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SAR for Low Enriched Fuel Conversion of Rhode Island Nuclear Science Ctr Research Reactor
	ML20086D214
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Site:	Rhode Island Atomic Energy Commission
Issue date:	11/30/1991
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p _1.g STALE OF RHODE l$1.AND AND PROVIDENCE PLANTATIONS

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RHODE ISL\ND ATOMIC ENERGY COhlNilSSION Ntulcar 5(ient e Center south Ferry Road Narraeansett. R I n;gs;.119;

" SAFETY ANALYSIS REPORT FOR THE LOW ENRICHED FUEL CONVERSION OF THE RHODE ISLAND NUCLEAR SCIENCE CENTER RESEARCH REACTOR" NOVEM BER, 1991 wn

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SAFETY ANALYSIS REPORT PART A LEU CO!1 VERSION ANALYSIS P AGE (S )

I Introduction 1-2 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

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Figure 3 10 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 Figure 5 19 X References for Beryllium Refleeftor Use 20-21 XI Design Basis Accident 22 XII Appendix A 23

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Part A LEU Conversion Analysis O INTRODUCTION This safety analysis report is submitted pursuant to 10CFR 50.64 which requires the Rhode Island Atomic Energy Commission to convert its open pool research reactor from the use of high enriched uranium (HEU) fuel to the use of low enriched (LEU) fuel. The studies required for the preparation of this report have been a joint project of the Reduced Enrichment for Research and Test Reactor (RERTR) group at Argonne National Laboratory -

and the staff at the Rhode Island Nuclear Science Center (RINSC). The Rhode Island Atomic Energy Commission is responsible for the contents of this report. .

The operating license for this reactor was issued on July a 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.

The reactor is multipurpose with capabili$1,es usually associated with open pool facilities. Because of staffing and funding limitations, utilization has concentrated in two areas-

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neutron scattering and neutron activation analysis. To meet the needs of the research programs, tne reactor operates one shift, five days per week. As of Septente r 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 follows:

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 cycle. This duty cycle allows for operation with an excess reactivity less than that O- required for continuous operation. Bece.use of the control blade 1

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configuration, this duty cycle also requires special start-up considerations when converting to an LEU core, i The studies performed for the LEU conversion have included l calculations for operation at power levels above 2 MW and for advanced core designs. This was done to insure that the conversion process did not compromise the future capabilities of the reactor. This safety analysis report however contains only information necet?ary for the initial conversion to LEU. 1 DESCRIPTION OF REACTOR SYSTEMS The reactor s" stem is described in the initial safety analysis report /1/. Only a synopsis of system components important to the convers'.on will b1 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 box 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 installed 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. The boral regulating rod, however, while fixed in the reflector region of the HEU core, can be relocated.

While some grid positions are shown vacant for clarity, during operation each grid position 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

r 0.508 mm ( 020 inches) and consists of 934 enriched uranium in a UA1x matrix. The clad is 0.508 mm (.020 inches) and consists of aluminum. The spacing between fuel plates is 2.54 mm (0.1 inches). When new, each plate contains 6.889 grams of Uranium 235 for a total Uranium 235 element weight of 124 gm.

The operating HEU core is made up of these fuel elements and consists of between 28 and 35 elements surrounded on four sides 'by graphite reflector pieces. This core may be characterized as large with a very low power density resulting in a low thermal flux per unit power. The lightly loaded fuel elements makes the core large enough to encompass the fixed control blades. Even with extra ordinary techniques, the maximum burn-up achievable is abcut 14%. Positions in the grid l plate not containing a fuel element or a reflector piece are filled with an irradiation basket. Figure 2 presents a typical 30 element HEU core.

Operation is also permissible with water reflection of the reactor. During the initial startup of the reactor, many l

measurements were made of the characteristics of a water i 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 tc 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 ,

thre- 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 thraugh tube, a graphite thermal column through which a hole has bein cut creating an additional radial beam tube, and- the termitals of two pneumatic irradiation (rabbit) systems. Between 1he graphite thermal column and the core is a permanently insttiled slab of lead serving as a l- thermal shield. The thermal s>teld'is cooled by water which is 3

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I currently forced around the shield using the pressure difference between the inlet and outlet primary coolant lines.

The control of make-up water to the pool is automatic using a float activated motorized valve. This system activates for a drcp in pcol 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 t he control and safety system. Although most of this system is the original equipment, it has been well maintained and is reliable.

Using a primary pump the core is cooled at 2 megawatts by downward flow of 0.109 m3 /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 accident. The controlled release is produced by a blower and is through HEPA and charcoal filters and a stack. The release creates a negative pressure differential between the atmosphere and the building insuring 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, control blade calibrations and the core thermal flux distribution were experimentally determined.

CONVERSION CRITERIA AND OBJECTIVES There are six basic criteria and objectives of the LEU i 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 . LF" core and an operating scheme to achieve tua. up greater than the current 141,

3. Design an LLU core which will optimize the thermal neutron flux in the beam tubes and will allow 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 approprir.te 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 amo mt allocated for the core by the g U. S. Department of Energy.

LEU NEUTRONIC CORE DESIGN The neutronic core design has been pe* formed using the standard fuel plate which the Department of Energy will provide for un *>ersity reactors. Figure 3 presents a compariso.. of this stcodard 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 iaportant 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.

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 reflectors

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 objactives, a LEU conversion design has emerged.

LEU CONVERSION CORE The LEU conversion core consists of a compact configuration using 22 standard plates per f~uel element and a combination of graphite and beryllium reflectors.

Figure 4 presents the startup version of the conversion 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 as to be adjacent to the compact core. The LEU fuel used is the uranium silicide-aluminum dispersion fuel approved for use by the NRC under NUREG-1313.

Table 3 presents reactivity data on this core. The core is graphite and beryllium reflected with an excess reactivity of 3.1% a regulating rod worth of 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 cyclic.,1 and this core can be operated as long as it is possible to vp3 rate on Friday morning. Using computer simulation, this core has ceen "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 l products, control, and the cold-hot swing is approximately 3%

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l 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 reactivity. Figure 4 also presents this second core showing the fuel remaining in each fuel element after the initial 14 weeks of operation. The reactivity balance is shown in Table 1 and it allows for an additional 70 weeks of operation.

Following 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 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 p la c..n g the remaining used fuel elements in the remaining positions with those elements containing the least fuel nearest the center of the core. This process provides the flatest 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 average discharge burn-up for this equilibrium core is about 211, which is 50% more burn-up than in the current HEU core. The LEU 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 Graphite Reflector Elements Approx. U 235 Loadings . g per Element MMMMMMM j / / // /)

108 111 114 114 111 111 F

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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 UAix-Al U 3 Si 2 -Al Cladding Material 1100 Al l 6061 A12 Fuel Meat Dimensions 3 thickness, mm 0.508 0.508 width, em 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 boron 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 2 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 O

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I Fig. 4. Startup, Transition, and Equilibrium Cores (Lifetimes Based on Operation f or 8 Hr/D,5 DiWK)

Cong2 sTARTUP CORE Core Ufetime: - 70 Wks (- 2800 Full Power Hours)

Core Ufetime: - 14 Wks (- 560 Full Power Hours) 2 1 1 3 3 2 , ,

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Table 1 Calculated Data and BOC Excess Reactivity for First Ten Cores Core Lifetime Accum. Operation BOC Express Weeks Weeks hm iAk/k Startup 14 14 0.3 3.0 Core 2 70 84 1.6 4.1 Core 3 60 144 2.8 3.7 Core 4 33 177 3.4 3.0 Core 5 51 228 4.4 3.6 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 Core 10 57 515 9.9 4.0 O

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TAllt.E 2 Reactivity llalances on the Friday Morning of the I.ast Week of Operation for Ten Cores from Startup to Equilibrium Reflector Changes Only 4 11urned Elements Removed and 4 Fresh clernents Added in Corners .

Startup Core 2 Core 3 Core 4 Core 5 Core 6 Core 7 Core M Core 9 Core 10

% Ak/k M Ak/k % Ak/k % Ak / k % A k /k '7e Ak/L % Ak/L.9 Ak/L q_JkkJ1 % AL/(

3.00 5.19 6.92 6.92 6.92 6.92 6.42 6.92 6.92 6.92 Fresh Cold Clean U

Reactivity Losses 0.30 1.85 3. I 7 3.08 3.09 3.12 3.07 3.06 3.06 3.06 11 u r n u p 1.54 1.54 1.54 1.54 1.54 1.54 1.54 1.54 1.54 1.54 Xe 0.57 0.73 0.73 0.73 0.73 0.73 0.73 0.73 0.73 0.73 Sm 0.09 0.57 0.97 1.06 1.03 1.03 1.07 1.06 1.07 1.07 Long-Lived F.P.

Cold-liot Swing 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 Control G (L20 m M G (L20 6.92 M

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6.M9 0.20 6.90 OL29 6.90 3.00 5.19 6.91 6.91 6.89 1

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Table 3 Reactivity Data and Power Peaking Factors Start-up Core Core 2 Core 3 Core 10 Ex. cess Reactivity 3.1 4.1 3.' 4.0 iAk/k 6.7 6.1 - 6.;

Shutdown Margin iak/k (Blade 3 stuck out)

Worth of Reg rod .45 .41 -

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%Ak/k Total Power Peaking 2.64/06 2.60/D6 -

2.36/D6 Factor / Grid Position (Control Blades Full Out)

Total Power Peaking 3.06/D6 3.05/D6 -

2.81/06 Factor / Grid Position (Control Blades 50% Inserted)

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l For the LEU cores, additional kinetic parameters and reactivity i l

coefficients were calculated by ANL. The comparisons are shown in l Table 4.

Table 4 )

HEU LEU LEU LEU Ref. Startup Transition Equilib.

h h core 2 core 10 Minetics Parametern Delayed Neutron 0.762 0.782 0.776 0.764 Fraction, 0-eff, i Prompt Neutron 76.3 66.2 66.0 68.3 Generation Time, ps Eeactivity coefficients: 20-40CC Change in Water Temperature Only tak/k x 10-4/oC Coolant -1.51 -0.80 -0.86 -0.89 Change in Water Density Only

%Ak/k x 10-4/oC h Coolr7t -0.44 -0,R? -0,75 -0,69 Coolant Coeff.,

1Ak/k x 10-4/oC Coolant -2.0 -1.6 -1,6 -1.6 Doppler Coeff.,

%Ak/k x 10-4/oC Fuel 2.2 -0.18 -0.18 -0.18 Temperature Coeff*,

iAk/k x 10-4/oC -2.0 -1.8 -1.8 -1.8 1Ak/k/% Void 0.0015 0.0027 0.0025 0.0023

Fuel and coolant temperature changes were assumed to be the same here. The fuel temperature rise will be larger than that of the l

coolant. Change in Reactivity = (Coolant Coefficient) x ATcoolant l + (Doppler Coefficient) x ATruel-i l

l It can be seen that there are no significant differences between the HEU and LEU kinetic and reactivity parameters, m

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A START-UP ACCIDENT g

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.

ii. The servo . egulating rod is withdrawn, followed by continuous withdrawal of all safety blades in succession at their maximum rate, iii. Period scram protection fails.

iv. The reactor is scrammed by the high flux sensor instrumentation when the power level reaches 2.4 MW (20%

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 channel) reaches 67.3oC (1530F) for the HEU fuel and 88. loc (191or) for the LEU fuel. Thus, it can be concluded that this accident results in no harm to the reactor.

If assumption "111" is modified to - " period an'd high flux scram protection fails" - then recctor 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 tne 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 (1080 F) nelting 16

temperature of 6061 cladding. The core in each case would operate in the nucleate boiling range without physical damage until the h

accident could be terminated by a manual scram.

Table 5 Peak Power and Cladding Temperatures Peak Cladding M Peak Pcwer. 5"N Temperatura. 20 HEU Equilibrium 32.1 149.1 LEU Startup 14.9 148.3 LEU Equilibrium 16.2 1

48.5 REFERENCES

(1) Atomic Energy Commission, Facility License No R-95, Docket No. 50-193, July 21, 1964 and Construction Permit No. CPRR-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, September, in press (3) Obenchain, C.F., "PARET - A Program for the Analysis of Reactor Transients"" 1D0-17282 (1969) i 1

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REPLACEMENT REGULATING ROD The current regulating rod in the HEU core is located in the D-1 grid position (refer to Figure 2 in the " Description of the Reactor System" section of this SAR). The rod is fabricated from beral plate and aluminum.

Calculations from ANL(1) indicate that the regulating rod worth in its present core position (D-1) is reduced irom its present value of .48% Ak/k to .2 .3% Ak/k which is too little. If the present rod were to be relocated to the D-2 position, it f- increases to .8 .9% which is too large. The regulating rod. limit l by :echnical specification 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 p regulating rod as part of the core arrangement.

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l (1) Memo from James Matos, ANL to RINSC, Eugene Spring, l

September 16, 1991 P

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?[ USE OF BERYLLIUM REFLECTORS IN THE RINSC-LEU CORE

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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 o2) has been using Beryllium and has conducted tests to determine a lifetime limit based upon a fast fluence level. Reference d0) in.iica t e s that embrittlement of Beryllium is first noted often approximately 3x1021 NVT. As a result of HFIR cetermination of small cracks occurring at a 1. bx10 f 2?! NVT level, a proposed changeout level of 1x 1021 is proposed. Using a maximum flux of 3.3x1013 and a 5 day, 7 hrs / day reactor operating cycle, a l 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.c2:m m m m m Gamma heating has been reviewed") G 3' 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.Mll Final drawings are due shortly. A 1

standard element is proposed similar to the graphite elements. A l

l 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 calculated 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 RINSt would require a technical specification change for use of Beryllium reflectors.

Table 5 shows the average midplane flux in the Be of the central flux trap.

