ML20101U191

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Forwards SAR for Low Enriched Fuel Conversion of Rhode Island Nuclear Science Ctr Research Reactor
ML20101U191
Person / Time
Site: Rhode Island Atomic Energy Commission
Issue date: 07/17/1992
From: Tehan T
RHODE ISLAND, STATE OF
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9207220276
Download: ML20101U191 (1)


Text

{{#Wiki_filter:- - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - - - E STATE OF RHODE ISt.AND AND PROVIDIJ:Ct: PL\\NTATh >NS Rhode Ir. land Atomic Energy Commission NUCLEAR SCIENCE CENTER Sou:h Ferry Road Narragansett, R.I. 02882-1:47 July 17, 1992 3 U. S. Nuclear Regulatory Commission Attn: Document Control Desk washington, DC 20555 Gentlemen: Encloscre (1), " Safety Analysis Rer?rt for the Low Enriched Fuel Conversion of the Rhode Island Nuclear Science Center Research Reactor" dated November 1991, is foruarded per conversation between Mr. Marvin Mendonca (MRC) and Mr. Terry Tehan (31AEC) as a description of the Low Enriched Fuel Conversion project, i Very truly yours, ~ 4 Terry Trc6n Direct RIEAC TT:cd Enclosure (1) / \\ i [ f doll,t 9207220276 920717 k PDR ADOCK 05000193 P PDR

r ,_..,1 STATE OF RHODE ISI.AND AND PROVIDENCE PIANTATIONS '- ~ ~ RHODE ISLAND ATOMIC ENERGY COMMISSION Nuclear Scit'nce Center South Ferry Road Narragansett, R.I. 02882-1197 l " SAFETY ANALYSIS REPORT FOR THE LOW ENRICHED FUEL CONVERSION OF THE p RHODE ISLAND NUCF. EAR SCIENCE CENTER RESEARCH REACTOR" 10-i i y L NOVEMBER, 1991 _O s - + gg;..2 w~ - 1 f P a Qi y . = ~ ~ - --

i SAFETY ANALYSIS REPORT PART A LEU CONVERSION ANALYSIS P AGE (S ) I Introduction 1-2 II Description of Reactor Systems 2-4 III Conversion Criteria and C'bjectives 4-5 IV ' LEU Neutronic Core Design 5-6 1 V LEU Conversion Core 6-7 Figure 1 8 Figure 2 9 Figure 3 10 Figure 4 11 Table 1 12 Table 2 13 l 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 Reflecftor Use 20-21 XI Design Basis Accident 22 XII Appendix A 23 [ \\

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Y: m Part A LEU Conversion Analysis INTRODUCTION t This safety analysis report is submitted pursuant to ICCFR 50.64 which requires the Rhode Island Atomic Energy Comission to convert its open pool research reactor from the use of nigh enriched uranium (HEU) fue2 to the use of low e n r i c h e.d (LEU) fuel. The studies required for the preparation of this report have been a joint project of the Reduced Enrichment f or Er searen and Test Reactor (RERTR) group at Argonne National Laboratory r and the staff at the Rhode Island Nuclear Science Center (RINSC). The ,< h o de Island Atomic Energy Commission is responsible for the contents of this report. The operating license for this reactor was issued on July 21, 1964 with an expiration date of August 27, 2002. The original license permitted operation at a power level of 1 MW. An amendment to the license was issued on September 12, 1988 and permitted operation as 2 MW. Since that time the reactor has operated at 2 MW. The reactor is multipurpose with capabili;i,es usually associated with open pool facilities. Because of staffing and funding limitations, utili::ation has concentrated in two areas-neutron scattering and neutron activation analysis. To meet the ^ needs of the research programs, the reactor operates one shift, five days per week. As of September 1, 1991, the accumulated operation of the reactor was 47066.3 megawatt-hours. This operation has required the use of HEU fuel elements distributed as 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 ?.ess than that O required for continuous operation. Because of the control blade

configuration, this duty cycle also requires special start-up considerations when converting to an LEU core. The studies performed for the LEU conversion have included calculations for operation at power levels above 2 MW and for advanced core designs. This was done to insure that the conversion process did not ccmpromise the future capabilities of the reactor. This safety analysis report however contains only information necessary for the initial conversion to LEU. DESCRIPTION OF REACTOR SYSTEMS The reactor system is described in the initial s a f e t ;. analysis report /1/. Only a synopsis of system components important to the ccnversion will be presented here. The reactor core sits on a 7x9 grid plate with the four corner grid positions occupied by the suscension frare corner posts. These corner posts connect the grid plate to the reactor bridge which spans the open pool. The hcIlow corner posts each contain a neutron detector required for the operation of the reactor. The grid plate is suspended about 8 meters (26.33 feet) below the pool water surface. This grid plate is insralled 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

1 0.508~mm (.020 inches) and consists of 931 enriched uranium in a -UAix 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 fcur sides by graphite reflector pieces. This core may be characte ri::ed - 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 about 141. Positions in the grid 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 l reflected core. However, such a core has never been operated above 100 KW because of the requirement -o f the experimental program and the - lack of suf ficient irradiation baskets to plug tgrid 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 j three - section,: interconnected pool. 6peration using forced convection is only possible in the circular end section where a connection can be _ made to cooling pipes. In this high power i section, the core ; neutrons are available to six radial' beam -tubes (three ~ in use), a through tube, a graphite thermal column through: whi'ch a hole has been cut creating an additional radial beam -tube, and -the terminals of two pneumatic-irradiation (rabbit) systeos. Between the graphite thermal column and the core :is a permanently installed slab of lead serving as a l-O _ thermal shield. The thermal shield is cooled by water which is 3 l l;.a ..u =

i g l' 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 autcmatic using a float activated motorir.ed valve. This system activates for a drop in pool level of 2.54 cm (1 inch). E'o r a drop of 5.08 cm (2 inches), the reactor, if running, will scram. The neutron detectors in ench cernet post provide signals to the control and safety system. Although most of this system is the original equipment, it has been well maintair ci and is reliable. Using a primary pump the core is cooled at 2 megawatts b ', 3 dcwnward flow of 0.109 m /sec (1730 gpm). Using a - nless steel heat exchanger, the heat in this primary w. .r is transferred to a secondary cooling system operating with a nominal flow of.0631 m3/sec (1000 gpm) and using a forced draft ccoling tower. The reactor is housed in a semi gas-tight, windowless building which uses the confinement concept for the controlled relesse 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 thrcugh 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 distribation were experimentally determined. CONVERSION CRITERIA AND OBJECTIVES There are six basic criteria and objectives of the LEU conversion program. These are: 9 4

I I 1. Convert the reactor to the use of LEU using the standard fuel plate which '?ill be provided to university research reactors by the U. S. Department of Energy. 2. Design a LEU core and an operating scheme to achieve burn-up greater than the current 14), 3. Design an LEU core which will optimize the thermal neutron flux in the beam tubes and will allow for further imprevement. 4. Design a reactor core with a flux trap for small e sample irradiation. 5. Design a core which does not preclude future operation at power levels up to 5 MW with the appropriate primary coolant flow. 6. Design a LEU core whose initial cost is about the same as the cost of 30 HEU fuel elements since this is the amount allocated for the core by the U. S. Department of Energy. LEU NEUTRONIC CORE DESIGN The neut ror, core design has been performed using the standard fuel plate which the Department of Energy will provide for university reactors. Figure 3 presents a comparison of this standard LEU plate with the current HEU plate. Also showi a rt. 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 bludes are an important consideration for LEU neutronic core design. Because of the more heavily loaded standard LEU fuel plate, the core may become so small that the control blade looses effectiveness. 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

d. graphite and beryllium reflectors e. relocation of the regulating rod positicn, if necessary f. use of stainless steel as the regulating rod. The neutronic calculations have been performed by Argonne Naticnal Laboratory using the EPRI Cell, C:E 3D, and */: ' Wnte Carlo Codes. Incorporating all of the infornatica gathered during these sccping studies and remerbering the si: :

r. ve z i n criteria and objectives, a LEU conversi:n design has emerged.

