ML20086E405
| ML20086E405 | |
| Person / Time | |
|---|---|
| Site: | Rhode Island Atomic Energy Commission |
| Issue date: | 11/18/1991 |
| From: | Seidel G RHODE ISLAND, STATE OF |
| To: | Michaels T Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9112020074 | |
| Download: ML20086E405 (1) | |
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/f[ S'! ATE OF RllODE ISI.AND AND PROVil)ENCE l'i.ANTNrlONS Rhode Island Atomic Energy Commission NUCLEAR SCIENCE CENTER South Ferry Road Narragansett, R.I. 02882-1197 NoverrJ)er 18, 1991 U. S. Nuclear Regulatory Cortnission Ther;dore S . Michaels, Project Manager Non-Power Reactor, Decommissioning and Environmental Project Directorate Mail Stop 1820 Washington, DC 20555
Subject:
Dear Mr. Michaels:
Pursuant to 10 CFR 50. 64 (c) (2) (iii) , we are enclosing a copy of the Safety Analysis Report for your review for the proposed low enrichement fuel conversion of the Rhode Island Nuclear Science Center research reactor.
Very truly yours,
.t>A$eu l.0 ',.
George Seidel "hairman, RIAEC C :cd Enclosure i
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STATE OF RHODE ISLAND AND PROVIDENCE PLANTATIONS RHODE ISLAND ATO::ilC ENERGY COhlhilSSION Nuclear Science Center South Ferry Road Narraeansett, R I 028821197
" SAFETY ANALYSIS REPORT FOR THE LOW ENRICHED FUEL CONVERSION OF THE RHODE ISLAND NUCLEAR SCIENCE CENTER RESEARCH REACTOR" NOVEMBER, 1991 Q%Lo0VAoy
4 O SAFETY ANALYSIS REPORT L)
PART A LEU CONVERSION ANALYSIS PAGE(S)
I I Introduction 1-2 ;
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II Description of Reactor Systems 2-4 III Conversion Criteria and Objectives 4-5 IV LEU Neutronic Core Design 5-6 V LEU Conversion Core 6-7 Figure 1 8 Figure 2 9 Figure 3 10 i
Figure 4 11 Table 1 12 Table 2 13 Table 3 14 Table 4 15 VI Start-Up Accident 16-17 Table 5 17 VII References 17 VIII Replacement Regulatory Rod 18 IX Use of Beryllium Reflectors in the RINSC-LEU Core 19 I
Figure 5 19 i X References for Beryllium Refleeftor Use 20 XI Design Basis Accident 22 XII Appendix A 23 ,
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Part A LEU Conversion Analysis INTRODUCTION This safety analysis repart is submitted pursuant to 10CFR 50.64 which requirm the Rhode Island Atomic Energy Commission to convert its open pool research reactor from the use of high enriched uranium (HEU' fuel to ' he use of Icw enriched ( 1.EU )
fuel. The studies required for the preparation of this report have bee a joint project of the Reduced Enrichment for Research and Test Peactor (PERTR) group at la tonne National Laboratory and the staff at the Rhode Isla Nuclent Science Center (RlMSC). The Rhode Island 4.uomic Energy Commission is responsible for the contents et this report. .
The operating license for this reactor was issued or. July 21, 1964 with an expiration date of August 27, 2002. The original license permitted operation at a power level of 1 MW.
An amendment to the license was issued on September 12, 1988 and permitted operation as 2 MW. Since that time the reactor has operated at 2 MW.
Th^ reactor is multipurpose with capubilit i.e s usually ,
associated with open pool facilities. Because of staffing and funding limitations, utilization has concentrated in two areas-neutrcn scattering and neutron activation analysis. To meet the needs of the research programs, the reactor operates one shift, five da/s per week. As of September 1, 1991, the accumulated operation of the reactor was 47066.3 megawatt-hours. This operation has required the use of HEU fuel elements distributed as tollows:
returned to reprocessing 110 spent, awaiting shipment 26 to reprocessing currently in use 35 new, available for use 13 There are no plans to change this duty aycle This duty cycle allows for operation with an excess reactivity less than that required for continuous operation. Because of t's control blade i
._. _ _ . . __ _ _ _ _ _ . . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ __ _ - . _ _ _ _ _
b configuration, this duty /cle also requires special start-up '
considerations when converting to an LEU core.
The studies performed for the LEU conversion have included calculations for operation at power levels above 2 MW and for advanced core designs. This was done to insure that the l l
conversion process did not ccmpromise the future capabilities of the reactor. This safety analysis report however contains only information necessary for the initial conversion to LEU.
DESCRIPTION OF REACTOR SYSTEMS l
The reactor system is described in the initial safety ;
analysis report /1/. Only a synopsis of system components important to the conversion will be presented here.
The reactor core sits on a 7x9 grid plate with the four corner grid positions occupied by the suspension frame corner posts. These corner posts connect the grid plate to the reactor bridge which spans the open pool. The hollow corner posts each contain a neutron detector required for the operation of the reactor. The grid plate is suspended about 8 meters (26.33 g
feet) below the pool water surface.
This grid plate is installed near the bottom of a grid "]x whose four sides are enclosed, top is open to the pool and bottom connects to an enclosed plenum for coolant flow. The grid box also contains two permanently instal. Led shrouds in which four boral control-safety blades (rods) move. This arrangement is shown in Figure 1.
The grid location of the four boral control-safety blades cannot be moved. Ti' boral regulating rod, however, while fixed in the reflector re. ton of the HEU core, can be relocated.
While some grid positio.s are shown vacant for clarity, during operation each grid cosit._ n must contain a fuel element, a reflector piece, an irradiation basket, or a plug. Otherwise the coolant flow will by-pass the core through the vacant grid position.
The HEU fuel element consists of 18 flat aluminium plates with a thickness of 1.52 mm (.060 inches). The fuel meat is 2
'e' l 0.508 mm (.020 inches) and consists of 93% enriched uranium in a ]
UA1x matrix. The clad is 0.508 mm (.020 inches) and consists of i aluminum. The spacing between fuel plates is 2.54 mm (0.1 inches). When new, each plate contains 6.889 grams of ffranium 235 for a total Uranium 235 element weight of 124 gm.
The operating HEU core is made up of these fuel elements 1
and consists of between 28 and 35 elements surrounded en four i sides by graphite reflector pieces. This core may be characterized as large with a very low power density tesulting in a low thermal flux per unit power. The lightly loaded fuel elements ma i the core large enough to encompass the fixed ,
control blaces. Even with extra ordinary techniques, the l maximum burn-up achievable is about 14%. Positions in the grid J
plate not containing a fuel element or a reflector piece are filled with an irradiation basket. Figure 2 presents a typica'l 30 element HEU core. !
Operation is also permissib e with water reflection of the reactor. During the initial startup of tne reactor, many measurements were made of the characteristics of a water reflected core. However, such a core has never been operated :
above 100 KW because of the requirement of the experimental ,
program and the lack of sufficient irradiation baskets to plug grid positions vacated by graphite reflectors. (Operation using natural convection cooling is limited to a maximum power level of 100 KW).
The core may be positioned anywhere on the center line of a three section, interconnected pool. 6peration using forced convection is only possible in the circular end section where a connection can be made to cooling pipes. In this high power section, the core neutrons are available to six radial beam tubes (three in use), a through tube, a graphite thermal coiumn through which a hole has beer. cut creating an additicnal radial beam tube, and the terminals of two pneumatic irradiation (rabbit) systems. Between the graphite thermal column and the ccre is a permanently installed slab of lead serving as a thermal shield. The thermal shield is cooled by water which is >
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currently forced around *he shield using the pressure difference g
between the inle- and outlet primary coolant lines. l The control of make-up water to the pool is automatic using a float activated motorized valve. This system activates for a drop in pool level of 2.54 cm (1 inch). For a drop of 5.08 cm (2 inches), the reactor, if running, will scram.
The neutron detectors in each corner post provide signals to the control and safety system. Although most of this system is the original equipment, it has been well maintained and is j reliable.
Using a primary pump the core is cooled at 2 megawatts by downward flow of 0.109 m 3 /sec (1730 gpm). Using a stainless steel heat exchanger, the heat in this primary water is transferred to a secondary cooling system operating with a nominal flow of .0631 m3/sec (1000 gpm) and using a forced draft cooling tower.
The reactor is housed in a semi gas-tight, windowless building which uses the confinement concept for the controlled release of radioactivity in the unlikely event of a reactor g
accident. The controlled release is producad by a blower and is through HEPA and charcoal filters and a stack. The release creates a negative pressure dif ferential between the atmosp'lere and the building i..suriag that leaks through the building are inward During the initial start-up phase for the reactor, criticality determinations were made for 17 graphite and water reflected cores. Excess reactivity measurements were made as the core size increased towards the operating core. In addition, cc.itrol blade calibrations and the core thermal flux distribution were experimentally de* . mined.
CONVERSION CRITERIA AND OBJECTIVES There are six basic criteria and objectives of the LEU conversion program. These are:
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- 1. Convert the reactor to the use of LEU using the standard fuel plate which will be provided to university research reactors by the U. S.
Department of Energy.
- 2. Design a LEU core and an operating scheme to achieve burn-up greater than the current 141.
- 3. Design an LEU core which will optimi::e the *. .ernal neutron flux in the beam tubes and will *'2cw for further improvement.
- 4. Design a reactor core with a flux trap for small sample irradiation.
- 5. Design a core which does not preclude future operation at power levels up to 5 MW with the appropriate primary coolant flow.
- 6. Design a LEU core whose initial cost is about the same as the cost of 30 HEU fuel elements since this is the amount allocated for the core by the U. S. Department of Energy.
LEU NEUTRONIC CORE DESIGN l
The neutronic core design has been performed using the standard fuel plate which the Department of Energy will provide for university reactors. Figure 3 presents a comparison of this standard LEU plate with the current HEU plate. Also shown are the characteristics of a LEU direct replacement _ plate . The standard plate is thinner and contains more Uranium-235 than the HEU or direct replacement LEU plate.
The not-readily movable control safety blades are an important consideration for LEU neutronic core design. Because of the more heavily loaded standard LEU fuel plate, the core may become so small that the control blade looses effectiveness.
l During extensive scoping studies, many core configurations were considered /2/. These studies included consideration of:
- a. 18 fuel plate elements
- b. 22 fuel plate elements
- c. several fuel element arrangements 5
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- d. graphite and beryllium reflectsrs
- e. relocation of the regulating rod position, if necessary
- f. use of stainless steel as the regulating rod.
The neutronic calculations have been performed by Argonne National Laboratory using the EPRI Cell, DIF 3D, and V!M Monte Carlo Codes. Incorporating all of the information gathered during these scoping studies and remembering the six conversion criteria and objectives, a LEU conversion design has emerged.
LEU CONVERSION CORE The LEU conversion core consists of a compact configuration using 22 standard plates per fuel element and a combination of graphite and beryllium reflectors.
Figure 4 presents the startup version of the convercion core which consists of 14 fuel elements. The elements contain a total of 275 grams of U-235 each. A central beryllium piece with a 38mm hole is incorporated as a flux trap. The regulating rod is stainless steel and has been moved one grid position so ll as to be adjacent to the compact core. The LEU fuel used is the uranium silicide-aluminum disper'sion fuel approved for use b-j the NRC under NUREG-1313.
Table 3 presents reactivity date on this core. The core is graphite and beryllium Jeflected with an excess reactivity of 3.1% a regulating rod worth or 0.45%, a shutdown margin with blade 3 struck out of 6.7% and a total power peaking factor of 2.64. This design allows the use of existing graphite reflectors along with newly acquired beryllium reflectors.
Because of the one shift operator, the xenon behavior of this core is cyclical and this core can be operated as long as it is possible to operate on Friday morning. Using computer simulation, this core has been "run down" until a Friday morning startup is no longer possible. The reactivity balance is shown in Table 2.
The reactivity requirements for Xe, Sm, long lived fission gg i products, control, and the cold-hot swing is approximately 3%
6
4 which will allow for approximately 14 weeks of operation before it ;
will not be possible to start up on Friday morning. >
After this initial operation, ten beryllium and ten graphite reflector pieces will be reconfigured to provide additional i reactivity. Figure 4 also presents this second core showing the ruel remaining in eacn fuel element after the iritial 14 weeks of operation. The reactivity balance is shown in Table 1 and it allows for an additional 70 weeks of operation. ,
rollowing this second phase of operation, the graphite and beryllium reflectors will again be reconfigured. This third core ,
is shown in Figure 4 which also shows the fuel in each element at the start .of this phase. Table 1 again presents the reactivity balance which now allows for an additional 60 weeks of operation.
Note - that the core is now almost completely beryllium reflected. The core has operated for about 3 years and refueling i is now required. .
Refueling consists of removing the four elements with the most burn-up, placing four fresh elements in the core corner
, positions, and placing the remaining used fuel elements in the t
l remaining positions with'those elements containing the least fuel
- nearest the center of the core. This process provides the flatest t flux and greatest neutron leakage. Eventually an equilibrium core will be reached.
Figure 4 presents this eventual equilibrium core where the
-four elements with the most burn-up have been discharged and four fresh elements have been added to the edge of the core. The
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average discharge burn-up for this equilibrium core is-about 21%,
which is 50% more burn-up than in the current HEU core. The LEU 1.
i fuel used is the uranium silicide-aluminum dispersion fuel approved for-use-by-the-Nuclear-Regulatory-Commission under NUREG-1313.
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O HEU CORE February 24, 1986 30 Fuel and 23 Graphlto Rollector Elements Approx. U 235 Loadings , g per Element 3 2 1 I E E E MMMMM O CIC G .
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Figure 3 -
Description of HEU and LEU Fuel Elements HEU LEU tiumber of Fueled Plates / Element 18 22 Fissile Loading / Element, g 235U 124 275 Fuel Meat Composition UA1x-Al U 3 Si;-Al l Cladding Material 1100 Al l 6061 A1 2 Fual Meat Dimensions 3 thickness, mm 0.508 0.508 j ,
width, mm 52.1 - 61.0 62.7 -
71.1 length, mm 559 -597 572 - 610 Cladding Thickness, mm 0.508 0.381 1 10 ppm natural boren was added to the composition of the cladding and all fuel element structural materials to represent the alloying materials, boron impurity, and other impurities in the 1100 Al of the HEU elements.
O , 20 ppm natural boron was added to the composition of the cladding and structural materials of the LEU elements to represent the alloying materials, boron impurity, and other impurities in 6061 A1. Aluminum with no boron or other impurities was used in the fuel meat of both the HEU and LEU elements.
3 Reference Drawings:
HEU : EG&G #411647
. Plate LEU " EG&G #422873 Plate v
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r Fig. 4. Startup, Transition, and Equilibflum Cores (Lifetimes Based on Operation for 8 Hr/D,5 D,Wk) cons ,
ST ARTUP CORE Core Ufetime:- 70 Wke (- 2000 Ful1 Power Pours)
Core Ufetime: - 14 Wke (- 660 Full Power Hours) '
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CORE 3 EQUluBRIUM CORE Core Ufetime:- 57 Wks (- 2300 Full Power Hours)
Core Ufetime: - 60 Wks (- 2400 Full Power Hours) 0 CC 0 O Cc 0
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l Calculated Data and BOC Excess Reactivity for First Ten Cores l 1
6 Core Lifetime Accum. Operation BOC Express Weeks h'eeks Year: SAk/k l Startup 14 14 0.3 3.0
! Core 2 70 B4 1.6 4.1 f 1
i Core 3 60 144 2.8 :.7 i
Core 4 33 177 3.4 3.0 I Core 5 al 228 4.4 3.6 i Core 6 66 294 5.7 4.0 Core 7 54 348 6.7 3.9 Core 8 53 401 7.7 3,9 Core 9 57 458 8.8 4.0 l '
Core 10 57 515 9.9 4.0 I
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TABLE 2 i Reactivity Balances on the Friday Morning of the I,ast Week of Operation for Ten Cores from -
Startup to Equilibrium :
t t
Reflector Changes Only 4 Burned Elements Removed and 4 Fresh elements Added in Corners !
Startup Core 2 Core 3 Core 4 Core 5 Core 6 Core 7 Core B Core 9 Core la !
t
% Ak/k % ak/L % ak/L % ak/k % AL/k 4 Ak/k 4 Ak/k 9 Ak/L 4 3k/L G ak/L !
I Fresh Cold Clean 3.00 5.19 6.92 6.92 6.92 6.92 6.92 6.92 6.92 6.92 e :
Reactivity Losses
- Burnup 0.30 1.85 3. I 7 3.08 3.09 3.12 3.07 3.06 3.06 3.06 Xe 1.54 1.54 1.54 1.54 1.54 1.54 1.54 1.54 1.54 1.54 i Sm 0.57 0.73 0.73 0.73 0.73 0.73 0.73 0.73 0.73 0.73 Long-Lived F. P. 0.09 0.57 0.97 1.06 1.03 [
1.03 1.07 1.06 1.07 1.07 Cold.Ilot Swing 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 [
Con trol Q,20 n.30 ;
020 010 Q20 020 Q20 020 Olf! .Ul0 Dl0 3.00 5.19 6.91 6.91 6.89 6.92 6.91 6.59 6.90 6.90 l
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() Table 3 Reactivity Data and Power Peaking Factors Start-up 'o re Core 2 Core 3 Core *O 31 41 37 4*O r, y.c e s ; R e a c t i v i t y
- Ak/k Shutdown Margin 67 6.1 -
6.4 lak/k (Blade 3 stuck out)
Worth of Reg rod .45 .41 -
47
'ak/k t
Total Power Peaking 2.64/D6 2.60/D6 -
2.36/D6 Factor / Grid Position (Control Blades Full Out)
Total Power Peaking 3.06/D6 3.05/D6 -
2.81/D6 Factor / Grid Position (Control Blades 50% Inserted)
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s For the LEU cores, additional kinetic 7arameters and reac ivity coefficients were calculated by A!;L. 'l e cortpa ri s ons are shown in Table 4.
Table 4
gna r n r. c o r r. , enra nn Minetics P a r re t e r n Delayed !Ieutron 0.762 0.782 0.776 0.764 Fraction, 0-eff, %
Prompt IJeut ron 76.3 66.2 66.0 68.3 Generation Time, ps Reactirity coefficients: 20-40oC Change in Water Temperature Only VAk/k : 10-4/oC Coolant -1.51 -0.80 -0.86 -0.89 Change in Water Density Only
%Ak/k x 10 ~ / oC Coolant .- 0. 4 4 .0.5 -0,75 -0,64 Coolant Coeff.,
iAk/k x 10-4/oC Coolant -2.0 -1,6 -1.6 -1.6 Doppler Coeff.,
iAk/k x 10~4/oC Fuel OLJ1 -0.18 -0.18 -0.18 Temperature Coeff*,
IAk/k x 10-4/oC -2.0 -1.8 -1.8 -1.8 1Ak/k/1 Void 0.0015 0.0027 0.0025 0.002
- Fuel and coolant temperature changes were assumed to be the same here. The fuel temperature rise will be larger than that of the coolant. Change in Reactivity = (Coolant Coefficient) x ATeoclant
+ (Doppler Coefficient) x AT uel- f i
1 It can be seen that there are no significant differences between the HEU and LED kinetic and reactivity parameters.
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START-UP ACCIDENT This accident was analyzed using a digital computer program PARET/3/. The accident is postulated to proceed under the following assumptions:
- 1. The reactor is in the cold clean condition with power at source level,
- 11. The servo regulating rod is withdrawn, followed by continuous withdrawal of a'l safety blades in succession at their maximum rate.
iii. Period scram protection fails.
iv. The reactor is scrammed by tae high flux sensor i
instrumentation when the power level reaches 2.4 MW (205 overpower).
O v. The delay time from generation of a high flux scram signal ,
to the instant when the safety blades are free to drop is ?
conservatively taken as 0.5 seconds.
The analyses indicate that the maximum fuel temperature (i.e., hot spot in'the hottest enannel) reaches 67.30C (153cF) for ,
the HEU fuel and 88. loc (191oF) for the LEU fuel. Thus, it ccn be concluded that this accident results in no harm to the reactor.
If assumption "111" is modified to - " period and high flux scram protection falls" -
then reactor power would continue to rise-beyond the trip point (2.4 MW) until the negative reactivity introduced by the void and temperature coefficients is greater that the net positive reactivity inserted by blade withdrawal.
Table I provides the peak power and the maximum cladding temperature reached in the cladding for both HEU and LEU fuel cases. In each case, the maximum cladding temperature is less that 1500C (3020F) -
much lower than the 5820C (1080oF) melting 16
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accident could be terminated by a manual scram.
Table 5 Peak Power and Cladding Te.ipera.ures Peak Cladding M Peak Pcwer. 51 Te.mr e r a t u r e . 20 lieu Equilibrium 32.1 149.1 LEU Startup 14.9 146.3 !