I Figure 5 l

l Grp Upper Lower l Un, Energy Energy Startup Core 2 Core 10 1 14.0 MeV 0.821 MeV 2. 5 x1013 2.4x1013 2.4x1013 2 0.821 MeV 5.531 MeV 3 . 3 x1013 3.3x1013 3.2x1013 3 5.531 kev 1.855 eV 2.7x1013 2 . 7 x1013 2.7x1013 4 1.855 eV 0.625 eV 3.7x1012 3 ,7 x i o 12 3.6x1012

'C- 5 0.625 eV 0.251 eV 3.3x1012 3.3x1012 3.3x1012 6 0.251 eV 0.057 eV 1.8x1013 1.8x1013 1.9x1013 7 'O.057 eV 0.00025 eV 2 . 5 x1013 2 . 5 x 1013 2 . 6 x1013 19

REFERENCES FOR BERYLLIUM REFLECTOR USE (1) ASME, 74-PUP-44, " Stress and Deformation Analysis or Irridation Induced Swelling" by B. V. Winkel, 1974

-(2) " Surveillance Testing and Property Evaluation of Beryllium in Test Reactors", J.M. Beeston, M.R. Martin, C.R. Brinkman, G,E. Korth, and W.C. Francis, Aerojet Nuclear Company, Idaho 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. Beeston, EG&G Idaho, October 1976 (Q

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(5) Missouri University TM-ERS-62-1, MOU-30203, June 22, 1962, l

i

" Stress and Thermal Analysis of the Beryllium Reflector for l- the University of Missouri Reactor" 1

1 l (6) General Electric. Co. Atomic Power Equipment Department Standard 788,-" Beryllium, Hot Pressed, Nuclear Grade" l

l (7) "The Effects of Neutron Irradiation on Beryllium Metal", B.S.

Hickman, The Institute of Metals Conference on the Metallurgy i

of Beryllium, October 1961 l

l l (8) "The Effect of High-Temperature Reactor Irradiation on Some

, Physical'and Mechanical Properties of Beryllium, J.R. Weir, l.

j The Institute of Metals, Conference on the Metallurgy of l Beryllium, October 1961 l

t (9) "The Behavior of -Irradiated Beryllium, R.S. Barnes, The Institute of Metals, Conference on the Metallurgy of i G Beryllium, October 1961 O

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(10) University of Missouri, Inter-Department Correspondence, g Gerald Schapper; Beryllium Reflector Changeout, December 16, 1975 (11) EG&G 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 Beryllium Reflector, Drawing No. 193, October 6, 1988 (13) FAX memo from Argonne National Laboratory to Rhode I31and Nuclear Science Center, W.L. Woodruff to Eugene Spring;

Subject:

Gamma Heating in Beryllium Reflectors, March 19, 1991 (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 O

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DESIGN BASIS ACCIDENT The design basis accident for this reactor has been a loss of coolant accident with the water draining through a beam port containing no plugs. Recall 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 which the bottom 21 cm of fuel remains in water. It has been possible to show that the low power density HEU core will not melt after this hypothetical loss of coolant accident.

The LEU core has a higher power density than the HEU core, Using the same accident sequence and calculations which were used for the HEU core, it is not possible to conclude that the LEU core will not suffer some melting following a loss of coolant accident.

The LOCA assumes 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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#* i APPENDIX A LEU FUEL SPECIFICATIONS AND DRAWINGS EG&G IDAHO INC.

(A) TRTR-6 Specification for Test hesearch Training Reactor LEU Silicide U 3 Si; Fuel Plates Rev. 4, 20 May 1988 (B) TRTR-11 Specification for Low Enriched U Metal for Reactor Fuel Plates Rev. 1, 1 April 1937 (C) TRTR-14 Specification for Reactor Grade Uranium Silicide U Si2 3 Powder Rev. 2, 1 July 1987 (D) TRTR-15 Specification for Aluminum Powder for Fuel Plate Core Matrix Pev. 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 Training Reactor 5 Fuel Element Assembly No. 411650 0

23

I q_ SAFETY ANALYSIS REPORT v

PART B THERMAL HYDRAULIC ANALYSIS P AGE (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 Rhode Island Nuclear Science Center Water Supply 15-16 IX Loss of Coolant Analysis 17

() X Emergency Core Cooling System Operation 18 V

XI Water Supply Analysis 19-20 XII Appendices Appendix A/ LEU Thermal Conductivity Calculation 21-22 Appendix B/ Critical Velocity for Fuel Plate 23-24 Deformation Appendix C/ Loss of Coolant 25-27

. Appendix D/ Decay Heat Calculations 28-32 Appendix E/ Maximum Heat Flux 33 Appendix F/ Maximum Core Specific Power 34 b>

u

1 e Part B Thermal Hydraulic Analysis

's INTRODUCTION The thermal hydraulic studies for the LEU core have been a joint effort by the Rhode Island Nuclear Science Center (RlNSC) 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 PROGRAMS USED IN THE THERMAL HYDRAULIC ANALYSIS

,m - The computer programs used by the Rhode Island Nuclear

, Science Center Steady-State Analysis, Hot Channel Analysis and Natura] 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 utili::e in-house computers .

The program entitled "PLTEMP" was-used to perform the

" Steady-State" and single Hot Channel Analyses.

The program entitled "NATCON" was use to perform the Natural Convection Calculations.

l l

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p LEU PARAMETERS G

The parameters used in the "PLTEMP" and "NATCON" 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 from 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.

CORE COVDONENT DRAWING NUunpp t

LEU Fuel Plate EG&G ~#411650 Radiation Basket (w/ orifice plate) GE #7980413 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 i

LJ 2

l l

HOT SPOT FACTORS O

The use of the LEU fuel element necessitated an evaluation of the engineering hot spot factors to be used in the single hot channel analysis. The Rhode Island Nuclear Science Center has prepared a report entitled " Report on the Determination of Hot Spot Factors for the Rhode Island Nuclear Science Research Using LEU Fuel." The report was performed in August 1989. The results are shown below:

Fb (Bulk Water Temperature Rise) = 1.62 Fq (Heat Flux) = 1.46 Ph (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.

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3

.. =,

4 e .

- STEADY STATE FULL CORE ANALYSIS The PLTEMP Program was used to analyze 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 elements, reflectors, baskets etc.) as shown in Figure 4. This analysis 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 plate surface temperature vs flow rate is shown in Table B. It is 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 ) 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 operation case is shown in Table C. The axial peaking factor vs. relative blade position for the core is tabulated for both the .!ades 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 (l) (see Table D). This data was input to the PLTEMP Program to calculate 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

6 -

REFERENCES (1) Memo from Bill Woodruff, Argonne National Laboratory to Eugene Spring, Rhode Island Nuclear Science Center, 1/30/91 (2) Memo from James Matos, Argonne National Laboratory to A.

F. DiMeglio, Rhode Island Nuclear Science Center, 1/22/91 O

l 1

0 3

l LEU FUI.L CORE ANALYSIS - 14' ELEMENT' CORE-2 MW .

TABIE A DP (MPA) Core Flow Hy Pass Total Flow Core Flow Hy-Pass Total I' low (k g/s) Flow . (kg/s) (GPM) (GPM) (GPM)

(kg/s) i .0025 43.81- 25.49 69.30 698.00 407.00 1105 l

.0030 48.56 28.15 76.71 774.00 449.00 1223

.0035 52.98 30.63 83.61 844.5 488.50 1333

}

.0040 57.14 33.66 90.80 910.8 536.20 1447

.0045 61.07 35.13 - 96.20 973.5 559.50 1533 ,

.0050 64.83 37.21 102.04 1033.4 593.60 1627

.0055 68.43 39.19 107.62 1090.4 624.20 1715

.0060 71.89 41.09 I I 2.98 1146.00 654.00 1800

.0065 75.23 24.77 ~ l 18. I 6 1190.00 684.00 1883 TAHIE H Ap fly -Pa ss Core Flow Total Flow Outict Hulk - Plate Plate Flow { Outlet Hulk (MPA)

Temp 0C Surf act Temp 0C Surface !

(GPM) (GPM) (GPM) Temp OC Temp OC F axial =1.32 F axial =1.32 F axial =1.536 F axial =1.536

.0025 407.0 698.0 1105 54.70 74.06 54.70 77.52 '

.0030 449.0 774.0 1223 53.49 71.49 53.44 74.71

.0035 488.5 844.5 1333 52.56 69.69 - 52.56 72.49 l

.0040 536.2 910.8 1447 51.81 68.03 51.81 70.69 r

.0045 559.5 973.5 1533 51.20 66.65 51.20 69.I8

.0050 593.6 1033.4 1627 50.69 65.29 50.69 67.89 '!

.0055 624.2 1090.8 1715 50.25 64.28 50.25 66.78

.0060 654.0 1146.0 1800 49.86 63.56 49.86 65.80

.0065 684.0 1199.0 1883 49.53 62.77 49.53 64.94

.0070 719.0 1250.0 1969 49.23 62.07 49.23 64.I6 NOII5: (1) Normal Primary Pump Operation 1730 GPM (2) Calculations Based on Inlet Temp. to Core of 42.30C

. . O O O - - - -

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TABLE C The PLTEMP full core analysis for each fuel element calculates additional parameters, A typical output value is shown for these parameters. (Element #8 data)

PARI 4!ETER VALUE Maximum-Surface Temp. oC 64.45 Clad Temp cC 64.78 Peak Axial Heat Flux 2 .156 (MW/M )

Channel Flow Rate (kg/s) .2298 Velocity (M/S) 1.605 .

Outlet Pressure (MPA) .1727 2

CHF (Critical Heat Flux) (MW/M ) 2.39 Flow Instability (MW/M2 ) .889 Exit Saturation Heat Flux 2 1,og7 l (MW/M )

{ Minimum DNB Ratio 8.461

! F axial 1.32 O

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TAlllE D AXIAL, DISTRIBlTrION INTliRI ACE Rlil.AllVE DISTANCE IILADE-OITF BLADE-IN 50%

l 0.00 4105 .0553 j 2 0.05 .5506 .2682 3 0.10 .6836 4759 4 0.I5 .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 !.4032 10 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.I850 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 .5306 .2682 21 I.00 4105 .0553 l

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SINGLE CHANNEL (HOT CHANNEL) ANALYSIS Computer runs using PLTEMP were run for the single channel analysis using the derived hot channel factors. Flow and power levels were varied to provide sufficient

r. f r ation for " limiting power level and core flow nations.

The tabulated results for axial factors of 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 i.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 v,1ues are more restrictive than the present trip ,

levels of 1301 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 effeet than the increase in number of fuel plates from 18 to 22.

The maximum surface temperature of the fuel resulting at the 1580 GPM pump _ flow is from Tabel E about 1100C.

The- corresponding coolant velocity from the PLTEMP output for 1533 GPM ( AP=. 004 5) = 1.44 M/S and for 1627 GPM (AP=.0050) = 1.53 M/S. An extrapc %t.ed va '.ue for the 1580 GPM condition is about 1.48 M/S.

These safety settings will require a technical specification change upon URC approval, n

TAltLE E IIOT CII ANNEL ANALYSIS F AXIAL = 1.536 Ap Total l low T Surface T Surface T Surface T Surface T SAT. T onb 4 y pg) (&W 2 2.2 2.4 2.6 MW MW MW MW

.0025 1105 122.26 122.57 122.86 123.10 115.32 122.1

.0030 1223 122.27 122.57 122.87 123.15 '

115 R2 122.1

.0015 1333 119.27 122.5R 122.RR 123.16 115.82 122.1 0040 1447 114.99 121 60 122R4 123.17 115 S2 122.1

.0045 1533 111.44 117.71 122.90 123.19 115.82 122.1

.0050 1627 108.40 114.41 120.36 123.20 115.82 122.1

.0055 1715 105.14 111.54 117.24 122.93 115.K2 122.1 0060 1K00 103 39 109.00 114.53 120.00 115.82 122.1

.0065 18K3 101.30 106.75 112.10 117.39 115.82 , 122.1

.0070 1963 49 42 104.71 109.91 115.06 II5 82 l 122.1 IlOT CIIANNEL ANALYSIS F AXIAL = 1.32 Ap Total Flow T Surface T Surface T Surface T Surface T SAT. T onb (MPA) (GPM) 2 2.2 2.4 2.6 oc oC MW MW MW MW 0025 1105 120.80 122.09 122.36 122.62 115.82 122.1 0030 1223 115 01 121.52 122.37 122.63 115.R2 122.1

.0035 1333 110.16 116.52 122.1R 122 f*1 115.82 122.1 0040 1447 106.55 112.40 118.16 122.65 115.82 122.1

.0045 1533 103.3R 108 44 114.44 11936 115.82 122.1 j 0050 1627 100.67 105 49 111.26 116 4K 115 82 122.1 0055 1715 4R 30 103.41 108.49 113.51 115 R2 122.1

.0060 iR00 46 21 101.18 106 07 110 40 115.82 122.1 0065 ISR3 ud. 36 49.17 101.01 10R.5% 115.82 122.1

. .0070 1063 42 69 47.36 101.47 106 51 115 R2 122.1

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I 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 Cperation).

The Matural Convection Analysis for the LEU core was performed using the NATCCN Program. The program was run for both the regclar 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 nataral cor t'e c t i on power level of 100 kw.

O 13

TAllLE F NATURAL CONVE(TION REGUI.AR CII ANNEL FAXIAL = 1.536 I

Power Level Exit Maximum Incipient T sat-T wall Radial Margin to (kw) Te n. p. W all Boiling "C (I l 7.34-TW) Peakin g incipient Boiling Temp. "c Factor oc 10 51.93 52.01 117.57 65.34 2 65.56 100 68.41 71.22 11R.28 46.12 2 46.06 200 77.29 R4.71 118.88 32.63 2 34.17 300 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 l 119.M9 119.89 -2.55 2 0 00 _

IlOT Cil ANhT1 10 59.18 58.89 117.72 58.45 2 5R.83 ,

4 100 86.63 92.X1 iiR.56 24.53 2 25.75 209I 102.I1 119.34 119.20 -2.00 2 -0.40 NATURAL CONVECTION RiiGULAR CilANNEL FAXIAI, = 1.32 Power 1.cvel Exit Maximum Incipient T sat-T wall Radial Margin to (kw) Temp. 0c Wall Boiling "C "C Pe a k i n g incipient Boiling Temp. "c Factor og 10 523kt 52.26 117.60 65.08 2 65.34 100 68.72 ! 71.5R 118.19 45.76 2 46.61 200 77.69 K4.25 118.72 33.09 2 34.47 300 84.26 95.11 119.00 22.23 2 23.89 400 89.61 105.28 119.35 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 I10r CilANNil 10 59.3X 59.80 117.65 57.54 2 57.X5 100 87.12 92.92 118.43 24.42 2 25.51 217.3 103.66 119.14 119.03 - 1.X O 2 O txi 14

. . O O O

RHODE ISLAND NUCLEAR SCIENCE CENTER WATER SUPPLY The Wakefield Water Company supplies water to the University of Rhode Island (URI) Narragansett Bay Campus.