LEU CONVERSION CORE The LEU conversion core consists of a coinpact configuration using 22 standard plates per fuel element and a cctbination of graphite and bery'.lium reflectors. Figure 4,; resents the startup version of the conversion core which c5nsists of 14 fuel elements. The elements contain a total of 175 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 ccre The core is graphite and beryllium reflected with an excess reactivity of 3.1% a regulating rod worth of 0.451, 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 wit -ly acquired beryllium reflectcrs. Because of the te stil f t operator, the xenon behavior of this core is cyclical and this core can be operated as long as it is possible to operate on Friday morning. Using computer simulation, this core has been "run down" until a Friday morning startup is no longer possible. The reactivity balance is shown in Table 2. The reactivity requirements for Xe, Sm, long lived fission

products, control, and the cold-hot swing is approximately 31 6

I;r 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 ce reconfigured to provide additional reactivity. Figure 4 also presents this sacond core showing the fuel remaining in each fuel element af er the in;ttal 14 weeks ;f operation. The reactivity balance is shown in Table 1 and it allows f or an additional 70 weeks of operat ion. Following this second phase of operation, the graph to a n :1 { beryllium reflectors wi21 again be reconfigured. This th.rd core is shown in Figure 4 which also shows the fuel in eacn 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 berylliua 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, plecing four fresh elements in the core corner positions, and placing 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 fl'x 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 21%, 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. O 7

-.. ~ _ - - - CONTDM BLAN 1 8" </ Gul0E TUSE - i [ ]. c NEN N l , g-9 w 1 SEMW CONTROL ELEWENT ] j-g '} a RA!XAT)ON SASKET g q. 4 // FUEL EllutNT 4 g / C) i., 'q 9' /q K 2 / .y a f l* K / MGRIO 90X I, x,- 1 m k ' g<p 'L; i l ! h!k i s. ,o 4 7 es, N a i 4se' m-a 2 % V' Figure 1 8

O HEU CORE February 24, 1986 30 Fuel and 23 Graphite Reflector Elements Approx. U 235 Loadings, g per Element 3 2 1 I I E I E I MMMMMMM l' l / R__100 F 111 114 114 111 111 _,_1-m_ = 108 115 120 120 118 110 E 3 d 114 119' 124 124 118 114 D 10s 115 119 118 115 109 C mammuusens 2 umaammuunu gn ueuunummes 1 aummunutsus 110 112 117 112 110 108 8 I.e m2F M. / a.re. 2 - 1 2 3 4 5 6 7 8 9 E 1 I E I E 4 5 6 Figure 2 9

__-,_ ~.._. p. Description of HEU and LEU Fuel Elements t Figure 3 HEU LEU Number of Fueled Plates / Element 18 22 215U 124 275 Fissile Loading / Element, g U 3i:-Al Fuel Meat Composition UAix-Al 3 Cladding Material 1100 All 6061 A12 Fuel Meat Dimensions 3 thickness, mm 0.50S 0.509 i 61.0 62.7 - 71.1

width, mm 52.1
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 element.s.
t

- 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 Y LEU " EG&G #422873-1 l~ Plate Lp ~ 10 (- t

e Fig 4. Startup, Transition,and Equilibrium Cores (Lifetimes Based on Operation for 8 Hr/D,5 DM/k) ST ARTUP CORE cent 2 Core Ufetime: - 14 Wks (. 560 Fult Power Hours) Core Ufetime:- 70 Wks (- 2800 Full Power Hours) Lb u_ O s G O

  1. s a m.-

e .mm-ex - m- =.,.m. l 272 271 271 271 m t 275 l 275278 8-E k 275 275

  • )

D na m 271 y 271 272 e D na 278 275 k B 275 275 e y 275 278 275 275 275 C m m 271 m C f -- nummass i e m-E mm-mme i ammmmme I 2 1 2 ' ~' O CORE 3 EQUtuBRIUM CORE Core de w - U h (- 22 X Power Hours) Cors Ufetime: - 60 Wks (- 2400 Full Power Hours) i i i i r i o x 0 O cic 0 Q] F B5 i m = = = m i mm m 2.. e 42 241 m D ma 252 247 244 251 a O E ~ j 154 249 tes 249 254 C mammamam a summmme amammes s umammame M= CIC FC A 1 2 3 4 5 e 7 9 1 ? 3 4 s a 7 e ' '~ O 11

Table 1 Calculated Data and 300 Excess Reactivity for First Ten Cores Core Lifeti:*e Accum. Operation BOC Express Weaks Weeks Years 4.dk/k Startup 14 14 0.3 3.0 Core 2 70 S4 1.6 4.1 Core 3 60 144 2.8 3,7 Core 4 33 177 3.4 3.0 Core 5 51 228 1.4 3.6 g 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 N s 12 l l ________________.___________.__.________.______________._______________________U

7 _7 / -,l TAllLE 2 Reactivity Italances on the Friday Morning of the Last Week of Operation for Ten Cores from Startup :o Equilibrium ) R< Hector Changes Only 4 Burned Eternents Reinoved and 4 1 resh clenients Added in Corners Sta: tup Core 2 Core 3 Core 4 Core 5 Core 6 Core 7 Core 8 Core 9 Core 10 SLyL % Ak/k % ak/k % ak /k % ak / k ii.Akl 'l ak/k 'i_a k Lk 4_akik %dkl1 Fresh Cold Clean .3.00 5.19 6.92 6.92 6.92 6.92 6.92 6.92 6.92 6.92 U Reactivity Losses Burnup 0.30 1.85 3.17 3.08 3.09 3.12 3 07 3.06 3.06 3.06 Xe 1.54 1.54 1.54 1.54 1.34 1.54 1.54 1.54 1.54 1.54 S rn 0.57 0.73 0.73 0.73 0.13 0.73 0.73 0.73 0.73 0.73 Long-Lived F.?. 0.09 0.57 0.97 1.06 1.03 1.03 1.07 1.06 1.07 1.07 Cold-ilot Swing 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 0.30 ConIr01 Ql0 OlD LL2D Dl0 LL20 LL20 0lQ LL10 0,29 OL29 3.00 5.19 6.91 6.91 6.89 6.92 6.91 6.59 6 90 6.90 i - w ~ ?

i J Table 3 l[ Reactivity Data and Power Feaking Factors i-Start-up Core. Core 2 Core 3 Core

  • .0 Excess Reactivity 3.1 4.1 3.7 4.0

%Ak/k r Shutdown Margin 6,7 6.1 6.4 i-lak/k l (Slade-'3 stuck out) j: Worth of Reg rod .45 .41 .47 %Ak/k j j-l- Total Pow'er Peaking 2.64/D6 2.60/06 2.36/D6 Factor / Grid Position (Control Blades Full Out)

9.

, l' Total Power Peaking 3.06/D6 3.05/D6 2.81/D6 i I Factor / Grid Position (Control Blades 50%-Inserted) 1-i l. I'- l i 6 lJ@ r g. a t; en ve r* # r ,, w w w.,,~, w "- ,c w.,w~w-,- ,,,-,.e.-, ....-.-.-....,_mm...ee_<_.. m. ...~-,---_ - -- ,-n--, ..,6- <----~4------.r-

For the LEU cores, additional kinetic parameters and reactivity coefficients were calculated by ANL. The comparisons are shown in Table 4. Table 4 HEU LEU LEU LEU Ref. Startup Transition Equilib. m re-a Core 2 core in Eineti g Parameters Delayed Neutron O.762 0.782 0.776 0.764 Fraction, p-eff, i Prompt Neutron 76.3 66,2 66.0 68.3 Generation Tine, ps Reactivity coefficients: 20-400C Change in Water Temperature Only SAk/k x 10-4/oC Coolant -1.51 -0.80 -0.86 2.89 Change in Water Density Only %Ak/k x 10-4/ C Coolant -la .- n, pp -Q,75 - n, 6 c. Ccolant Coeff., %Ax/k x 10-4/oC 3 Coolant -2.0 -1.6 -1,6 -1.6 l Doppler Coeff., i %Ak/k x 10-4/oC Fuel 0.0 -0.18 -0.18 -0.18 1 Temperature Coeff", %Ak/k x 10-4/CC -2.0 -1.8 -1.8 -1.8 l %Ak/k/i void 0.0015 0.0027 0.0025 0.0023 4