LEU Equilibrium 16.2 146.5 l
REFER'.NCES (1) Atomic Energy Commission, Facility License No.R-95, Docket No. 50-193, July 21, 1964 and Construction Permit No. CPPR-73 (2) DiMeglio, A.F., Matos, J.E., Freeese, K.E., and Spring, E.F.:
The Conversion of the 2 MW Reactor at the Rhode Island Nuclear Science Center. Proceedings of 1989 International Meeting on Reduced Enrichment for Research and Test Reactors, Berlin, West Germany, Sept embe r , in press (3) Obenchain, C.F., "PARET - A Program for the Analysis of Reactor Transients"" 100-17282 (1969)
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REPLACEMENT REGULATING ROD e
The current regulating rod in the HEU core is located in the f j D-1 grid position (refer to Figure 2 in the " Description of the l
- Reactor System" section of this SAR). The rod is fabricated from boral plate and aluminum. ,
1 Calculations from ANL(li indicate that the regulating rod ;
worth in its present core position (D-1) is reduced from its present value of .48% Ak/k to .2 .31, Ak/k which is too little. If ,
the present rod were to be relocated to the D-2 position, it ,
increases to .8 .9% which is too large. The regulating rod limit by techuical speelfication is .6% Ak/k. Therefore a new stainless steel regulating blade is necessary in the D-2 position. ;
Calculations indicate a satisfactory worth of .4 .5% Ak/k results.
The new regulating blade will be fabricated with the same ;
dimensions to. properly fit the core grid box. All-references in ,.
this SAR relating to the new LEU cores are made with the new '
I regulating rod as part of the core arrangement.
i (1) Memo from James Matos, ANL to RINSC, Eugene Spring, I. September 16, 1991 i
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r USE OF BERYLLIUM REFLECTORS IN THE RINSC-LEU CORE The proposed use of Beryllium reflectors in the LEU start up and equilibrium cores has been reviewed. The Beryllium reflectors are currently being designed by EG&G Idaho. The University of Missouri Reactor M25 has been using Beryllium and has conducted tests to determine a lifetime limit based upon a fast fluence level. ReferencedO indicates that embrittlement of Beryllium is first noted often approximately 3x1021 NVT. As a result of HFIR determination of small cracks occurring at a 1. 8x10 u2' NVT level, a proposed changeout level of 1x1021 is proposed. Using a maximum flux of 3.3x1013 and a 5 day, 7 hrs / day reactor operating cycle, a proposed changeout of 45.8 years is predicted.
Little change in other Beryllium properties at our operating temperatures and integrated fission neutron dose occurs up to the propose limit.'28*
- W *
- Gamma heating has been reviewed
- G3' and poses no problem. At present EG&G has reviewed the Beryllium materials available and has developed the specification for use in our element fabrication.MM Final drawings are due shortly. A standard element is proposed similar to the graphite elements. A special element, with a 1.25cm center hole, will be designed and used as a flux trap for special experiments (see Figure 3) . The maximum cal.culated flux in the Be portion of the flux trap is 3,3x1013 A removable plug would be used to fill the hole when not in use. The RINSC would require a technical specification change for use of Beryllium reflectors, e
' Table 5 shows the average midplane -lux in the Be of the contral flux trap.
Figure 5 Grp Upper Lower IL2 Energy Energy Startup core 2 core 10 l
1 14.0 MeV 0.921 MeV 2 . 5 x1013 2 . 4 x1013 2.4x1013 2 0.'821 MeV 5.531 MeV 3.3x1013 3.3x1013 3.2x1013 3 5.531 kev 1.855 ev 2.7x1013 2.7x1013 2.7x1013 4 1.855 ev 0.625 eV 3.7x1012 3.7x1012 3.6x1013 5 0.625 eV 0.251 eV 3.3x1012 3.3x1012 3.3x1012 O 6 0.251 eV 0.057 eV 1.8x1013 2.5x1013 1.8x1013 2.5x1013 1.9x1013 7 0.057 eV 0.00025 eV 2. 6x1013 19
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RETERENCES FOR BERYLLIUM REFLECTOR USE (1) ASME, 74-PUP-44, " Stress and Deformation Analysis or Irridation Induced Swelling" by B. V. Wi"kel, 1974 (2) " Surveillance Testing and Property Evaluation of Beryllium in I Test Reactors", J.M. Beeston, M.R. Martin, C.R. Brinkman, G.E. Morth, and W.C. Francis, Aerojet Nuclear Company, Idaho j Falls, Idaho (U. S. Atomic Energy Commission Idaho Operations Office under contract number AR(10-1)-1375)
(3) The Mechanical Properties of Some Highly Irradiated Beryllium, J.B. Rich, G.P. Walters and R.S. Barnes, Atomic Research Establishment, Metallurgy Division, March 1961 (4) Properties of Irradiated Beryllium Statistical Evaluation J.M. B eston, EG&G Idaho, October 1976 (5) Missouri University TM-ERS-62-1, M00-30203, June 22, 1962,
" Stress and Thermal Analysis of the Beryllium Reflector for the University of Missouri Reactor" (6) General Electric Co. Atomic Power Equipment Department-Standard 788, " Beryllium, Hot Pressed, Nuclear Grade" (7) "The Ef fects of Neutron Irradiation on in ryllium Metal", D.S.
Hickman, The Institute of Metals Conference on the Metallurgy of Beryllium, October 1961 (8) "The Effect of High-Temperature Reactor Irradiation on Some Physical and Mechanical Properties of Beryllium, J.R. Weir, The- Institute of - Metals, Conference on the Metallurgy of Beryllium, October 1961 (9) "The Behavior of Irradiated Beryllium, R.S. Barnes, The Institute of Metals, Conference on the Metallurgy of Beryllium, October 1961 20
_. . ; _ . _ . . , _ _ ..-_._u_... .;- .. - , - _ . _
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l (10) University of Missouri, Inter-Department Correspondence, Gerald Schapper; Beryllium Reflector Changeout, December 16, g 1 1975 (II) EGSG Idaho, Material Specification, Beryllium Pressings and Components for Nuclear Reactors and Reactor Systems, Document No. ANC-80005G, April 26 1978 (12) University of Missouri, Specification Drawing Seryllium Reflector, Drawing No. 193, October 6, 1988 1
1 (13) FAX memo from Argonne National Laboratory to Rhode Island '
Nuclear Science Center, W.L. Woodruff to Eugene Spring; l
Subject:
Gamma Heating in Beryllium Reflectors, March 19,
! 1991 l (14) FAX memo from Argonne National Laboratory to Rhode Island Nuclear Sciences Center, James Matos to Eugene Spring,
Subject:
Flux in Beryllium, September 19, 1991 0
O 21
l . .
i I DESIGN BASIS ACCIDENT i
The design basis accident for this reactor has been a loss of 3
coolant accident with the water draining through a heam port containing no plugs. Recal. that the core sits in a grid box and draining of this box is through a 1.25 cm hole drilled in the
' bottom. Because of this, about 17 minutes is required to complete the draining, after wh 3 c.h the bottom 2 '. cm of fuel remains in i water. It has been possible to show that the low power density HEU core will not melt after this hypothetical loss of coolant accident.
I' The LEU core has a higher power density than the HEU core.
Using the same accident sequence and calculations which were used r
will not suffer some melting following a loss of coolant accident.
The LOCA *ssumes a gillotine severence of the end of a beam port in the pool with water leaving an open beam port end to the
, reactor room main- floor level. The data, discussions and calculations are shown in the thermal hydraulic section of the report (Part B). !
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APPENDIX A LEU FUEL SPECIFICATIONS /dlD DRAWINGS i
EG&G IDAHO INC.
(A) TRTR-6 Specification for Test Research Training Reactor LEU Silicide U3Si: Fuel Plates Rev. 4, 20 May 1988 (B) TRTR-11 Specification for Low Enriched U Metal for Reactor Fuel Plates Rev. 1, 1 April 1987 (C) TRTR-14 Specification for Reactor Grade Uranium Silicide U3Si2 Powder j Rev. 2, 1 July 1987 I (D) TRTR-15 Specification for Aluminum Powder for Fuel Plate Core Matrix hev. 2, 1 July 1987 EG&G DRAWINGS (A) Test Research Training Reactor LEU Fuel Plate No. 422264 (B) Rhode Island Nuclear Science Center Test Research Training Reactor 5 Fuel Plate No. 422873 (C) Rhode Island Nuclear Science Center Test Research Training Reactor 5 Side Plate No. 432325-(D) Rhode Island Nuclear Science Center Test Research Training Reactor 5 End Box No. 411649 (E) Rhode Island Nuclear Science Center Test Research Trr.ining Reactor 5 Fuel Element Assembly No. 411650 c.
O 23
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4 SAFETY ANALYSIS REPORT PART B THERMAL HYDRAULIC ANALYSIS PAGE(S)
I Introduction 1 II Description of Computer Programs used 1 in the thermal-Hydraulic Analysis III LEU Parameters 2 IV Hot Spot Factors 3 V Steady State Full Core Analysis 4-9 VI Single Channel (Hot) Analysis 10-12 VII Natural Convection 13-14 VIII Phode Island Nuclear Science Center Water Supply 15-16 IX Loss of Coolant Analysis 17 X Emergency Core Cooling System Operation 18 XI Water Supply Analysis 19-20 XII Appendices Appendix A/ LEU Thermal Conductivity calculation 21-22 l
Appendix B/ Critical Velocity for Fuel Plate 23-24 Deformation Appendix C/ Loss of Coolant 25-27
, Appendix D/ Decay Heat Calculations 28-32 l
Appendix E/ Maximum Heat Flux 33 Appendix F/ Maximum Core Specific Power 34 O
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. 1 Part B Thermal Hydraulic Analysis INTRODUCTION The thermal hydraulic studies for the LEU core have been a joint effort by the Rhode Island Nuclear Science Center (RINSC) and Argonne National Laboratory. The proposed new fuel elements have been described in the main introduction of the Safety Analysis Report. Pertinent documents reviewed by the RINSC for LEU fuel use are referenced in Appendix A.
Fuel plate, channel dimensions and other parameters used in the thermal hydraulic studies are hereby referenced in the Appendix A documents.
DESCRIPTION OF COMPUTER PROGRAHS USED IN THE THERMAL HYDRAULIC ANALYSIS The computer programs used by the Rhode Island Nuclear O Science Center Steady-State Analysis, Hot Channel Analysis and Natural Convection Analysis were obtained from Argonne National Laboratory. The programs were supplied as a VAX/ Fortran Version and were subsequently converted for use on an Apple (Macintosh II) computer using the_ "Absoft Compiler". This was performed so that the staff could utilize in-house computers.
The program entitled "PLTEMP" was used to perform the
" Steady-State" and single Hot Channel Analyses, i
The program ent. i t le d "NATCON" was use to perform the Natural Convection Calculations.
i O
I
LEU PARAMETERS 6O The parameters used in the "PLTEMP" and "!MTC ON " programs were calculated using the " LEU Fuel Element" and the proposed core configurations. Previous sections address nuclear parameters.
The physical dimensions of core components used were obtained frcm current drawings.
In addition to the normal dimensions of core components used in the Thermal-Hydraulic Analysis, the LEU fuel thermal conductivity was calculated. This information is as shown in
, Appendix A.
Another parameter studied is the " Critical Velocity for Fuel Plate Deformation". This analysis- is shown in " Appendix B".
Below is a list of core components and their respective drawing numbers used as reference data.
COPE COMPO11E11? DRAWING 1: UMBER LEU Fuel Plate EG&G #411650 Radiation Basket (w/ orifice plate) GE #798D413 Control Blade GE #197E647 Servo Control Element (Regulating Blade) GE #612D964, 762D407 Graphite Reflector (also Be reflector) GE #985C248 Antimony Beryllium Source GE #655C430 Radiation Basket (center hole type) GE #798D413
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HOT SPOT FACTORS The use of the LEU fuel element necessitated an O
evaluation of the engineering hot spot factors to be used in the single het channel analysis. The Rhode Island flu cle a r Science Center has prepared a report entitled " Report on the Determination of Hot Spot Factors for the Rhode Island IJu cle a r Science Fesearch Using LEU Fuel." The report was performed in August 1989. The results are shown below:
Eb (Bulk Water Temperature Rise) = 1.62 Eg (Heat Flux) = 1.46 Fh (Heat Transfer) = 1,41 These factors were used in the single channel (Hot Channel) analysis to determine a " limiting power level" based upon incipient boiling utilizing the PLTEMP Program.
9 3
STEADY STATE FULL CORE ANALYSIS The PLTEMP Program was used to analy::e the full core for the axial peaking factors of 1.32 (blades out) and 1.536 (blades 50% in) core conditions. The full core analysis included the various components (fuel elemen*s, reflectors, baskets etc.) as sho,wn in Figure 4. This antlysis initially determines the flow rates for the fuel portion of the core, the by-pass flow through the other components and the total core flow versus the pressure drop across the core. This data is tabulated in Table A. The fuel platie surface l temperature vs flow rate is shown in Table B. It is I important to note the the maximum fuel plate surface temperature does not vary by more than 3.52 degrees centigrade for the two axial factors (F axial = 1.32 and 1.536). From the tabulated data and our pump flow of about 1730 GPM, the core flow (1100 GPM t ) and a by-pass flow (625 GPM t
) is determined. The corresponding AP .0055 MPA. The results are graphically depicted as LEU core " Flow vs. DP".
The output of the program also determines a number of other parameters. A list of these for the steady state 2 MW P
operation case is shown in Table C. The axial peaking factor vs. relative blade position for the core is tabulated for both the blades out condition (F axial - 1.32) and the blades 50% in (F axial = 1.536). This data was obtained from the nucleonic studies of the core (D (see Table D). This data was input to the PLTEMP Program to calculete the various parameters at a point by point basis along the axial plate length. The maximum plate surface temperatures shown in Table B reflects these values.
It should-be noted that the highest power peaking factor.
occurs in core position D-6(2), for both the blades out and blades 50% in situations.
O 4
. . _ _ . . _ _ _ _ . _ . _ _ . _ . . - _ . , _ . _ _ _ , . - ~ - . - . _ - - - - _ _ . - . - - .
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ERENCES s.
> s
]. ' I.g f]) Memo from Bill Woodruff, Argonne National Laboratory to
- j l Eug.ine Spring, Rhode Island Nuclear Science Center, J, f!$9 1/3
, ,2 ~q
.' 'S,%
"#~ 9i? ~. ', (2) Memo tram James Matos, Argonne National Laboratory to A.
c> . ,g Center, r' . OiMeglio, Fhode Island Nuclear Science 1/22/91 O
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I LEU FULL CORE ANALYSIS - 14 ELEMENT CORE 2 MW TABIE A By I' ass , Total Flow Core Flow Ily-Pass Total Flow DP (MPA) Core Flow (kg/s) Flow (kg/s) (GPM) (GPM) (GPM)
(k g/s) 69.30 698.00 407.00 1105
.0025 43.81 25.49 76.71 774.00 149.00 1223
.0030 48.56 28.15 30.63 83.61 844.5 488.50 12'3
.0035 52.98 90.80 910.8 536.20 '+7
.0040 57.14 3 3.e6 96.20 973.5 559.50 1533
.0045 61.07 35 .3 102.04 1033.4 593.60 1627 0050 64.83 37.21 107.62 1090.4 624.20 1715
.0055 68.43 39.19 4; 09 112.98 1146.00 654.00 1800
.0060 71.89 118.16 1199.00 684.00 1883
.0065 3 5.23 _ 24.71 TABLER Outlet Bulk Plate Outlet Bulk Plate fly-Pass Core Flow Total Flow Ap Temp "C Surface Temp 0C Surface Flow Temp "C (y p3 3 (GPM) (GPM) ,iPM) Temp "C F axial =1.32 F axial =1.32 F axial =1.536 F asial=1.536 54.70 74.06 54.70 77.52
.0025 407.0 69S.0 1105 71.49 53.44 74.71 ,
449.0 774.0 1223 53.49 0030 __ 52.56 69.69 52.56 72.49
.0035 488.5 844.5 1333 51.81 68.03 51.81 70.69
.0040 536.2 910.8 1447 51.20 66.65 51.20 69.I8
.0045 559.5 973.5* 1533 50.69 65.29 50.69 67.89
.0050 593.6 1033.4 1627 50.25 64.28 50.25 66.78
.0055 624.2 1000.8 1/15 49.86 63.56 49.86 65.80 654.0 1146.0 1800
.(1060 49.53 64.94 1199.0 1883 49.53 62.77
.0065 684.0 64.I6 1969 49.23 62.07 49.23
.0070 719.0 1250.0 N(7ES: (1) Normal Primary Pump t ation 1730 GPP r (2) Calculations llased on Intct Temp.1o Core of 42.30C 3
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TABLE C -
O The PLTEMP full core analysis for each fuel alement i
calculates additional parameters. A typical output value is shown for these parameters. (Element #8 data)
PARAMETER VALUE Maximum Surface Temp. C 64.45 Clad Temp CC 64.78 Peak Axial' Heat Flux (MW/M2 ) .156 Channel Flow Rate (kg/s) .2298 Velocity (M/S) 1.605 l Cutlet Pressure (MPA) .1727 l CHF (Critical Heat Flux) (MW/M2 ) 2.39 Flow ~ Instability (MW/M2 ) ,ggg E.vit Saturation Heat Flux (isW/M2 ) 1.087 Minimum DNB Ratio 8.461 F axial 1.32 h .
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TAlllE D AXIAL DISTRil1UTION INTliRI ACE REl.ATIVII DISTANCE IlLADE-OlJr IILADE-IN 50%
1 0.00 4105 .0553 2 0.05 .5506 .2682 3 0.10 .6836 4759 4 0.15 .8076 .6744 5 0.20 .9219 .8597 6 0.25 1.0228 1.0283 7 0.30 1.1111 1.1770 8 0.35 1.1850 1.3028 9 0.40 1.2435 1.4032
- 1O 0.45 1.2858 1.4764 11 0.50 1.3200 1.5360 12 0.55 1.2858 1.4764 13 0.60 1.2435 1.4032 14 0.65 1.1850 1.3028 15 0.70 1.1111 1.1770 16 0.75 1.0228 1.0283 17 0.80 .9219 .8597 18 0.85 .8076 .6744 19 0.90 6836 4759 20 0.05 .5506 .26h2 2I I.00 .4105 .0553 r
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SINGLE CRANNEL (HOT CHANNEL) ANALYSIS
.O Computer runs using PLTEMP were run for the single channel analysis using the derived hot channel factors. Flow rates and power levels were varied to provide sufficient j information for " limiting power level and core flow terminations.
The tabulated resul.s for axial factors or 1,32 (blades in position)'and for axial factors of 1.536 (blades 50% out) are shown in Table E. The results are also presented as a
" Hot Channel Fuel Surface Graph" depicting " Fuel Temperature" vs. Total Core Flow.
The normal primary flow rate for the Rhode Island Nuclear Science Center reactor is about 1730 GPM. From the data it can be seen that incipient boiling occurs at about 2.6 MW or 130% of the normal 2 MW power level. At a reduced flow of abut 1580 GPM incipient boiling is reached at about 2.4 MW or 120% of normal power. 'The proposed limiting safety settings are then chosen as shown'below:
Normal Power Level Over Power Trip (scram) 2MW 120% ( 2 . 4 MW)
Normal Flow Reduced Flow Trip (scram) 1730 GPM 1580 GPM These values are more restrictive than the present trip levels of 130% for overpower trip (2.6 MW) and 1260 GPM flow.
This is due to the fact that the compact core of 14 elements and higher fuel density have more effect than the increase-in number of fuel plates from 18 to 22.
The maximum surface temperature of the fuel resulting at the 1S80 GPM pump flow is from Tabel E about 110oC.
The corresponding coolant velocity from the PLTEMP output for 1533 GPM (AP=.0045) = 1.44 M/S and for 1627 GPM (AP=.0050) = 1.53 M/S. An extrapolated value for the 1580 GPM condition is about 1.48 M/S.
These safety settings will require technical O
V specification caange upon NRC approval.
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TAllLE E IlOT CilANNEL ANALYSIH F AX1AL = 1.536 !
AP Total How T Surface T Surface T Surface .T Surface T SAT. T onb -I (M PA) WP 2 2.2 2.4 2.6 MW MW MW MW
.0025 1105 , 22.26 122.57 122.86 123.10 115.82 122.1
.0030 1223 122.27 122.57 122.87 123.15 115.82 122.1
.0035 1333 119.27 122.58 122.88 123 16 115.82 122.1
.0040 1447 114.99 121.(0 122.89 123.17 115.82 122.1
.0045 -1533 111.44 117.71 122.90 123.19 115.R2- 122.1
.0050 1627 108.40 114.41 120.36 123.20 115.82 122.1
.0055 1715 105.14 I11.54 117.24 122.93 115.82 122.1
.0060 1800 103.39 109.00 114.53 120.00 115.82 122.1
.y65 IR83 101.30 106.75 112.10 117.39 115.82. 122.1
.0070 1963 94A42 104.71 109.93 115.06 115.82 122.1 IIOT CIIANNEL ANALYSIS i F AXIAL = 1.32 I Ap I ToIai Flow T Surface T Surface T Surface '
T Surface - T SAT. T onb (yp3) WPM) 2 2.2 2.4 I.6 T T MW MW MW MW
\
.0025 1105 120.80 122.09 122.36 C 2.62 - 115.82 122.1 !