The water is pumped to the 300,000 gallon water storage via a 8" feedline. Water is then distributed to ths. URI Bay Campus (including the Rhode Island !Juclear Science Center (RI!!SC) )

through a 12" main. Water is supplied to the RI!JSC through an 8" line which feeds fire protection (6" line) and potable supply (2" line) and a reactor building fire hose (4" line).

The entire pumping system has backup generators for total 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , even in the event of a power failure.

The Bay Campus can provide uninterrupted water supply in the event of a line rupture or planned shutdown by utilizing various cross connections and hydrant connections located in the system network.

LO 16 l -- . . . _ _ . _ . . _, . . - . ,, _ - - - -

" i ine un.,eesity of nnoce mano creavate scnavi at oceanogoir,

.h . s 'irra;r u" a # at 19 7 g ' r a ;2 s e" d a .- Oi +:

O July 12, 1991 Mr. Eugene F. Spring, Sr. Reactor Facility Engineer

'luelear Science Center 30uth 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.

l Presstire (PSI) Volume (Gal)

Condition w/o Fire Pump Available Duration Normal Town supply 4eservoir 800 ster Pumps 70 *

Reservoir With Booster Pumps 70 300,000 24 Hrs, Town Supply With Booster Pumps 70 *

_ Town Supply Only 30 * *

= Unlimited within present demand Campus (Max) 200 GPM Reactor (Min) 5 GPf1 Total 205 GPM m_,_ g///

Kenneth W. Morrill l Asst. Dir. Physical Plant l

O L .

l 16

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LOSS OF COOLANT ANALYSIS g rollowing a postulated loss of coolant accident, the pool drain time is calculated by using the falling head calculational method.(1) This is considered the maximum credible accident (refer to SAR Part A-Section XI) . It is assumed that water drajns from the beam tube (S") from the pool surface (el.139.417) to the p of the core box (115.916). Water then drains from the 1/2" diameter hole in the core box. Water cannot drain below tne 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 arsuming l

that the beam port shutter is up, no plugs are in the tube, and no l I

flange cover bolted over the outlet flange.

The original RINSC license amendment for 2 MW operation calculated the decay generation and heat removal 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 an emergency core cooling system is necessary, The design and operation of such a system is discussed in Section X.

(1) Handbook of Hydraulics, Ernest F. Brater, 6th Edition, McGraw-Hill Book Company, 1976 O

o l

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_______.r.- - _ _ . ~ . . _ _ . _ _ _ _ _ ._ _. . . _ _ _ _ _ _ _ _ _ _ _ . _ .

EMERGENCY CORE COOLING SYSTEM OPERATION i

l l

Under normal operating conditions, make-up water is i 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 the procedure for piping and valve alignment and procedures for normal and emergency filling of the pool. An emeigency 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 at the secretary's " desk alarm" box which notifies 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.

Modifications 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 obaerved in the make-up . system to be about 25 GPM. (Tests will be T 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/CA program.

O i

18 T+-vg ar e r* ---ger

_ err'ww--r- "M'-wm =-ep yT-Seggw <-ar re-v- + = - y 4$ www"-- e--Dyw r- eWew'- 'Wup'1 *se-w e- w:- -aT+"qr-w ge*P2*T-t'mr'u~79"*'" -fre^r#'"Fe**'tT

WATER SUPPLY ANALYSIS O

The analysis of the facility water supply syster was performed using a computer program called " Service Sizer",n' It calculates pipe size and demand. The program has built in piping tables, valve and fitting tables and fixture unit tables. Standard computation techniques are used to I determine losses. The program allows any fixture to be specified in either a public or private use situation. 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 and 1

fittings and also a permitted velocity. The program calculates pipe size, actual velocity, head loss and demand pressure.

The demand calculation includes options for fluthometer units; public use or private use. Input for the calculation includes numbers and types of fixtures, a continuous demand flow and additional fixture option. The program calculates total fixture units, continuous and fixture demand and total demand.

The enclosed report shows both the demand and calculated supply size for a proposed water line extension to insure an adequate supply to the reactor core in case of a LOCA.

The report shows that a 2" line will produce 42 GPM.

This line size is adequate for normal reactor pool supply and certainly for a 5 GPM supply in a LOCA situation.

(1) Parkcon, Inc., 250 N. Center Street, P. O. Box 5980, Woodland Park, Colorado 8086-5980 19

-- i . c I N G C A LCU LAT I C N-- --- - --------- ----- --- - - - - -- --- ---- P r i n t e d On : 7/18/1991 Supply 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 lower than supply) h; Pipe Length: 142.0 ft Other Loss In Equivalent Pipe Length: ft i

Number of Valvee & 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 Elbew : : :

Backflow Prev: 1.0 psi Water Meter: psi PRV: psi Other: psi

--Design Calculation----------------------------------------------------------

Permitted Velocity: -

1ps 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----------------------------------------------------------

Predominantly Flushometers: N Public Use:'N i --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.Pedest 2: Urinal Wa'1 . : Urinal Tank

Wash Sink :WC Flushometr 2:WC Tank :

~

l Additional: fixture units Total: 23.0 fixture' units

)

Continuous Demand: 25.0 gpm Fixture Demand: 17.0 gpm

( Total Demand: 42.0 gpm J

N

. 20

r l

)

l APPENDIX A l

LEU THERMAL CONDUCTIV.ITY CALCULATION Denrity of U 3212 The densities of the dispersants 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/"- (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%

O 11.28 From reference (3), page 11 Vp = .072 Vf - .275 y2 + 1,32 Vf3 therefore Vp = .072(.3068) -

. 275 ( .3068) 2 + 1.32 ( .3068) 3 = .0343 where Vp and Vf are volume fractions or porosity and fuel in the meat, respectively.

Thermal Conductivity of 03E12 Volume fraction of fuel plus voids = .3068 + .0343 = .3411 the thermal conductivity is obtained from Figure 6, page 16 of reference (3)

K = 88 W/m.k O

21

, o .

REFERENCES (1) R.F. Domagala, T.D. Wiencek, and H.R. Tresh, "Some Properties of U-Si Alloys in the Composition Range U3Si to U3S12," CCNF-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 l

O 22

i t f

l r . l i

l i !

] f APPENDIX D .

l 2

I r

CRITICAL VELOCITY FOR FUEL PLA'IE DEFORMATION i I

It has been shown that a critical flow velocity exists

] for a given plate assembly. m At this critical velocity, the plate becomes unstable and large deflections of the plate can f l

occur. These plate deflections can cause local overheating l l of the fuel plates and possibly a complete blockage of the f

! coolant flow, i Miller'2) derived a formula for the critical velocity f l based on the interaction between the changes in channel l

! cross-sectional areas, coolant velocities, and pressures in j j two adjacent channels. For design purposes, reference (3) l

\ .,

l and (4) recommends that the coolant velocity be limited to {

i 2/3 of the critical velocity given by Miller, W therefore for a flat plate; !

l !

V Critical = 2/3 n x ' 0i E f ty - ty) , tu):/2 l p . W.4 (1 -y ? ) ,

where '

E = Young's Modulus of elasticity, bar : 10.4 x 106 psi /14.5 (A1) j j tp = Fuel plate thickness, cm : .127

tm - Fuel meat thickness, cm: .0508 p = Density of water, kg/m3 : 1000.0 l i

tw = Water chLnnel thickness, cm, : .381 i i

l w = Fuel plate width, em : 6.096

-l 7 = Poisson's ratio, dimensionless W : .3 (A1) l !

I r l- :

V Critical = 16.6 m/sec !

i i The average core velocity of the 14 element LEU core calculated for l j the normal primary pump flow of 1730 gal / min ( .109 m3/sec) is about !

f 1.6 m/sec. For a projected 5 MW core, the velocity would be !

j increased to about 4 m/sec. This is well below the limiting value ;

l of 16.6 and therefore is not a problem in the proposed RINSC core. j

\

73 i

N.____ _ __ . . - . _ . - - - - - - - - - - - - - - - - - - - - - - 2

s i

REFERENCES (1) International Atomic Energy 1sgency, Research Reactor Core Conversion frcm the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium ruels. Guidebook IAEA-TECDOC-233 (1950)

! (2) Miller, D.R., " Critical Velocities for Collapse of Reactor Parallel Plate Fuel Assemblies", Trar's ASME, J. Eng,for Power, 62,83 (1960)

I l

(3) Mishima, K. and Shibata, T. "Thermall-Hydraulic Calculations ,

for KUHER with Reduced Enrichment Uranium Fuel, "KURRI-TR-223 i (1982)

(4) S. McLain and J. H. Martens, Reactor Handbook, Vol. IV, Interscience Publishers (1964)

O (5) Standard Handbook for Mechanical Engineers, Baumeister and Marks, 7th Edition, 1967 pg. 5-6 i

O 24

~ _ _ _ _ . . . . - . . . - _ . _ _ , . - . . _ . _ _ . _ _ . _ , . . _ . . _ _ _ _ _ . . . . _ , . _ .

. _ . ~ _ _ _ . _ . __ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ . -- .____.__ -

! i

. . l i

APPENDIX C l LOSS OF COOLA!!T. !

i

'^^^' ' " ^ ~ - ~ ~*~ ^ " - ~ % *

, y qry n ; I l !

i ! i I I

! es ne ro - s eo ,

l i

t i a; - !

q ---- u , u t .11%

ccu ___ t.w . % n t. ;

I e**

gL n to

- ---- {}4 et.....

l

.= .

( g ,.

/s *,6c 5"RFACE 7J.EAS (FREE FLOW AREA) 9 l

! Area of entire pool surface : 150 f* 2 l Area of core box : 5.06 ft2 Area of core (loaded) : .917 ft2 Area of 1/2 diameter hole in core box : .00136 ft2 '

E Area of 8" pipe : .349 ft i

The data elevation of 114.13 is used due to the assumption that l l water will not drain below this elevation in the event of shear of the 8" beam tube. l l

t The amount of wcter remaining in the core box after draining =

114.13 - 113.213 = .917' AsseMcTION:

Gravity draining of pool from pool surfaces to the top of the core box out the 8" beam tube which has no plug in place and the shutter in the up position. There is no cover flange.

O 25

I ccMPUTATicMAL METHcD: g Discharge under falling head (l) t = 2A ((H);'?

1 - ( H; ) '2 )

Ca ( 2 g ) 1/2 Datum is el.114.13 (invert of bottcm of 8" beam tube)

H1 - 139.417 - 25.287

!!2 = 115.713 - 114.13 - 1.583 C= .6 A- 150 ft2 a= .349 ft2 t- 2x150 [ (25.2871/2-(1.583)1/2 ) = 67 3.15 sec 6x.349 ( 2 x 3 2 . 2 . 2 ) 1/ 2 -11.219 min train Tire of core Box (loaded with components) t = 2x.917 [ (H )l/2-(H 2)1/2) 1 H1

- 115.713 - 114.130 - 1.583 g II: - 114.130 - 114.130 = 0 t2= 335.6 sec/60 = 5.59 min TOTAL TIME = 16.8 min Minimum required water ficw rate to keep the core box full From (1) a F= . 61 A (2g H) l/2 where H = 115.713-114.13 = 1.583 F= .61 x .00136 (64.4 x 1.583)1/2 P= .61 x .00136 x 10.097 = .008376 cfs r = .008376 h3 x 7. 4 8 gal x 60 s ec = 3. 7 6 gpm2 sec ft3 ,

(this assumes core is full)

O 26

- ~ . . .-_ - - - . . _ - . _ ~ ~ - - - - ~ ~ - . - . - - - - - - - - , ~ -

e

e .* e m er * ..st Dischstge under Failing Heed, ligure 44 shou t s t ryel - . .. , p.)@ d.e, gt',jf T T*-1-i%.2g Mied w it h w ster to a depth A.. The time required to icur tr.e i e .c ns @

e ster surinte to a depth A,is required a is the sets Of crifice. A

.. y)r.guAMu4//g havnl.y

- IAC'08Ml[N id A is the area of w6ter surfare for a derth y. C in t he ~ j yhy.: f, { pyre, A cefficient of docharge The incremerit of tiene di tequired to lower the n ater the inEtitesim61 t

{ ,g } ,, '?ppy/gl{9

, y ,4 f , ., p .g .g g dk%t6nre av zay 4~ p 6

,m

~~='*e ~'~ '4 dy mandy,, W q7~gI'0 3 4Vign dl = -

L4-15)

Ca s'.'ry m614).} '

e

_ [e r mmm 1.?u*Re D**Y *' --~Jm : 3 fgh) v ,3Rm Ytom ( bl5 '. sI A ca0 be en reued ^ ^ ]: v

0 99 U9A : y::' '. 9 y,c m;p l7 ~

s y

in terms isi y. by antegratics be-tween linuts A 6nd A,. the time

, y r.Ac.pya@

-i ; .pe, A +,,urra g 3.p.,pgj,.,gg w g. ,j ,, ,,g T

I needed to nomer the meter surface

} -]c'Wg'.cMede'

. * - - 3 *B h h m,teec5B t he distene, b - A, cio be gottex u.e 4 Mp=

Plactr e A, = 0 gives the time of empt3 :ng the veuel. Equation

- %0 . 1-.dj W

] b) ml (4-15) apphes to horisontal or in. ,, p, , ,g g .9 , , %,

flo. 44. Dischstge under clined on6ces provided the unter surface does Det fell below the top p r .p tlpph gg _ g __] f . y.g falling 1.ead, - q.-- a : vf;, n f As.e d?tAj@ 'lj'~

J re g @

<.f the nrifit e. For a r> Hn. lor or remn with vertirrd ins.1 in 5 d' A ***"'# I h

_..r! {._ h_ .y a nnatant, and Ih. ( t.1%) nf ter u.tegrat...n, hernmm d '.*#" U"#'# .d- g. - -f*ih @-.. f.