  • Fuel and coolant temperature changes were assumed to be the same r

here. The fuel temperature rise will be larger than that of the coolant. Change in Reactivity = (Coolant Coefficient) x ATcoolant (Doppler Coefficient) x AT uel- + f t i It can be seen that there are no significant differedces betweer i the HEU and LEU kinetic and reactivity parameters. t 15 l 4

_. ~. _ _ _ _ _ _. - - _. _. _. _ _ _ _ _.. - _ START-UP ACCIDENT-This accident was analyzed using a digital computer program PARET/3/. The accident is postulated to proceed under the following assumptions: i ) 1. The reactor is in the cold clean condition with power at source' level, o I l -11. The servo regulating rod is withdrawn, followed by continuous withdrawal of all safety blades in succession at their maximum rate. i lii. Period scram protection tails, t iv. -The reactor is scrammed by the high flux sensor instrumentation when the power level reaches 2.4 MW (201 overpower). 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 ta' ken as 0.5 seconds. The analyses indicate that the maximum fuel temperature L (1.e., hot spot in the hottest channel) reaches 67.33C (1530F) for the HEU fuel-and 88.10C (191'3F) for the LEU fuel. Thus, it-can be concludeci that this accident results in no harm to the -reactor, io L If_ assumption "lii" is modified to-- " period and high flux l p -scram protection fails" then reactor power would continue _to rise beyond the trip _ point (2.4 MW) until the negative reactivity y introduced by the void and = temperature coefficients is greater that; the net. positive ' reactivity inserted by blade withdrawal. Table -I provides the peak power and the maximum cladding temperature reached in the cladding for both HEU and LEU fuel f' cases. In' _ each ; case, - the maximum cladding temperature is less thatil500C - (3020F) - much lower than the 5820C (1080ct) melting L 16 I' L..

temperature of 6061 cladding. The core in each case would operate in the nucleate boiling range without physical damage until the accident coula be terminated by a manual scram. Table 5 Feak Power and Cladding Terperatures Peak Cladding M Peak Prwer, Md Terperature. 2; HEU Equilibrium 32.1 149.1 LEU Startup .i.9 148.3 4 LEU Equilibrium 16.2 148.5 I REFERENCES (1) Atomic Energy Commiss!00, 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, S ept embe r, in press A Program for the Analysis of (3; Obenchain, C.F., "PARET Reactor Transients"" 1DO-17282 (1969) ) 4 4

-... ~ - -. . ~ l r a REPLACEMENT REGULATING ROD The current regulating rod in the HEU core is located in the f D-1 grid position (refer to Figure 2 in the " Description of the l Reactor System" section of this SAR). The rod is fabricsted frcm i boral plate and aluminum, Calculations from AN L (1 ) indicate that the regulating rod -i worth in its present core position (D-1) is reduced from' 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 i increases to .8 .9% which is too large. The regulating rod limit j by technical specification is.6% Ak/k. Therefore a new stainless i steel regulating blade is necessary in the D-2 position. Calculations indicate a satisfactory worth of .4 .5% Ak/k results. i -. 1 The new regulating blade will be fabricated with the same j dimensions to properly f! t

t. h e core grid box.

All references in this SAR relating to the new LEU cores are mede with the nev regulating rod as part of the core arrangement. O -(1) Memo from James Matos, ANL to RINSC, Eugene Spring, Septeter 16, 1991 l l 1 9 0-18 {~ J

USE OF BERYLLIUM REFLECTORS IN THE RINSC-LEU CORE The proposed use of Beryllium reflectors in the LEU sta:t up and equilibrium cores has been reviewed. The Beryllium reflectors are currently being designed by EG&G Idaho. The Universit." of Missouri Reactor'W has been using Beryllium and has ccnducted tests to determine a lifetime limit based upon a fast fluence level. Referencet.;) indicates that embrittlement of Beryllium is first noted often approximately 3x10" NVT. As a result of HF:R determinatien of small cracks occurring at a

1. '3 x 10 G )

NVT level, a proposed changeout level of l x 1 0 <' ' is proposed. Using a maximum flur of 3.3x1013 and a 5 day, 7 hrs / day reactor operating cycle, a proposed changeout of 45.8 years is predicted. Little change in other Beryllium properties at our operating temperatures and integrated fission neutron dose occurs up to the propose limit, m m m m m m Gamma heating has been reviewed

  • C' 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. M Final drawings are due shortly. A standard element is proppeed siellar to the graphite elements. A \\ special e lerce n t, witv a 1.25cm enter hole, will ha designed and used as a flux trap for-+,p dial experiments (see Figure ?) The gaxinum -epiculat ed flux in the Be portion of the flux trap is \\ 3.3x1013. ' A removable plug would be used to fill the hole when not j L ~in use. The RINSC__would_ require.a technical-. J pecification change for use-of Beryllium reflectors. Table 5 shows the average midplane flux in the Be of the central flux trap. Figure 5 Grp Upper Lower H2 hergy r = "r y Startur core 2 core 10 1 14.0 MeV 0.821 MeV 2.5x1013 2.4x1013 2.4x1013 2 0.821 MeV 5.531 MeV 3.3x1013 3.3x1013 3.2x1013 3 5.531 kev 1.855 ev 2. 7 x1013 2. 7 x1013 2.- 4 1.855 eV 0.625 eV 3.7x1012 3,gx1912 3.( O 5 0.625 eV 0.251 eV 3.3x1012 3.3x1012 3, 6 0.251 eV 0.057 eV 1.8x1013 1.8x1013 1.' 7 0.057 eV 0.00025 eV 2. 5 x1013 2.5x1013 2.i a

~. -. - -. -. ~.....-. - - ~ REFERENCES FOR _ BERYLLIUM REFLECTOR USE i '( 1 ) ASWE,- 74-PUP-44, " Stress and Deformation Analysis or 1 Irridation Induced Swelling" by B. V. Winkel, 1974 j -- (2) " Surveillance Testing and Property-Evaluation of Beryllium in Test Reactors", J.M.

Beeston, M.R.
Martin, L.R.
Brinkman, G.E.

Korth, and W.C. Francis, Aerojet Nuclear Company, !daho Falls, Idaho (U. S, Atomic Energy Comission 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 Beryl-lium Statistical Evaluation J.M. Beeston, EG&G Idaho, October 1976 (5) Missouri Unit ersity TM-ERS-62-1, MOU-30203, June 22,

1962,

" Stress and Thermal Analysis of the Beryllium Reflector for l the University of Missouri Reactor" l (6) General Electric Co. Atomic _ Power Equipment Depar_tment Standard 788, " Beryllium, Hot Pressed, Nuclear Grade" p -(7) "The Effects-of Neutron Irradiation on Beryllium Metal", B.S. i-Hickman, The Institute of Metals Conference on the Metallurgy of_ Beryllium, October 1961 -(8)_ "The-Effect-of High-Temperature Reactor. Irradiation on Some Physical and-Mechanical Properties of Beryllium, J.R.

Weir, l-The Institute of Metals, Conterence on the Metallurgy of l

Beryllium,-October 1961 (9): "The : Behavior of" Irradiated Beryllium, R.S.