.0030 1223 115.01 121.52 122.37 122.63 115.82 122.1
.0035 1333 110.36 - 116.52 122.38 122.64 115.82 122.1
.0040 1447 106.55 112.40 118.16 122.65 115.82 122.1
.0045 1533 '03.38 108.94 114.44 119.86 115.R7 122.1
.0050 1627 n 00.67 105.99 '111.26 116.48 115.82 122.1
.0055 1715 98.30 .03.43 108.49 113.51 115.82 122.1
.0060 1800 96.21 101.18 106.07 110.90 115.82 122.1 0065 1883 94.36 99.17 103.91 108.58 115.82 122.1 '
.0070 1963 92.69 97.36 101.47 106.51 115.82 122.I I1 i 9 .
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12
NATURAL CONVECTION The present HEU core has a licensed limit of 100 kw operation for the reactor in the natural convection mode (No Primary Pump Operation).
The 11a t u ral Convection Analysis for the LEU core was performed using the NATCON Program. The program was rur. for both the regular channel and the " hot channel" conditions.
Both cases were run for the blade out situation (F axial =
1.32) and the blades 50% in (F axial = 1.536).
The results are shown in Table F. Note that for the most conservative case, (hot channel) the power level is 217.3 kw, using incipient boiling as the limiting parameter.
The maximum wall temperature was calculated as a function of axial length and the value was tabulated from the data. The program run terminates when the fuel surface temperature reaches incipient boiling.
Since the 217.3 kw exceeds the current licensed power level of 100 kw for natural convection, no change is deemed necessary in the licensed maximum natural convection power level of 100 kw.
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TABLE F NATURAL CONVECTION ,
REGUI.AR Cil ANNEL FAXIAL = 1.536 Power Level Exit Maximum Incipient T sat-T wall Radial Margin to (kw) Te m p. Wall Boiling "C (I l7.34-TW) Peaking Incipient Boiling Temp. "c Factor o c
- 10 's9 52.01 117.57 65.34 2 65.56 ,
100 63 71.22 118.28- 46.12 2 46.06 l 200 77.29 84.71 118.88 32.63 2 34.17 l 300 83.80 96.48 119.21 20.86 2 22.73 500 93.69 117.85 119.85 .51 2 2.00 520?4 94.56 119.89- 119.89 , -2.55 2 0.00 IIOTClIANNEL 10 59.18 5 R.89 117.72 58.45 2 58.83 ,
100 86.63 ' 92.81 118.56 24.53 2 -
25.75 209.1 102.11 I19.34 j 119.20 -2.00 2 -0.40 i NATURAL CONVECTION RECULAR CilANNEL FAXIAL = 1.32 Power Level Esit Maximum Incipient T sat-T wall Radial Margin to (kw) Temp."C : Wall Boiling "C "C Peaking incipient Boiling Temp. Oc Factor og r a r M
10 52.1M 52.26 117.60 63.08 2 65.34 100 ( .72 .71.58 118.19 45.76 2 46.61 200 77.69 84.25 118.72 33.09 2 34.47 300 84.26 95.11 119.00 22.23 2 23.89 400 89.6I I05.28 119.35 12.06 2 14.07 500 94.24 114.45 119.67 2.89 - 2 5.22 -
558.45 96.69 119.80 119.80 -2.46 2 0.00 i IIOT CilANNEL 10 59.38 59.80 117.65 57.54 2 57.85 100 87.12 92.92 118.43 24.42 2 25.51 i
'17.3 103.66 119.14 119.03 -1.80 2 0.00 14 .
. - - _ - . ._ . . . . - --- - -. ..- - - . ~.
RHODE ISLAND NUCLEAR SCIENCE CENTER WATER SUPPLY
.O5 The Wakefield Water Company supplies water to the University of Rhode Island (URI) !!a r ra gan s et t Bay Campus.
The water is pumped to the 300,000 gallon water storage via a 8" feedline. Water is then distributed to the URI Bay Campus I (including the Rhode Island flu clea r Science Centet (RI!!SC ) )
through a 12" main. Water is supplied to the RIf1SC through an 8" line which feeds fire protection (6" line) and potable supply (2" line) and a reactor bu11 ding fire hose (4" line).
The entire- pumping system has backup generators for t.otal supply reliability.
The enclosed letter from the URI Graduate School of Oceanography which oversees the Bay Campus water supply, can meet a minimum demand of 5 GPM or greater for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, even in the event of a power failure.
The Bay Campus can provic'e uninterrupted water rupply in the event of a line rupture or planned shutdown by utilizing t variaus cross connections and hydrant connections located in the system network.
O 15
The Vn yerSily Of Rhode Isthna Graduate Senoor of Ocet.nogrsn, D .am;rse aa, :a :_; Nwa;ame a neesT .
OL July 12, 1991 Mr. Eugene F, Spring, Sr. Reactor Facility Engineer fluclear Science Center South Ferry Road Narragansett, RI 02882-1197 Subject : Emergency Water Supply Dear Eugene.
The chart below indicates that our water system, which includes a stand-by generator, will fullfill your cooling water requirements under various conditions.
Pressure (PSI) Volume (Gal)
Condition w/o Fire Pump Available Duration flormal Town supplydeservoir 8 coster Pumps 70 * *
- Reservoir With Booster Pumps 70 300,000 24 Hrs.
O l U- Town Supply l With Booster Pumps 70 *
- Town Supply 0nly 30 * *
- = Unlimited within present demand ,
Campus (Max) 200 GPM l Reactor (Min) 5 GPfi Total 205 GPM -l
&Je#
Kenneth W. Morrill ,
Asst. Dir. Physical Plant 16
LOSS OF COOLANT ANALYSIS Following a postulated loss of coolant accident, the pool drain tire is calculated by using the falling head calculational met hod . (1) This is considered the maximum credible accident (refer to SAR Part A-Section XI). It is assumed that water drains from the beam tube (8") from the pool surface (el.139.417) to the tcp of the core box (115.916). Water then drains from the 1/2" diameter hole in the core box. Water cannot drain below the bottom (invert) of the 8" beam tube and therefore about .7' of water remains in the core box above the active fuel plate edge.
Appendix C shows the schematic and the calculations determining the pool drain time and the flow rate to keep the core box full. This is the minimum drain time, conservatively assuming that the beam port shutter is up, no plugs are in the tube, and no flange cover bolted over the outlet flange.
The original RINSC license amendment for 2 MW operation calculated the decay generation and heat removp1 following a postulated LOCA for the HEU core. The proposed HEU core has a higher heat density per plate.
Appendix D shows the same simplistic calculation. The results show that the heat removal is not sufficient to remove the decay heat after a LOCA. The conclusion is that un emergency core cooling system is necessary.
The design and operation of such a system is discussed i Section X.
(1) Handbook of Hydraulics, Ernest F. Brater, 6th Edition, McGraw-Hill Bock Comparv, 1976 O
o
EMERGENCY CORE COOLING SYSTEM OPERATION Under normal operating conditions, make-up water is supplied to the reactor pool via the automatic make-up system. In an emergency, for a pool level drop of 2", the automatic value (NC) opens to supply water at the pool level at a design rate of 20 GPM. A manual by-pass can be opened to supply additional flow. Present operating procedures describe le procedure for piping and valve alignment and procedures for normal and emergency filling of the pool. An emergency core cooling line will be installed directly to the reactor core grid box to provide a water supply directly to the fuel elements in the core.
The Emergency Plan (4.1.5 Utilities Failure) directs specific actions to be taken following a drop in water pressure (Implementing Procedure 3.3.1). At present, the detection of a loss of/or drop in city water pressure alarms O at the secretary's " desk alarm" box which notifiss the operator at the reactor console with other alarms lumped together as a " vital access alarm". The operator must check with the " desk alarm" in order to take appropriate action.
It is proposed, as part of the ECCS, that a low city water -
pressure signal be directly connected to the reactor scram circuitry.
Modific.itions to the current Emergency Plan and ,
Implementing Procedures would need to be performed.
To insure that such a proposed ECCS be adequate, the RINSC has conducted a water supply analysis to calculate the expected flow. The water flow to the pool has been observed in the make-up system to be about 25 GPM. (Tests will be conducted to verify the actual flow rates and pressures).
The design and installation of the ECCS would be performed in accordance with the RINSC QC/QA program.
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WATER SUPPLY ANALYSIS O
The analysis of the facility water supply systen. was
-performed using a computer program called " Service Sizer".M)
It calculates pipe size and demand. The program has built in piping : a b '. e s , valve and fitting tables and !ixture unit tables. Standard corputation techniques are used to determine losses. The program allows any fixture to be specified in either a public or private use situatica. Input to the pipe size calculation includes demand flow, demand pressure elevation difference, supply pressure, pipe length, other equivalent pipe length losses, numbers of valve an.
fittings and also a permitted velocity. The program calculates, pipe size, act ual velocity, head loss and demand pressure.
The demand calculation includes options for flushometer units; public use or private use. Input for the calculation includes numbers and types of fixtares, a continuous demand flow and additione.1 fixture option. The orogram calculates total fixture units, continuous and fixture damand and total demand.
The encicsed report shows both the demand and calculated supply size for a proposed water line extension to insure an adequate supply tc the reactor core in case of a LOCA.
T h e. report shows that a 2" line will produce 42 GPM.
This line size is adequate for normal reactor cool supply and certainly for a 5 GPM supply in a L3CA situation.
(1) Parkcon, Inc., 250 N. Center Street, F. O. Box 5980, Woodland Park, Colorado 8086-5980 ,
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--SIZING CALCULATION------------------------------------Printed On: 7/18/1991 Supp]y Location:
60.0 psi, supply pressure available during demand Demand Location:
42.0 gpm demand flowing at 40.0 psi pressure
--Head Loss Data--------------------------------------------------------------
Elevation Difference: 30.0 ft (minus if demand location lcwer than supply)
Pipe Length: 142.0 ft Other Loss In Equivalent Pipe Length; ft Number of Valves & Fittings:
- Corp Stop : Curb Stop 3: Gate Valve : Globe Valv : Angle Valv
- Bfly Valve : Swing Chk Side Tee : Straight T 13:Std Elbow
- Long Elbow 3:45 Elbow : : :
Backflow Prev: 1.0 pai Water Meter: pai PRV: psi Other: psi
--Design Calculation-------------------------------- -------------------------
Permitted Velocity: fps Pipe Type: CUM Calculated Pipe Size: 2 in
() Actual Velocity: 4.2 fps Head Loss: 17.1 psi Pres at Demand: 42.9 psi
--DEMAND CALCULATION----------------------------------------------------------
l Predominantly Fluehometers: N Public Use: N
--Number of Fixtures----------------------------------------------------------
- Bathtub :Bar Sink : Bidet : Clothes Washr
- Cuspidor : Dishwasher 1: Drinking Ftn : Hose Bib ,
1: Kitchen Sink : Lavatory : Laundry Tub 1: Shower Head 1: Service Sink : Urinal Fedest 2: Urinal Wall : Urinal Tank
- Wash Sink :WC Flushometr 2:WC Tank :
Additional: fixture units Total: 23.0 '.1xture units Continuouc Demand: 25.0 gpm Fixture Demand: 17.0 gpm
(} Total Demand: 42.0 gpm 20
(^)h
\m . APPENDIX A LEU THERMAL CONDUCTIVITY CALCULATION renrity of U3112 The densities of the d' persants are taken from reference (1) with the volume fraction related to the uranium density, Pu, in the fuel by:
Pu = 1.28Vf where Vf is the volume fraction of the dispersant for the purposes fuel loading of 12.5 g/cr (22 plate element, 275 g U-235) plate the U density is 3.4682 per reference (2) the volume fraction of U3Si2 in fuel meat is U3Si2 Vf = 3.4692 = .3068 or 30.68%
() 11.28 From reference (3), page 11 l
vp = .072 Vf - .275 Vf2 + 1.32 Vf3 tnerefore Vp - .072(.3068) - ,275 ( .3063) 2 + 1.32 ( .3068) 3 = .0343 I where Vp and Vf are volume fractions or porosity and fuel in the meat, respectively.
Therral Conductivity of U 3 Sig Volume fraction of fuel plus voids = .3068 + .0343 = .3411 L the thermal conductivity is obtained from Figure 6, page 16 of
! reference (3)
I-l K = 68 W/m.k l-($) '
21 l
yy--yr -
, y% y
I REFERENCES (1) R.F. Domagala, T.D. Wiencek, and H.R. Tresh, "Some Properties of U-Si Alloys in the Composition Range U3Si to U3Si2," CONF-8410173, ANL, RERTR/TM-6, 47, July 1985.
(2) Memo from W. Woodruff (B17681 at ANLOS) to Eugene Spring (RMA101 at URI MUS), Sept 5, 1989.
(3) J.L. Snelgrove, R.F. Domagala, G.L. Hofman, T.C., Wiencek, G.L. Copeland, R.W. Hobbs, and R.L. Senn, "The Use of U3Si2 Dispersed in Aluminum in Plate-type Fuel Elements for Research and Test Reactors," Argonne National Laboratory (ANL/RERTR/TM-11), October 1987 O
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- 9 22 l
. . - . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - . _ _ . ~ _ . . _ _ _ _ . _ . . _ . _ _ _ _ _ _ _ _ _ . _ _ - - . -
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4-1 --
I- -
APPENDIX B f
}
f CRITICAL VELOCITY FOR FUEL PLATE DEFORMATION k
It has been shown that a critical flow velocity exists j for a given plate assembly, m At this critical velocity, the j plate becomes unstable and large deflections of the plate can l jj - occur. These plate deflections can cause local overheating of the fuel plates and possibly a complete blockage of the i coolant flow.
4 l Miller (2) derived a formula for the critical velocity based on the interaction between the changes in channel i cross-sectional areas, coolant velocities, and pressures in two adjacent channels. For design purposes, reference (3) l and (4) recommends that the coolant velocity be limited to :
1 2/3 of the critical velocity fiven by Miller, (2) therefore for ,
a flat plate; I l- -V Critical - 2/3 ( 15 x 101 E (t al- ta l) tw ] 1/2
! p . W.4 (1 -y 2 )
l where E = Young's Modulus of elasticity, bar : 10.4 x 106 psi /14.5 (A1) tp = Fuel plate thickness, cm : .127 tm = Fuel meat thickness, cm : .0508 3
p = Density of water, kg/m : 106 .0 tw = Water channel thic). ness, cm : .381 1
w = Fuel plate width, em : f.096 y =' Poisson's rati'. dimensionless m : ,3 (A1) t t Crinical = 16.6 m/sec The average core velocity of the 14 element LEU core calculated for j_ the normal primary pump flow of 1730 gal / min ( .109 m 3 /sec) is about i 1.6 m/sec. For a projected 5 MW core, the velocity would be i
- increased to about 4 m/ .3 ec . This is well below the limiting value i of 16.6 and therefore is not a problem in the proposed RINSC core.
10 l-23
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-l REFERENCES (1) International Istomic Energy Agency, Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use l of Low Enriched Uranium Fuels, Guidebook IAEA-TECDOC-233 (1980)
(2) Miller, D.R., " Critical Velocities for Collapse of Reactor Paralie) Plate Fuel Assemblies", Trans ASME, J. Eng.fot Pcwer, 82,83 (1960)
(3) Mishima, K and Shibata, T. "Thermail-Hydraulic Calculaticns for KUHFR with Peduced Enrichment Uranium Fuel, "KURRI TR-223 (1982)
(4) S. McLain ana J. H. Martens, Reactor Handbook, Vol. IV, Interscience Publishers (1964)
O (5) Standard Handbook for Mechanical Engineers, Baumeister and Marks, 7th Edit! an, 1967 pg. 5-6 9 ,
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4
- .. . .. . . . = .- . . - ~ . . -.-. .
t .
APPENDIX C LOSS OF COOLANT _. _
.e--.- - - - - - - _ , - __ cn,.. l'a . n I f.ti. kr. t'o . o n o ,,,
L w_ _
_ -. n.us,t) c os e _ tt. IN 1%
' ~ ----
Q- e. t a q . ,, . i -
m.m q O,, M\ Yo me.
- st o < a u . , .
SURFACE AREAS (FREE FLOW AREA)
Area of entire pool surface : 150 ft2 Area of core box : 5.06 ft2 l
Area of core (loaded) : .917 ft2 l
I Area of 1/2 dia' meter hole in core bo:t : .00126 ft2 Area of 8" pipe : .349 ft2 The data elevation of 114.13 is used due to the assumption that water will not drain below thin elevation in the event.of shear
- of the 8" beam tube.
l The amount of water remaining in the core box after draining =
114.13 - 113.213 = .917' ASSUw m :
'iravity draining of pool from pool surfaces to the top of the core box out the 3" beam tube which has ne plug in place and the shutter in the up position. There is no cover tlange.
1 .
25
CO!*PnTATIC!!AL ETHCD :
Discharge under falling head il) t =
2A ((H)/21
-(H 2);/2 )
Ca (2g) 2 Datum i,s el.114.13 (invert of bottcm of 8" beam tube)
H1 = 139.417 = 25.287 H2 = 115.713 - 114.13 = 1.593 C= .6 A= 150 ft2 a = .349 ft2 t= 2x150 [ (2 5. 2871/2- (1. 5 8 3 ) 1/2 ] = 67 3.15 sec 6x.349 (2x32.2.2);2 =11.219 min 2 rain Time of Core Bas (loaded with components) t2 = 2x 917 [ (H 1) 1/2- (H2 ) l'2 Hi= 115.713 - 114.130 = 1.583 llh l H2= 114.130 - 114.130 = 0 t2 = 335.6 sec/60 = 5.59 min TOTAL TIME = 16.8 min Minimum required water flow rate to keep the core box full From (2) a F= . 61 A (29H) 1/2 where H = 115.713-114.13 = 1.583 F= .61 x .00136 (64.4 x 1. 583) :/2 F= .61 x .00136 x 10.097 = .008376 cfs F= .000376 113 x 7.48 cal x 60 sec = 3.76 gpmt sec ft3 (this assumes core is full) l 26
, . -..~ - .- .-.- - . . . _ - - - - . .- --- - - . ~ - ..
. .. 3 Mcm r k,n
= Disebarge under Falling Heat Figure 44 shou a s mel. =. ..,, y.r.Qd9df_O. Tr==s,b g fille,i with wher to a depth h., The time required to ioner the : r io 935 @ m Hed32Flya
~
ter surfare to a depth A,is required A..
is the ares Of Orihee. gj f.05/AuM6/AS' t 9
- h.MAu'03.'AII5 d .1 is the area of water surf ace fot'a depth y. C s the A
.a giu ,a
. etheient Of discharge The increment of time dt required to j1: pj#* I- M *,w ed lower .he water the . inEnitesimal g,g,, g.y .' iv>
4 - f rassect.on g H distance dy is f} bnA meswed -
(a) - , - . b ;4 ro y v > 0 6.,v,,s,9A y 2- = .r.e- A dy U) Fah a dl = a 0 15) :06tAyQ co wn e Laa From s'4-151 if A esa be espressed
=g - . . . P en!4 (vW g ;,f 9,,gy yy',,
- :1; v . 0 951/f 4 p .g y,ggg.jp l iT' N
l r
in terms el y, by integrating be-tween limits k and A., the time .
3 r,gy.oggAyp r- ; ; ,,A,.,,asured .
3g
-- e 4 g g Q g
' l needed to loser the unter surface KG,non.7echon 3 D __y_,fjysano,,ng.y I I t he distance hi - he eso be gotten. - N W t%W 2 '
Placing hi = 0 gives the time of '
O
. r --- empt3i ng the veuet. Equation g) B (i)
N (415) apphes to horizontal or in- .: p, ,, m-.. _m #u, .
n~~7'22=__ g; ges,g g]
Fia. 44 Dinharge under chned on6ces provided the water , _ , q y.orfgfa surf ace does not (60 below the top _ __,".p.,A,,ayj g falhos head. - - , - - - - - jf 9I of the ntifice,- For a cylimler or pri m with vertieni nsis .t is N d8 A d83u"d 5'
- 0'0
,M conat....t. and 1:4. ( l 15), af te r mtegratmn, hernmes LI ,_ m_r.
=
u kk 37----- W . Q. .
.-.d. w '."#"^
& ~...,l [. , D
4 f = , "g (vi - v'E) i ( t. I'y
"((....
La \g4 .s 9 he,s A reasured (c) (g) enfersv3 S Orit!ce Coemcients. One of s he iarhrst experimenters on sharpelgral orificn m linnnlinn 8,nith, .f r> lib. velin,. of
__ w -- - v 1,gr a ,5 n o.E # # M#M - - - . . - -
Cr yst. .
kisP CUCllil'IUfat Of tlivl arge fair rntigt,l jing l MjtlNrP t or k ftret nry **7*****'.'*.a p ]#
d? -
f *[S'k.4 ka given jn Table 4 3. '- + -
- * = " z/0 h dres A mrowed
, 4 ocriensS 3 1p'" f,17:09]Al2 3;9 h k NI5.6 SQq ", -3. ~~p..
fd U p si n la
.t.. -a ArnfA mekseed B ga} o<Juron3 D Table 4-3. Smith's Coefficienta of Discharge for Circular and Square Oritices with Full Contraction Diameter of eiecular erinces. tut Side of square erinem. feet Ile.J.