4 N

t a.

ca iu s jg (d - f) ( 1. ps) u

[ .-

lh a::4 Arend reonred (C) (9) ***"*#*O Orince Coemclents. One of ihr rathni nperin enters on

+hnrp. edge.1 nr6. n a. linnohnu sn n h, n,a ui,vaine,ut m,., ,j p,f #

"*. g, .,

._f (g 2zl the euciljent of shwharge f..r roun.1 anel mpmre e. rih.w nre given in Inhle 41 f 3 .[], r fgj @ ~ ]'p 't}g 7,Q;g Ugg, e j,,, j ,,,3,,,

. aA_ w en Je< 6eeBS d-L*4W 1

%.::.~.p.7"*

~. ~. I * * %*hg%si . , ,

. . , " ' *. i!~ *.i' *~ :[ <=

"~

~-A N i W

(

.?y,3

? ' .ShW(

1 Aren A mrowed (d) p) " #" 6 *d ' # ;

Tnine 44 Sndth's Coefficienta 0f Discharge for Circular and Squero OriGces with Full Contraction Diameter of Wewist on6ees, fe,g Side of soworo en6ees. feet llud, feet *

0. ,r3 0 04 0 07 0.1 0.3 0. 6 1.0 0.02 0 04 0,07 0.1 01 06 1.0 0 627 0 ott 0 616 04 0 643 0.626 0.628 0.651 0 630 0 6tt 0 643 0 001 0 493 06 0.660 0 636 0.623 0 tif 0 604 0 $96 0.646 0 til 0.614 0 610 0.608 0 594 0 400 0.6 0,651 0 631 0 430 0.611 0 M5 0 600 0 697 0 H4 0.623 0 611 0 tot 0.600 0 804 0.591 1 0.646 0,626 0,446 0 613 0.604 0 501 0 699 0 637 0 616 0.606 0 605 0.600 0 596 0 H3 1.5 0.641 0 623 0 614 0.610 0.60s 0.601 0 601 0.632 0.614 0 606 0 604 0.596 0 597 0 808 3 0.637 0.616 0 tit 0.666 0 604 0 604 0 601 0 629 0 613 0 604 0 603 0.696 0 694 0 $90 36 0 634 0 $lf 0 680 0.607 0 605 0.604 0 602 0 677 0 611 0 re4 0 003 0.8p6 0 $96 0.197 3 0.632 0 646 0 606 0 60r 0 60S 0 604 0.603 0.673 0 000 0 603 0 602 0.809 0 897 0 806 4 0 626 0 614 0 tot 0 606 0 605 0 603 0 602 0.616 0. W7 06*:2 0.600 0,606 0 597 0 496 6 0.623 0 612 0.607 0 606 0 604 0 603 0 603 0.614 0.605 0 601 0.600 0 SOS 0 696 0 506 6 0 619 0 610 0 606 0,608 0 604 0 603 0 603 0 ott 0 turt a t.00 0 506 0,507 0 600 0 r06 to o ate 0 606 0 nos 0.004 0 603 0 601 0.601 0 601 0 5V9 0 SW7 0 $96 U apo O Gb6 0 694 20 0 600 0.004 0,601 0 602 0 607 0 604 0 600 0 406 0 495 0 24 0 $04 0 504 0 594 0 L93 40 0 602 0 Sol 0 601 0.600 0.600 0 896 0 696 0 803 0.6e1 0.&v2 0.592 0.592 0 &vt 0 692 100 0 699 0 196 0 896 0 696 0 $96 0 396 0 696 l

27

- - . ~ _ . - - - - . . - . - - . _ - - . - . . - - - . - - - - . - - - . - . . ~ . ---

i APPENDIX D DECAY HEAT - CALCULATIONS (1) ASSUMPTIONS A. The reactor has been operating for 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months continuously.

B. The reactor scrams when the loss of coolant sequence begins at time zero.

C. neactor 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 temperature reaching the melting point of aluminum.

(2) DECAY HEAT GENERATION

() Q 2 MW = 1,96x 106 Dru/hr 14 el x 22 plates / element x H12/3600 sec = 6.07 Btu /sec From Table 5.1 at Ti:ae = 16.8 min (1008.75 sec) ~ 103 sec E_ - .0185 Po P= .0185 x 6.187 = .114 Dtu/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 0111 = Cmax = Omnx ' Btu /ft 3 (1) volume 1.w.t where-Qmax = Btu /sec and 1 = fuel plate length w = fuel plate width t - fuel plate thickness l

28

For an average sine '

Cmax = Qave x D/2 (2)

Substituting equation (2) into equation (1) we obtain:

l Olli = Qave . R/2 (3)

For conduction downwards (-x direction) and using the heat conduction equation from reference (1)

(4) d2 = --nlli dx2 k' l

kf = fuel plate thermal conductivity I i

for a sinusoidal heat generation I Qlll = Omax . sin Ox/l (5) substituting equation (3) into equation (5) we obtain i

)

0111 = Oave . R/2 . sin Ox/l (6) I l.w.t substituting (6) into (4) g d2 t/dx2 =

-Qave/l.w.t . U/2 . 1/kf . sin Ux/2 (7) integrating (7) dt/dx = -D/2 . Qave/2.w.t.kf (-cos Dx/l) . 1/U +C1 evaluating 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 = 12000F at top of fuel plate x = 2' T - 2120F at surface of water in core box x= .7' then Qave = .013 Btu /sec O

29

O Since this value is lecs than the decay heat generaticn of .114 Btu /sec, it is assumed that melting will occur.

(5) DECAY TIME TO HAVE GENERATIO!J EQUAL TO REMOVAL The length of time that core cooling would be needed t o have the decay heat to reduce to .049 Btu /sec can be calculated using Table 5.1.

the Power Ratio = .049/6.187 = .00792 and time = 2x104 sec = 5.56 hrs therefore emergency core cooling is required for at least 5.56 hours6.481481e-4 days 0.0156 hours 9.259259e-5 weeks 2.1308e-5 months . Water supply can be supplied for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(6) iOTAL CORE COOLId:T LOSS UllDER LOCA CC!JDITIO!JS Total' loss would include evaporation of the water from the core box ( a s s urred to be the rate at the maximum value associated with the initial LOC at time equal to the drain time, or 16.3 min).

From Table 5.1 - P3 /Pa =

.0185 @ T= 1008 sec The heat generation then - .0185 x 6.87 Btu /sec

.1144 Btu /sec The maximum 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/ft 3 x 7.48 gal /ft3= .000851277 gal / min (liquid)

This is added to the drainage loss and the total loss is still about 3.76 1 gpm.

O 31

(4) HEAT CCNDUCTICN TO THE WATER IN CCRE BOX FisCM THE NON FUEL ALUMINUM IN THE ELEMENT Calculation Basis - Per Plate A. Non Fuel Plate Cross Section Plate Cross Section = .05" x 2.79" a .1395"2 Max Fuel Cross Section = .02" x 2.47" = .0494"2 Non Fuel Plate Cross Section = .1395 - .0494 = .0901"2 22 plates x .0901 = 1.9822"2 B. Side Plates of the Element Average width = .187" 22 grooves (.187 .088) x .058" Cross Section 2 Side Plates x [(.187" x 3.045")-22 x (.87 .088) x .058))

Area = .8862"2 C. Total Area for the Element Area = 1.9822 + .8862 = 2.8684 Per Plate Basis A = 2.8684/22 = .13038"2/144 = .0009054'2 Heat Conducted from the Aluminum to the Water Q = kae A dt/dx Q = kala (Tmax-Tsat/l) o = 131x0009x (1200-212) = 89.604 Btu /hr x 1/3600 1= (2 .7) = 1.3' O = .02489 Btu /sec total heat conducted = fuel + aluminum

= .013 + .02489 = .03789 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 Btu /sec and the total heat removal .03789 + .011367

=

llh

= .049 Btu /sec n

_ _ _ _...._ .____._ _ _ _ _ - ... _ _ _ _ _ _ -.m-.

5.d .

Table 5.1

' The Ratio, P(ts) ./ Pn, of the Fission Product Decay Power to Reactor Operating Power as a Function of Time, tn, After Shutdown (ANS, 1968) l Time After Time After i Shutdown, t 3 Power P.atio Shutdown, t 3 Power Ratio (seconds) P(tej / Pn _ (seconds) P(tn) / Pn 0.0675 6 X 104 0.00566 1 X 10-1 0.0625 0.00505 1 X 10 0 8 2 0.0590 1 X 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 0.00282 l 1 X 101

.0.0450 1 X 106 0.00267 2

0.0396 2 0.00215 4

6 0.0365 4 0.00166 0.0746 6 0.00143 5

0.00130 1 X 10 2 0.0331 8

-h 0.0275 1 X 10 7 0.00117 -

2 4 0.0235 2 0.00089 0.0211 4 0.00068 6

0.0196 6 0.00062 8

1 X 10 3 0.0185 8 0.00057 2 0.0157' 1 X 10 8 0.000550 e 4 0.0123 2 0.000485 6 0.0112 4 0.000415-0.0105 6 0.000360 8

0.00965 0.000303 1 X 10 4 8 2- 0.00795 1 X 10 9 0_.000267 4 0.000625 e

32 _ -

m y +- :ew,< 9%-ry:.ww---r-r%+-,-,w%u----y,ww,.,#w,.r._ ww.m.,re.w--..,w--w,.....-. .---,-,...,..wn,-.-.,,..=,.e---ww-----.--..,.m.mv----

j i

APPENDIX E MAXIMUM HEAT FLUX The Rhode Island Nuclear Science Center Technical Specifications Section K,3, e, (2) specifies the maximum heat flux.

Since it is not specific in regard to how this was originally calculated using an overall hot spot factor of 2.3, the LEU het channel analysis for 2 MW calculates two conditions resulting in slightly different values.

I care 1 i 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.

case 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.

e O

I I 33 l

-

w'w ,--wn--- vg--v-= w -ver rmV w evr ,,i- *--e y--=r-t- ees ~ +-ev-4L-g ,--ec p-v * ' - te r'-3i f"'*1" 't it * 'T'"#F m-- "*

l APPENDIX F MJsXIMUM CORE SPECIFIC POWER The Rhode Island Nuclear Science Center Technical Specifications Secticn M,3,e(2) cpecifies the naximun ;pecific power. For a 14 element LEU core we are cia,.ly calculating maximum core specific power as 2 MW divided by the number of fuel elements having a maximum loading 275 g U235 Therefore, the specific power i s 2xi@ wa t t s - 519.48 IL 14x275 gU235 Since burnup increases this value, this is the limiting value which cannot be reduced.

O O

34

._ -. ~ .. -. --_-- -. .-.-. -. . _-.-- ._..--__ .-.=____...- -.

c .

SAFETY ANALYSIS REPORT O

PART C TECH 11ICAL SPECIFICATIO!1 REVIEW A!JD MODIFICATIO!J I :ntroduction II Appendix A/Rhode Island 11uclear Science Center Reactor Technical Specifications Appendix A to +

Facility License R-95 Dated July 21, 1964 Revised Through Amendment #16 III Appendix B/ Proposed Rhode Island !Juclear Science Center Reactor Technical Specifications O

l l

l I

l

'O i

i

. - . . - . ..-- - .-. - . - . - - - _--- - - - - . . - . - . . - - - - . - - . ~ . - - . - - _ _ . -

Part C Technical Specifications Review and Modification INTRODUCTION D

(V 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 the Safety Analysis Report 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 Technical Specification version. Appendix A i s a copy of the Rhode Island Nuclear Science Center current Technical Specifications. Appendix B is a copy of the Technical Specifications with the changes included as a result of the SAR and review process. The double vertical 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 Nuclear Science Cent er . Conditions outside the control of the licensee, such as key staff retirements, budget cuts, small operating staff etc., increase the difficulty and will L

curtail the operation of the facility during the conversion process. The Rhode Island Nuclear Science Center f acknowledges the assistance of Argonne National Laboratory in the preparation of the Safety Analysis Report.

) '%

--- y,w-'-ww gy y>g----,-+-- -we-eq y q.,,e-m.e__,,.y+g,--.,i..,,,,.-w_ .,-ww.,...mg-.%v-w-- ,-- y- .,,mm,m-_..-wn-,,--e y m% ,m.,-,.e-m.,--,m,-- w- , , , . ,_g, ,,,y.,,--e.a.,

. s O

APPENDIX A RHODE ISLAND NUCLEAR !iCIENCE CENTER REACTOR TECHNICAL S?ECIFICATIONS APPENDIX A TO FACILITY LICENSE R-95 DATED JULY 21 la64 REVISED THROUGH AMENDMENT #16 O

, s.

3. . .

TABLE CF CCNTENTS p- - PAGE

( ,f _ A, SITE 1

1. Location 1

2. Exclusion Area 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 .Y'*' -

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

f. Clean-up System for Primary Coolant System 5 D. SECONDARY COOLANT SYSTEM 6 E. REACTOR COPE AND CONTROL ELEMENTS 7

1. Principal Core Materials 7

2. Fuel Eler.onts 7

3. Reflector Elements 8

4. Control Slements 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 (FC) 9

2. Jesign Features 10

a. The 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

b. Gaseous Radioactive Waste Disposal System 14

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

r.

[

[ -

!.-- TABLE OF CONTENTS (CONTINUED) p.

1:

l -H. FUEL STORAGE 17

+. -

1.. New Fuel Storage 17 l= 2. Irradiated Fuel Storage 17 17 I. EXPERIMENTAL FACILITIES J. AOMINISTRATIVE AND PROCELURAL SAFEGUARDS 18

1. Organization 18 j 2. Qualifications of Pesonnel 19-

3. Responsibilities of Personnel 19

, a. Director 19

[ 'b.- Senior-Reactor Operators 20 L c. Reactor Operators 20

!- d. Health Physicist 21 l 4. Written Instructions and Procedures. 21 i =- 5; -Site Emergency Plans 21 I: K. OPERATING LIMITATIONS 22

1. General 22

2. Experiments 23 3, Operations 24

a. Site 24 b .' Containment 24 p[ . c. Primary Coolant System 25 j- d. Secondary Cooling System 26 i .- e. Reactor Core and Control Elements 26

!. f. Reactor Safety _ Systems 29

g. ' Waste Disposal and Reactor Monitoring Systems 30

h. Fuel Storage 31

4. Maintenance 31 9-f.

t 11

^=

O

1. L:ALLCD The coactor 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 University of Rhode Island. The reservation is in the Town of Narragansett, Rnode Island on the west shore of Narragansett Bay approximately 22 miles south of Providence, Rhode Island and approximately six miles north of the entrance of the Bay from the Atlantic Ocean. The Rhode Island Nuclear Science Center and various buildings used for research, education 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 CFR 100.

3. Restricted Area Figure A.1 also shows the location of the reactor building on the three acre area. The reactor building and attached offica laboratory wing shall be considered the restricted area _as defined in 10 CFR 20.

4. Princirsi Activities The principal activities carried on within the restricted and exclusion area shall 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 interfere with the operation of the reactor.

