~
l
-'s. :. .
. f - UO 2 Fuel Pellet 1 i L -
'~
Stainlese Steel Claddisq Spert Fuel Pal C!taracteristics U02 Fuel Pellet Density 10.08 quVa1 3 l U-235 Enriciment 4.81 + 0.15w/o U-235 Content per Ikx1 35.2 qin Pellet Outside Dianeter 0.420 in lenoth of Fuel Pod 41.75 in Cladding Outside Diareter 0.466 in Claddino 'lliickness 0.020 in
,. Active Fuel Intigtli 36.00 in '
's. -
Cladding Ic xith 41.62 in s -
~., Ibd overall Intxith 41.75 in s ~.
Figure 4.7 SPERT FUEL Pill _-, _-~ _'- _ - - . . ~ . , . . i- _ - . - - -
s. 4 I\ 1 Control l'al Basket
/' l,
.h
, If '
,f ' t-k' k
10 h, ' D N)sorber Sections
}.'
l-3 0 . i
' control l'al Guide Figure 4.8 CollTROL ROD ASSDfBLY I
__y,_7._ - - _ _ _ _ _ _ , _ _ _ _ _ _ - _ - - - _ - - - - - - - * ~~ '
1, & f '3i# ,- , . i p , ai .i i , lit; l'
~ i t i i i!;} j i l l 4
.I ! I '!I 5 e
!!!) l l !~ l- I II i 1!;i p 'I!!!
i Il ! l. ll I~ l . ! IIII h. p' l-5- fl-l-- REACTIVITY INSERTION TRANSIENT 4 -._ .1.1- -- ._- Il _f-...------ , { -t*
]
I .i ['I ll - --__. . _._ b .
.1k. . _
.ll l- q_. . - -
,l l 1
- j. . . ._. - - ___-- - -- - - -- - - .
- - -- J
\
s
/
/ '
3 1 - ---
.I 1 L--
(
,q j.
10 x . , ,, l l 1
~
e .-- . . _ ! I _L_ I q--- ,
.L _
r_ . . a. __.-. . - . -. l 6_ ,
---- ---- ~
s_ ..I I I ' /\
% M 34_ ___
M y .. .. [ ___.. .- . . . _.._.j. 3 i . 4 3 . . _ _ . _. ...._. 6_ { } i E. a t=5, SCRAM and all rods bottomed - -- ~~ ~ ~ ~~ 2_ . . -
. ~'. . _
^
t 2 10 x , , _ I- h ' I + l N4 ,
.I I i i FT li
~-
_ l- . ._- t--t' 6 .l l}
- 1.
s .... t=0, Insertion of 60c reactivity l
- 3. .._ _. .
- ~
m .- m 1 i t e ,' e$ .T.'[.I",,7.'.a. 0 1 2 3 4 5 .6 7 8. 9 10 11 12
~
1 l TIME (Seconds) Figure 5.1 1
\
+ ,\
' LEII CORE A T
SOLID CORE
/s , g IS OTilERFAL TFJtPERATURF. ' COEFFICI ENT (for 421 pin core) $
a 5stoT- . fit i
.fH.
tre o g 18 */o
, 4 70o /.e G'
- 1
- 1 5 s to < - . / e<'k (*c)=1.8 2.s to'8T'-d.8 e lo' T + (.132 v 10
-Ic> < 10 ' -
p II.*c- Go.f*F
~lS < la' --
t 2s Id' -- O Data point derived from LEOPARD and DIFXY connuter code analysis ! - t i e t6 f - E Data point plotted from nuadratic fit to computer generated csefficients
- 3o (/6 ..
Figure 5.2 I
, . _ , . . ,, -. _ - _ . _ , - _ _ , _ , . . ~ . . , _ , - _ . - - . . _ _ _ -
n
+/5 , faf -
LEU CDP.E B l J ANNULAR CORE ISOT13EIUfAL TUtPERATURE COEFFICIEtir (for 424 pin core)
+ oo s a f -
.U.. , I.
fj[ dy d)= 2,ll 16 s 16 T* s.03'/r a io 'T' + l.'/2 33 < fo Y 4 s ad .
~
M9,
- t i
d . i g n p .. ,, ,, - rfe) A
-Sw5 .
L9 y "T5 3 2,8 *6 - 91.o*F 4
-is s u'~ -
8 Data point derived from LEOPARD and DIFXV f 3g __ computer code analysis
@ Data noint clotted from nundradic fit to computer generated coefficients
*)*a i fb - - - - - . - - - - - - - ~ ----~
Figure 5.3 s}}