Barnes, The

-Institute--of

Metals, Conference _on the Metallurgy of

[ ~ Beryllium, Octocer 1961 t l 20 ~

1 (10) University of Missouri, Inter-Department Correspondence, g Gerald Schapper; Beryllium Peflector Changeout, December 16, a s i b-(11) EG&G Idaho, Material Specification, Beryllium Pressings and j Components for Nuclear Peactors and Reactor Syctems, Document !;o. A!!C-8 0 0 0 5 G, April 26 1MB (12) University of Missouri, Specification Draw :.g Beryl. u~. Reflector, Drawing No. 193, October 6, 1988 (13) FAX memo from Argonne 'ational Laboratory to Rhode !sland Nuclear Science Center, W.L. Woodruff to Eugene Spring;

Subject:

Gamma Heating in Beryllium Beflectors, '4 a r c h 19, 1991 (14) FAX remo from Argonne Nat ici.a i Laboratory to Phode Island Nuclear Sciences Center, James Matos to Tagene Spring,

Subject:

Flux in Beryllium, September 19, 1991 O l I { t O u 1

DESIGN BASIS ACCIDENT The design basis accident for this reactor nas 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 2.25 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. i -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, discussiona and = -calculations are shown in the thermal hydraulic section of the report (Part B). I I l 0 22

APPENDIX A LEU FUEL SPECIFICATIONS AND DRAWINGS EG&G IDAHO INC. (A) TRTR-6 Specification for Test Research Training U Si2 Fuel Plates Reactor LEU Silicide 3 Rev. 4, 20 May 1988 m) TRTR-11 Specification for Low Enriched U '<etal for Reactor Fuel P':ites Bev. 1, 1 April 1987 (C) TRTR-14 Specification for Reactor Grade Uranium Silicide U Si2 Powder 3

Rev, 2,

1 July 1987 (D) TRTR-15 Specification for Aluminum Powder for Fuel Plate Core Matrix Rev. 2, 1 July 1987 EG&G DRAWINGS (A) Test Research Tri.ining Reactor LEU Fuel Plate No. 422264 (B) Rhode Island Nuclear Science Center Test Research Training Reacto: 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 4 23

SAFETY ANALYSIS REPORT PART 3 THERMAL HYDRAULIC ANALYSIS PAGE(S) I introduction 1 II Description of Computer Programs used 1 in the thermal-Hydraulic Analysic III LEU Parameters 2 IV Hot Spot Factors 3 V Steady State Full Ccre Isn a l ys i s 4-9 VI single Channel (Hoti Analysis 10-12 VII Natural Convection 13-14 VIII Rhode Island Nuc2 ear Science Center Water Supply 15-16 IX Loss of Coolant Analysis 17 X Emergency Core Cooling System Ope.n. tion 18 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 4 Appendix F/ Maximum Core Specific Power 34 t O

l Part B Thermal Hydraulit Analysis INTRODUCTION The thermal hydraulic studies for the LEU core have seen a joint effort by the Rhode I61and Nuclear Science C. iter (RINSC) and Argonne National Laboratory. The proposed new fuel elements have been described in the main introduction of o, l}; the Safety Analysis Report. Pertinent documents reviewed by a !)! the RINSC for LEU fuel use are referenced in Appendix A. Fuel plate, channel dimens wns and other parameters used in p the thermal hydraulic studies are hereby referenced in the Appendix A documents. l DESCRIPTION OF COMPUTER PROGRAMS USED IN -THE THERMAL HYDRAULIC ANALYSIS The computer programs used by the Rhode Island Nuclear '\\ Science Center Steady-State Analysis, Hot Channel Analysis l and Natural Convection Analysis were obtained from Argonne National Labofatory. 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 utl.11::e in-house computars. l The program entitled "PLTEMP" was used to perform the t [ "Strady-State" and single Hot Channel Analyses. The program entitled "NATCON" was use to perform the Natural Convection Calculations, i 1 m. 1._- my i~9 .e n ~

- j I LEU PARAMETERS i The parameters used in the PLTEMP" and "NATCCN" programs were calculated using the " LEU Fuel Element" and the proposed core c r, figurat ions. Previous sections address nuclear parameters. L T..u physical dimensions of acre components used were obtained frcn f current drawings. In addition to the normal dimensions of ccre cceponents used' l 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 referer.ce data. c^RE cmpcMENT ORAWING M"Mnrk LEU Fuel Plate EG&G d411650 } Radiation Basket (w/ orifice plate) GE #7980413 Control Blade GE #197E647 Servo Control Element (Regulating Blade) GE #612D964, 762D407 Graphite 3eflector (also Be reflector) GE #985C248 Antimony Beryllium Source GE #655C430 Radiation Basket (center hole type) GE #798D413 P n I f,- t f t ...p, 9-.93, --y_,,- ,7-,..w.wn..,y,m,- ,,,,,,,,--,,w,._y-4.,-,,-ess.w- -,,u.+ -.e -sw----ew.+n ..-e-en me-wes-6==.m a1.-am-'.ev-wen s-~+rr-

1 HOT SPOT FACTORS S: 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 11uclear Science Center has prepared a report entitled " Report , the l D e t e rmi n a t. i o.1 of Hot Spot Factors for the Rhode Island !;uclear Science Research Using LEU Fuel The report was performed in August 1989. The results are shown telcw: 1.62 Eb (Bulk Water Terperature Rise) = 1.46 Fq (Heat Flux) = 1,41 Th (Heat Transfer) = 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. G: 1 l l l 3 ~

STEADY STATE FULL CORE ANALYSIS The PLTEMP Program was used to analy:e the full core for the axial peaking factors of 1.32 (blades out) and 1.536 (blades 50i in) core conditions. The full core analysis included the various components (fuel elements, reflectorc, baskets etc.) as shcwn in Figure 4. This analysis initially determines the flow rates for the fuel port ion of the care, the by-pass flow through the other ccrpenents and the total core flow versus the pressure drop across the core. This data is tabulated in Table A. The fuel plate surface t e rap e r a t u r e 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 3.51 degrees centigrade for the two axial factors (F axial 1.32 and 1.536). From the tabulated data and our pump flow of about 1 1730 CPM, the core flow (1100 GPM ) and a by-pass flow (625 1 GPM ) 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 lis 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 blades out condition (F axial = 1.32) and the blades 5 0 't 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's in situations. O 4

l { RETE:tENCES (1) Memo from Bill Woodrutf, Argonne !!aticnal Laboratory to Eugene Spring, Rhode Island !J u c le a r Science Center, 1/30/91 (2) Memo from James Matos, isrgonne !;ation al Laboratory t o A. F.

DiMeglio, Rhode Island !;uclear Science
Center, 1/22/91 0

O 1 5 l

LEU FULL CORE ANALYSIS - 14 ELEMENT CORE l

2 MW TABLE A i

DP. (MPA)

Core Flow By Pass Tet:1 How Core Flow B y-Pass Total How (kg/s) Flow (kg/s) (GPM). (GPM) p (kg/s) (GPM) i .0025 '43.81 25.49 69.30 698.00 407.00 1105 .0030 48.56-2R.15 76.7I 774.(X) 449.00 1223 .0035 --52.98 30.63 83.til 844.5 489.50 1333 i .0040 57.14 33.66 90.80 910.8 536.20 1447 .0045 61.07 35.13' ' 96.20 973.5 559.50 1533 j'- .0050 64.33 37.21 102.04 1033.4 593.60 1627 .0055 = 68.43 39.19 107.62 1000.4 624.20 1715 { .0060-71.89 41.09 I I 2.98 II46.00 654.00 1800 i I- .0065 75.23 24.77 118.16 1199.00 684.00 1883 i, TABLEB { ap-By. Pass Core How Total How l Outlet Hulk Plate Outlet Bulk Plate How g y pA) Temp 0C Surface Temp 0C Surface -(GPM) (GPM) (GPM) Temp UC Temp UC F asial=1.32 F' axial =1.32 F axial =1.536 F axial =1.536 L .0025 407.0 698.0 1105 54.70 74.06 54.70 77.52 I h- .0030 449.0-774.0 1223 53.49 71.49 53.44 74.71 i t i' 0035 48M.5 844.5 1333 52.56 69.69 52.56 72.49 .0040 536.2 910.R 1447 51.R1 68.03 51.81 70.69 l .0045 559.5 973.5 1533 51.20 66.65 51.20 69.1R l 2 .0050 593.6 1033.4 1627 50.69 65.29 50.69 67.89 f f .0055 624.2 1090.R I715 50.25 64.28 50.25 66.78 [ j~ 0060 654.0 I I46.0 1800 49.86 63.56 49.86 65.80 l l .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 64.16 1 4 - 1 NOTES: (i) Normal Primary Pump Operation 1730 GPM j~ (2):LCalculations Based on Inlet Temp. to Core of 42.30C [ li k + i h r O G G