Amendment 15 O

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B. MMT A MMENT

1. Pmt e r auilM M 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 telease of radioactive material, the confinement building air shall be exhausted through a clean-up system and stack creating a flow of air into the building with a negative differential pressure between the building and the outside atmosphere. The building shall be gas tight in the sense that a negative differential pressu're can be maintained dynamically with all gas leaks occurring inward. The confinement and clean-up systems shall become operative when a building evacuation button 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- pa rt 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 the site to insure the availability of power to operate the clean-up system.

The reactor building exhaust blower, which is designed to exhaust at least 4000 c f m, operates in conjunction with

.i 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 l into the reactor building at less than 4000'cfm. The total exhaust rate through the stack is at least 14000 cfm. During ncrmal 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 roaghing filter, an absolute particulate filter, a charcoal filter for removing radiciodine, and an absolute filter for removing charcoal dust which may be contaminated with radiciodine. 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 j 0.3 micron diameter diocty1phthalate smoke. The minimum l- removal efficiency of the charcoal filters shall be 99%, based

! on ORNL data and measurements performed locally.

i Gases from the beam ports, thermal column, pneumatic system, i

and all other radioactive gas exhaust points shall be exhausted

] to tha stack through a roughing and absolute filter system.

Change 4 Amendment 16

_4 C. PEACTOR POOt A'ID PRD'3RY C20 TANT FYSTEM

1. % n a 'm '

The primary coolant system shall consist of the reactor pool, delay tank, heat exchanger, coolant pump, and the associated valves, piping, flew channels and sensors. During forced ccnvection 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 core to a plenum below the grid box. The coolant water shall then flow into the outlet flow channel on the opposite side of the suspension frame and then through a discharge line to the delay tank, coolant pump, heat exchenger and then return to the coolant inlet line.

2. Peacter Pool 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. Sbielef f ng The reactor pool and primary system shielding shall be adequate to meet the applicable personnel radiation protection requirements of 10 CFR 20.

4. Pr % ry coolant System The primary coolant system shall conform to the following:

a. Heat Ev-hancer The heat exchanger shall be designed to remove heat at the rate generated by t l.e reactor at maximum licensed steady state power from the primary water and shall be designed to perform under the maximum primary system operating temperature and pressure.

Replacement heat e.< changer 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 Pmn Number of pumps 1 Type Horizontal mounted, Centrifugal, Single Suction O

Change 3,4

- . -_ ~. - ~. .- -. .

L) Materials of construction Rating Worthite 1500 gpm Head 59 feet Design Pressure 75 psig minimum Design Temperature 1500F minimum-Motor Type Drip proof, induction, 410 v, 3-phase, 60 cycle

c. Delay Tank Number of tanks 1 Material of constructicn Aluminum Association Alloy 5083 and 5086 Material Thickness Walls 0.25 inch Dished Heads 0.375 inch

-Capacity 3000 gal., minimum d, pr a m rv Recircuintien pioina Material and thickness Sch. 40 A1, type 3003 aluminum Size 8 and 10 inch Design cemperature 1500F, minimum Design pressure 100 psig, minimum ,

l

, e. Make-uc 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 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. Cleanun System for Prim rv Coolant Water i

Cleanup pump Capacity 40 gpm Head 100 ft l Cleanup demineralizer l Type Mixed-bed, single shell,

[ regenerative Cleanup demineralizer capacity Normal 40 gpm

'. Emergency 50 gpm Change 1

-f-l D. SECT '_iDARY COOL A$iT SYSTEM The seccndary coolant system shall carry the heat rejected from the 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 from the heat exchanger through a control valve to the cooling towers. From the cooling tower basins, the water is then pumped back to the heat exchanger.

Change 4 0

9

7 I

E. REACTOR CORE A'm ccNTR0t E' EME'!U The reactor core and contro. elements shall have the following characteristics and nominal dimensions:

1. Princiral core Msterials Fuel matrix Alloy, UAlx,U 309 U-235 enrichment Approximately 931 Fuel clad 1100 and/or 6061 aluminum Fuel element side plates 6061 aluminum End fittings 356-T6 or 6061 aluminum Moderator Water Reflector AGOT grade for equivalent) graphite and/or water Control elements Mixture of B4C and aluminum, clad with aluminum Servo Element Mixture of B C4 and aluminum, clad with aluminum.

2. Fuel rierents Plate width overall 2.8 inches Active plate width 2.2 inches Plate length overall 25 inches Active plate length 24 inches Plate thickness 0,06 inch Clad thickness 0.024 inch Fuel matrix thickness 0.012 inch Water gap between plates 0.1 inch O

Amendment 8,11

a .

_g.

Nurter of plates per fuel element 19 U-235 per fuel element 124 grams, nominal Overall fuel element dimensions 3 in x 3 in. x 40 in.

3. Reflector Ela-ants overall reflector element 3 in x 3 in x 40 in.

dimensions, nominal Nominal clad thickness .1 in.

Nominal graphite dimensions 2.8 in. x 2.9 in. x 29.7 in.

4. Centrol Ela-ants Width 10.6 in.

Thickness 0.38 in.

Overall length 54.1 in.

Active length 52.1 in.

5. Servo Reculatine Flamant Shape Square boral tube Width 2.1 in.

Overall length 28.8 in.

Active 24.9 in.

6, Centrol Elamant Drive Type , Electromechanical screw Drive to safety element Electromagnet connection Streke 32 in maximum O

Amendment 11

. r .

g.

s/ - 7, serve Regula*in3 Ele ent Drive Type Electromechanical screw Drive to element corn'.ection Lock screw (no scram)

Stroke 26 in, maximum Position indication accuracy ! 0.02 in.

B. Neutren seurces i

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 F. PEACTCR 9AFETY SYSTEM 9 1, Medes of Power Ceeration There shall be two modes of power operation:

a. Power Oeeration - 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 Operaticn - Forced Circulation (FC)

Power operation - FC shall by any reactor _ operation performed with reactor cooling provided by forced l

circulation. The reactor power shall not exceed 2 MW during FC operation.

l Change 4

. +

i

2. Desien Fea*ures

a. The Re32 tor C OM TO I SY nC3 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. Process Inetrr~ntation Process instrumentation with readout in the control ronc shall be provided to permit measurement of the Stw ran, _

i temperature, and conductivity of the primary ( colant snd the flow rate of the secondary cos ant. In addition, a second primary flow indicating device with readout in the control room shall be located betseen the reactor outlet plenum and the reactor outlet header.

After normal working hours, an independent protection system, separate from the system described in Section Y, . 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, (5) An alarm condition from the radiation monitors -

reading out in the control room, (6) An alarm condition from any other selected feature.

c. tin te r Switch A key lock r i ? *. e r switch shall be provided with three positions: " . " test", and "on". These positions shall have the fo, s no M ctions:

(1) The "off" position shall de-energize the reactor control circuit.

(2) The " test" position shall energize the reactor control circuit exclusive of the control blade magnets.

(3) The "on" position shall energize the reactor control circuit including the control blade magnets,

d. Lwg r tevel Selectcr Ewitch A power level selector switch shall be providea with four positions: "9.1 MW", "1 MW " , "2 MW", and "5 MW" These positons shall have the following functions:

(1) The "0.1 MW" position shall activate all safety system sensors except those indicated in Table F.1.

(2) The "1 MW" and "2 MW positions shall activate all safety system sensors.

(3) The "5 MW" position shall scram the reactor.

e. Centrol Ela-ant Withdrawal Interlech Interlocks shall prevent control rod withdrawal unless all of the following conditons exist:

(1) The master switch is in the "on" position, (2) The safety system has been reser, (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 possible to withdraw more than one control element at a time,

f. Servo System Control Interlock Irterlocks 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 following a scram the Servo control shall automatically return to the manual mode of operation.

Change 4 O

1 1

I l

a 9 F 4 1'

12

!~

TA9tr r.1 PEAc?cR SArtTY SYSTEM

!-[ .. Trip Set Alarm Set

' - \s Sensor or Trip Device No. of Switches or Sensors Point Point

, Short Period 1 3 sec, min. 7 sec. min.

i High Neutron Flux 2 Max, of 130% of 110% max.

- - full scale with a 2.6 MW max.

4 High_ Temperature of Primary ll30F max.

Coolant Entering Core During Forced Convection Cooling

i High Temperature of Primary 12 5 cF ma x . 123cr max.

p Coolant Leaving Core During Forced Convection Cooling

Low Flow Rate of Primary 1 1200 gpm, 1350 gpm,

!- Coolant

min, min.

l 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 s ' \m/ Coolant Gates Open* 1 per gate X S X Neutron Detector High 1 per Decrease of Voltage Failure in Linear power 50 volta max.

Level Safety Channels supply Manual Scram (Switch at 2 X X bridge and on console)

High Conductivity of 1 Equivalent Primary Coolant to 24mho/cm at 250C, max.

i Safety Blade Disengaged 1 X Log N - Period Amplifier 1 X X Failure 4

Regulating Rod at Either 1 X

-Limit of Travel Low Flow Rate of Secondary 1 800 gpm, Coolant

min, i

Bridge Movement 1 X X

No Flow Thermal Column

1 X X g

These functions are bypassed when the Power Level Selector Switch is in the "0.1 MW" position.

Change 3,4,5

i TAP. LE F.2 REACTOR NUCLEAR INSIRUliEflTATIO!!

Channel Detector Sensitivity Range Information Information Information FeCorded ;

to to to Information 1 j

Operator Logic Element Servo System

( Sc r a.')

Retractable Neutrons- Source Neutron Relative gas filled approximately level Flux power level Start-up B-10' filled 12 counts /nv to full None None on log proportional power scale Neutrons- Source Power level Power level I Log N Fixed fission approximately level to Period Period scram None log scale counter 7 cps /nv 3x106 and period watts Linear Compensated Neutrons- I watt Powtr level level ion chamber approximately to Power level Level scram Power level linear scah safety 4x10-14 amp /nv 3x106 (either watts (channel)

Linear Compensated Neutrons- I watt level ion chamber approximately to Power level Level scram None safety 4x10-14 amp /nv 3x106 watts I

Change 4

G. WAETE DIf POE A' DTD FACILITY M?MITOPIM3 Evc'r"C 1, Maste Dirreeni Systems Design Fea M es

a. 1.12*.ir d d As:tive WArte Di*p2fal Sy2 M All liquid wasto 'exet pt sanitary waste) from the teactor building shall flew to retention tanks. These tanks shall be located either underground with ' dirt ecver or in a locked room (s) in the reactor building,

b. Gasecus Radinct ive M1** a M ag :A1_Iyner, All gaseous radicactive waste from the beam ports, thermal column, pneumatic irradiaion system and all other _

radioactive gas exhaust points associated with the reactor itself shall be collectm in a manifold and discharged to the reactor stack througt. an absolute filter, blower and damper.

c. Melid Radicact ive Waste Sternce 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 and Fxh9ust Can Meniter 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 10.01 to 10 mr/hr, 0.1 to 100 mr/br, o r 1 t o 1,0 0 0 m1/hr) . -

2) A radiation dose rate output indicated in the control room.

3) An adjustable high radiation alarm which shall be annunicated in the control room.

Amendment 12 Change 2,3 0

=:._ .

. e ..

4) The three fixed gamma monitors shall be lo c a*t e d

\

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 elements, and at the reactor bridge,

b. A gamma menitor shall be provided near the primary coolant 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 monitors shall be as stated in a. above.

c. Six additional direct reading area monitors employing Geiger tube detectors . shall be provided to monitor the pneumatic system receiver stations, the beam port areas, and other areas as required. Each of these shall have the following characteristics:

1) A range conslatent 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 botn an audible and visual signal.

O d. A stack exhaust gas monitor system shall be provided which

( draws a representative semple 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 Rndiation Mrnitorinc Muiment

a. Portable survey instruments for measaring 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 O

4

' s

-:5

c. A radiation monitor shall be provided to monitor all persons leaving the reactor room for beta-gamma contamination.

4. Mich h diatien Area During reactor operation, the dose rate from the delay tank may be i .. 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 posssible through a door equipped with a lock.

i O

l l

G

- ._ _ ~ ,

H. FUEL S TOPME 1, !!ew Fuel Storag 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 number of fuel elements which can be placed in each box shall not exceed three. For all conditions of moderation possible at the site Kef f shall be less than 0.8.

2. Irradiated ruel sterage 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 d'imensions of 35.5 in, wide by 26 in, high by 6.25 in, thick, and shall be fixed to the pool wall. At least two of these racks shall be provided. The second type of rack shall consist of two of the nine fuel assembly racks described above attached together with a minimum space between the centerlines of fuel assemblies in adjacent racks of 12 inches. This 18 fuel assembly rack shall be covered on the two 35.5 x 26 in, outside faces with a neutron 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 components other than fuel assemblies. The irradiated fuel storage racks shall have ' a maximum Kegf of 0.8 for all O, conditons of moderation possible at the site. Storage spaces shall be provided for at least 36 fuel assemblies.

I. N ER rNTAL FAcTttiTES 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 diaP9ter through port.

4. Radiation bas its .

5. A two-tube pneumatic tube system.

6. Dry gamma cave, i

i t

l LO

l !

J. A?MI!!! E T* ATIVE A?!S PROCEDURAL EAFEO M

1. crenniration 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 five members, as follows:

(1) The Director of Operations (2) The Reactor Facility Health Physicist (3) A qualified representative f rom the f aculty 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, s

The Director and Health Physicist are not eligible for chairmanship of the Committee. The Reactor Utilization Committee shall have the following functions:

a. Review proposale for the use of the reactor O considering the suitability of the reactor for the proposed use and the safety factors involved.

7 O

i

b. Approve or disapprove prcposed use of the reactor,

c. Review at least annually the operating and emergency procedures and the everall radiation safety aspects of the facility.

The Reactor Utilization Committee shall maintain a written record of its findings regarding the above.

2. Ous11ficatiens of Persennel

a. The Director of Operations 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 staff Health Physicist shall ce professionally trained and shall have at least a bachelors degree in one of the physical or biologi.al sciences or engineering. He shall have experience such as may have been gained throigh 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 alternatea appointed by the RIAEC.

3. Fesconsibilities of Personnel

a. D i re ct o r (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 cc.itrol 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 operating and emergency procedures are performed in accordance with the license. Where uncertainty exists, the Director shall refer the decision to the Reactor Utilization Committee.

Change 4 0

-;o.

b. sam er Reacter Or r a t en (1) A licensed senior reactor cperator 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 en call. The senior operator shall be present 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 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. Reacter Orernters I 1

l (1) The responsible senior operator shall designate for his shift a licensed operator (hereafter 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 ope rator 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) When operations are performed which may af fect core reactivity a licensed operator shall be stationed in the control room. When it is 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 relief, who shall also be a licensed operator. In such instances, the change of duty shall be definite, clear, and explicit. The relief shall acknowledge his entry on duty by proper notation in the reactor log.