(vu t 0.02 0.04 0 07 01 0.2' O.6 1.0 0.tM 0 64 0 of 0.1 02 06 1.0 0 637 1 6.14 0 618 04 .. . 0.643 0.628 0 621 0 635 0 630 0 615 0 GI3 0 Got 0.503 -0.6 0 660 0 636 0 623 0,617 0 603 - 0 S98 0 648 0 626 0.615 0 610 0 601 0 594 0 $00 0.6 0 632 0 631 0 620 0.615 0 603 0 600 0 S97 0.644 0 623 0 012 0 t,08 0.600 0 SDS 0 591 1 0.64 8 0 628 0.618 0 613 0 603 0 601 0 599 06P 0.618 0.608 0.604 0.600 0 S96 0 $93 1.3 0 641 0.622 0,614 0.610 0.603 0.602 0 601 0.632 0 644 0.606 0 6M 0.599 0 $97 0.30$ 2 0.637 0 619 0.612 0 608 0 603 0 6M 0 602 0.629 0 812 0 605 0.60'l 0.599 0.398 0 Soo 2.5 0.634 0.617 0.610 0 607 0 405 0.604 0 602 0 627 0.611 0 r.M 0 003 0 509 0 508 0 $07 3 0.632 0 886 0.600 0. 60r 0 603 0 6M 0.603 0.623 0 600 0_003 0.002 0.500 0 507 0.506 4 0 628 0.614 0.ros 0 606 0 603 0 603 0 602 j 0 018 0.e07 0.602 0.600 0.308 0 597 0.S96 6 0.623 0.612 0 607 0 603 0 6M 0 603 0 602 l
O.614 0 6AS 0.601 0.800 0 SO4 0,506 0 506 8 0.619 0.6l0 0.606 0.606 0 604 0 603 0.602 l 0 oli 0 fim o r.no - 50s 0.507 0 390 0 SUS to o eso 0 Me 0 603 0 604 0 603 0.602 0.601 0.604 0 500 0 Sui - Jo u 500 0.500 0 Sui 20 0 006 0 604 0.60'2 0 602 0,602 0 604 0.600 0 596 0 SOS 0 504 .404 0 504 0 $94 0 503 SO O noa 0 604 0 60t 0.600 0.600 0.509 0 S99 0 Sc3 0.502 0.502 0.592 0 SW2 0 SU2 0 392 800 0 599 0 Ses 0.508 0 394 0.508 0.S94 0 506 r
l 27 I-t
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4 APPENDIX D i 6O DECAY HEAT - CALCULATIONS ,
(1) ASSUMPTIONS J
A. The reactor has been operating for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> continuously.
- B. The reactor scrams when the loss of coolant sequence l begins at time zero.
C. Reactor pool water level reaches its lowest level as shown in Appendix C.
D .. Following the LOCA, the decay heat is conducted away to the remaining water in the core box.
. E. The conduction loss will occur with the plate i
j temperature reaching the melting point of aluminum.
(2) DECAY HEAT GENERATION H( ) Q 2 MW = 6.a6x 106 Bru/hr x H12/3600 sec = 6.87 Btu /sec 14 e1 x 22 plates / element From Table 5.1 at Time = 16.8 min (1008.75 sec) ~ 103 sec E_ = .0185 Po P= .0185 x 6.187 = .114 Btu /sec (3) HEAT CONDUCTION DOWN THE PLATE (FUEL SECTION)
It is assumed that the heat generation sine distribution The volumetric heat rate Q111 is defined a Q111 = Cmu =
Qn Btu /ft3 (1) volume 1 w.t where Omax = Btu /sec and 1 = fuel plate length w = fuel plate width t- fuel plate thickness O
28
e l
Ecr an average sine Cmax = Cave x fl/2 (2)
Substituting equation (2) into equation (1) we obtain:
Q 111 =
Qave R/2 (3)
For conduction downwards (-x direction) and using the heat conduction equation from reference (1) 2 d" = -n111 (4) dx2 kf kf = fuel plate thermal conductivity for a sinusoidal heat generation 0111 =
Cmax . sin Rx/l (5) substituting equation (3) into equation (5) we obtain n111 . n ,..a . R/2 . sin Rx/l (6) 1.w.t substituting (6) into (4) d2 t/dx2 = -Qave/l.w.t . R/2 . 1/kf sin Rx/2 (7) integrating (7) dt /dx = -M/2 . Qave/l.w.t.kf (-cos Rx/l) .
1/R 4C1 eviluating C1 dt/dx = 0 AT x = l (l = top of fuel)
C1 =
Qave/2 w.t.kf then dt/dx = Qave/w.t.kf . cos Rx/l - Qave/2 wt .kf then integrating with the limits T= 1200oF at top of fuel plate x= 2' T= 212cF at surface of water in core box x= 7' l
l then Qave = .013 Btu /sec 1
l l
l 29 l
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4(f%,t M N' g O IMAGE EVALUATION [jj/g// ,9 4,
\///7
[' +4ff TEST TARGET (MT-3) [f&'f 4pg j #
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%\+sp r, l
1.0 lf r 2 m I na i % p" "2.2 h tLt \
l,l f ,IA bN ls-1.8 l
1.25 1.4
_1.6 4 150mm . >
4 6" >
t 4W[% / !b
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g IMAGE EVALUATION j/// (/ [^'0?
,ir, TEST TARGET (MT-3)
\gV/ %.
/// !?<{N!
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% 7////
+ 'k 4
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',s m-m Q. 2 2 l;~'o
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l t (A %c -lMAGE EVALUATION ///gs &o g+%4,
\//o/7%[- $$/4* - TEST TARGET (MT-3) / f $ ' 4g'
///g a,jp, Yo4*p 4 (4'k I.0 E2 M n e a p<=un:
.g l
l' tu l_2.0 1.1 L 1 I
l 1.8 l
1.25 1.4 1.6 4 - 150mm >
- 6" >
s% ++bsA _
e>A>y,,,,Szz,,;g ,
c,--
y$ /f/4 4e +. ,
r ,
'N L - ,sihi ^
l- :. . ]
C 4
Since this value is less than the decay heat generation .
of .114 Btu /secz it is assumed that melting will occur.
(5) DECAY TIME TO HAVE GENERATION EQR.L TO REMOVAL The length of time that core cooling would be needed to have the decay heat to reduce to .049 Btu /sec can be calculated using Tab]e 5.1 the Power Ratio = .049/6.187 = .00792 ',
and time = 2x104 sec = 5.56 hrs therefore emergency core cooling is r-quired for dt 2 east 5.56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br />. Water supply can be supplied for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(E) TOTAL CORE COOLANT LOSS UNDER LOCA CONDITIONS Total loss would include evaporation of the water from the Core box (assumed to be the rate at the maximum value associated with the initial LOC at time equal to the drain O time, or 16.8 min),
from Table 5.1 - P3 /Po =
.0185 0 T = 1008 see The heat generation then = .0185 x 6.87 Btu /sec
= .1144 Btu /sec The maximun evaporation rate for water at atmospheric -
saturation (1000C) = .1144 Btu /sec x 1/970 Btu /lb Then: .00011794 lb/sec x 60 sec/ min x 1 ft3/59.8 lb/ft3 x 7.48 gal /ft'3 = .000851277 gal / min (liquid)
This is added to the drainage loss and the total loss is still about 3.76 i gpm.
O 31 l
l Mws_ . .
(4) HEAT CCNDUCTIC1N TO THE WATER !N CORE BOX FROM THE NCN FUEL ALUMINU'; IN THE ELEMENT Calculatiot. Basis - Per Plate A. Non Fuel Plate Cross Section P3ste Cross Section = .05" x 2.79" = 1395"2 Pax Fuel Cross Section = .02" x 2.47" - .0494"2 Non F4el Plate Cross Section = .1395 - .0494 = .0901"2 22 p) iter x .0901 = 1.9822"2 B. Side r tes of the Element Average width = .187" 22 grooves (.187 .088) . .058" Cross Section 2 Side Plates x tr ( .18 7 " x 3. 0 4 5") -2 2 x (.87 .088) x .058)
Area = .8862"2 C. Total Area for the Element Area = 1.9822 + .8862 = 2.86S4 Per Plate Basis A = 2.8684/22 = .13038"2/144 e .0009054'2 Heat Conducted from the Aluminum to the Water Q = kae A dt/dx Q = ka2A (Tmax-Tsat/l)
Q = !31x0009x(1200-212) = 89.604 Stu/hr x 1/3600 1= (2 .7) = 1.3' O = .02489 Btu /sec total heat conducted = fuel + aluminum
= .013 + .02489 = .03799 Btu /sec From the original SAR it was assumed that about 30% of the heat was used in steam formation therefore .3 x .03789 = .011367 Stu/sec and the total heat removal = .037S9 + .011367 llh
= .049 Btu /sec 30
_ . - . _ = . . - - . .
Tacle 5.1 g The Ratio, P(ts) / Pn, of the Fission Product Decay
-W Power to Reactor Operating Power as a Function of Rme , ts, After Shutdown ( /W S , 1968)__
Time After Time After Power Racio Shutdown, t s P wer Ratio Shutdown,.t (seconds) 3 F(te) / Pn (seconds) P(tn) / Pn 1 X 10~1 0.0075 6X 104 0.00566 0.0E25 0.00505 1 X 10 0 8 2 0.0590 1X 10 5 0.00475 0.0552 2 0.00400 4
0.0533 4 0.00339 6
0.0512 6 0.00310 8
0.0500 8 0.30282 1 X 101 0.0450 1 X 106 0.00267 2
0.0396 2 0.00215 4
0.0365 4 0.00266 6
0.0346 6 0.00143 8
0.0331 8 0.00130 1 X 10 2 2 0.0275 1 X 10 7 0.00117 0.0235 2 0.00089 4
0.0211 4 0.00068 6
0.0196 6 0.00062 8
3 0.0185 8 0.00057 1 X 10 0.0157' 1 X 10 8 0.000550 2
0.0123 2 0.000485 4
0.0112 4 0.000415 6
0.0105 6 0.000360 8
l.
0.000303 1 X 10 4 0.00965 8 0.00795 1 X 10 9 0.000267 2
4 0.000625 6
32
_ ._ . - . _ - ~ - .
- - ~
APPENDIX E MAXIMUM HEAT FLUX The Rhode Island Nuclear Science Center Technical Specifications section K. 3, e, (2 ) specifles the maximum heat flux.
Since it is not specifAc in regard to how this was originally calculated using an overall hot spct factor of 2.3, the 1.E U het channel analysis for 2 MW calculates two conditions resulting in ,
slightly different values, r.ea 1 This calculation determines the maximum heat flux of .365 MW/M2 when using an axial peaking factor of 1.32. This is the case when the blades are out of the core. The hot spot factors-cited in Section IV are used, i
O c"' 2 This calculation determines the maximum heat flux of .424 MW/M2 when using an axial peaking factor of 1.536. This is the case when the blades are 50%
inserted in the core. Again the standard hot spot factors were included.
t l
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t
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l t
APPENDIX F l MAXIMUM CORE-SPECIFIC POWER l The Rhode Island Nuclear Science Center . Technical l Specifications Section K,3,e(2) s pe c i f .t e s the maximum specific i- powe*r. For a 14 element LEU core we are si.Tply calculating p maximum core specific power as 2 MW divided by the number of fuel elements having a maximum loading 275 g U235, n- ,
a i I
- Therefore, the specific pcwer is 2x19t6 watts = 519.48 IL, 6 14x275 oU235 i .
I Since burnup increases this value, this is the limiting value r
which cannot be reduced, i i Lo 1
1 k
a 0
P i l I
i i
4 e i 34
.. . - . . , - . - _ . . . . .- - - _-. - - _=. - . _ _ - - . . . . .
i 4
SAFETY ANALYSIS REPORT PART C TECHNICAL SPECIFICATION REVIEW AND MODIFICATION I Introduction i
II Appendix A/Rhode Island Nuclear Science Center Peactor Technical Specifications Appendix A to Facility License R-95 Dated July 21, 1964 Revised Through Amendment #16 III Appendix B/ Proposed Rhode Island Nuclear Science Center Reactor Technical Specifications l
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, Part C Technical Specifications Review and Modification INTRODU,CJION There are numerous Technical Specification changes required as a result of the use of the LEU fuel in the Rhode Island Nuclear S-ience Center reactor.
Parts A and 1 of the S r. f e t y Analysis Beport touch on many of them. As a result of the Rhode Island Nuclear Science Center review process, additional changes which reflect current conditions or clarifications of some Technical Specification sections are also included in the final Technica' Acecification versien. Isppe n d i:s A is a copy of the Rhode Island Nucler.t Science Center current Technical Spec;fications. Appendix B is a copy of Technical Specifications with the changes included as a .esult of the SAR and review process. The double vertical lines adjacent to a section designates the section which ha r, the proposed changes.
!=plementation of the final approved Safety Analysis Report will be a difficult task for the Rhode Island Nuclear Science Center. Conditicns our. side the control of the l i c e r. s e e , such as key staff r2tirements, budget cutc, small
! operating staff etc., incrence the difficulty and will curtail the operation of the facility during the conversion orocess. The Rhode Island Nuclear Science Center acknowledges the assistance of Argonne Nat! nal 1.aboratory in the preparation of *he Safety Analysis Report.
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O APPENDIX A RHODE ISLAND NUCLEAR SCIENCE CENTER REACTOR TECHNICAL SPECIFICATIONS APPENDIX A TO FACILITY LICENSE R-95 l
DATED JULY 21 1964 REVISED THROUGH AMENDMENT #16 i
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- TABLE OF CCNTENTS PAGE
, LA. SITE- 1 1
Location' l.
'2. Exclusion Area 1
- 3. Restricted Area 1 .)
- 4. Principal Activities 1 Figure A,1 2,2a B. CCNTAINMENT 3
- 1. Reactor Building -3 C. REACTOR POOL AND PRIMARY COOLANT SYSTEM 4
- 1. -General 4
- 2. ' Reactor Pool 4
- 3. Shielding 4
- 4. Primary Coolant System 4
- a. Heat Exchanger 4
- b. Primary Pump. 4
- c. Delay Tank 5
- d. Primary Recirculation Piping 5
- e. -Make-up System 5 fe . Clean-up System for Primary Coolant System 5
'D, SECONDARY COOLANT SYSTEM 6 I
E. REACTOR CORE AND' CONTROL ELEMENTS 7 i
- 1. Principal Core Materials 7 l
- 2. Fuel Elements 7; l
- 3. Reflector Elements 8 .,
'4. Control Elements 8
- 5. Servo Regulating Element 8 l
- 6. Control Element Drive 8 1
- 7. Servo Regulating Element Drive 9 ,
- 8. Neutron Sources' 9 l F. _ REACTOR SAFETY SYSTEMS 9 l
- 1. Modes of Power Operation 9 !
a, Power Operation - Natural Circulation (NC) 9
- t. Power Operation - Forced Circulation (FC) 9
- 2. -Design-Features. 10
- a. 4'he Reactor Control System 10
- b. Process-Instrumentation 10
- c. Master Switch 10
- d. ' Power Level Selector Switch 11
- e. Control Element-Withdrawal Interlocks 11
- f. Servo System Control Interlock- 11 Table F.1 Reactor Safety System 12 Table F,2 Reactor Nuclear Instrumentation 13 G. WASTE DISPOSAL AND FACILITY MONITORING SYSTEMS 14
- 1. Waste Disposal Systems Design Features 14
- a. Liquid Radioactive Waste Disposal'. System 14 i b. Gaseous Radioactive waste Disposal System 14 l c. Solid Radioactive Waste Storage 14
- 2. Area and Exhaust-Gas Monitor Design Features 14
- 3. Other Radiation Monitoring Equipment 15-
~ 4. -High Radiation Area 16 1
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- - TABLE CF CONTENTS,-(CCNTINUED) -
i
,. i H. FUEL STORAGE 17
' - : 1. New Fuel Storage 17
'2. Irradiated Fuel Storage 17-I. EXPERIMENTAL FACILITIES 17 J. AOMINISTRATIVE AND PROCEDURAL SAFEGUARDS 18
- 1. Organization 18
- 2. Qualifications of resonnel 19
- 3. Responsibilities of Personnel 19 l
- a. Director 19 9 b. Senior Reactor Operators 20
- c. -Reactor Operators 20 l
- d. Health Physicist 21
- 4. Writt'en Instructions and Procedures 21 .
- 5. Site Emergency Plans 21 ;
K. OPERATING LIMITATIONS 22
- l. General 22 ,
- 2. Experiments 23
- 3. Operations 24
- a. Site 24
- b. Containment 24
.c. Primary Coolant System 25
- d. Secondary Cooling System 26
- e. Reactor Core and Control Elements 26 i
- f. Reactor Safety Systems 29
- g. Waste Disposal-and Reactor Monitoring Systems 30
- h. Fuel Storage 31 9 4 Maintent ce 31 ,
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- 1. Locatien The reactor shall . be located at the Rhode Island Nuclear Science Center on three acres of a 27-acre former military reservation, originally called Fort Kearney and now called'the Narragansett Bay Campus of the University of Rhode Island. The University of Rhode Island is a state' agency. The 27-acre reservation is controlled by the State of Rhode Island through the Unf*?stsity of Rhode Island. The reservation is in the Town of Narragansett, Rhode Island on the west shore cf Narragansett Bay approximately 22 miles south of Providence, Rhode Island 4
and approximately six miles north of the entrance of the Bay frem the Atlantic Ocean. The Rhode Island Nuclear Science Center and various buildings used for research, educaticn and training purposes are located on this 27-acre campus.
- 2. Exclusien Area Figure A.1 is a drawing of the Narragansett Bay Campus showing.
the three acre Nuclear Science Center site. The boundary of this area shall be posted with conspicuous signs to delineate the area. This three acre area shall be the exclusion area as '
defined in 10 CTR 100.
3, Rest ricted Area f'
> Figure A.1 also shows the locetion of the reactor building on the three acre area. The reactor building and attached office laboratory wing shall be considered the restricted area .as defined in 10 CFR 20.
- 4. Princinni A-t ivit ie s The principal activities carried on within the restricted and exclusion area shall be those associated with operation and utilize. tion of the reactor. It shall be permissible to locate additional Nuclear Science Center or University of Rhode Island buildings within the exclusion area provided that -these additional buildings are capable of timely evacuation and do-not interfere with the operation of the reactor.
Amendment 15
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B. CcN?A' M MT 1' . Pescte - euiMim The reactor shall be housed in a building capable of meeting the following functional requirements:
In the event of an accident which could involve the release of radioactive material, the confinement building air shall be exhausted through a clean-up system and stack cresting a flow of air into the building with a negative differential pressure between the building and the outside atmotphere. The building shall be gas tight in the sense that a negative I differential pressure can be maintained dynamically I with all gas leaks occurring inward. The confinement I and clean-up systems shall become operative when a i building evacuation butten is pressed. This action ]
shall: (1) turn off all ventilation fans and the air conditioner system and (2) close the dampers on the ventilation and air conditioning system intakes and exhaust, other than those which are a part of the clean-up system. No further action shall be required to establish confinement and place the clean-up system in operation. An auxiliary electrial power system shall be provided at tne site to insure the availability of power to operate the clean-up system.
The reactor building exhaust blower, which is designed t o q exhaust at least 4000 cfe, operates in conjunction- with Q additional exhaust blower (s) which provide an additional exhaust of at least 10000 cfm from non-reactor building sources and in conjunction with the air handling unit which takes air into the reactor building at less than 4000 cfm. The total exhaust rate through the stack is at least 14000 cfm. During normal operation, the building is at a pressure somewhat belong atmospheric. The control room air conditioner shall be a self-contained unit, thermostatically controlled, providing constant air temperature for the control room. If it is installed with a penetration through the wall of the reactor building, it shall have a damper at this penetration which closes when an evacuation button is pressed.
Upon activation, the clean-up system shall exhaust air from the reactor building through a filter-and a 115 foot high-stack, creating a pressure less than atmospheric pressure. The clean-up filter - shall contain a roughing filter, an absolute particulate filtet, a charcoal filter fre removing radioiodine, and an absolute filter for removing chnacoal dust which may be contaminated with radioiodine. Each absolute filter cartridge shall be individually tested and certified by the manufacturer to have an efficiency of not less than 99.97% when tested with 0.3 micron diameter dioctylphthalate smoke. The minimum removal efficiency of the charcoal filters shall be 99%, based on ORNL data and measurements performed locally.
Gases from the beam ports, thermal column, pneumatic system, and all other radioactive gas exhaust points shall be exhausted g to the stack through 6 roughing and absolute filter system.