TABLE C O; <r PLTEMP full core analysis for each fuel element 1 at ',es additional parameters. A typical output value is shown for these parameters. (Element #8 data) PARAMETER VALUE Maximum Surface Temp. OC 64.45 Clad Temp OC 64.78 2 Peak Axial Heat Flux (MW/M ) .156 Channel Flow Rate (kg/s) .2298 Velocity (M/S) 1.605 Outlet Pressure (MP A) .1727 2 CHF (Critical Heat Flux) (MW/M ) 2.39 2 Flow Instability (!M/M ) .889 2 Exit Saturation Heat Flux (MW/M ) 1.087 Minimum D!1B Rat 10 8.461 FJaxial 1.32 O t 7

TABLE D A XI AI. DISTRIBl TTION INFliRFAW RELATIVE DISTAN& IIIADE-OtJT BLADE-IN 50% 1 0.00 .4105 .0553 2 0.05 .5506 .2682 3 0.10 .6836 4759 4 0.15 .8076 .6744 5 0.20 .9219 .R597 5 0.25 1.0228 1.0283 7 0.30 1.1111 1.1770 8 0.35 1.1M50 1.3028 9 0 40 1.2435 1.4032 1" 0.45 1.2K58 1.4764 __.,,18 0.50 1.3200 1.5360 12 0.55 1.2858 1.4764 13 0.60 1.2435 14032 14 0.65 I.IR50 1.302X 15 0.70 1.1III I.1770 16 0.75 1.0228 1.0283 17 0.XO 4219 .8597 1M O.85 .8076 .6714 1u O 90 6M16 4759 20 0.05 .5506 .26R2 21 I.00 4105 .0553 0 9 9

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-SINGLE CRANNEL (HOT CHANNEL) ANALYSIS O-Computer runs using PLTEMP were run for the single channel analysis using the derived hot channel factors. Flow rates' _ and power levels were varied to provide sufficient information for "1imiting power level and core flow terminations. The tabulated results for axial factors of 1.32 (blades in positicn) and for axial factors of 1.536 (blades 50% out) are shown in Table E. The results are also presented as a " Hot Channel Fuel Surface Graph" depicting " Fuel Temperature" ~ vs. Total Core Flow. The normal primary flow rate for the Rhode Island Nuclear Science Center reactor is about 1730 GPM. From the -data it can be seen that incipient boiling occurs at about 2.6 MW or 130% of the normal 2 MW power level. At a reduced flow of abut 1580 GPM incipient boiling is reached at about 2.4 MW or 120% of normal power. The proposed limiting safety settings are then chosen as shown below: Normal. Power Level Over Power Trip (scram) c 2MW. 120% (2. 4 MW) Normal Flow Reduced Flow Trip (scram) 1730 GPM 1580 GPM These values are more restrictive than the present trip-levels of 130% for overpower trip (2.6 MW) and 1260 GPM flow. This is due to the fact that the compact core of 14 elements and higher fuel density have more effect than the increase in number of fuel plates from 18 to 22. The maximum surface temperature of the fuel resulting at the 1580 GPM pump flow is from Tabel E about 1100C. The corresponding coolant velocity from the PLTEMP output for 1533 GPM (AP=.0045) 1.44 M/S and for 1627 GPM = (AP=.0050) 1.53 M/S. An extrapolated value for the 1580 v GPM condition-is about 1.48 M/S. These safety settings will require a technical specification change.upon NRC approval. 10 l

4 TAHLE E IIOT CIIANNEL ANAI YSIS F AXIAL = 1.536 { Ap Total flow 'T - Surface T Surface T - Surface T Surface T SAT. T cub (MPA) (GPM) 2-2.2-2.4 2.6 [ d MW MW MW MW i .{ .0025 1105 122.26 122.57 122.R6 123.10 115.82 122.1 .0030 '1223 122.27 122.57 122.87 123.15 115.R2 122.1 i .0035 1333 119.27 122.58 122.RR 123.16 115.R2 122.1 ) .0040 1447 I14.99 121.60 - 122.89 123.17 115.R2 122.1 .0045 I533 111.44 117.7i 122.90 123.19 115.82 122.1 f 0050 1627 108.40 114.41 120.36 123.20 115.82 122.1 .0055 I715 105.14

  • 11.54 117.24 122.93 II5.22 122.1

[ .0060 1800 103.39 109.00 114.53 120.00 115.R2 122.1 I .0065 IRR3 101.30 106.75 112.10 117.39 115.R2 122.1 .0070 1963 99.42 104.71 109.93 I I 5.06 I I 5.R2 122.1 l 7 IIOT CIIANNEI ANALYSIS 1 F AXIAL = 137 l Ap Total llow T Surface T Surface T Surface T Surface T SAT. T onb j (MPA) IGPM) 2 2.2 2.4 2.6 oC OC ~ l MW MW MW MW l .0025 1105 120.80 122 09 122.36 122.62 115.32 122.1 i j .0030 1223 115.01 121.52 122.37 122.63 115.82 122.1 I 0035 1333 110.36 116.52 122.3R 122.64 115.82 122.1 l } .0040 1447 106.55 112.40 118.16 122.65 115.R2 122.1 .0045 1533 103.3$( 108 94 114.44 119.86 115.R2 122.1 i 0050 1627 100.67 105.99 I I 1.26 I I 6.4R I15.R2 122.1 [ .0055 1715 98.30 103.43 108.44 113 51 115.R2 122.1 f .0060 3800 96.21 10i.18 106.07 110.W) 115.32 122.1 l .0065 1883 94.36 99.17 101.91 108.58 115.R2 122.1 .0070 1963 92.69 97.36 101.97 106.51 115 R2 122.1 i i f it e 9 G-i - - t

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1 p NATURAL CONVECTION j_ i l i The present HEU core has a licensed limit of 100 kw i l operation fc.; the reactor in the natural convection mode (No l l Primary Pur p Opert. tion), j i The Natural Convection Analysis for the LEU core was i performed using the NATCON Program. The program was run for j both the regular channel and the " hot channel" conditions. Both cases were run for the blade out situation (F axial = l 1.32) and the blades 50% in (F axial = 1.536). l The results are shown in Table F. Note that for the f [ _ most conservative case, (hot channel) the power level is j 217.3 kw, using incipient boiling as the limiting parameter. The maximum wall temperature was calculated as a i l function of axial length snd the value was tabulated from the l data. The program run terminates when the fuel surface l r temperature reaches incipient boiling. l Since the 217.3 kw exceeds the current licensed power -/ level of 100 kw for natural convection, no change is deemed necessary in the licensed maximum natural convection power - level of 100 kw. [ F F K P k 13 .~n- -. - - - -., - - +. + - - - -. - -. ,..- ~.,----. -,..-,.. -- - _,,. _ n. ..,~-,,,_,n. ,,n.-wnm,,-,nn- .---,rnm- ..-m

i-i TABLE F t NATUltAL CONVECTION REGUI.AR CilANNEL FAXIAL = 1.536 f P Power Level . Exit Maximum - Incipient T sat-T wall Radial Margin to (kw) Temp. W alt Boiling "C (117.34-TW) Peaking Incipient Boi!ing 1 Factor o Temp. ' Oc g i t i 10 31.93 52.01 117.57 65.34 2 65.56 100 68.41 71.22 118.28 46.12-2 46.06 200 77.29 84.71 '118.88 32.63 2 34.17 [ 4 300 83.80 96.48 119.21 20.86 2 22.73 i 500 93.69 117.85 119.85 .51 2 2.00 520.4 94.56 119.89 1I9.89 -2.55 2 0.00 i l. T IlOTCilANNI3. ( '10 59.1M 58.89 117.72 58.45 2 58.83 [ l00 86.63 92.81 118.56 24.53 2 25.75 [ 209.1 102.11 I19.34 119.20 -2.00 2 -0.40 I NATURAL CONVECTION REGULAR CilANNEL FAXI/.L = 1.32 Power Level Esit Maximum Incipient T sat-T wall Radial Margin to (Lw) Temp. OC Wall Boiling o c "C Peaking Incipient Boiling [ Temp. Oc Factor oc j r t {- 10 52A1 52.26 117.60 65.08 2 65.34 i L 100 bX.72 71.58 11H.I9 45.76 2 46.61 { 200 77.69 84.25 11 H.72 33.09 2 34.47 [ 300-54.26 95.11 I19.00 22.23 2 23.89 400 89.61 105.2R 119.35 12.06 2 14.07 f 500 94.24 I I4.45 119.67 2.89 2 5.22 [ 558.45 % 69 119.80 119 80 -2.46 2 0.00 i 1 ] IK71 CIIANNEI, I ~ lO 59.38 59.80 117.65 57.54 2 57.R5 l 100 87.12 92.92 118.43 24.42 2 25.51 1 4 217.3 103.66 119.14 119.03 -1.80 2 0.00 [ l f O. O O ? (