- . ~ -- .- . -

l (4) The operator, under the senior operator on duty, shall be responsible for the operation of the reactor according to the approved operating schedule. l l

(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 i person working on the reactor bridge shall be I similarly authorized to scram the reactor by I pressing a scram button located on the bridge.

d. Health Phvsicist The Health Physicist shall be responsible for assuring l that adequate radiation montoring and control are in !

effect to prevent undue exposure of individuals to !

radiation. l

4. written Instructions and Precedures Detailed written operating instructions and procedures shall be )

prepared for all normal operations and naintenance and for 1 emergencies. These procedures shall be reviewed and approved )

p by quali#ied personnel before use. Each member of the staff <

5 s shall be familiar with those procedures and instructions for l which he has responsibility.

5, Site r*,=rcencv Plans l N Rhode Island Nuclear Science Center shall have available the services of other state agencies for dealing with certain l types of emergencies. The RIAEC shall enter into an agreement with the Rhode Island Civil Defense Agency whereby the Civil Detense Agency will maintain a r. 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 *'11 tors, respirators, and a particulate air sampler. Co- anications using the statewide emergency network shall be av Mble.

Personnel of the Civil Defense Agency and of local fire 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

K. cPEFATIMO tIMI?A?!_M5

1. General The following administrative controls shall be employed to O assure the safe operation of the facility:

a. The reactor shall not be operated whenever there are any significant defects in fuel elements, control rods, or control circuitry,

b. The reactor control and safety system must be turned en 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 be stationed in the control room.

Ccemunications between the control room and the senior reactor operator directing the cperation 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 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 operating the reactor shall be determined using unirradiated fuel elements or elements containing fission products in which the effect of xenon poisioning on 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 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 concurrence of both. A written record shall be rade at the time of each fuel element addition or other core change which could significantly affect core reactivity.

h. The basic operating principles for the assembly and reloading of cares whose nuclear properties have been previously determined f rom 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 the designated senior reactor operator at least one other technically qualified person who shall act as an observer.

Change 6

4 i The e" Jct . procedure to be followed for a G parti ular reloading cperation will be determined by tis 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. Evreriments

a. '" Experiments" as. used in this section shall be construed as any apparatus or device installed in the core region which is not a ccmponent of the core.

b. The Reaci.or Utilization 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 Committee.

The documentation shall include at least:

(1) The purpose of the experiment, (2) A description of the experiment, and (3) An analysis of the posskble hazards associated with the performance of the I

y. experiment.

l 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 their combined reactivity could be added to the core simultaneously, their ' cortined 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 corbined

, reactivity could be added to the core simultaneously, their cortined reactivity worth shall not exceed l

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 l 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 I O with the insertion of a reactor control element.

l l

h. No experiment shall be performed involving materials used in such a way that they might credibly result in an explosion,

i. 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 damage,

k. There shall be no more than one vacant fuel element position within the periphery of the active section of the core.

3. eperatirns

a. SLLa Control of access to the reactor facility shall be the responsibility of the Director of Operations.

b. Centain-ent, (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 automatically on actuation of the evacuation button shall be sealed, except that doors other than the truck door may be gened 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 clean-up system is operable.

(2) Re niireent s for Retest of confino-ant (a) Method of Retest The building cleanup system shall be retested by pressing an evacuation button and observing that the following functions occur automatically:

1. Evacuation horn blows.

2. air conditonc.;g 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

l

-;5 G 5. The negative differential pressure b9 tween the inside and outside of the building is at least 0.5 inches of water. This shall be determined by reading the differential raanome t e r located in the control rocm.

(b) rr ecun;ni_Ingn The b u l '. li n g cleanup eveters including the auatliary electrical pcwer system shall be tetested at least weekly.

(? The exhaust rate through the cleanup syJtem shall not exceed 4500 efm with not more an 1500 cfm ecming from the reactor building ar- ssing through the charcoal scrubber. The te ning air will be provided by a separate bic.er 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. Prim ry coolant Evstem (1) The minimum depth of water above the top of the active core shall be 23 feet.

(2) No piping shall be pla;ed in the pool which could cause or fail so as to cause a siphon of the pool watar to below the level of the ten inch coolant line penetratiens.

(3) M eue 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, cleanup system The effluent water of the primary coolant water clean up system shall be of a quality to insure compli.ance . tith K.3.c.(5) and (6) e below. l (5) The primary coolant shall be sampled at a minimum f requer$cy of once per week and the samples ana3yzed for gross radioactivity, pH, and conductivity in accordance with written procedures. Oorrective action shall oe taken to avoid exceeding the limits listed below:

Amer.dment 10 Change 4

~:6-pH 5.5 to 7.5 conductivity 2 Hmho/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 pool shall be less than 10 mrem /hr.

(7) During the forced circulation mode of cperation, the primary coolant flow rate shall not be less than 1200 gpm. During determinations of reactor power by coolant heat belances, the coolant flow rate may be reduced to 600 gpm providing all cther aspects of these Technical Specifications are met.

d. fa nndary Coeling system (1) The secondary coolant shall be sampled 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 conctntration 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 a radionuclide concentration in excess of the values in 10 CFR 20, Appendix B, Table I, Column II, above background, the reactor shall LJ shutdown and the condition corrected before operation using the secondary cooling system resumes.

(4) The secondary coolant system shall be placed in operat'.on as required during power operatien utilizin forced convection in order to maintain a primary 'cs ' ant core cutlet temperature of 1250r or below,

e. Reactor Core a nd Os_ *rol Elements (1) The reactor shall ns contain in excess of 35 fuel elements. There shaal be a minimum of four operable control ele nents.

Amendment 6,14 Change 3,4,7

s (2) The limiting thermal and hydraulic core l characteristics based on a 28 element, graphite ;

reflected core are specified below:

(a) Maximum Heat riux 47,200 BTU /hr ft2 (b) Maxirnum Core Specific Power 1,120 watt /gm U235 (c) Maximum Fuel Sarface 1970F Temperature (d) Coolant Velocity during 2.65 ft/sec, min.

Forced Convection Cooling (e) Coolant Inlet Temperature 1150F max.

(f) Ave nge Coolant Ten.pe rat u re 10er max.

e (g) Primary System Dulk Outlet 12$0F max, >

Coolant Temperature (h) Temperature Margin in Primary 430r

, Coolant (Tsat-Tsurf)

(1) Number of Coolant Passes 1 Through Core (3) Principal Nuclear Characteristics of the core (a) Core and contrel System Reactivity Worth l

1. The reactor shall be subcritical by at least 1% Ak/k from the cold, Xe-free, critical condition with the most reactive control element and the servo regulating element fully

, withdrawn.

l .

2. The maximum worth of the servo

'- regulating element shall be 0.7%

Ak/k.

(b) Ha v i mm Reactivity A+14 tion Rate - Ak/k/see

1. By servo regultting element ' maximum of 0.0002

2. Manual by control element maximum of 0.0002 l

l O

Amendment'13 Change 4 t-1:

I___._..______. _._._.._..,.~.a. . _ , - _ _ ~ . _ . _ . - ~ - - _ - - . - - - -

- _ _ . _ - _ --- _ _ -. _ ~.

8 .

Q (c) p,gntivity prefficients

1. Temperature coefficient approximately

-0.5 x 10-4 0C

/

(calculated)

2. Void coefficient approximately (core average) -1.9 x 10-3/ i void (calculated)

(4) Princiral core operating tir.itations (a) Maxi *um POO1 'TP"pgrature t?PltAtlanS The pool water temperature shall not exceed 1300r.

(b) Reactivity L4-4tatfens

1. ExcesM Penctivity The cold, clean excess reactivity for any core used in the reactor shall not exceed 0.047,

2. Minimum shutdown Margin All reactor cores used shall be such that they would be suberitical if any single W

control element and the servo regulating element were withdrawn.

(c) Peactivitv csefficient Limitation The reactor power coefficent (as inferred by the contro) rod movements required to compensate for changes . power) ahall be negative.

(d) centrol Ele-ent Drive Perforamnce Re nfro-onts 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 =" ram 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 2 months that the above specifications are met.

Change 4,7

. r .

(e) ferve Reculatine El_ ment Drive r e r f e r m a n c c-l Pereire-ante If in use during eperation, the servo r e gul a t i r.g element drive shall meet the follewing specifications:

1. The drive withdrawal rate shall not be more than 7B inches per minute.

2. It shall be demonstrated at least once per ~; nth that the above specification is ret.

(f) E1:21; G ensity ti-it The fission density limit for alloy, . r a r. l u r aluminide, and uranium oxide fuel shall eet the following specificatiens:

1. The fission density limit stall be 0.5 x 1021 fissions /cc.

2. The fission density of all fuel elements which have burnup shall be calculated at least quarter.3

f. Resctor safety Evste m (1) The reactor safety syttem shall be operable during all reactor operation. The safety system shall be checked out before each start-up and functionally tested for calibration at least monthly.

(2) It shall bis permissible to continue operations with one or more of the safety system functions that produce only an alarm temporarily disabled providing that additional 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 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 operable and shall provide a neutron count rate of at least 3 counts per s e c o r.d 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 alarm 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 tl provided with written instructions on the steps to be taken following an alarm.

Amendment 8: Change 4

. . -_ -- = . -

g, warte rinnes31 and Rearter Menitorin2 f y r t res l (1) The 11guld 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 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 MPC stated is for individual isotepes and mixtures contained in Column 1, Table ::, Appenttix 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 Parti.culate Matter and Halogens with half-lives 140 X MPC (ue/ce) 14 x MPC (uc/cc) icnger than 8 days All other Radioactive 105 X MPC (ue/ce) 104 X MFC (ue/cc)

Isotopes 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 from the site by a commercial licensed organi:ation.

(4) The exhaust gas monitor shall be calibrated to alarm at an instantaneous release rate which instantaneously exceeds the limitt' 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

-w

O (6) The area and the primary and secondary coolant system monitors shall be adjusted to alarm at a maximum reading of 2 mr/hr or 2004 of the normal radiation levels in their area, whichever is larger. -

( ~1 ) The door which controls entrance to the " maze" 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 *adiation area shall require the presence of the Health -

Physicist or a licensed senior operator and the use of direct reading portable radiation monitoring equipment.

h. f.2el Storagg (1) New fuel 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 tne 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 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 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, it shall be performed before reactor sta t-up proceeds.

(b) . Maintenance shall be performed with the approval of the Director. Equipment and system maintenance records shall be kept to f acilitate scheduling and completion of all necessary maintenance.

(c) Routine maintenance on all control and . process system components shall be performed in accordance with written schedules and with written procedures.

O

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Part C 1echnical Specifications Review and Modification INTRODUCTION I

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. j Parts A and B of the Safety Analysis Report touch on many of them. As a result of the Rhode Island !1uclear Science Center review process, additional changes which reflect current conditions or clarifications of some Technical Specification sections are also included in the final Technical Specification version. Appendix A is a cop'y of the Rhode Island Nuclear Science Center current Technical Specifications. Appendix B is a copy of the Technical Specifications with the changes included as a result of the SAR and review process. The double vertical lines adjac?nt 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 Nuclear Science Center. Conditions outside the control of the i licensee, such as key staff retirements, budget cuts, small

)

l operating staff etc., increase the difficulty and will curtail .the operation of the facility during the conversion process. The Rhode Island Nuclear Science Center 9 acknowledges the assistance of Argonne National Laboratory in l

l the preparation of the Safety Analysis Report, i

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f i

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9 9

P y O

APPENDIX B PROPOSED RHODE ISLAt;D NUCLEAR SCIENCE CENTER G REACTOR TECHNICAL SPECIFICATIONS i

)

4 L__._....-...-

TABLE CF CCNTENTS l PAGE l A. SITE 1

1. Location 1

2. Exclusion Area 1

3. Restricted Area 1

4. Principal Activities 1

. Figure A.1 2,2a l

B. CONTAINMENT 3 1.Peactor Building 3 C. PEACTOR 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

f. Clean-up Syst%m for Prinary Coolant Syst em 5 D. . SECONDARY COOLANT SYSTEM 6 E. REACTOR CORE AND CCNTROL ELEMENTS 7

1. Principal Core Materials 7

2. Fuel Elements 7 ,

3. Reflector Elements 8 O 4.

5.

Control Elements Servo Regulating Elemer.t 8

B ,

6. Control Elemer.t Driva 8 7 Servo Regu16 ting Element' Drive 9 8 ~. Neutron Sources 9 t

F. REACTOR SAFETY SYSTEMS 9

1. Modes of Power Operation 9

a. Power Operation.- Natural Circulation (NC) 9

b. Power Operation - Forced Circulation (FC) 9

'2. Design Features 10 a.. The Reactor Control System 10 c .- 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 r a. Liquid Radioactive Waste Dioposal System 14

!. 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 Monitorina Equipment 15

4. High Radiation Area 16 l . _

l . _ , - . - _ . , . _ ., .

. e s

. o .

TABLE OF CONTENTS (CONTINUED) l H. FUEL STORAGE 17 1

1. New Fuel Storage 17

2. Irradiated Tuol Storage 17 I. EXPERIMENTAL FACILITIES 17 ,

, J. ADMINISTRATIVE AND PROCEDURAL SAFEGUARDS 18

1. Organization 18

2. Qualifications of Personnel 19

3. Responsib111 ties of Personnel 19

) a. Director 19

b. Senior Peactor Operators 20 I

c. Reactor Operators 20

d. Health Physicist 21

4. Written Instructions and Procedures 21

5. Site Emergency Plans 21 :

K. OPERATING LIMITATIONS 22

1. 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

f. Reactor Safety Systems 29

g. Waste Disposal and Reactor Monitoring Systems 30

h. Fuel Storage 31

4. Maintenance 31 I

11

3 O x. m

1. Ltcatien The reactor shall be located at the Rhode Island Nuclear Science Center on three acres of a 27-acre former military

eservation, 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 University of Rhode Island. The reservation is in the Town of Narragansett, Rhode Island on the west shore of Narragansett Bay approximately 22 miles south of Providence, Rhode Island and approximately six miles north of the entrance of the Bay from the Atlantic Ocean. The Rhode Island Nuclear Science Center and various buildings used for research, education and training purposes are located on this 27-acre campus.