- Change 4 Amendment 16
C. PE ACCOR PCCt MD PR A RY C 0 0LA'!* EYETEM
- 1. General The primary coolant system shall consist of the reactor pool, delay tank, heat exchanger, coolant pump, and the associated valves, piping, flow channels a r.d sensors. During forced convection cooling, coolant water shall be supplied to the core by an aluminum line connected to the inlet flow channel which is on one side of the suspension frame. The coolant water shall flow from the inlet flow channel downward through the l core to a plenum below the grid box. The coolant water shall i then flow into the ou*. l e t flow channel on the opposits side of I the suspension frame and then throu'h a discharge line to the l delay tank, coolant pump, heat exchangor and then return to the ;
coolant inlet line. l
- 2. Rearter Pec1 The reactor pool shall be constructed of ordinary cencrete with 1/4" thick 6061-T6 aluminum liner and shall have a volume of approximately 36,300 gal.
- 3. shielding The reactor pool and primary system shielding shall be adequate to meet the applicable personnel radiation protection requirements of 10 CFR 20.
- 4. Prim ry ceslant ..:; 1 tam The primary coolant system shall conform to the following:
- a. lient Exchancer The heat exchanger shall be designed to remove heat at _ the rate generated by the reactor at maximum licensed steady state power from the primary water ar.d shall be designed to perform unoer the maximum primary system cperating temperature and pressure.
Replacement heat exchanger shell and tube bundles shall be constructed from. stainless steel according to the requirements of Section III, Class C of the ASME Boiler and Pressure Vessel Code,
- b. Pr4-*ry P"?
Number of pumps 1 Type Horizontal mounted, Centrifugal, Single Suction O
Change 3,4
Materials of construction Worthite Rating 1500 gpm Head 59 fett Design Pressure 75 psig minimum Design Temperature 150CF minimum Motor Type Drip proof, induction, 440 v, 3-phase, 60 rycle
- c. ;rin Tank Number of tanks 1 Material cf construction Aluminum As sociat ion Alloy 5083 and 5096 Material Thickness Walls 0.25 inch Dished Heads 0.375 inch Capacity 3000 gal., minimum l
- d. Prf-nrv Perirculatirn Pirina Material and thickness Sch. 40 A1. type 3003 aluminum N;
Site 8 and 10 inch Design temperature 1500F, minimum Design pressure 100 psig, minimum O e. Mnke-ur Eystem A check valve shall be installed in the line between the potable water supply and the make-up and cleanup demineralizer to prevent entry of potentially contaminated water into the potable water supply.
Water source Potable water from city main Make-up demineralizer type Mixed-bed single shell, regenerative Make-up demineralizer capacity Normal 25 gpm Emergency 50 gpm Water softener capacity Normal 50 gpm f, ciennun system for pr4-nry cec 13nt water Cleanup pump Capacity 40 gpm Head 100 ft Cleanup demineralizer Type Mixed-bed, single shell, regenerative Cleanup demineralizer capacity Normal 40 gpm Emergency 50 gpm Change 1
6- .
- ) . secrNrARY COOLANT SYSTEM The seccndary coolant system shall carry the heat rejected from the primary coolant at the heat exchanger to the atmc here at cooling towers. It shall be compcsed of the heat exchanger,. cooling towers, pu.mpt and associated valves, piping and sensors. In this system, ^
water flows from the heat exchanger through a control valve to the s cooling towers. From the cooling tcwer basins, the water is then purped back to the heat exchanger.
Change 4 4
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9 E, prAc?cP ccPr Mm ccNTPet ELEMENTS s
The reactor core and control elements shall have the following characteristics and nominal dimensions:
- 1. Princiral core Materials Fuel matrix Alloy, UA1x,U300 U-235 enrichment Approxicately 93t Fuel clad 1100 and/or 6061 aluminur.
Fuel element side plates 6061 aluminum End fittings 356-T6 or 6062 aluminum Moderator Water Reflector AGOT gride (or equivalent) graphite and/or water Control elements Mixture of B 4C and aluminum, clad with aluminum Servo Element Mixture of B 4C and aluminum, clad with aluminum.
- 2. Fuel Ele-ants Plate width overall 2.8 inches Active plate width 2.2 inches Plate length overall 25 inches l
Active plate length 24 inches I
Plate thickness 0.06 inch Clad thickness 0.024 inch Fuel matrix thickness 0.012 inch l
l Water gap between plates 0.1 inch l
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Amendment 8,11 i
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.g Nu:-ter of plates per f uel element 18 U-235 per fuel element 124 grams, ncminal overall fuel element dimensions 3 in x 3 in. x 40 in.
- 3. F.eflerter EIa::em Cverall reClector element 3 in x 3 in. x 40 in, dimensiens, neminal Nominal clad thickness .1 in.
Nominal graphite dimensicns 2.8 in. x 2.9 ir x 29.7 .
- 4. crntrol E i n-a nn Width 10.6 in.
Thickness 0.39 in, b Overall length 54.1 in.
Active length 52.1 in.
- 5. Serve Regulatine Ela-ant Shape Square boral tube 3 Width 2.1 in.
Overall length 28.8 in.
Active 24.9 in.
- 6. Centrol Ele-ant Drive Type Electromechanical serev Drive to safety element Electromagnet connection Stroke 32 in. maximum t
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Amendment 11 l
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- 7. serve Reguistine rierent Drive Type Electromechanical screw Drive to element connection Lock screw (no scram)
Stroke 26 in maximum Position indication accuracy + 0.02 in.
- 8. !:eut ren seurces Start-up Source Number 2 Type Plutonium-beryllium Unit Source Strength 1 x 106 neutrons /sec. minimum Maximum Power Level with Plutonium-beryllium sources installed 10 Kw operational Source Number 1 Type Antimony-beryllium Source Strength 2 x 106 neutrons /sec. minimum l
- 1. Medes of Power Orerntlen There shall be two modes of power operation:
- a. Power Oneratien - Natural circulation (NC)
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( Power operation -
NC shall be any reactor operation performed with the reactor cooling provided by natural circulation. The reactor power shall not exceed 0.1 MW during NC operation,
- b. Power creration - rerend circularien frci Power operation - FC shall by any reactor operation i performed with reactor cooling provided by forced I
circulation. The reactor power shall not exceed 2 MW during FC operation, l'
Change 4
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a, The F.eact o r centrol Symm The reactor safety system shall consist of sensing devices and associated circuits which automatically sound an alarm
- and/or produce a reactor scram. The systems shall be
> designed on the fail-safe principle (de-energizing shall cause a scram). Table F.1 and F.2 describe the arrangement and requirements of the safety system,
- b. Pro e3s instrumentatien Process inst rumentation with readout in the control roem
< shall be provided to permit measurement of the flow rate, temperature, and conductivity of the primary coolant and
-the flow rate-of the secondary coolant. .In addition, a second primary flow indicating device with readout in the control room shall be located between the reactor outlet plenum and the reactor outlet header.
After normal working hours, an independent protection system, separate from the system described in Section K.3 a, shall be used to monitor certain items in the reactor building and alarm in the event of an abnormal condition, The alarm channels provided are:
(1) A fire in the reactor rocm, 5
(2) A fire in a location other than the reactor room, (3) A decrease of 2 inches in reactor pool water level, (4) A power failure in the reactor building, (5) An alarm condition from the radiation monitors reading out in the control room, ,
- l. (6) An alarm condition from any other selected feature.
C, Master Switch i
l A key lock master switch shall be provided with three positions; "off", " test", and "on". These positions shall have the following functions:
(1) The "off" position shall de-energize the reactor control circuit.
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(2) The " test" position shall energize the reactor control circuit exclusive of the control blade magnets.
(3) The "on" position shall energize the reactor l cont rol circuit including the control blade l magnets.
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- d. Emr tevel E & m I._iwit;L A power level selector switch shall be provided with four l positicns; "C.1 MW", "1 MW", "2 MW", and "5 MW" These positens shall have the following functions:
1 l (1) The "0.1 MW" position shall activate all safety j system sensors except those indicated in Table F.1.
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(2) The "1 MW" and "2 MW positions shall activate all safety system sensors.
l (3) The "5 MW" position shall scram the reactor.
- e. Orntrol Zla-ant Withdrawal Interlo;ks i
l l Interlocks shall prevent control rod withdrawal unless all l of the following conditons exist:
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(1) The master switch is in the "on" position, (2) The safety ' tem has been reset, l (3) The Log N amplifier switch is in the " operate" position, (4) The startup channel neutron count rate is three counts per second or greater, and (5) The start-up counter is not being withdrawn.
It shall not be possiDie to withdraw more than one control element at a time.
- f. In xa Evata- contrn! nterleck t
l Interlocks shall prevent switching to servo control unless the period as indicated by the Log N channel is thirty seconds or greater. The Servo control system shall be designed so that immediately f ollowing a scram the Servo control shall automatically return to the manual mode of
( operation, i
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Change 4 l
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-. ._.-._.~._-._.._..-._____-_.-_-_.-__---m . _ . _ . . . . _ _.
l taste E.1 l PEACTOP. SAFE?Y SYETEM l
.k Sensor or Trip Device No. of Switches Trip Set Alarm Set or Sensors Point Point Short Period i 1 3 sec. min. 7 sec.: min.
High Neutron Flux 2 Max. of 130% of 110% rax.
full scale with a 2.6 MW max.
High Temperature of Primary 1130F max.
Coolant Entering Core During F0rced Convection Cooling
- High Temperature of Primary 12507 max. 123cF max.
Coolant Leaving Core During Forced Convection Cooling
- Low Flow Rate of Primary 1 1200 gpm, 1350 gpm, Coolant
- min. min.
Low Pool Water Level 1 2" max. decrease 2" max. decrease Seismic Disturbance 1 IV on Modified Mercalli Scale max.
Bridge Misalignment
- 1 X X Coolant Gates Open* 1 per gate X X Neutron Detector High 1 per Decrease of Voltage Failure in Linear power 50 volts max.
Level Safety Channels supply Manual Scram (Switch at 2 X X bridge and on console)
High Conductivity of 1 Equivalent l Primary Coolant - to 24mho/cm at 250C, max.
Safety Blade Disengaged 1 X
- Log N - Period Amplifier" 1 X X Failure Regulating Rod at Either 1 X Limit of Travel Low Flow Rate of Secondary 1 800 gpm, Coolant
- min.
Bridge Movement 1 X X No Flow Thermal Column
- 1 X X
- These functions are bypassed when the Power Level Selector Switch is in the "0.1 MW" position.
Change 3,4,5 i
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I r TABLE P2 -
j REACTOR NUCLFAR ItlSTRUMENTATICl{
l Channel- Detector Sensitivity Range Information Informat. ion Information
- Fecorded to to to Information
~
Operator Logic Element Servo System '
(Scram) ,
, Retractable Neutrons- Source Neutron Relative '
gas filled approximately level Flux power level i Start-up B-10 filled 12 counts /nv to full None None on: log- i proportional power _
scale Neutrons- Source Power level i
Power level i Log N Fixed fission approximately level to Feriod Period scram None log Scale counter .~1 cps /nv 3x106 and period watts '
Linear Compensated Neutrons- I watt Power level I level ion chamber approximately to Power level Level scram Power level linear scale:
safety 4x10-14 amp /nv 3x106 '
(either watts (channel) ;
Linear Compensated Neutrons- I watt level ion chamber approximately ta Power level Level scram None -i safety 4x10-14 amp /nv 3x106
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.atts t Change 4 1
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G, WAETE S!!PCEAL AMS F AS ILITY M2!!ITIRI!!] EYETFME 1, 'fAs t e Dispc?al Svgem Desien Fearm
- a. M 2uid Padi230tive W3Ste DiFE2 s 31. E%EC All liquid waste (except sanitary waded from the reactor building shall flow to retention tanks. These tanks shall be located either underground with a dirt cover or in a locked rocm(s) in the reactor building.
- b. Oaseous Radioactive WAfte O i ? G111_12n!=
I All gaseous radioactive waste f rcm the beam ports, thermal column, pneumatic irradiaion system and all other radioactive gas exhaust points associated with the reactor it se lf shall be collected in a manifold and discharged to the reactor stack through an absolute filter, blower and damper.
- c. Mild Radicartive WAete Steragg Solid Radioactive wastes shall either be stored in radioactive waste storage containers located within the reactor building or removed from the site by a commercial licensed organization.
2, Area snM Fxhaust an, Meniter resien Features
- a. Three fixed gamma monitors employing suitable detectors shall be employed in the reactor building. Each of these shall have the following characteristics:
- 1) A range consistent with the expected radiation levels in the area to be monitored (0.01 to 10 mr/hr, 0.1 to 100 mr/hr, or 1 to 1,000 mr/hr).
- 2) A radiation dose rate output indicated in the control room.
- 3) An adjustable high radiation alarm which stwll be annunicated in the control room.
Amendment 12 Change 2,3 ,
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b.-
V 4) The thret fixed gamma monitors shall.be loca'ted to detect radiation as follows: At the pool biological shield between a beam port and the thermal column, above the storage container for new fuel elemen*s, and at the reactor bridge.
- b. A gamma monitor shall be provided near the primary tcolant system, and an additional one shall be provided near the secondary coolant system for use in determining the presence of abnormally high concentrations of radioactivity in these systems. The characteristics of these monitora shall be as stated in a. above,
- c. Six additional oirect reading area monitors employing Geiger tube dete :t o rs shall be provided to monitor the pneumatic system receiver stations, the be am . po rt areas, and other areas as required. Each of these shall have the following characteristics:
- 1) A range consistent with expected radiation levels in the area being monitored (0 to 10 mr/hr or 0 to 50 mr/hr).
- 2) A radiation dose rate output at the instrument,
- 3) An adjustable high radiation alarm to alarm at the inst rument and create both an audible and visual signal, e
- d. A stack exhaust gas monitor system shall be provided which draws a representative sample of air from the exhaust gas.
The monitor with indicators and alarms in the control room, shall have the following characterics:
- 1) A beta particulate monitor with an alarm.
- 2) A gas monitor incorporating. a scintillation detector with high level alarm-and a sensitivity for an Argon-41 concentration in air of 10-6 c/cc. The monitor shall have a range of at least four decades.
- 3. Other omdistien Meniterine Ecuinsnt
- a. Portable survey ' instruments for measuring beta-gamma dose rates in the range from .01 mr/hr to 250 r/hr shall be available at the facility. Portable instruments for measuring fast and thermal neutron fluxes in the range from 1 n/cm2 see to 25,000 n/cm2 sec shall also be available to the facility,
- b. Reactor excursion monitors shall be placed in the facility for measuring-gamma and neutron doses in the event of an accident.
Amendment 5 l
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-!6-
- c. A radiation monitor shall be provided o monitor all persons leaving the reactor room for beta-gamma contamination, a
4, 1112h Radiation Area During reactor operation, the dose rate from the delay tank may te in excess of 100 millirem per . cur. On three sides, the tank shall be shielded. On the fourth aide, the tank is shielded using a " maze" so that access to the tank is posssible threugn a door equipped with a lock.
O O
I O "r===
- 1. New ruel Storag1 New fuel shall be stored in a security container in " egg crate" boxes. Sheet cadmium at least 0.020 inches thick shall be fastened around the outside of the boxes in the region which contains the fuel. The aumber of fuel elements which can be placed in each box shall not exceed three. For all conditions of moderation possible at the site Kegg shall be less than 0.9.
- 2. nradiated ruel Su m Two types of irradiated fuel element storage racks shall be provided. One type of rack shall contain spaces for nine fuel assemblies and shall have approximate over-all dimensions of 35.5 i n '. wide by 26 in, high by 6.25 in. thick, and shall be fixed to the pool wall. At least two of these rack, shall be provided. The second type of rack shall consist of two of the nine fuel asserbly racks des.'ribed above attached together with a minimum space between the centerlines of fuel assentlies in adjacent racks of 12 inches. This 18 fuel asserbly rack shall be covered on the two 35.5 x 26 in, outside faces with a neutrcn absorbing material. At least one 18 fuel assembly rack shall be provided, and the rack may be moved within the pool.
The fuel storage racks may also be used to store core compot.ents other than fuel a s s emb lie s . The irradiated fuel storage racks shall have a maximum Keff of 0.8 for all O conditons of moderation possible at the site.
st.all be provided for at least 36 fuel assemblies.
Storage spaces I. EvrrRIMENTAL PAcTLITirs The permanent experimental facilities shall consist of the following:
- 1. Thermal column.
- 2. Beam ports; two 8 inch dia, and four 6 dia.
- 3. A six inch diameter through port.
- 4. Radiation baskets.
- 5. A two-tube pneumatic tube system.
- 6. Dry gamma cave.
O
J. A;MiMir?tt?!'/r A' 2 PROCED" PAL fl FE 0" APfli 1, 3pnizatirn The Rhode Island Atcmic Energy Ccesission (RIAEC) shall have the responsibility for the s&le operation of the reacter. The R7AEC shall appoint a D1:ector of Operations and a Peacter Utilization Committee censisting of a minimum of five cerrore, as fallows:
(1) The Director ot 0,neraticns (2) The Anactor Facility Health Thsicist (3) A qualified representative frem the faculty of Brown University (4) A qualified r e p r e s e n'.a t i ve frem the fa ty of Frovidence College
($) A qualified representative from the faculty of the University of Rhode Island.
A qualified alternate may aerve in lieu of one of the above.
The Director and Health Physicist are not eligible for chairmanship of the Committee. The T<e a c t o r Utilizaticn Cornittee shall have the following functions:
- a. Review proposals for the ris e of the reactor j
considering the suitability of the reactor for the proposed use and the safety factors involved.
9
_ . _ _ _ _ . . ___m_..._.__. _ _ _ . _ . . _ . . _ _ . _ . . _ . ___ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . .
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- b. Approve or disapprove proposed use of the reactor.
- c. Peview at least annually the operating and emergency procedures and the overall radiation safety aspects of the facility.
The Reactor Utilism. ion Committee shall maintain a written record of its findings regarding the above, i
- 2. - 22alificatirnn cf Perarnnel
- a. The Director of Operations shall have at least a bachelors degree in one of the physical sciwnces or engineering, and ,
he shall be trained in reactor technology and be a licensed senior cperator.
- b. The staff Health Physicist shall be professionally trained and shall have at least a bachelors degree in one of the l- physical or biological sciences or engineering 9 shall l have experience such as may have been gair 5 rough ernployment in a responsible technical positior in the fleid of health physics.
- c. The reactor operators ano senior operators shall be licensed in accordance with the provisiorss of 10 Crn 55.
4 l d. In the -event of t erepo r a ry vacancy an the position of Director of Operations or the Health Physicist, the '
O' functions of that position shall be assumed by qualified alternates appointed by the RIAEC.
- 3. Peenensibilities of Persennel
- a. D.1 Lector (1) The Director shall have responsibility for all activities in the reactor facility woich asy affect rea c e. or operations or involve radiation hazards, ina.uding centro 111ng the admission of i
personnel to the building. This responsibility l
shall encompass administ rative ' cont rol of all
- experiments being performed in the facility ,
.ncluding those of outside agencies.
(2) It shall be the responsibility of t he Director to insure that all proposed experiments, design modifications, or changes in operating and emergency procedures are performed in e cordance with the license. Where uncertainty exists, the Director shall refer the decision to the Reactor Utilization Committee. ,
h Change 4 i
i jO 4
- b. L.euter po uc t cretsters (1) A licensed senier reactor cperator shall be assigned each shift and be responsible for all activities during his shift which may affec' reactor Operation or involve radiation herards.
The reactor operators on duty shall be respcnsible directly to the senior cperator.
(2) The reacter operations which affect core reactivity shall not be performed without the senior cperator en duty or readily available en call. The senior eperator shall be present at the facility during initial startup and approach to power, recovery frem an unplanned or unscheduled shutdcwn cr significant reduction in pcwer, and refueling. The name of the person serving as senior operator as well as the tire ne a s s urt.e s the duty shall be entered in the reactor log. When the senior operator is relieved, he shall turn the operation duties over to another licensed senior operator. In such instances, the change of duty shall be logged and shall be definite, clear, and explicit. The senior operator being relieved of his duty shall insure that all pertinent information is logged. The senior operator assuming duty shall check the log for information or instructions.
- c. Rearter Orersters (1) **he respons ible senior operator shall designate for his shift a licensed operdlot (hereaf ter called " operator") who shall have primary responsibility under thL senior operator for the operation of the reactor and all associa,ed control and safety devices, the proper functioning of which is essential to the safety of the reactor or personnel in the facility.
The operator shall be responsible directly to the senior operator.
(2) Only ont cperator shall have the above duty at any given time. Each operator shall enter in the reactor log the date and time he assumed duty.
(3) Wher, operations are perf ormed which may af fect core reactivity a licensed operator snall be ststioned in the control room. When it is necessary for him to leave the control room during such an opuration, he shall turn the reactor and the reactor controls over to a designated relief, who shall also be a licensed operator. In such instances, the change of duty shall be cefinite, clear, and explicit. The relief shall acknowledge his entry on duty by proper notation in the reactor log.