L 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 watar storage via a 8" feedline. Water is then distributed to the URI Bay Campus (including the Rhode Island !!u clea r Science Center RI!iSC) ) through a 12" main. Water is supplied to the RI!;5C through an 9" 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 hours, c ven 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. L O 15 -4v--- y ~, ,y,--- y. -,w._ .,.y,,, y ,v,,,.__ .-.my-p. _..,_.y_.9_., ,_w..,4 w,_ ,,.9

The University of Rhode t' land Graduate Sch%t t)1 Oceanograpn. ( ,I h

  • r a;rve" B ay Cam A

' afra;am:W M 72 5 92 t

  • ia ?

h 9 July 12, 1991 Mr. Eugene F. Spring, Sr. Reactor f acility Engineer Nut ear Science Center South Ferry Road Narragansett, RI 02882-1197

Subject:

Emergency Water Supply

Dear Eugene:

The chart below indicates that our water system, which includes a stand-by generator, will fullfill your cooling water requirements under various conditions. Pressure (PSI) Volume (Gal) Condition w/o Fire Pump Available Duration tiormal Town supply 4eservoir.,8 coster Pumps 70 Reservoir With Booster Pumps 70 300,000 24 Hrs, (.3 f / Town Supply With Booster Pumps 70 l Town Supply Only 30

  • = Unlimited within present demand Campus (Max) 200 GPM

-Reactor (Min) S GPfi i Total 205 GPM nnink Kenneth W. Morrill Asst. Dir. Physical Plant 1 16 .r., ,,m...

I i j i l i LOSS OF COOLANT ANALYSIS rollowiag a postulated loss of coolant accident, the pcol drain time is calcu Tted by using the falling head calculational f method, m This is considered tne maximum credible accident (refer to SAR Part A-Section XI). It is assumed that water drains frcm l the beam tube (8") from the pool surface (el.239.417) to the

  • op j

of the core box (115.916). Water then drains frc~. the 1J2" I diameter hole in the core box. Water cannot drain below the i bottom (invert) of the 9" beam tube and therefore about. cf I water remains in the core bcx above tue active fuel plate edge. l Appendix C shows the schematic and the calculations j determining the pool drain time and the flow rate to keep the core l box full. This is the minimum drain time, conservatively assuming I that the beam port shutter is up, no plugs are in the tube, and no flange cover bolted over the outlet flange. The original RINSC license amendment for 2 MW operation g 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 I L (1) Handbook of Hydraulics, Ernest F. Brater, 6th Edition, McGraw-Hill Book Company, 1976 O o

EMERGENCY CORE COOLING SYSTEH OPERATION Under normal operating conditions, make-up water is supplied to the reactor pool via the automatic make-up system. In an emergency, for a pool level drop of 2", the automatic value (NC) opens to supply water at the pool level at a design rate of 20 GPM. A manual by-pass can be opened to supply additional flow. Present operating procedures describe the procedure for piping and valve alignment and procedures for normal and emergency filling of the pool. An emergency core cooling lire 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 diractly 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 propcsed ECCS be adequate, the RINSC has conducted a water supply analysis to calculate the expected flow. The water flow to the pool has been observed in the make-up system to be about 25 GPM. (Tests will be conducted to verify the actual flow rates and pressures). The design and installation of the ECCS would be performed in accordance with the RINSC QC/QA program. O 19 I

WATER SUPPLY ANALYSIS g The analysis of the facility water supply system was performed using a computer prcgram called " Service S i:e r". 'i' It calculates pipe size ant demand. The program haa bu!!t in piping tables, valve and fitting tables and fixture unit tables. Standard computation techniques are used to ) determine losses. The program allows any fi xt ur e to Le 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, i other equivalent pipe length losses, numbers of valve and fittings and also a permitted velocity. The progran calculates pipe size, actual velocity, head loss and demand pressure. The demand cciculation includes options for flushometer units; public use or private use. Input for the calculation includes numbers and types of fixtures, a continuous demand gg 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

~ _... _ _ _ ) 'jr -- Si Z I NG CA LCU LATI ON------------------------------- -- --- P r in t e d On : 7/18/1991 r Supply Location: 1 (. i 60.0 psi, supply pressure available during demand Demand Location: 42.0 gpa demand flowing at 40.0 psi pressure 1 - - H e a d _ Lo s s D a t a - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ) -Elevation Difference: 30.0 ft (minus if demand location Icwer than supply) i Pipe-Length: 142.0 ft Other Loss In Equivalent Pipe Length; ft Number of Valves & Fittings:

Corp Stop
Curb Stop 3: Gate Valve
Globe Valv
Angle Valv
Bfly Valve
Swing Chk
Side Tee
Straight T 13:Std Elbow
Long Elbow 3:45 Elbow 1

Backflow Prev: 1.0 psi -Water Meter: poi PRV: psi Other: psi --Design Calculation------------- -------------------------------------------- Permitted Velooit/:~ fps Pipe Type: CUM Calculated Pipe Size: 2 in () _ Actual Velocity: 4.2 fps Head Loss: 17.1 psi Pres at Cemand: 42.9 poi t --DEMAND CALCULATION---------- --------------------------------- ------------- a Predominantly Flushometers: N Public Use: N --Number of Fixtures-----------------------------------~~~-------------------- r i

Bathtub
Bar Sink
Bidet
Clothes Washr
Cuepidor?
Dishwasher 1: Drinking Ftn
Hose Bib 1: Kitchen Sink
Lavatory
Laundry Tub 1: Shower-Head 1: Service Sink
Urinal Fedest 2: Urinal Wall
Urinal Tank

.: Wash l Sink

WC Flushometr 2:WC Tank Additionali fixture-units Total:

23.0 fixture units e LContinuous Demand: 25.0 gpm Fixture Demand: 17.0 gpm {} Total Demand: 42.0 epm 20 - i -.w.

.-.m i t APPENDIX A 4 f LEU THERMAt CO!3DUCTIVITY CALCULATIC!J 2 n31t'e. 21..'llS! 2 l l The densities of the dispersants are taken frcm reference (1) with the volume fraction related to the uranium density, Pu, in the fuM by: Pa = i 28Vf where Vf is the, volume fractzon of the daspersant P for the purposes fuel loading of 12.5 g/m (22 plato element, 275 g U-235) plate i the U density is 3.4682 per reference (2) the volume fraction of ')3Si2 in fuel meat is i U3Si2 J.46A2 .3068 or 30.68% = Vf = = 11.28 - From reference (3), page 11 vp-=.072 Vf -.275 Vf2 + 1.32 Vf3 therefore -Vp 072 (.3068 L -.275 (.3068) 2 + 1.32 (.3068) 3 =.0343 where Vp and Vf are volume fractions or porocity and fuel in tha meat, respectively. Ihytrm l Conductivity ofJ)3Eig [ 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 21

REFERENCES (1) R.F.

Domagala, T.D.
Wiencek, and H.R.
Tresh, "Some Properties of U-Si Alloys in the Corposition Range U3Si to U3Si2," OCMF-8410173, ANL, PERTR/TM-6, 47, July 1985.

(2) Memo from W. Woodruff (B17631 at ANLCS) to Eugene Spring i (RMA101 at URI MUS), Sept 5, 1989. (3) J.L. Snelgrove, R.F.