2. rxclu s it_.n A rea Figure A.1 is a drswing 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 bo the exclusion area as defined in 10 CFR 100.

3. Rest ricted Area Figure A.1 also shows the location of the reactor building en 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. Princiral Activities The principal activities carried on within the restricted and exclusion area shall 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 interfere with the operation of the reactor.

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B. CONTMMMT

1. Peactor Building The reactor shall be housed en a cuilding capable of meeting the following functional roo"irements:

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 creating a flow of air -into the r building with a negative diffsrential pressure between the building and the autside atmosphere. The building shall be gas tight- in the sense thht a negativo differential pressure can be mainta3ned dynamically with all gas leaks occurring inward. The confinement ,

and clean-up systems shall becomo operative when a r building evacuation button is pressed. .This ' action shall: (1) turn off all ventilation fans and the air conditioner system and (2) close the dampers on_the r 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 confiner..ent and place the clean-up system in operation. An auxiliary electrical-power system shall be provided at the site to insure the availability of power to operate the clean-up system.

The reactor building exhaust blower, which is designed to exhaust at least 4000 c f m, operates in conjunction. with O

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 tarough 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 filter, a charcoal filter for. removing radiciodine, and an absolute _ filter for removing charcoal dust which may be.

contaminated with radiciodine. .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 O. 3 ' micron - diameter diocty1phthalate smoke. The minimum removal efficiency of the charcoal filters shall be 99%, based on ORNL data and measurements performed locally.

Gases f rom the beam ports, thermal column, pneumatic system, and all other radioactive gas exhaust points shall be exhausted to the stack through a roughing and absolute filter system.

4 C. ,nnc r 2 R PCC L AS" FPHLA.RY 2 DLA'l? f YITI.li *

1. Ocne:Al The primary coolant system shall consist of the reactor pool, delay tank, heat exchanger, coolant p urtp , and the associ.'ted 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 fIcw channel which is on one side of the suspension f r arte . The coolant water shall flow from the inlet flow channel downward througn the core to a plenum below t he grid box. The coclant watet snall then flow into the outlet flow channel On the .pposite aido of the suspensien frame and then through a discharge line to the delay tank, coolant pump, heat exchanger and then return to the coolant inlet line.

2. Pe_ictor Pool The reactor pool shall be const ructed of ordinary concrete with 1/4" thick 6061-T6 aluminum liner and shall have a volume of approximately 36,300 gal.

3. ShLeldLag The retctor pool and primary system shielding shall be adequate to meet the applicable personnel radiation protection requirements of 10 CTR 20.

4, Prim 9ry Coolant System The primary coolant system shall conform to the following;

a. Heat rehancer 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 perform 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 A3ME Boiler and Pressure Vessel Code.

b. Prim ry Pumn Numoer of pumps 1 Typt Horizontal mounted, Centrifugal, Single Suction O

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, Materials of censtruction Worthite Rating 1100 gpm i

Head 59 feet Design Pressute 75 psig minimum .

Design Temperature 1500f' minimum l Moto! Type Drip proof, induction, !

440 v, 3-phase, 60 cycle

c. Delay 2 nh Number of tanks 1 Material of construction Aluminum Association Alloy 5083 and $006 Material Thickness Walls 0.25 inch Dished Heads 0.375 inch Capacity- 3000 gal., minimum

d. Prim,rv pecirculatien Pleine Material and thickness Sch. 40 A1. type 3003 aluminum size 8 and 10 inch Design temperature 1500F, minimum Design pressure 100 psig, minimum

e. Mske-un evstem A check valve shall be installed in the line between the potable water cupply and the make-up and cleanup deminerallcer 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. Cleanun System for Primarv Occlant Water Cleanup pump

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Capacity 40 gpm Head 100 ft

-Cleanup dominera11ter Type Mixed-bed, single shell, o regenerative .

Cleanup demineralizer capacity Normal 40 gpm Emergency 50 gpm

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'I D. EE00?!rARY COSLMIT S YFTL11 l

The secendary coolant system shall carry the heat rejected from the I primary coolant at the heat exchanger to the at mosphe r e at cooling towers. It shall be composed of the heat exchanger, cooling towers, pur.ps and associated valves, piping and sensors. In this system, water flows frcm the heat exchanger through a control valve to the cooling towers. From the cooling tower basins, the water is then pumped back to the heat exchanger.

E. EMER0r!!CY CCRE JOSLIt!3 SYETEM 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 duraticn of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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( E. PEACTOP CORE A'!D CONTROL ELEMENTE The reactor core and control elements shall have the following characteristics and nominal dimensions:

1. Prineiral cere Materials ruel matrix U3S 12-Al di3Persion l U-235 enrichment Approxit,ately 20 4 Fuel clad 60,61 aluminum ruel element side plates 6061 aluminum End fittings 356-76 or 6061 aluminum Moderator Water Reflector-Graphite AGOT grade (or equivalent graphite and/or water Reflector-Beryllium Beryllium-aluminum clad Control elements Mixture of 84C and aluminum, clad with aluminum

[ Servo Element Stainless steel 314 ll

2. Enel Elements Plate widt.5 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 inch Clad thickness 0.02 1 h reel matrix thickness 0,02 inch Water gap between plates 0.1 inch

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Nar:1,or of plates per fuel element .-

U-235 per fuel element 2 5 g r a r.s , nerni na l Overall fuel element dirensions 3 in x 3 in. x 40 in.

3. F.cilectr;LIle::.ents - OrsItitrud_forylliu::.

Overall reflector element 3 in x 3 in. x 40 in.

dimensions, nominal Nominal 714d thickness .1 in.

bominal graphite dimensions 2.9 in. x 2.8 in. x 29.7 in.

Nominal Ber yllium dirnensions 2.94 in. x 2.94 in. x 29 in

4. Centrol riements Width 10.6 in.

Thickness 0.3B in.

Overall length 54.1 in.

Active length 52.1 in.

5. servo Pagularing Element Shape Square stainless steel ll l

Width 2.1 in.

Overall length 29.8 in.

Active 24.9 in

6. Control El e, ant Drivg Type Electromechanical screw Drive to satety element Electromagnet cor nectic.a Stroke 32 in. maximum i

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7, ferve peculatina Element Drive Type Electremechanical screw _

Drive to element - nnection Lock screw (no scram)

Stroke 26 in, maximum Position indication accuracy 0.02 in. 1 S, rieut r en sources Start-up Source Numner 2 Type Plutonium-beryllium Unit Source Strength 1 x 106 neutrons /sec. minimum )

Maximum Power Level with Plutonium-beryllium sources installed 10 Kw l

operational Source l

Number 1 Type Antimony-beryllium Source Strength 2 x 106 neutrons /sec. minimum F, REACTOR MAFETY EYSTEMS

1. Moden of ~ Power Ocerhtien There shall be two n. odes of power operations

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. Pcwer Oeerat Sn - Forced Circulatlen frC)

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. Desien restures

a. The Reacter Crntrol Systes The reactor safety system shall consist of sensing decices and associated circuits which automatically sound an alarm and/or produce a reactor scram. The systems shall be designed on the fal.-safe principle (de-energizing shall cause a scram). Table F.1 and F.2 describe the arrangement and requirements of the safety system.

b, Frerers In?tr"-nntatirn Process instrumentation with readout in the control room shall be provided to permit measurement of the flow rate, temperature, and conductivity of the prima ry. 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 teactor outlet plenum and the reactor outlet hender.

After normal working hours, ar 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 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, (5) An alarm condition from the radiation monitors reading out in the control room, (6) An slarm condition from any other selected feature.

C. Master Switch A key lock master switch shall be provided with three positions; "off", " test", and "on". These positlens 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 cont rol L1ade magnets.

(3) The "on" position shall energize the reactor control circuit including the control blade magnets.

d. roaor Lovel relector switch l

A power level selector switch shall be provided with four ;

positionst "0.1 MW " , al MW " , "2 MW", and "5 MW". These l positions shall have the following functions: i (1) The "0.1 MW" position shall activate all safety system sensors except those indicated in Table F.1.

(2) The "1 MW" and "2 MW positions shall activate all safety system sensors.

(3) The "5 MW" pcsition shall scram the reactor,

e. Osntrol Elamant Withdrawal Inter 1cckM Interlocks shall prevent control rod withdrawal unless all of the following conditions exist:

(1) The master switch is in the "on" position, (2) The safety system has been reset, !

l (3) The Log N amplifier switch is in the " operate" i position,-

l (4) The startup channel neutron c ot.nt rate is three !

counts per second or greater, .and I

(5) The start-up counter is not being withdrawn.

It shall not be possible to withdraw more than one control element at a time.

f. Servo Svstem Centrol Interlock Interlocks shall prevent switching to servo control unless the period as indicated by the Log N channel is thirty

- seconds or greater, the - Servo cont rol- system - shall be designed so that immediately.following a scram the Servo control shall automatically return to_the manual mode of operation. ,

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- ay . r pi . or,fyre carr+v eve-ru Sensor or Trip Device No. of Switches Trip Set Alarm fet er Sensors F: int Point Short Period 1 3 sec. min. 7 sec. min.

High Neutron Flux 2 Max. of 100% of 110% max. Il full scale with a 2.4 MW rax.

High Temperature of Primary 113cr max.

Coolant Entering Core Ouring Forced Convection Cocling*

High Temperature of Primary 1250r max. 1230F max.

Coolant Leaving Core During Forced Convection Cooling

Low ricw Rate of Primary 1 1580 gpm, 1650 gpm, ll Coolant

min. min.

Low Pool Water Level 1 2" max. decrease 2" max. decrease Seismic Disturbance 1 IV on Modified Mercalli Scale max.

High Pool Temp 1 125 r 120er ll Bridge Misalignment

1 X X Coolant Gates Open* 1 per gate X X Neutron Detector High 1 per Decrease of Vo1*. age ra31ure 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 Primary Coolant to 24mho/cm at 250C, max.

Safety Blade Disengaged 1 X Log N - Period Amplifier 1 X X railure Regulating Rod at Either 1 X Limit of Travel Low riow Rate of Secondary 1 900 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.

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O TAta tE F.2 PEACTOR NUCLEAR INSTRtmENTATICH Channel Detector Sensitivity Range Information Information Information Recorded f to to to Information Operator Logic Elernent Servo System (Scram) pelative Hetractable E utrons- Source Neutron power level gas filled approximately level Flux B-10 filled 12 counts /nv to full None None on log Start-up proportional power scale ticutrons- Source Power level Power level Fixed fission approximately level to Period Period scram None log scale Log N counter 7 cps /nv 3x106 and period watts Compensated 1 watt Power level Linear tJeut r on s-ion chamber approximately to Power level Level scram Power level lineer scalo level safety 4 x10-14arnp/ nv 3x106 (e itt'e r watts (channel)

Linear Compensated Neutrons- I watt ion chamber approximately to Power level Level scram None level safety 4x10-14amp/nv 3x10 6 watts

t G. WAETE DIEP09AL AND PACILITY MSNI?ODING EYSTEMA .

1. Maste Discessi Systems resign Features

a. Liquid Radleactive Waste Disg sal System All liquid waste (except sanitary waste) from the reactor building shall flow to retention tanks. These tanks shall be located either underground with a dirt cover or in a locked room (s) in the reactor building.

b. Garecus Radleactive Waste Disposal System All gareous radioactive waste from the-beam ports, thermal ]

column, pneumatic irradiation system and all other radioactive gas exhaust points associated with the reactor i itself shall be collected in 0 manifold and discharged to the reactor stack through an absolute filter, blower and damper.

c. selid Radionetive 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 commercial licensed organization.

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_ O 2. Area and Exhaust 039 M niter resien reatures

a. Three fixed gamma monitors employing suitable detectors shall be employed in the reactor building. Each of these

'shall have the following characteristics:

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1) A range consistent with the expected radiation levels in the area to be monitored (0.01 to 10 l 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 shall be annunciated in the control room.

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The thtee fixed garres monitors shall be located d 4) to det ect tadiation as follows: At the pool biological shield between a beam port and tne t he r al c o lu rt.n , above the storage container for new fuel elements, and at the reactor bridge,

n. A gamma nonitor shall be provided near the primary coolant 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 monitors shall be as stated in a. ebove.

c. Six additional direct reading area monitors employing Geiger tute detectors shall be provided to monitor the pneumatic system receivet stations, the beam port areas, and other areas as required. Each of these shall have the following characteristics:

g 1) A range consistent with expected radiation levels j 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 O draws a representative sample of air from the exhaust gas.

The monit or with indicators and alarms in the control room, 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 10-6 c/cc. The monitor shall have a range of at least four decades.

3. Other Padiatien Msnito ri nn Eeuirment

a. Port able survey inst ruments 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 frcm I n/cm2 see to 25,000 n/cm2 see shall also be available to the facility.

b. Reactor excursion monitors shall be placed in the facility for measuring garna and neutron doses in the event of an accident.

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c. A radiation monitor shall be provided to monitor all persons leaving the reactor room for beta-gamma centamination.

4. Hiah Padisticn Area During reactor cperation, the dose rate from the delay tank may be in excess of 100 millirem per hour. On three sides, the tank shall te shielded. On the fourth side the tank is shielded using a " maze" so that access to the tink is possible through a door equipped with a lock.

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.Q EL . E 2Fl3E 1 iir y Fuel Storag

(,, Jew fue; shall be stored in a security container in " egg crate" b

" &cxes sheet cadmium at least 0.020 inches thick shall be a fastens i t.he outside of the boxes in the region which i b'A> cantaint '

el. The nv'rbe r of fuel elements which can be h "" .[ place; ;n s box shall not exceed three. For all conditions

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ci moderation possible at the site Kegt shall be less than 0.9 4

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d ated ruel Storan Twc types e' irradiated fuel element storage racks shall be y provided. One type vf rack *: hall contain . spa:es for nine fuel -

sssemblies and shall have approximate over-all dimensions of 55.5 in, wide by 26 in, high by 6.25 in. thick, and shcIl be fixed to the pool wall. At least two of these rackc sr cll be provided. The second type os rack shall consist of two of the nine fuel ass mbly racks described above attached together with a minimum space betwet.n the center lines of fuel assemblies in adjacent racks of 12 inches. Th! s 18 fuel asserbly rack shall be covered on th two 35.5 x 26 in, outside faces with a neutren absorbir.g materias. At least one 18 fuel assembly rack shall be previoed, cnd the rack may be moved within the pool.