~ 'l-i s
(4) The operator, under the tenior operator on duty, f shall be responsible for the operation of the reactor according to tl e approved operating schedule.
(5) The operator th a '. ; be authorized at any time to !
reduce the power of the reactor or to scram the reactor without reference to higher authority, when in his judgement such action appears advisable or necessary for the safety of the reactor, related equipment, or personnel. Any person working on the reactor bridge shall to similarly authorized to scram the reactor by ,
pressing a scram button located on the bridge. l d, gealth rtysirini The Health Physicist shall be responsible for assuring t b - *. adequate radiation montoring and control are in et act to prevent undue exposure of individuals to rac etion.
- 4. hritten -L. s t r u c t l a n
- and Precedures Detailed written operating instructions and procedures shall be prepared _for all normal operations and maintenance and for emergencies. These procedures shall be reviewed and approved by qualified personnel before use. Each member of the staff O .shall be familiar with those procedures and instructions for which he has responsibility.
- 5. Mite rmargenev Plans The Rh oc'e Island Nuclear Science Center shall have available the services of other state agencies for dealing with certain types of emergencies. The RIAEC shall enter into an agreement with the Rhode Island Civil Defense Agency whereby the Civil Defense Agency will maintain an emergency monitoring and communications vehicle which they shall make available to the Nuclear Science Center in the event of an emergency involving
-release of fission products or other radioactive isotopes to the atmosphere. The emergency vehicle shall contain equipment such as portable radiation monitors, respirators, and 4
-particulate air sampler. Communications using the statewide j
emergency network shall be available. !
Personnel of the Civil Defense Agency and of local f e departments shall have received training from the Civil Defense Training Officer in the use of certain radiological instruments. Future training shall be augmented by including orientation on the reactor facility.
O
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M. r m.Tiin MM1?A?ic!:;
- 1. E1111 The following administrative centrols shall te erployed ta O
assure the safe cperation of the facility:
- a. The teactor shall not te cporated whenever there are any signif3 cant defects in fuel elements, control rods, or centrol circuitry.
- b. The reactor centrol and safety syster must be turned en and functioning properly and an apptcpriate neutrcn scurce must te in the core during any change which can affect core reactivity,
- c. During cperations which could affect core react ivit y, a licensod cperator shall te static.ed in the cont rol recm.
Cormanications between the centrol room and the senior reactor operator directing the operation shall be maintained,
- d. The operator shall not attempt to start up the reactor following an automatic scra*n or unexplained power decrease until the senior Operator has determined the cause of the scram or pcwer decrease and has authorized a start-up.
- e. The initial start-up of the reactor shall be performed in conjunction with personnel of the General Electric Company,
- f. The reactivity of all core loadings to be utilized in cperating the reactor shall be determined using unirradiated fuel elements or elements containinq #issien products in which the effect of xenon poisioning L.. total core reactivity has decayed to 0.05% delta k/k or less,
- g. Critical experiments shall be performed under the supervision of the Director or other competent supervisory scientist licensed as a senior reactor operator. During the experiment there shall be present, in additon to this licensed supervisor, at least one other technically qualified person who sha'l act as an independent observer.
Each step in the procedure shalt be considered in advance by both persons, each calculation shall be checked by both persons, and no step shall be taken without the concurrence of both. A written record shall be made at the time of each fuel element addition or other core change which could significantly affect core reactis ly.
- h. The basic operating principles for the assembly and reloading of cores whose nuclear properties have been previously determined from critical experiments shall be as follows:
All core loading changes sht11 be performed under the supervision of a person having a senior operator's license. During the operation there shall be present in addition to the designated senior reactor operator at least one other technically qualified person who shall act as an observer.
Change 6
The exact procedure to be fc11 owed for a particular reloading creration will be determined by the ebserver and the senior reactor operator in charge of the operation before the cperation l begins. Each step in the procedures shall be considered by both persons, and no step shall be taken without the concurrence of bcth.
- 2. Exrerinnts i l
1
- a. " Experiments" as used in this section shall te l construed as any apparatus or device installed in the ;
core region which is net a O c:tpo ne nt of the core,
- b. The Pecctor Utilization CcrTittee shall teview and I I
approve all experiments Lefore initial performance at the facility. New types of experiments or experiments ;
of a type significantly different from those previously perf ormed shall Le described and doeurnented for the study of the Peactor Uti11 ration Ccmmittee.
The documentation shall include at least:
(1) The purpose of the experiment, (2) A description of the experiment, and (3) An analysis of the possible hazards associated with the performance of the experiment.
- c. All use of experimental facilities shall be approved by the Director of Operations,
- d. The absolute value of the reactivity worth of any single independent experiment shall not exceed 0.006. i If such experiments are connected or otherwise related so that their combined reactivity could be added to the core simultaneously, their combined reactivity shall not exceed 0.006.
- e. The calculated reactivity worth of any single independent experiment not rigidly fixed in place shall not exceed 0.0008. If such experiments are connected or otherwise related so that their combined reactivity could be added to the core simultaneously, l their combined recctivity worth shall not exceed 0.0008.
- f. No experiment shall be installed in the reactor in such a manner that it could shadow the nuclear instrumentation system monitors and thereby give erroneous or unreliable information to the control system safety circuits.
- g. No experiment shall be installed in the reactor in such a manner that it could fail so as to interfere with the insertion of a reactor control element.
. s.
-:4 .
- h. No experiment s h & J 1 1: e perfermed involvir g materials u s e:d in such a way that they might credibly result in an exploei:n i= No experir,ent shall Le perforned involving materials !
wh'ch eculd credibly contaminate the reactor pool l causing corrosive action on the reactor ccmponents.
- j. L.v.perimente shall not be perforned involving equipment
.hase failure could credibly result in fuel element damage,
- k. There shall be no more than one vacant fuel element positicn within the periphery of the active section of ;
the core.
- 3. 3nA;.una i
- a. m '
Control of access to the reacto facility shall be the responsibility of the Director of Operations, b, centain-ent (1) During any operation in which the control rods are withdrawn frcm the core containing fuel, the foll? wing conditions shall be satisfied;
- a. Confinement building penetrations whien are not designed and set to close automatically on actuation of the evacuation button shall be sealed, except that doors other than the truck door may be opened during reactor operation. If a coor is to remain open, an individual from the reactor operations staff is continuously in attendance at the door,
- b. The building clean-up system is operable.
(2) Recyfremants fer Retent of confinement (a) Methed ef Retent The building cleanup system shall be retested by pressing an evacuation button and observing that the follow!nq functions occur automatically:
- 1. Evacuation horn blows.
- 2. air conditoning and normal ventilation has turned off.
- 3. Dampers on all ventilating ducts leading to the outside have closed.
- 4. Building cleanup system-air scrubber and fresh air blower come on.
Change 4
, l 1
O 5. The negative differential pressure between the inside and outside of the building is at least 0.5 inches of water. This shall be determined by reading the differential manometer located in the control roem.
(b) ragnency cf Re t.sutt t
The building cleanup system including the auxiliary electrical power system shall be retestt'd at least weekly.
(3) The exhaust rate through the cleanup system shall not exceed 4500 cfm with not more than 1500 cfm coming :
from the reactor building and passing through the charcoal scrubber. The remaining air will be ,
provided by a separate blower from an uncontaminated source. This shall create a pressure in the building ,
which is equivalent to at least 0.5 inch of water below atmospheric pressure.
- c. Prfr3rv ceriant system (1) The minimum depth of water above the top of the active core shall be 23 feet.
() No piping shall be placed in the pool which could cause or fail so as to cause a siphon of the pool water to below the level of the ten inch coolant line ;'
penetrations.
(3) Makeup Eystem The effluent water of the primary coolant water makeup system shall be of a quality to insure compliance with K. 3.c . ( 5) and (6) below.
(4) ciennun system The effluent water of the primary coolant water clean up system shall be of a quality ,
to insure co.tpliance with K.3.c. (5) and (6) below.
g5) The primary coolant shall be sampled at a minimum frequency of once per week and the samples analyzed for gross radioactivity, pH, and conductivity 'in accordance with written procedures. Corrective action shall be taken to avoid exceeding the limits listed below:
Atr.endment 10
. Change 4 O.-
s
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pH 5.5 to 7.5 conductivity 2 kmho/cm (6) The radioactive materials contained in the pool water and in the primery coolant water shall te such that the radiation level one meter above the sut' face of the pool shall be less than 10 mrem /hr.
(7) During thu forced circulation mode of operation, j the primary coolant flow rate shall not te less than 1200 gpm. During determinations af t e s :: t o r power by coolant heat balances, the coolant flow i rate may be reduced to 600 gpm providing all other j aspects of these Technical Specifications are met.
- d. 122;;mtary ceriing_.fy'".es I (1) The secondary coolant shall be s a rtpl e d at a minimum frequency of once per week and the samples analyzed for pH in accordance with written procedures. Corrective action shall be taken to avoid exceeding the pH limit given below:
pH 5.5 to 9 (2) The concentration of radionuclides in the secondary water shall be determined at least once each day the reactor cperates using forced convection cooling. The concentration shall be determined at least once per week when not being eperated using forced convection cooling.
(3) If the radioactive materials contained in the secondary coolant exceed a radionuclide concentration in excess of the values in 10 CFR 20, Appendix D, Table I, Column II, above background, the reactor shall be shutdown and the condition corrected before operation using the secondary cooling system resumes.
(4) The secondary coolant system shall be pla:ed in operation as required during power operation utilizing forced convection in order to maintain a primary coolant core outlet temperature of 12500 or below.
- e. Egactnr ce rta nd control Eleonta (1) The reacter shall not contain in excess of 35 fuel elements. There shall be a minimum of four operable control elements.
Ie.endment 6,14 Change 3,4,7
-27 (2) The limiting thermal .a n d hydraulic cote O cha ract e rist ic s based on a 28 element, graphite reflected core are specified telew:
(a) tsaximum Heat riux 47,200 BTU /hr ft2 (b) Maximum Core Specific F;wer 1, ; 2 0 wat t / 7m U235 (c) Maximum Fuel Surface 197Dr Temperature (d) Coolant Velocity during 2.E5 ft/re:, min.
Forced Cenvection C ling (e) Coolant Inlet Temperature 1150r max.
(f) Average Coelant Terrp e rat u r e 1000 max.
Rise (g) Primary System Dulk Outlet 125Cr max.
Coolant Temperature (h) Ter*perature Ma rgin in Pritnary 430F Coolant (Tsat-Tsutt)
(1) Number cf Coolant Passus 1 Through Core (3) Princiral Muclear cha r a cte rinh Cf g;;ta (a) Cern a n d C e nt r o l E ys t em " n actigity__W;nt
- 1. The reactor s h r.11 be suberitical by at least it Ak/k frem tne cold, Xe-free, critical condition <i t h the most reactive centro; element and the servo regulating element fully withdrawn.
- 2. The maximum dorth of the servo regulating element s h a l '. be 0.7%
ak/k.
(b) Maxi ~um Neactivity Additien Ratu - Ak/k/sg;
- 1. By servo regulating element traximur of 0.0002
- 2. lianut.1 by control element n.aximum of 0.0002 O
Amendment 13 Change 4
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(c) Eca;11vity Czeffirignis l
- 1. Tenperature coefficient approximately j
-0.5 x 10-4 /cc (calculated)
- 2. Void ccefficient approximately (core average) -1.9 x 10-3/* void
( c ' '. c u l a t e d )
(4) E.rin;1;;al ca r e c; staling _Limitst hns (ai li.LKE.2*- poi IC0&RIALMIS ih.ikathna The pool water t e rt pe r a t u r e shall not exceed 1300r.
(b) Reactivity limitaticta
- 1. Exce.?s Reactivity The cold, clean excess reactivity for any core used in the reactor shall not exceed 0.047 2, Mini-o- Shutdrwn Marcin All reactor cores used shall be such that they would be suberitical if any single control element and the servo regulating element were withdrawn.
(c) Reactivity Ceefficient Limitation The reactor power coefficent (as inferred by the control rod movements required to compensate for changes in power) shall be negative.
(d) e ntrel rie-nnt Drive perfer-3nce Requi r-onts l
All control element drives shall meet the following specifications:
- 1. The control drive withdrawal rate shall not j be more than 3.6 inches per minute.
- 2. For the electronic scram system, the time from initiation of a scram condition until control element release shall not exceed 100 milliseconds.
- 3. The time from initiation of a scram condition until the control element is fully inserted shall not exceed 900 milliseconds.
- 4. It shall be demonstrated at least every 3 months that the above specifications are met.
Change 4,7
(e) gerv- Reuuistine ricrent Drive r' e nf e r- a n c e Fecei*e*ents ,
If in use during operation, the servo regulating element drive shall meet the following specifications:
- 1. The drive wit .rawal rate shall not be more than l*
78 inches per minute.
I
- 2. It shall be demonstrated at least cnce per month that the above specification is met.
(f) rission rensity limit The fissicn density limit for alloy, uranium aluminide, and uranium oxide fuel shall meet the following specifications:
- 1. The fission density limit shall be 0. 5 ' x 1921 fissions /ce. ,
- 2. The fission density of all fuel elements which have burnup shall be calculated at least ,
quarterly.
- f. Reacter Safety Evsteam (1) The reactor safety system shall be operable during all reactor operation. The safety system shall be checked cut before each start-up and functionally tested for calibration at least monthly.
'2) It shall be permissible to continue operations with one or more of the saf ety system functions that produce only an alarm temporarily disabled providing that additicnal procedural controls are instituted to replace the lost safety system alarm function (s).
(3) The control element withdrawal interlocks and the servo system control interlocks shall be functionally tested at Isaat once per month.
(4) During reactor startup or during mechanical changes that could - af f ect core reactivity, the startup range neutron '
-monitoring channel shall be operable and shall provide a neutron count rate of at least 3 counts per second with a signal to noise ratio at least 3 to 1.
(5) The linear level safety channels shall not read less than 15% of full scale when the reactor is operating at power levels above 1 watt.
(6) Following a - reduction in power level, the operator shall adjust the servo power schedule to the new power' level before switching to automatic operation.
-(7) - An - sla rm conditon from any one- of the items - listed in Section F.2.b. after working hours shall transmit coded information to a continuously manned central station in Providence, Rhode Island. The central station shall be i
O provided with written instructions on the steps to be taken following an alarm.
Amendment Bt Change 4
- g. girte Dirrr*31 and peneter Meniu rins syst m l (1) The liquid waste retention tank discharge shall flow to a monitor station in the reactor building where the effluent shall be batch sartpled and the gross activity per unit volume determined before release. All off-site releases shall te directly into Narragansett bay.
(2) Gaseous radioactive waste shall be disposed of using the reactor stack. Disposal limits shall conform to the following table. In this table, the MFC stated is for individual isoteres and mixtures centained in Column 1, Table II, Appendix B of 10 CFR 20.
1 2 Type of Activity Maximum Curies Curies por second to per second to be ce released averaged released cver one year Particulate Matter and Halogens with half lives 140 X MPC (ue/ce) 14 X Mi c (uc/cc)
Icnger than 9 days All other Radic ctive 105 X MFC (ue/cc) 104 X MFC (uc/ce)
Isetopes O
(3) All radioactive liquid and solid wastes disposed of off-site shall be within the limits established by 10 CFR 20 or shall be removed frcm the site by a cor:nercial licensed organization.
(4) The exhaust gas monitor shall be calibrated to alarm -
at an instantaneous release rate which instantaneously exceeds the limits stated in Column 2 for the annual average release rate. If the maximum permissible stack release rate stated in Column 1 is exceeded, the reactor shall immediately be placed in the shutdown mode of operation and the situation investigated.
(5) The area, primary and secondary coolant system and the exhaust gas monitors shall be in operation at all times when control elements or the servo regulating elements are withdrawn; however, indivdual area coolant system monitors may be taken out of service for maintenance and repair if replaced with portable radiation detection equipment. Adequate spare parts shall be on hand to allow necessary repairs to be made during the maintenance or calibration outages of the monitors.
O Amendment 12
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(6) The area and the primary and secendary coolant system monitors shall be adjusted to alarm at _ a maximum reading of 2 mr/hr or 200% of the normal tadiation levels in their area, whichever is larger (7) The door which controls entrance to the " mare" leading to the delay tank shall be locked with the key in the possession of the Director or a licensed senior operator. Entrance to the delay tank high radiation area shall require the presence of the Health rhysicist or a licensed senior eperator and the use of dirqct- reading portable radiation monitoring equipment.
- h. EacL.Etr.LGa (1) New futi shall be stored in egg crate boxes located in a security container. Access to the security container shall be restricted, through use of a lock, to the Director of Operations and the licensed senior reactor operators.
(2) Irradiated fuel, not in use in the reactor core, shall be stored in the criticality safe storage racks described in Section H. Only one fuel assembly may be inserted or moved from a storage rack at a time.
(3) Safety against inadvertent criticality shall be O provided by limiting the number of fuel assemblies per rack to nine and then positively securing such racks at least 30 cm. apart, or by limiting the number of fuel assemblies to 18 per rack and then covering the two large faces of each rack with a sheet of aluminum covered cadmium.
- 4. Maintenance (a) The elescronic control and the process control system shall be i checked for proper operation and calibration before each reactor start-up. If maintenance or recalibration is required, it shall be performed before reactor start-up proceeds.
(b) Maintenance shall be performed with the approval of the Director. Equipment and system maintenance records shall be kept to facilitate wheduling and completion of all necessary naintenance.
-)
(c) Routine maintenance on all control and process system '
components shall be performed in accordance with written schedules and with written procedures.
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Part C Technical Creci11 cations Review and Modification INTRODUCTION There are numerous Technical Specification changes required as a result of the use of the LEU fuel in the Rhode Island Nuclear Science Center reactor.
Parts A and B of t h te Safety Analysis Report touch on many of them. As a result of the Rhode Island Nuclear Science Center review process, additicnal changes which reflect current conditions or clarifications of some Technica] Specification sections are also included in the final Technical Specification version. Appendix A is a copy of the Rhode Island Nuclear Science Center current Technical Specifications. Appendix B is a copy of the Ttchnical
- Specifications with the changes included as a result of the SAR and review process. The double tertical lines adjacent to a section designates the section which has the proposed changes.
Implementation of the final approved Safety Analysis Report will be a difficult task for the Rhode Island tJuclear Science Center. Conditions outside the control of the l licensee, such as key staff retirements, budget cuts, small j cperating staff etc., increase the difficulty and will curtail the operation of the facility during the conversion process. The Rhode Island Nuclear Science Center acknowledges the assistance of Argonne National Labort. tory in the preparation of the Safety Analysis Report.
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APPENDIX B I
e PROPOSED RHODE ISLAND NUCLEAR SCIENCE CENTER REACTOR TECHNICAL SPECIFICATIONS P
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e ,
TABLE CF CCNTENTS PAGE O' A. SITE
- 1. Location 1
1
- 2. Exclusion Ates 1
- 3. Restricted Area 1
- 4. Principal Activities 1 Figure A.1 2,2a
[
B. CONTAINMENT 3 -
- 1. Reactor Building 3 ;
C. REACTOR POOL AND PRIMARY COOLANT SYSTEM 4
- 1. General 4
- 2. Reactor. Pool 4
- 3. . Shielding 4 i
- 4. Prinary coolant System 4
- 2. Heat Exchanger 4
- b. Primary Pump 4
- c. Delay Tank 5
- d. Primary Recirculation Piping 5
- e. Make-up System 5
- f. Clean-up System for Primary Coolant System 5
)
D. SECCNDARY COOLANT SYSTEM 6 ,
E. REACTCR CORE AND CONTROL ELEMENTS 7 1.- Principal Core Materials 7
- 2. Fuel Elements 7
- 3. Reflector Element s 8
- 4. Control Elements. 8
- 5. Servo Regulating Element 8
- 6. Control Element Drive 8 7 Servo Regulating Element Drive 9
- 8. Neutron Sources 9 F. REACTOR SAFETY SYSTEMS 9
- 1. Modes of Power Operation 9
- a. Power Operation <- Natural Circulation'(NC) 9
=b. Power Operation - Forced Circulation (PC) .9-
- 2. Design Features 10
- a. The. Reactor Control System 10
- b. Process Instrumentation 10
- c. Master Switch 10
- d. Power Level Selector Sw. h 11
- e. Control Element Withdrawal Interlocks 11
- f. Servo System Control Interlock 11 Table F.1 Reactor Safety System 12 .