Domagala, G.L.
Hofman, T.C.
Wiencek, j

G.L Cc,reland, R.W. Hobbs, and R.L. Senn, "The Use of U3Si2 Dispersed in A10mfnum in Plate-type Fuel r.lements for Research and Test Redc M r.*, " Argonne National Laboratory ( AN L/ RERTR/ TM-11), October 1987, J O' l O n

t i APPENDIX B CRITICAL VELOCITY FOR FUEL FLATE DEFORMATION i l It has been shown that a critical flow velocity exists for a given plate a s sembly. m 'At this critical velocity, the j plate becomes unstable and large deflections of the plate can occur. These plate deflections can cause local overheating j of the fuel plates and possibly a complete blockage of the coolant flow. j -- M111er m derived a formula for the critical velocity based on the interaction between the changes in channel I cross-sectional areau, coolant velocitics, and pressures in two adjacent channels. For design purposes, reference (3) and (4) recommends that the coolant velocity be limited to l r 2/~4 of the critical velocit/ given by Miller,(2) therefore for a flat plate; V Critical = 2/3 15 x 10 L E tgJ-tgi.tw)U2 p. N.4 (1 -7 ) 2 where j' E =-Young's Modulus of elt.sticity, bar ; 10,4 x 106 psi /14.5 (A1) l i tp u Fuel plate thickness, cm .12' l-tm a Fuel meat P.hickness, cm .0?J i p = Density of water, kg/m3 1000.0 l 1 tw = Mater channel thickness, cm-i .381 ~ f j w a' Fuel plate width, cm 6.096 l 7= P61sson's ratio, dimensionless W .3-(A1) l-V Critical = 16.6 m/sec i TheLaverage core velocity of the 14 element LEJ core calculated for the normal primary pump-flow of 1730 gal / min (.109 m /sec) is about 3 l '. 6 m / s e c. For a nrojected 5 MW core, the velocity would be increased-to about -4 m/sec. This is well below the limiting.value of 16.'6 and therefore la not a probleni in the proposed RINSC core. 23 ,, ~,.. _... - _. _.,.. _.,. _. _ _ _ _ _.. _ _ _.. _ _ _.. _ _ _ _. - _., _ _ _ _..

REFERENCES (1) International Atomic Energy Agency, Research Reactor Core Conversion trom the Use of Highly Enriched Uranium to the Use of Lcw Enriched Uraniu:T. Fuels, Guidebool IAEA-TECDCrC-233 (1980) (2)

Miller, D.R.,

" Critical Velocities for Ccllapse of P,e a c t e r Parallel Plate Fuel Accerblier", Trans ASME, J. Eng.for Power, 82,83 (1960) (3)

Mishima, K.

and Shibata, T. "Thermall-Hydraulic Calculations for KUHFR with Reduced Enrichment Uranium fuel, "MURRI-TR-22J (1982) (4) S. McLain and J. H.

Martens, Reactor Handbook, Vol.

IV, Interscience Publishers (1964) O (5) S'.andard Handbook for Mechanical Engineers, Baumeister and Marks, 7th Edition, 1967 pg. 5-6 i l i l l l l O 24

. ~... ~. APPENDIX C I, s LOSS OF CCOLAUT. _ g, m y., '^^ - -v - --- u _ g g,gy.,. n c, f ee -c eo.. .co. I -- f,u tt.tg v2 r c.v __ %. is. tw

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-{} -1 u...., - u. i. 3 i ~~ ~ cv us.m ( 8,, e.mus ru na - v %,4 e SURFACE AREAS (FREE FLOW AREA) Area of entire pool surface 150 ft2 Area of core box 5.06 ft2 Area of core (loaded) .917 ft2 Area of 1/2 diameter hole in core box .00136 ft2 Area of 8" pipe .349 ft 7.te data elevation of 114,13 is used due to the assumption ti.at water will not drain below this elevation in the event of shear of the 8" beam tube. The amount of water remaining in the core box after draining = 114.13 - 113.213 =.917' ASSUMPTION: Gravity draining of pool from pool surfaces to the top of the core box out the B" beam tube which has no plug in place and the shutter in the up position. There is no cover flange. O 25

ccMPUTAricMAL MET C : Discharge under falling head ill -(H ) U2 ) [ (H ) U2 2A t 2 = 1 Ca (2g) U2 Datum 1,s el.114.13 (invert of bottom of 8" beam tube > H1 = 139.417 - 25.287 H2 - 115.713 - 114.13 = 1.583 C= .6 A = 150 ft2 a= .349 ft2 ( (25. 2871/2-(1. 58 3) 1/2 ) = 673.15 sec t-2x150 6x 349 (2x32.2.2)U2 ~11.219 min Orsin Tira of core Box (loaded with components) t2 - 2x.917 ( (H ) U2-(H )"# ] 1 2 Hi = 115.713 - 114.130 = 1.583 llh 11/.130 - 114.130 = 0 H = 335.6 sec/60 = 5.59 min t2 = TOTAL TIME = 16.8 min Minimum required water flow rate to keep the core box full From (1) a F= . 61 A(2 H) u2 where H = 115.713-114.13 = 1.583 9 F=.61 x.00136 (64.4 x 1.583) v2 F=.61 x.00136 x 10.097 =.008376 cfs F=.008376 113 x 7.48 gal x 60 sec = 3.76 gpm2 see ft3 (this assumes core is full) 9 26

H o**!.st Distbarge undet Fallitig llead. Ugure 44 shows 's sessel re t -, O g i.v! M I Tt-H u, -.g fWed with u nter t a depth A.. tlc time required to loact tte

r.o ns h*B d e' O dlliy' 1

r f' Ar08!.4 I E 5 6ter surf ace to a apth A is required. o is the stes o' ordee, j g r.guAv >064)S { f t d.4 is the are6 of unter sutisce for a depth s. C ts the ~ L'.S w h 3 +,. a oefheient of discLarge. The increment of time di required to }{ %y lfydg louer the u tter the ID6tatesimal .] g ju p,g y y ~% 7 dister ee O'y is g _g j ~,,ggjp syy.- r my.,- A dy W '14 dl = (4.lb) wA)}p x i Ca V.'ry g \\__ . [e, "qm 9rs-w Qw m p;;. g )ung From (4-lh if A ess h espressed 37 0 0 Hlip ~ -=.p. .g,g ppg l 4T' }*J, :hgk y r.A,.0pg.4)?p 'l in termt ei y, by epaticg t-e. 3 7,g.g 4 Q[@Q N -f tween limas Ai ann .4, the time .9,, p e,w es a,,, g.,,,g I needed to lower the noter surface M'.sa Jec hen 3'B .gJ,e,. d 3 _I i d t he dist st ce A - As eno be gotten. t i t ' l'!acitig A, = 0 gives the time of 3 U) p:s emptyitig - the veuel. Equation g) Al (415) apphes to horisontal er in. ,,.,.,.,,,.g, g F60. 44. Disthatge under Ch!'ed ottfttes pr0vided the u stit ]

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a. 02 0.04 0.0f 0.1 0.1 06 l.0 0.637 0 624 0 618 04 0.643 0.626 0.918 0.651 0 630 0.016 0 013 0 601 0 303 06 0 660 0 636 0 623 0 817 0 60s 0 Ses O M5 0,616 0 6t!

0.840 0.604 0 804 0 $90 0.3 0.6At 0 634 0 820 0.614 0 603 0 600 0 691 0 644 0.623 0.611 0 tot 0.600 0.608 0 Set 1 0,645 0 628 0.618 0.613 0 eos 0.601 0 896 7, 0 637 0 ell

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APPEllDIX D DECAY HE AT - CALCULAT10!!S (1; AS SUMPTIO!JS A. The reactor has been operating f c-r 40 hours continuously. B. The teactor scrams when the loss of ecolant sequence begins at time zero. C. Beactor pool water level reaches its icwest level as shewn in Appendix C. D. Following the LOCA, the decay heat is conducted awey to the remainittg water in the core box. E. The conduction loss will occur with the plate temperature reaching the melting point of aluminum. (2) DECAY HEAT GEtJERATIC!J Q 2 MW = 5. M 106 Btu /hr x H12/3600 sec = 6.8'l ata/ cec 14 e1 x 22 plates / element 103 From Table 5.1 at Time = 16.8 min (1008.75 sec) see E_ = .0185 Po P= .0185 x 6.187 =.114 Stu/sec s (3) HEAT CCtJDUCTIO!J DOW!J THE PLATE (FUEL SECT 10!J) It is assumed that the heat generation sine distribution The volumetric heat rate 0111 is defined a Qlll = CI:1. Omax Btu!ft3 (1) = volume J.w.t where Omax = Btu /sec and 1 = fuel plate length w = fuel plate width t = fuel plate thickness 28

For an average sine h Omax = Qave x [l/2 (N Substituting equation (2) into equation (1) we cbtain: Ou= Cave D/2 (3) l For conduction downwards (-x direction) and using the heat conduction equation from reference (1) db Alli (41 a dx2 kf kf = fuel plate thermal conductivity for a sinusoidal heat generation 0111 = Cmax sin Ox/l (5) substituting equation (3) into equation (5) we obtain 0111 = Oave U/2 sin Ux/2 (6) 1.w.t h substituting (6) into (4) d t/dx2 .Qave/l.w.t U/2 1/kf sin Ox/2 (7) 2 integrating (7) dt/dx = -U/2. Oave/l.w.t.kf (-cos Ux/1) 1/U +C1 evalenting C1 dt/dx = 0 AT x = 1 (l - top of fuel) C1 = Cave /2 w.t.kf then dt/dx = Qave/w.t.kf cos Ux/l - Cave /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 /see O 29 L.