The fuel s*orage racks may also be used to store core components other than fuel assemblies. The irradiated fuel storage racks shall have a maximum Keff of 0.8 for all conditions of moderation possible at the site. Storage spaces shall be provided for et least 36 fuel use:rt_les .

I. EXELL1 MENTAL FACTLTTIES The permhnent experimental facilities shall consist of the following: ,

1. Therma 1 cc;,. n.

2. Beam '

two 8 inch dia, and four 6 dia.

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3. A six luch diameter through port.

4. Radiation baskets.

5. A two-tube pneumatic tube system.

6. Dry gamma cave.

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J. ArMIMIS?cATIVE AND PPfCEDURAt SAFESUAPfS l

1, orgmintien The Rhode Island Atomic Enercjy Commission (RIAEC) shall have the responsibility for_the safe operation of the reactor. The ;

l RIAEC shall appoint a Director of Operations and a Reactor Utilization Committee consisting of a minimum of five members, l I

as follows:

I (1) The Director of Operations 1

(2) The Reactor Facility Health Physicist :

i (3) A qualif i representative from the faculty of Brown Universit 2 l

(4) A qualifi<.d representative from the faculty of Providence College i l (5) A qualified representative from the faculty of the University cf Rhode Island.

l A qualifhd alternate may serve in lieu of one of the above.

i The Director and Health Physicist are not eligible for

! chairmanship of the Committee. The Reactor Utilization Committee shall have the following functions:

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a. Review proposals for the use of the reactor considering the suitability of the reactor for the proposed use and the safety factors involved, i

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b. Approve or disapprove proposed use of the reactor,

c. Review at least annually the operating,and emergency procedures and the overall radiation safety aspects of the facility.

The Peactor Utilization Committee shall maintain a written re: ;d of its finf.ings regarding the above.

2. Ounlificatiens of Pers nnel

a. The Director of Operations shall have at least a bachelor.,

degree in one of the physical sciences or engineering, and he shall be trained in reactor technology and ta a licensed senior operator.

b. The staff 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 cperators shall be licensed in accordance with the orovisions of 10 CTR 55.

d. In the event of temporary vacancy in the pos3; ion of Director of Operations or the Health Physicist, the functions of that position shall be assumed by qualified alternates appointed by the RIAEC.

3. Rescensibilities of Parannnel

a. Director (1) The Director shall hava 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 operating 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. f.ani^r Re30 tor crerators (1) A licensed senior reactor ope ra t o r 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 en duty or readily available on call. The senior operator shall be present at the f acility 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 relieved, he shall turn the operation dutias 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 j his duty shall insure that all pertinent information is logged. The senior operator

assuming duty shall check the log for A infomation or instructions.

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c. Remotor crerators l (1) The responsible senicr operator shall designate l for his shif t a licensed operator (hereaf ter l caAled " operator") who -shall have primary I; responsibility under the senior operator for the operation of the reactor and all associated control and safety. devices, the proper fur.ctioning of which is essential to the safety of the reactor or personnel in the facility.

The operator shal' be responsible directly to the senior operatot (2) Only one operator 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) When operations are performed which may affect core reactivity a licensed operator shall be stationed in the control room. When it is necessary for him to leave the control room during such an operation, he shall turn the reactor and the reactor controle over to a designated relief, who shall also oe a licensed operator. In such instances, the change of duty O shall be definite, clear, nd explicit.

relief shall acknowledge his entry on duty by proper notation in the reactor log.

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f (4) The operator, under the senior operator on duty, shall be responsible for the operation o; the 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 9 similarly authorized to 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 rad 3ation.

4. u ten Instructions and Trecedures 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 shal] be familiar with those procedures and instructions for which he has responsibility.

5. Site Frarcency Plans The Rhode Island Nuclear Science Centar s hr. ll 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 a particulate air sampler. Communications using the statewide emerger.cy network shall be cvailable.

Personnel of the Civil Defense Agency and of local fire departments shall have received training from the Civil Defense

_ Training Officer .in the use of certain radiological instruments. Future training shall be augmented b-f including crientation on the reactor facility.

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K. OPERATING LIMITATIONS

- 1, General ,

The' following a d. .inis t ra t ive 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 rods, or control circuitry,

b. The reactor control 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 af f ect ' core reactivity, a licensed operator shall be stationed in the control room.

Communications between the control room an'd 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 ccre loadings to be utilized in operating the reactor shall be determined using

- unieradiated fuel elements or elements containing fission ,

products in which the effect of xenon poisoning on total 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 I 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 cot.'.d significantly af fect core. reactivity.

l g. The basic operating principles ' for the assembly and l~ reloading of' cores - whose nuclear properties have been l .-

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 l: license. During the operation there shall be present in addition to the designated senior-reactor operator at least one other technically qualified person who I ehall 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 bo*.h.

2. EG2LL" 9:LLH

a. " Experiments" as used in this section shall be construed as any aponatus or device installed in the core region whi"h is not a component of the core,

b. The Reactor Utilization Committee shall review and approve a'.1 experiments before initial performance at the facility. New types of experiments or experiments of a type significantly different from those previously periormed shall be described and documented for the study of the Reactor Utilization Committ(e.

The documentation shall include at leas -

(1s 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 their ccmbined reactivity could be addeu to the core simultaneously, their combined reactivity shall not exceed 0.006.

e. The calculated reactivity worth of any single indenendent experiment not rigidly fixed in place sha] not exceed 0.0008. If such experiments are cont eted or otherwise related so that their combined reactivity could be added to the core simultaneously, their combined reactiviti rth shall not exceed 0.0008.

f. No experiment shall be i n s t a .' l e c' 'n the reactor in such a manner that it could 1adow the nuclear instrumantation system mon' tori and thereby give erroneous or unreliable in*ormation to the control j system safety circuits,

c. No experiment shall b installed in the reactor in l such a manner that it muld fail so as to interfere with the insertion of a react:- control element.

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h. No experiment shall be performed involving materials p used in such a way that they might credibly result in

() an explosion.

1. No experiment shall be performed involving materials which could credibly contaminate the reactor pool causing corrosive action on the reactor components.

3 Experiments shall not be perfortned involving equipment l whose failure could credibly result in fuel element damage,

k. There shall be no more than one vacant fuel element ,

position within the periphery of the active section of the core.

3. cseu icas

a. Eit.2 Control of access to the reactor facility shall be the responsibility of the Director of Operations.

b. Centainment (1) During any operation in +hich the control rods are withdrawn from the core containing fuel, the following conditions shall be satisfied:

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a. Confinement building penetrations which are not designed and set to close automatically on actuation of the evacuation butten shall be sealed, except j 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 clean-up system is operable.

l (2) Remiremants fer Pete-t of csnfinemant t

l (a) Method of Retest i

The building cleanup system shall be

! retested by pressing an evac aation button l and observing that the following functions occur automatically:

1. Evacuation horn blows, i

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2. air conditioning and normal ventilation has turned off.

3. Dampers on all ventilating ducts leading to the outside have c'esed.

! 1 V 4. Building cleanup system-air scrubber and basement chem lab blc 'er come on.

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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 magn.:

helic gauge located in the control room.

(b) Ernquency_2L_ Retest The building cicanup system including the auxiliary electrical power system shall be retested at least weekly.

(3) The exhaust rate through the cleanup system shall not exceed 4500 cfm with not more than 1500 cfm coming f r c i. 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 cressure in the building which is equivalent to at least 0.5 inch of water below atmospheric pressure,

c. Pr4merv Coolant System (1) The minimum depth of water above the tcp 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 line penetrations.

(3) Makeup System The effluent water of the primary coolant water makeup system shall be of a quality to insure compliance with K.3.c.(3) and (6) below.

(4) Cleanup System The effluent water of the primary coolant water clean up system shall be of a quality to insure compliance with K.3.c. (5) and (6) below.

(5) 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:

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n pH 5.5 to 7.5

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conductivity 2 mho/cm (6) The radioactive materials contained in the pool water and in the crimary coolant water shall be such that the radiation level one meter above the surface of the pool shall be less than 10 mrem /hr.

(7) During the forced circulat ion mode of cperation, i the primary coolant flow rate uhall not be less l l than 1500 gpm.

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d. Secondarv cenline Syster; (1) The secondary coolant shall be sampled at a minimum frequency of once per week and the samples l analyzed for pH in acccrdance with written i procedures. Corrective action shall be taken to l avoid erneeding the pH limit given below: )

pH 5.5 to 9 l

l (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 i A determined at least once per week when not being I

Q, operated 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 B, Table I, Column II, above background, the reactor shall be shutdown and the condition corrected bc* ore operatica using the secondary cooling system resumes.

(4) The secondary coolant system shall be placed in operation as required during power operation I utilizing forced convection in order to maintain a l primary coolant core outlet temperature of 1250F l

or below.

e. Em etor core and centrol E l e-e n u 11 the reactor shall not contain in excess f 35 fuel elements. There shall be a minimum of four operable control elements.

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(2) The limiting thermal and hydraulic core cha racte rist ics based on a 14 element, graphite and berylliam reflected core are specified below:

(a) Maximum Heat Flux .424 MW/M2 (b) Maximum Core Specific Power 519.48 W/gU235 (c) Maximum Fuel Surface 11000 Temperature (d) Coolant Velocity during 1.48 M/sec Forced Convection Cooling (e) Coolant Inlet Temperature 11';OF max.

(f) Average Coolant Temperature 100F max.

Rise (g) Primary System Bulk Outlet 1250F max.

Coolant Temperature (h) Temperature Margin in Primary 5.80C ll Coolant (Tsat-Tsurf)

(1) Number of Coolant Passes 1 Through Core (3) Princiral Nuclear Characteristics of the Cere (a) Core and Centrcl System Reactivity Worth

1. Tne reactor shall be subcritical by at least 1% Ak/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.

(b) Max!=nm Reactivity Additien Rate .ik/k';; g

1. By servo regulating element maximum of 0.0002

2. Manual by control element maximum of 0.0002 0 ,

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, b (c) peactivity Coefficients

1. Temperature coefficient approximately

.82 x 10-4 /3C (calculated) density only

2. Void coefficient approximately (core average) -2.7 x 10-3/% void (calculated)

(4) Principal Core Operating Limitations (a) MaMienm Pool Ta*nerature limitatiens The pool water temperature s h =.11 not exceed 1250F. The pool water ' e.tp shall be monitored with readout in the Control Room. A trip and alarm shall be included in the system.

(b) Reactivity Limitatiens

1. Excess Peactivity The cold, clean excess reactivity for any core used in the reactor shall not exceed 0.047.

( 2. Minimum Shutdowfi Marcin All reactor cores used shall be such that they would be subcritical if any single control element and the servo regulating element were withdrawn.

(c) Reactivity Ceefficient Limitation The reactor power coefficient (as inferred by the control rod movements required to compensate for changes in power) shall be negative.

(d) Centrol Element Drive Perfe mance Requireannts I

! All control element drives shall meet the following specifications:

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1. The control Grive withdrawal race 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 l control element release shall not exceed l

milliseconds.

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3. The time from initiation of a scram condition

( until the control element is fully inserted abM1 not exceed 900 milliseconds.

4.

shall be demonstrated at least every 3 l

.tanths that the above specifications are met.

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a e 3 29 (e) 3rvo Pegulatine Element Drive performnne.

Recuiremega If in use during operation, the servo regulating element drive specifications:

shall meet the following &

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1. The drive withdrawal rate shall not be more than 78 inches per minute.

2. It shall be. demonstrated at least once per month that the above specification is met.

f. Reactor safety syst m (1) The reactor safety system shall be operable during all reactor operation. The safety system shall be checked out before each start-up and functionally tested for calibration at least monchly.

(2) It shall be permissible to continue operations with one or mo e of the safety system functions that produce only an alarm temporarily disabled providing that additional procedural controls are instituted to replace the lost safety system alarm function (s) .

(3) The control elettent withdrawal interlocks and the s?rvo system control interlocks shall be ful.ctionally tested at least ence per month.

(4) During reactor startup or during mechanical changes that could affect 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 aoise ratio at least 3 to 1.

(5) The linear level safety channels shall not read lese 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 alarm 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 station in Providence, Rhode Island. The central station shall be provided with written instructions on the steps to be taken following an alarm.

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g. wete Dismsal and Penctor McMtorina systres (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 following table. In this table, the MPC stated is for individual isotopes and mixtures contained in Ca' umn 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) 14 X MPC (uc/cc) longer than 8 days All other Radioactive 15- X MPC (uc/cc) 104 X MPC (uc/cc)

/7 Isotopes (3) All radioactive liquid and solid wastes dispossd of off-site shall be within the limits established by IC CFR 20 or shall be removed from the site by a commercial Jicensed organization.

(4) The eFhaust gas monitor shall be cL11hrated to alarm at an instantaneous release rate which instantaneously exceeds the limits stated in Cclumn 2 for the annual average ralease 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, individual area coolant system monitors may be taken cut of service for maintenance and repair if replaced with portable radiation detection equipment. Adequate spare parts shall be on hand to allow necesrary repairs to be made during the maintenance or calibration outages of the monitors.

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(6) The area and the primary and secondary coolant system monitors shall be adjusted to alarm et a maximum reading of 2 mr/hr or 200% of the no rma l radiation levels in their area, whichcver is larger.

(7) The door which controls entrance to the " maze" leading to the delay tank shall be locked with the key in the possession of the Director or a licensed senior operater. Entrance tc 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 portable radiation monitoring equipment.

h. ruel Storage (1) New fuel shall be stored in egg crate boxes located in a sacu.-ity 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 srcred in the criticality safe storage racks descriced 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 provided by limiting the number of fuel 1ssemblies per 9-rack to nine ard then positively securing such racks at least 30 cm, apart, or by limiting the number of fuel essemblies to 18 per rack and then covering the two large faces of each rack with'a sheet of aluminun.

covered cadmium.

4. Maintenanta (a) The slectronic control and the process control system shall be checked for proper operation and calibration before each reactor start-up. If maintenance or recalibratica is required, it shall be performed before reactoz 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 naintenance.

(c) Routine maintenance on all control and process system comoonents shall be performed in accordance with written schedules and with writteh procedures.

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ML20086D214

Contents

Text

48.5 REFERENCES

Subject:

Subject:

Dear Eugene:

.1144 Btu /sec The maximum evaporation rate for water at atmospheric saturation (1000C)

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ML20086D214

Text

48.5 REFERENCES

Subject:

Subject:

Dear Eugene:

.1144 Btu /sec The maximum evaporation rate for water at atmospheric saturation (1000C)

Navigation menu

Search