Table F.2 Reactor Nuclear Instrumentation 13:
G. WASTE DISPOSAL AND FACILITY MONITORING SYSTEMS 14
- 1. Waste Disposal Systems Design Features 14 *
- a. Liquid Radioactive Waste Disposal System 14
- b. _ Gaseous Radioactive Waste Disposal System 14
- c. Solid Radioactive Waste Storage 14 2.. Area and Exhaust Gas Monitor Design reatures 14
- 3. Other Radiation Monitoring Equipment 15
- 4. High Radiation Area 16 i
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l TABLE OF CCNTENTS (CONTINUED) 3 H, FUEL STORAGE 17 l 1. New Fuel Storage 17
! 2. Irradiated Fuel Storage 17 )
l l
1 1. EXPERIMENTAL FACILITIES 17 l t i J. ADMINISTRATIVE AND PROCEDURAL SAFEGUARDS 18 l 1. Organization IB i
! 2. Qualifications of Personnel 19 j 1
- 3. Responsibilities of Personnel 19 5
- a. Direct or 19 l l
- b. Senior Reactor Operators 20
- c. Reactor operators 20
- d. Health Physicist 21 ,
- 4. Written Instructions and Procedures 21 !
i S. Site Emergency Plans 21 l 1 i I K. CPERATING LIMITATICNS 22 l
- j. ,
- 1. General 22 l
i 2. Experiments 23 ,
- 3. Operations 24 l
- a. Site 24 l
- b. Containment 24 ,
! c. Primary Coolant System 2$ 1 l d. Secondary Cooling System 26 i i e. Reactor Core and Control Elements 26 L
- f. P.eactor Safety Systems 29 i' g. Waste Disposal and Reactot Monitoring Systems 30 [
- h. Fuel Storage 31
- 4. Maintenance 31 l l i
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- 1. M:.RL:n The reactor shall be located at the Rhode Island Nuclear science Center on three acres cf a c7-acre former military reservation, originally called Fort rearney and now called the Narragansett Bay Car pus of the University of Rhode Island. The University of Rhode Island is a state agency. The 27-acre reservation is controlled by the State of Rhode Island through the University of Rhode Island. The reservation is in the Town I of Narragansett. Rhode Island on the west shore of Narragansett Bay approximately 22 miles south of Providence, Rhode Island and approx 2mately six miles north of the entrance of the Day f - c rt +5e Atlantic Ocean. The Rhede Island Nuclear Science Center and various bu!.1 ding.i used for research, education and training purposes are located on this 27-acre campus.
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- 2. rxclusien Area rigure A.1 is a drawir.g of the Narragansett Bay Campus showing the three acre Nuclear Science Center site. The boundary of this area shall be posted with conspicuous signs to delineate the area. This three acre area shall be the exclusion area as defined in 10 CFR 100.
- 3. Rest ricte d AltA rigure A.1 also shows the location of the reactor building on the three acre area. Ti.e reactor building and attached of fice laboratory wing shall be considered the rostricted area as defined in 10 CrR 20,
- 4. EI i n-ba l P.ctivities The principal activities carried on within the restricteo and exclusion area enall be those associated with operation and utilization of the reactor. It shall be permissible to locate additional Nuclear Science Center or University of Rhode Island buildings within the exclusion area provided that these additional buildings are capable of timely evacuation and do not inte J fere with the operation of the reactor.
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j 1. pi~ m Nil m g the reacter shall be housed in a 1uilding capable of meeting the following functienal r equi r e ment s :
In the event of an accident which c uld involve the release of radioactive ruterial, the confinement building air shs11 1e exhausted t b reuah a clean-up Syst(m and stack creating a flow of air intc the building with a negative diftetential pressure t+ tween the l'uilding and tho s u t s i c'e atmeschete. *I h e building shall te ans tight in the sense that a negative differential pressure can be taintainmi dynamically with all gas leaks occutting inward. The centinicent and clean-up syst em' shall becerm operative when a 1;u11 ding evacuation but t cn is pressed. This action shall: (1) turn eft all ventilation fans and the eat conditicner system snd (2) close the d wpets cn the ventilation and ait conditiching system Antakes and exhaust, other than these which are a part of the clean-up system. Na further action shall te required to establish e c n f i n er*,e n t and pleco the clean-up system in operativn. An auxiliary elect rical power system shall be prov1Jed at the site to insure the availability of power to cperate the clean-up system.
The reactor building exhaust blower, which is designed to exhaust at least 4000 c f m, eperates in conjunction with
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i3 additional exhaust blewer (s ) which provido an additional
() exhaust of at least 10000 cfm item non-teactor building searcon and in conjunction wit h the air handling unit which takes air into the reactor building at less than 4000 cfm. The total exhaust rate through the stack is at least 14000 cfm. Daring normal epotation, the building is at a pressur9 semewhat 1.elong atmospheric. The control room air (Inditioner shall t o a self-contained unit, thermostatically co tolled, providing ccnstant air temperature for the ccntrol tn- If it is Installed with a penetration thrcugh the wall t( the reactor bu11dinge it shall have a damper at this penett ' ion which closes when an evacuation button is prossed. Y t'pon a ct ivat ion, tne clean-up system shall exhaust air f rcm t he reactor building through a filtet and a 115 foct high stack, creating a pressure less than atmnspheric pressure. The clean-up filter shall contain a roughing filter, an absolute particulate filter, a charcoal filter for removing radiciodine, and an absolute filter for temoving charcoal dust which may te contaminated with radiciodine Each absolute filter ca rt ridge shall be individually tested and certified by the manufacturer to have an efficiency ci nct less than 99.97% when tested with 0.3 micron diameter diocty1phthalate smoke. The minimum removal efficiency of the charcoal filters shall be 99%, based on CP.NL data and measurements performed locally.
Gases frcm the beam ports, thermal column, pneumatic system, and all other radioactive gas exhaust points shall be exhausted to the stack throuah a roughing and absolute filter system, v
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- 1. Octmal The primary ecolant sys*em shall censist of the reactor pool, delay tank, heat e xch arg , coolant p urr p , and the associated valves, piping, flow channels and sensors. During forced convection cooling, coolant water shall be supplied to the core by an aluminum line connected to the inlet flew channel whten is on one side of the suspension frame. The ecolant water shall flow frem the inlet flow channel downward through the cere to a plenum below the grid box. The coelant watet shall then flow into the outlet flow channel en the epposate side f the suspension frame and then threugh a discharge line to the delay tank, coolant pu p, heat exchanger and then returt to the
, coolant inlet line.
- 2. Peseter reel The reactor pool shall be constructed of ordinary concrete with 1/4" thick 6061-T6 aluminum liner and shall have a volume of approximately 36,300 gal.
- 3. Phleiding The reactor pool and primary system shielding shall be adequate to meet the applicable personnel radiation protection requirements of 10 CTR 20.
y 4. Erf ra ry cociant system The primary coolant system shall conform to the following:
- a. Heat Ex hancer The heat exchanger shall be designed to remove heat at the rate generated by the reactor at maximum licensed steady state power from the primary water and shall be designed to perforn under the maximum primary system operating temperature and pressure.
Replacement heat exchanger shell and tube bundles shall be constructed from stainless steel according to the requirements of Section III, Class C of the ASME Boiler and Pressure Vessel Code,
- b. Prim ry P p Number of pumps 1 Type Horizontal mounted, Centrifugal, Single Suction O
_ _ _ _ . _ - - - - - - =--
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Materials cf ecnstruction Werthite Rating 1500 gpm Head 59 feet Design Pressure 75 psig minimum Design Temperature 150er minimum Motor Type Drip proof, induction, 440 v, 3-phase, 60 cycle
- c. [3 13 y_. M r.
Number of tanks 1 Material of construction Aluminum Association Alloy 5083 and 5086 Material Thickness i
Walls 0.25 inch Dished Heads 0.375 inch capacity 3000 gal., minimum d .. Pr b rv Recirculation Pining Material and thickness Sch, 40 A1. type 3003 aluminum Sire 8 and 10 inch Design temperature 150er, minimum t
, Design pressure 100 psig, minimum i
e, Mke-un System A check valve shall be installed in the line between the potable water supply and the make-up and cleanup
, demineralizer to prevent entry of potentially contaminated water into the potable water supply.
Water source Potable water from i city main !
Make-up domineralizer type Mixed-bed _ single shell, regenerative Make-up demineralizer capacity Normal 25 gpm Emergency 50 gpm Water softener capacity l Normal 50 gpm t i
- f. Clennuo System for Prirary ccolant Water +
Cleanup pump Capacity 40 gpm Head 100 ft
!- Cleanup cemineralizer Type Mixed-bed,_ single shell, regenerative Cleanup demineralizer capacity Normal 40 gpm Emergency 50 gpm 1
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-t-D. SrcSNDARY c30LANT SYSTEM The secondary coolant system shall carry tho heat rejected from :.h e primary coolant at the heat exchanger to the atmosphere at cooling towers. It shall be composed of the heat exchanger, cooling towers, pumps and associated valves, piping and sensors. In this system, water flows frem the heat exchanger through a control valve to the ceoling towers. From the cooling tower basins, the water is then purred back to the '; eat exchanger.
e ruercr!!?Y COPr cort ?:0 Erz;rM An emergency core cooling system shall be in place to provide a minimum of 4 GPM directly to the core grid box for a minimum caration of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. l 9
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E. REAcTcn EcRr ANS ccliTROL EttMEMTL j The reactor core and control elements shall have the following characteristics and nominal dimensions: i
. ruel clad 6061 aluminum :
Fuel element side plates 6061 aluminum ,
End fittings 356-T6 or 6061 aluminum Moderator Water ,
Reflector-Graphite AGOT grade (or equivalent graphite and/or water Reflector-Beryllium- Beryllium-aluminum clad Control elements Mixture of B 4C and aluminum, clad with aluminum ,
4 Servo Element Stainless steel 314 ll
- 2. E.acLElemen:,a Plate width overall 2.81 inches Active plate width 2.4 inches maximum ;-
Plate length overall 25 inches Active plate length 23,5-inches ll Plate thickness 0.06 ines.
Clad thickness 0.02 inch Fuel matrix thickness 0,02 inch Water gap between plates 0.1 inch r
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Nu:Ti:er of plates per fuel element 22 U-235 per fuel element M 5 grams, nemins1 Overall fuel element dimensinns 3 in x 3 in. x 40 in.
- 3. Ec11c a;,;t.Ilenent - 2:g hi+c_,and_Ituty111 =
Overall ref lect er element 3 in x 3 in. x 40 in, dirensions, ncminal No.tinal clad thickness .1 in.
bominal graphite dimensions ~.6 in. x 2.8 in. x 29.7 in.
Nominal Beryllium dimensions 2.94 in. x 2.94 in. x 29 in
- 4. m uroi ri e nn+a Width 10.6 in.
Thickness 0.38 in.
Overall length 54.1 in.
Active length 52.1 in.
- 5. fervc P.c;glating E.' engnt Shapo Square stainless steel ll Width 2.1 in.
Overall length 28.8 in.
Active 24.9 in.
- 6. Centrel rieront Brire Type Electromechanical screw Drive to safety element Electromagnet coanection Stroke 32 in, maximum i
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. -9 7 Serve peculatin, Elerent rrive Type Electrcmechanical screw Drive to element connection Lock screw (no scram)
Stroke 26 in, maximum Position indicatien accuracy 1 0.02 in.
- 9. "eutrrn searcqs Stut-up Source Nurter 2 Type Plutonium-beryllium Unit Source Strength 1 x 106 neutrons /sec minimum Maximum Power Level with Plutonium-beryllium sources installed- 10 Kw Cperational Source Nurbe r 1
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' Type Antimony-beryllium Source Strength 2 x 106 neutrons /sec. minimum T. REACTOP. ?AFE?Y EYSTEMS
- 1. Medom of Power Creration
-There shall be two modes of power cperation:
(
j a. Pcwer Operation - Natural Circulation (NC)
Power operation -
NC shall be any reactor operation performed with the reactor cooling provided by natural circulation. The reactor power shall not exceed 0.1 MW during NC operation.
( b. Power Ceeratien - Perced Circulation (FC)
Power operation -
FC shall by any reactor, operation performed with reactor cooling provided by forced circulation. The reactor power shall not exceed 2 MW during-FC operation.
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- 2. resi~n restures
- a. The Rescter C =trel Iystem The reactor safety system shall consist of sensing devices and associated circuits which automatically sound an alarm and/or produce a reactor scram. The systems shall be designed on he fail-safe principle ( de -e n e rgi z i r.g shall cause a sciam). Table " .1 and F.2 describe the arrangement and requirements of the safety system,
- b. Preress In r t rure nt s r ie n Frocess instrumentatien with readout in the control rocm shall be provided to permit measurement of the 110w rate, temperature, and conductivity of the primary coolant and the flow rate of the secondary coolant. In addition, a second primary flow indicating device with readout in the centrol room shall be located between the reactor outlet pler.um and the reactor outlet header.
After nt. mal working hours, an independent protection system, semarate from the system described in Section K.3.a, shall be used to monitor certain items in the reactor building and alarm in the event of an abnormal condition. The alarm channels provided are:
(1) A fire in the reactor room, (2) A fire in a location other than the reactor room, (3) A decrease of 2 inches in reactor pool water level, (4) A power failure in the reactor building, An alarm condition from the radiation monitcrs reading out in the control room,
( alarm condition frcm any other selected feature,
- c. Mit s t e r swit 3 A key lock master switch shall be provided with three positions; "off", " test", and "on". These positions shall have the following functions:
(1) The "off" position shall de-energize the reactor control ~.rcuit.
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l (2) The " test" position shall energize the reactor O control circuit exclusive of the control bl-de magnets.
(3) The "on" position shall energi;:e the reactor control circuit including the control blade magnets.
- d. E;En d evel Selc1^r " Wit 2h A power level selector switch shall be provided with four positions; "O.1 MW", "1 MW", "2 MW", and "5 MW". These positions shall have the f ollowing f t . -tior.3:
(1) The "O.1 MW" po,ition shall activate all safety system sensors except those indicated in Table F.1.
(2) The a' MW" and "2 MW positions shal* activats e. l l safety system sensors.
(3) The "5 MW" position shall scram the reactor,
- e. Centrol Element Withdrawnl Interirrks Interlocks shall prevent control rod withdrawal unless all of the following conditions exist:
(1) The master switch is in the "on" position,
) The safety system has been reset, (3) The Log N amplifier switch is in the " operate" posi.icn, (4) The startup channel neutron count rate is three counts per second or greater, and (5) The start-up counter is not being withdrawn.
It shall not be possible to withdraw more than cone control element at a time,
- f. Servo S ys t e.n C-snt r ol InterlSek Interlocks shall prevent switching to servo control unless the period as indicated by the Log N channel is thirty seconds or greater. The Servo control system shall be designed so that imme lla tely f 011c t- 4 scram the Servo control shall automatically t3 turn 'o no manual mode of operation.
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- r r.1 - REArTTP IArr?Y SY;;ng I Sensor or Trip Device No. of Swit ches T r i .o Set Alarm Set er Sensors Point Point '
Short Feriod 1 3 sec. min. 7 sec, min.
High Neutron Flux 2 Max. of 1:01 ef 110i max. ll full scale with a 2.4 MW max.
High Temperature of Primary 113sF max.
Ccolant Entering Core During Forced Convection Cocling*
High Terperature of Primary 1250F rax. 1230F max.
Coclant Leaving Core During Forced Convection Cooling
- Low Flow Rate of Primary 1 1580 gpm, 1650 gpm, ll Coolant
- min, min.
Low Pool Water Level 1 2" max. decrease 2" max. decrease 3eismic Disturbance 1 IV on Modified Mercalli Scale max.
High Pool Temp 1 1250F 1200F ll Bridge Misalignment
- 3 ,
X X Coolant Gates Open* 1 per gate X X Neutron Detector High 1 per Decrease of voltage Failure in Linear power 50 velts max.
Level Safety Channels supply Manual Scram (Switch at 2 X X bridge and on console)
High Cenductivity of 1 Equivalent Primary Coolaat to 24mho/cm at 250C, man.
Safety Blade Disengaged 1 X Log N - Feriod Amplifier 1 X X Failure Regulating Aod at Either 1 X Limit of Tra'el Low Flow Rate et secondsry 1 800 7pm, Coolant
- min.
Bridge Movement 1 X X No Flow Thermal Column
- 1 X X
- These f unctions are bypassed when the Power Level Selector Switch is in tPc "O.1 MW" position.
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f TABLE F.2 REACTOR NUCLEAR IfGIlW1*ENTATIQ1]
- Channel Detector Sensitivity Range. Informatron Information Information Recorded to to to Information Operator Logic Element Servo System (Scram) ,
Retractable Neutrons- Source Neutron Felative gas filled approximately level Flux powe- level Start-up B-10 filled 12 counts /nv to full None None on log' proportional power scale !
Neutrons- Source: Power level Power level Log N Fixed fissi.sn approximately level to Period Period scram None log scale counter 7 cps /nv 3x106 and period watts Linear Compensated Neutrons- I watt Fower level .
level ion chamber approximately to Power level Level scram Power level linear scale safety 4x10-14 amp /nv 3x106 (either watts (channel)
Linear Compensated Neutrons- I watt i level ion charrber approximately to Power level Level scram None safety 4x10-14amp/nv 3x106 watts h
p.. _____ ___-m_m.- ----------- - - -
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e G. WASTE DISFOSAL AND FACIt!TY MONITORING SYSTEMS
- 1. Waste Disposai Systems Design Features
- a. Liquid Radioactive Waste Dispesal Jystem all liquid waste (except sanitary waste) from the reactor building shall flow to retention tanks. These tanks ehall be located either1 underground with a dirt cover or in a locked room (s) in the reactor building.
- b. Gaseous Radiegetive Waste Disposal System All gaseous radioactive waste from the beam ports, thermal column, pneumatic irradiation system and all other radioactive gas exhaust points-associated with the reactor itself shall be collected in a manifold and discharged to the reactor stack through an absolute filter, blower and damper,
- c. Solid Pndicactive Waste Storace Solid Radioactive wastes shall either be stored in radioactive waste storage containers located within the reactor building or removed from the site by a commerefal licensed organization.
s 2, Area and Evhaust Gas MSnitor Desien Features
- a. Three fixed gamma monitors employing suitable detectors shall be employed in the reactor building. Each of these shall have the-following characteristics:
- 1) A range consistent with the expected radiation levels in the area to be monitored (0.01 to 10 mr/hr, 0.1 to 100 mr/hr, or i to 1,000 mt/hr).
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- 2) A radiation . dose rate output indicated in the control room.
- 3) An adjustable high radiation alarm which shall be-annunciated in the control room.
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- 4) The three fixed garna monitors shall be located to aetect r. a d i a t i o n as follows: At the pool biclogical shield between a beam port and the thermal column, abOve the storage container for new tuei elements, and at the reactor bridge.
D. A gamma monitor shall be provided near the primary coolant syster, and an additional one shall be provided near the secancary coolant system for use in determining the presence of abnormally high concentrations of radios:tivity in these systems. The characteristics of these m:nitors shall be as stated in a. above.
- c. Six additional direct reading area monitors employing Geiger tube detectors shall be provided to monitor the pneumatte system receiver stations, the beam port areas, and other areas as required. Each of these shall have the following characteristics:
- 1) A range consistent with expected radiation levels in the area being monitored (0 to 10 mr/hr or 0 to 50 mr/hr).
- 2) A radiation dose rate output at the instrument.
- 3) An adjustable high radiation alarm to alarm at the instrument and create both an audible and visual signal.
- d. A stack exhaust gas monitor system shall be provided which draws a representative sample of air from the exhaust gas.
The monitor with indicators and alarms in the control r:cm, shall have the following characteristics:
- 1) A beta particulate monitor with an alarm.
- 2) A gas monitor incorporating a scintillation detector with high level alarm and a minimum detectability level for an Argon-41 concentration in air of 10-o c/cc. The monitor shall have a range of at least four decades.
- 3. Other RMiarirn Monitorine Equi ~wnr
- a. Fortable survey instruments for measuring beta garna dose rates in the range from .01 mr/hr to 250 r/hr shall be available at the facility. Portable instruments for measuring fast and thermal neutron fluxes in the range fr:m 1 n/cm2 see to 25,000 n/cm2 ser shall also be available to the facility,
- b. Reactor excursion monitors shall be placed in the facility for measuring gamma and neutron doses in the event of an accident.
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- c. A radiatlan monitor shall be provided to . monitor - all 4
persons- leaving the . reactor room for beta-gamma contaminatic.n.
. 4. High Radiatien Area During teactor operation, the dose rate from the delay tank may be in excess of 100 millirem per hour. On three sides, the tank shall be shielded. On the fourth side, the tank is
- _. shielded using a " maze" so that access-to the tank is possible through-a door equipped with a lock.
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- 1. h_DelEtr:Ee tieu fuel shall be stored in a security container in " egg crate" b o x e +. . Sheet cadmium at least 0.020 inches thick shall te fastened around the outside of the boxes in the region which centains the fuel. The nu:Tle r of fuel ele: ent s which can be placed in each box shall not exceed three. FC: all renditio.s cf moderation possible at the site Eeff shall he less than U.S.
- 2. 1;:1;ii 1 c d Fue1 St0Ia;c Twc types of irradiated fuel element sterage racks sha;l be prcvided. One type of rack shall centain spaces for nine tue.
asserblies and shall have approximate over-all dimensions of 35.5 in. wide by 26 in. high by 6.25 in, thick, and shall be fixed to the pocl wall. At least two of these racks shall be prcvided. The second type of rack shall consist of two cf t r.e nine fuel assembly racks described above attached together with a minimum epace be: ween the center lines of fesi assemblies in ad$atent racks of 12 inches. This 16 fuel assembly rack shall be covered on the two 35.5 x 26 in, outside faces with a neutrcn absorbing material. At least one 18 fuel assently rack shall be provided, and the rack may be coved within the pool.
The fuel storage racks may also be used to store core ccmponents other than fuel assemblies. The irradiated fuel storage racks shall have a maximum Koff of 0.8 for all ccnditions of moderation possible at the site. Storage spaces shall be p;cvided for at least 36 fuel assemblies.
I. EXPERIMF' tat FACILI* Irs The permanent experimental facilities shall consist cf the following:
- 1. Ther:ral column.
- 2. Beam ports; two 8 inch dia, and four 6 dia.
- 3. A six inch diameter through port .
- 4. Radiation baskets.
- 5. A two-tube pneumatic tube system.
- 6. Ory gamma cave.
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) J. Ar_MINisTRATm AN'S P o oc E S"o A'- SAFET3 ADDS 1, or,ani-stien <
l The Rhode Island Atomic Energy Commission (RIAEC) shall have the responsibility for the safe operation of the reactor. The RIAEC shall appoint a Director of Operations and a Reactor Utilization Committee consisting of a minimum of fi ve merrters, as follows:
(1)- The Director of Operations (2) The Reactor Facility Health Physicist (3) A qualified representative from the faculty of Brown University (4) A qualified representative from the faculty of Providence College (5) A qualified representative from the faculty of the University of Rhode Island.
A qualified alternate may serve in lieu of one of the above.
The Director and H+)alth Physicist . are not eligible ior chairmanship of the Committee. The Reactor Utilization Committee shall have the following functions:
O a. Review proposals for the use of the reactor considering the suitability of the reactor for the proposed use and the safety factors involvad.
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- b. Approve or disapprcve proposed use of the reactor.
- c. Review at Jeast annually the cperating and emergency procedures and the overall radiatien safety aspects of the facility.
The Peactor Utilization Committee shall maintain a written reccrd of its findings regarding the above.
- 2. 02slifirstiens sf Persennel
- a. The Director of Operatiens shall have at least a bachelors degree in one of the physical sciences or engineering, and he shall be trained in reactor technology and be a licensed senior operator,
- b. The statf Health Physicist shall be professionally trained and shall have at least a bachelors degree in one of the physical or biological sciences or engineering. He shall have experience such as may have been gained through employment in a responsible technical position in the field of health physics.
- c. The reactor operators and senior operators shall be licensed in accordance with the provisions of 10 CFR 55.
- d. In the event of temporary vacancy in the position of Director of Operations or the Health Physicist, the functions of that position shall be assumed by qualified alternates appointed by the RIAEC.
- 3. Rescensibi!ities of Persennel
- a. ht.OI (1) The Director shall have responsibility for all activities in the reactor facility which may affect reactor operations or involve radiation hazards, including controlling the admission of personnel to the building. This responsibility shall encompass administrative control of all experiments being performed in the facility including those of outside agencies.
(2) It shall be the responsibility of the Director to insure that all proposed experiments, design modifications, or changes in ope rating and emergency procedures are performed in accordance with the license. Where uncertainty exists, the Director shall refer the decision to the Reactor Utilization Committee.
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. b. Fenior Reacter Orerators k
(1) A licensed senior reactor operator shall be assigned each shift and be responsible for all activities during his shift which may affect reactor operation or involve radiation hazards.
The reactor operators on duty shall be responsible directly to the senior' operator.
(2) The reactor operations which affect core reactivity shall not be performed without the senior operator on duty or readily available on call. The senior operator shall be pre'ent at the facility during initial startup and approach to power, recovery from an unplanned or unscheduled shutdown or significant reduction in power, and refueling. The name of the person serving as senior operator as well as the time he assumes the duty shall be entered in the
, reactor log. When the senior operator is l relieved, he shall turn the operation duties over to another licensed senior operator. In such instances, the change of duty shall be logged and shall be definite, clear, and explicit. The senior operator being relieved of his duty shall insure that all pertinent information is logged. The senior operator assuming duty shall check the log for information or instructions.
- c. Re a ct o r Oreraters (1) The responsible senior operator shall designate for his shift a licensed operator (he reaf ter called " operator") who shall have primary responsibility under the senior operator for the operation of the reactor and all associated control and safety devices, the proper functioning of which is essential to the safety
! of the reactor or personnel in the facility.
The operator shall be responsible directly to the senior operator.
(2) Only one operator shall have the above duty at any given time. Each operator shall enter in the reactor leg the date and time he assumed duty.
(3) When operations are performed which may affect core reactivity a licensed operator shall be stationed in the control room. When it is l
necessary for him to leave the control room during such an operation, he shall turn the reactor and the reactor controls over to a designated rellof, who shall also be a-licensed operator. In such instances, the change of duty
, shall be definite, clear, and explicit. The
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relief shall acknowledge his entry on duty by proper notation in the reactor log.
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1 (4) The operator, under the senior operator on duty, shall be responsible for the ope r at ion of the 1 reactor according to the approved operating schedule.
(5) The operator shall be authorized at any time to reduce the power of the reactor or to scram the reactor without reference to higher authority, when in his judgement such action appears advisable or necessary for the safety of the reactor, related equipment, or personnel. Any person working on the reactor bridge shall be similarly authorized .o scram the reactor by pressing a scram button located on the bridge, d, Health Physicist The Health Physicist shall be responsible for assuring that adequate radiation monitoring and control are in effect to prevent undue exposure of individuals to radiation.
4, written Instructions and Prrredures Detailed written operating instructions and procedures shall be prepared for all normal operations and maintenance and for emergencies. These procedures shall be reviewed and epproved by qualified personnel before use. Each member of the staff shall be familiar with those procedures and in st ruc t ior. s for which he has responsibility.
- 5. Sito Emarcenev Plans The Rhode Island Nuclear Science Center shall have available the services of other state agencies for dealing with certain types of emergencies. The RIAEC shall enter into an agreetaent with the Rhode Island Civil Defense Agency whereby the Civil Defense Agency will maintain an emergency monitoring and communications vehicle which they shall make available to the Nuclear Science Center in the event of an emergency involving release of fission products or other radioactive isotopes to the atmosphere. The emergency vehicle shall contain equipment such as portable radiation monitors, respirators, and a particulate air sampler. Communications using the statewide emergency network shall be available.
Personnel of the Civil Defense Agency and of local fire depr.rtments shall have received training from the Civil Defense Training Officer in tne use of certain radiological instruments. Future training shall be augmented by including orientation on the reactor facility.
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- y. . ;mATir:0 tiMIT AT Mlui i 1. Ccml The following administrative controls shall be employed to assure the safe operation of the facility:
- a. The reactor shall not be operated whenever there are any significant defects in fuel elements, control 4s, or control circuitry,
- c. The reactor contrcl and safety system must be turned on and functioning properly and an appropriate neutron source must be in the core during any change which can affect core reactivity,
- c. During operations which could affect core reactivity, a licensed operator shall t* stationed in the control v em.
Communications between the contr 1 room and the senior reactor operator directing the operation shall be maintained.
- d. The operator shall not attempt to start up the reactor following an automatic scram or unexplained power decrease until the senior operator has determined the cause of the scram or power decrease and has authorized a start-up.
- e. The reactivity of all core loadings to be utilized in operating the reactor shall be determined using unitradiated fuel elements or elements centaining fission products in which the effect of xenon poisoning on total O core reactivity has decayed to 0.05% delta k/k or less, f Critical experiments shall be performed under the supervision of the Director or other competent supervisory scientist licensed as a senior reactor operator. During the experiment there shall be present, in addition to this licensed supervisor, at least one other technically qualified person who shall act as an independent observer.
Each step in the procedure shall be considered in advance by both persons, each calculation shall be checked by both persons, and no step shall be taken without the n concurrence of both. A written record shall be made at the time of each fuel element addition or other core
-chance which could significantly affect core reactivity,
- g. The basic operating principles for the assembly and reloading of cores whose nuclear properties have been previously determined from critical experiments shall be as follows:
All core loading changes shall be performed under the supervision of a person having a senior operator's license. During the operation there shall be present in addition to tre designated Senior reactor operator at least one other technically qualified person whc shall act as an observer.
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The exact procedure to be followed for a particular reloading operation will be determined by the observer and the senior reactor operator in charge of the operation before the operation begins. Each step in the procedures shall be considered by both persons, and no step shall be taken without the concurrence of both.
- 2. E xpe r ire.n*J
- . " Experiments" as used in this section shall be construed as any apparatus or device installed in the core region which is not a component of the core.
- b. The Reactor Utiliration Committee shall review and approve all experiments before initial performance at the facility. New types of experiments or experiments of a type significantly different from those previously performed shall be described and documented for the study of the Reactor Utilization Cecmittee.
The documentation shall include at least:
(1) The purpose of the experiment, (2) A description of the experiment, and (3) An analysis of the possible hazards associated with the performance of the experiment.
- c. All use of experimental facilities shall be approved by the Director of Operations.
- d. The absolute value of the reactivity worth of any single independent experiment shall not exceed 0.006.
If such experiments are connected or otherwise related so that thej r combined reactivity could be added to the core simultaneously, their combined reactivity shall not exceed 0.006.
- e. The calculated reactivity worth of any single independent experiment not rigidly fixed in place shall not exceed 0.0008. If such experiments are connected or otherwiss related so that their combined reactivity could be added to the core simultaneously, their combined reactivity worth shall not exceed 0.0008.
- f. No experiment shall be installed in the reactor in such a manner that it could shadow the nuclear instrumentation system monitors and thereby give erroneous or unreliable intormation to the control system safety circuits.
l g. No experiment shall be installed in the reactor in l such a manner that it could fail so as to interfere with the insertion of a reactor control element.
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.. . h. No experiment shall be performed involving materials used in such a way t. hat they might credibly result in an explosion 4
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- 1. No experiment shall be performed involving materials which could credibly contaminate the reactor pool causing corrosive action on the reactor components.
- j. Experiments shall not be performed involving equipment
.whose failure could credibly result in fuel element darage,
- k. There shall be no more than one vacant fuel element position within the periphery of the active section of the core.
- 3. Operations
- a. Sit.e Control of access to the reactor facility shall be the responsibility of the Director of Operations, b, cenrainmant (1) During any operation in which the control rods are withdrawn from the core containing fuel, the following conditions shall be satisfied:
- a. Confinement building penetrations which are not designed and set to_ close I automatically on actuation of the evacuation button shall be sealed, except that doors'other than the truck door may be opened during reactor operation, If a door is to remain open, an individual from the reactor operations staff is continuously in attendance at the door.
- b. The building cleaa-up system.is operable.
(2) Re mli reman t s for Retest of Confinement (a) Methnd of Retent j The building cleanup system shall be L retested by pressing an evacuation button
- and observing that the following functions
, occur automatically:
- 1. Evacuation horn blows.
- 2. air conditioning and normal ventilation has turned off, L 3. Dampers on all ventilating ducts leading to the outside have closed.
p 4, Building cleanup system-air scrubber and basement chem lab blower come-on.
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- 5. The negative dif f e rer.t ial pressure between the inside and outside of the building is at least 0.5 inches of water. This shall be determined by reading the dif f e rer.tial magna helic gauge located in the control room.
(b) Frequenny of Roten The building cleanup system including the auxiliary electrical power system shall be retested at least weekly, tJ) The exhaust rate through the cleanup system shall not exceed 4500 cfm with not more than 1500 cim coming from the reactor building and passing through the charcoal scrubber. The remaining air will be provided by a separate blower from an u nc on t an.ina t e d scurce. This shall create a pressure in the building which is equivalent to at least 0.5 inch cf water below atmospheric pressure,
- c. Prim ry ceslant system (1) The minimum depth of water above the top of the active core shall be 23 feet.
(2) No piping shall be placed in the pool which could cause or fail so as to cause a siphon of the pool water to below the level of the ten inch coolant ine penetrations.
(3) Pakeup system The effluent water of the primary coolant water makeup system shall be of a quality to insure compliance with K . 3. c . ( 5 ) and (6) below.
(4) cleanun System The offluent water of the primary coolant water clean up system shall be of a quality to insure compliance with K . 3.c . ( 5) an1 (f) below.
(5) The primary coolant shall be sampled at a minimum frequency of ence per week and the samples analyzed for gross radioactivity, pH, and conductivity in accordance with written procedures. Corrective l action shall be taken to avoid exceeding the limits
! listed below:
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'[v pH 5.5 to 7.5 conductivity 2 mho/cm (6) The radioactive materials contained in the pool water and in the primary coolant water shall be such that the radiation level one meter above the surface of the p%; aball be less than 10 mrem /hr.
(7) During the forced circulation mode of operation, the primary coolant flow rate shall not be less than 1580 gpm.
d, facenderv Cooline Eyst m (1) The secondary coolant shall be sampled at a minimum frequency of once p6r week and the samples analyzed for pH in accordance with written procedures. Corrective action shall be taken to avoid exceeding the pH limit given below:
pH 5.5 to 9 (2) The concentration of radionuclides in the secondary water shall be determined at least once each day the reactor operates using forced convection cooling. The - concentration shall be determined at least once per week when not being operated using forced convection cooling.
(3) If the radioactive materials contained in the secondary coolant exceed e radionuclide concentration in excess of the values in 10 CFR 20, Appendix B, Table I, Column II, abeve background, the reactor shall be shutdown and the condition corrected before operation using the secondary cooling system resumes.
(4) The secondary coolant system shall be placed in operation as required during power operation utilizing forced convection in order to maintain a primary coolant core outlet temperature of 125cF or below.
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- e. Reacter Core and Centrol Elamarta-(1) The reactor shall not contain in excess of 35 fuel elements. There shall be a minimum of four operable control elements.
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(2) The limiting thermal and hydraulic core characteristics based on a 14 element, graphite and beryllium reflected core ate specified below:
(a) Ma.imum Heat Flux
. .424 MW/M2 (b) Maxi.num Core Specific Power 519.48 W/gU235 (c) Maximum Fuel Surface 11000 Temperature (d) Coolant Velocity during 1.46 M/sec Forced Convection Cooling (e) Coolant Inlet Temperature 1150F max.
(f) Average Coolant Temperature 10c-F max.
Rise (g) Primary System Bulk Outlet 1250F max.
Coolant Temperature (h) Temperature Margin in Primary 5.80C ll Coolant (Tsat-Tsurf)
(1) tiumber of Coolant Passes 1 Through Core (3) Prinrical nuclear characteristics cf the C; n (a) Core and Control System Reactivity Wor *h
- 1. The reactor shall _ . , subcritical by at least it ok/k from the cold, Xe-free, critical condition with the most reactive control element and the servo regulating element fully withdrawn.
- 2. The maximum worth of the servo regulating element shall be 0.7%
Ak/k.
tb) Maxi *"- Paartivity Addi*ien Rate - Ak/k/se;
- 1. By servo regulating element maximum of 0.0002
- 2. Manual by control element maximum of 0.0002 0
e (C) EqMetivity Ceefficients
- 1. Temperature coefficient approximately
,82 x 10-4/OC (calculated) density only
- 2. Void coefficient approximately-(core average) -2.1 x 10-3 /% void-(calculated)
(4) Prindpal Core Cperat ing Limitaticas (a) Mnximum 1 col 'reme r a t u r e ti rni t at ion s The pool water t empe r a t'tr e s ha l'1 not exceed 125oF. The pool water temp shall be ronitored with readout in the Control Room. A trip and alarm shall be included in the system.
(b) Reactivity Limitations
- 1. Excemn Fenctivity The cold, clean excess reactivity for any core used in the reactor shall not exceed 0.047,
- 2. Minimum Shutdown Margh All reactor cores used shall be such that they would be suberitical if any single centrol element and the servo regulating element were withdrawn.
(c) Reactivity Coefficient tim!tation The reactor power coefficient (as inferred by the control rod movements required to compensate for changes in power) shall be negative.
(d) Control Element r' rive' Performance Requirer m All control element drives shall meet the following specifications:
- 1. The control drive withdrawal rate shall not be more than 3.6 inches per minute.
- 2. For the electronic scram system, the time from initiation of a scram condition until i control element relesso shall not exceed 100 l milliseconds.
- 3. The time from initiation of a scram condition until the control element is fully inserted I shall not exceed 900 milliseconds.
- 4. It shall be demonstrated at least every 3 I months that the above specifications are met.
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e (e) feiVD Regj; M M 2 E 1 e me ri t D ii'Lg {'oThin;Q2 paguirementft
!i in use during operation, the servo r e g u l a t ir.g element drive shall meet the following specifacations:
- 1. The drive withdrawal rate shall not be more than 78 inches per minute,
- 2. It shall be demonstrated at least once per m nt h that the above specification is met.
- f. Reactor fafety fystomt (1) The reactor safety system shall be operable during all reactor operatien. The safety system rhall be checked out before each start-up and functionally tested for calibration at least monthly.
(2) It shall be permis*ible to continue operations with one or more of the safety system functions that produce only an alarm temporarily disabled providing that additional procedural contrels are instituted to replace the lost safety system alarm function (s),
(3) The control element withdrawal interlocks and the servo system control interlocks shall be functionally tested at least once per month, (4) During reactor startup or during mechanical changes that could affect core reactivity, the startup range neutron monitoring channel shall be cperable and shall provide a neutron count rate of at least 3 :ounts per second with a signal to noise ratio at l e a s t. 3 tc 1, (5) The linear level safety channels shall not read less than 15% of full scale when the reactor is operating at power -
levels above 1 watt, (6) Fcilowing a reduction in power level, the operator shall adjust the servo power schedule to the new power level before switching to automatic operation.
(7) An a.larm condition from any one of the items listed in SectiOn F.2,b, after working hours shall transmit coded information to a continuously manned central statien in Providence, Rhode Island. The central station shaL be provided with written instructions on the steps to be taken following an alarm.
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g, waste Disressi and Peact er M^ni t erino syst m (1) The liquid waste retention tank discharge shall flow to a monitor station in the reactor building where the effluent shall be batch sampled and the gross activity per unit volume determined before release. All off-site releases shall be directly into the municipal sewer system.
(2) Gaseous radioactive waste shall be disposed of using the reactor stack, Disposal limits shall conform to the tollowing table. In this table, the M?C stated is for individual isotopes and mixtures contained in Column 1, Table II, Appendix B of 10 CFR 20, 1 2 Type of Activity Maximum Curies Curies per second to per second to be be released averaged released over one year Particulate Matter and Halogens with half-lives 140 X MPC (uc/cc) ?4 X MPC (uc/cc) longer than 8 days All other Radioactive 105 X MPC (uc/cc) 104 X MPC (uc/cc[
Isotepes (3) All radioactive liquid and solid wastes disposed of off-site shall be within the limits established by 10 CFR 20 or shall be removed from the site by a coraercial licensed organization.
(4) The exhaust gas monitor shall be calibrated to alarm at an instantaneous release rate which instantaneously exceeds the limits stated in Column 2 for the annual average release rate. If the maximum permissible stack release rate stated in Column 1 is exceeded, the reactor shall immerliately be placed in the shutdown mode of operation and the situation investigated.
(5) The area, primary and secondary coolant system and the exhaust gas monitors shall be in operation at all times when control elements or the servo regulating elements are withdrawn; however, individual area coolant system monitcrs may be taken out.of service for maintenance and repair if replaced with portable radiation detection equipment. Adequate spare parts shall be on hand to allow necessary repairs to be made during the maintenance or calibration outages of the monitors, O
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C (6) The a:ea and the primary and secondary coolant system moniters shall be adjusted to alarm at a maximum reading of 2 mr/hr or 200% of the no rma l radiation levels in their area, whichever is larger, (7) The door which controls entrance to the " maze" leading to the delay tank shall be locked with the key in the possessicn of the Director or a licensed senicr operater. Entrance to the delay tank high radiation area shall require the presence of the Health Physicist or a licensed senior operator and the use of direct reading pcrtable radiation monitoring equipment.
- h. Fuel Storage (1) New fuel shall be stored in egg crate boxes located in l a security container. Access to the security r container shall be restricted, through use of a lock, l to the 04 rector of operations and the licensed senior reactor operators.
(2) Irradiated fuel, not in use in the reactor core, shall be stored in the criticality safe storage racks described in Section H. Only one fuel assembly may be inserted or moved from a storage rack at a t ime .
(3) Safety against inadvertent criticality shall be providec by limiting the number of fuel assemblies per rack t- nine and then positively securing such racks at l' 30 cm. apart, or by limiting the number of f _ssemblies to 18 per rack and then covering the l _ , large faces of each rack with a sheet of aluminum I covered cadmium.
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- 4. Maintenen e (a) The electronic control and the process control system shall be checked for proper operation and calibration before each reactor start-up. If maintenance or recalibration is required, I
it shall be perf ormed bef ore reactor start-up proceeds.
(b) Maintenance shall be performed with the approval of the Director. Equipment and system maintenance records shall be kept to facilitate scheduling and completion of all necessary maintenance.
(c) Routine maintenance on e?1 control and process system components aball be performed in accordance with written schedules and with written procedures.
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