+ f Since thl's value is less than.'the deca 2 heat generation v [- of.114-Btu /sec, it-is assumed that melting will occur. l !~ (5) DECAY TIME TO HAVE GENERATION EQUAL.TO REMOVAL The length of time that core cooling would be needed to. nave I the decay heat to reduce to.049 Btu /sec can be calculated i [ using Table 5.1. the Power Ratio =.049/6.187 =.00792 9 l 6 time = 2x104 dec = 5.56 hrs therefore emergency core cooling -is requireJ for l at least 5.56 hours. Water supp.-y can be supplied' l-for at least 24 hours. p

(6)

TOTAL' CORE COOLANT LOSS UNDER LOCA CONDITIONS 1 h, Total ~1oss would - include evaporation of the water from the l [ core box (assumed. to be nhe rate at the maximt value 1 associated-with the initial LOC at time equal to the drain 3 h time, or'16.8 min). P /Po - - '0185 6 T = 1008 see From Table 5.1 - 3 i The heate generation then .0185 x 6.87 Btu /sec 11 .'_144 Btu /sec = I-The: maximum : evaporation rate for water at-atmospheric I _ saturation (1000C) = J1144 Btu /sec x:1/970 Btu /lb 3 ft /59.8 lb/ft3 Then: .00011794 lb/sec-x 60'sec/ min x 1 i x 7,48 gal /ft3=.000851277 gal / min (liquid) l [ This.is added to the drainage loss and the cotal loss is li still about 3.76 i gpm. t i.. I 3 b 1 is IO 1: ^ I 31 {^ r m..

(4) HEAT CONDUCTICN TO THE WATER IM CCRE DOX FROM THE NON FUEL ALUMINUM IN THE ELEMENT Calculation Basis - Per Plate A. Non Fuel Plate Cross Section Plate Cross Section =.05" x 2.79" =.1395"2 .0494"2 Max Fuel Cross Section = .02" x 2.47" = .0901"2 Non Fuel Plate Cross Section =.1395 -.0494 = 1.9822"2 22 plates x.0901 = B. Side Plates of the Element Average width =.187" 22 grooves (.187.088) x.058" Cross Section .05E) ] 2 Side Plates x [ (.18 7 " x 3.045")-22 x (.87.088) x Area =.8862"2 C. Total Area for the Element Area = 1.9822 +.8862 = 2.8684 Per Plate Basis A = 2.8684/22 =.13036"2/144 =.0009054'2 Hea. Conducted from the Aluminum to the Water Q = kae A dt/dx Q = kala (Tmax-Tsat/l) Q = 131x0009x n200-212) = 89.604 Btu /hr x 1/3600 1.3' 1= (2.7) = O =.02489 Btu /sec total heat conducted = fuel + aluminum .013 +.02489 =.0J789 Btu /sec = From the original SAR it was assumed that about 30% of the heat was used in steam formation therefore .3 x.03789 =.011367 Stu/sec gl and the total heat removal =.03789 +.011367 .049 Btu /sec = 30

J.~ Tacle S.1 The Ratio, P(ts) / P of the Pission Product Decay n, O Power to Reactor Operating Power as a Function of Time, tn, After Shutdown (ANS, 1968)_ Time After Time After P wer Ratio Shutdown, t Power Ratio Shutdown, t s (seconds) P(tm) / Pn (seconds) P(tc) / Pn s 4 0.00566 1 X 10-1 0.0E75 6 X 10 1X 10 0.0625 8 0.00505 0 5 2 0.0590 1 X 10 0.00475 4 0.0552 2 0.00400 6 0.0533 4 0.00339 8 0.0512 6 0.00310 4 1 X 101 0.0500 8 0.00282 2 0.0450 1 X 106 0.00267 4 0.0396 2 0.00215 6 0.0365 4 0.00166 8 0.0346 6 0.00143 1 x to o.o:21 8 o.0012o 2 0 7 2 0.0275 1 X 10 0.00117 4 0.0235 2 0.00089 6 0.0211 4 0.00068 8 0.0196 6 0.00062 1 X 10 0.0185 8 0.00057 3 8 2 0.0157' 1 X 10 0.000550 4 0.0123 2 0.000485 6 0.0112 4 0.000415 8 0.0105 6 0.0003E0 1X 10 0.00965 8 0.000303 4 9 2 0.00795 1 X 10 0.000267 4 0.000625 O 32

. - _. _ _ _... _ _ _ _ _. _ _. _. _. ~.. 1 APPENDIX E MAX 1 MUM HEAT FLUX 3 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. Case 1 This calculation determines the maximum heat flux of.365 MW/M2 when using an axial peaking factor of 1.32. This is the case when the blades sre out of the core. The hot' spot factors cited in Section IV i are used. Case 2-This calculation determines the maximum heat flux i of 2 .424 MW/M. 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. l l I . O 33 ) .<*..wiJr -r. ~+w.. ..---..-r, .--.,.----.i....,~~.---,.r-.<vc- --e,----- ,-w-v-- 1-.,-,-w*


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.. _ _... _ _, _. _. _.... ~ _ _..... _ _ _ _ _. _ _ _. _ _ _ _ _ _. _. _ _ _. _. _ _ _ _. ~ _ _ _ _. _ _ d + q APPENDIX F MAXIMUM CORE SPECIFIC PONER i The Rhode Island Nuclear Science Center Technical Specifications Section K,3,e(2) specifies the maximum specific power. For a 14 element LEJ-core we are simply calculating I d~ maximum core-specific power as 2 MW divided by the number of fuel i elements having a maximum loading 275 g U235, i 6 wa t t s = 519. 4 8 li_ Therefore, the specific power is 27.12 t 14x275 gU235 Since burnup increases this value, this is the limiting value which cannot be reduced. O l 0 34

1 SAFETY ANALYSIS REPORT PART C TECHNICAL SPECIFICATION REVIEW AND MODIFICATION I Introduction II Appendix A/Rhode Island Nuclear Science Center Reactor Technical Specifications Appendix A to Facility License R-95 Dated July 21, 1964 Revised Through A.end:nent #16 m III Appendix B/ Proposed Rhode Island Nuclear Science Center Reactor Technical Specifications 4

Part C-Technical Specifications Review and Modification i INTRODUCTION- / \\ (V 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. 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 is 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 (Oj Science Center. Conditions outside the control of the licensee, such as key staff retirements, budget cuts, small operating staff etc., increase the difficulty and will curtail the operation of the facility during the conversion process. The Rhode Island Nuclear Science Center acknowledges the assistance of Argonne National Laboratory in the preparation of the Safety Analysis Report. 4 + 4 W 1 v i i

( APPENDIX A RHODE ISLAND NUCLEAR SCIENCE CENTE.R REACTOR TECHNICAL SPECIFICATIONS APPENDIX A TO FACILITY LICENSE R-95 DATED JULY 21 1964 REVISED THROUGH AMENDMENT #16 O

. ~ TABLE CF CCNTENTS v.y ~ PAGE- . A. ' SITE 1 1. Location-1 2. Exclusion Area 1 3. Restricted Area _ 1-4. Principal-Activities .1 Figure A.-l 2,2a 1 B. CONTAINMENT 3

1. Reactor Building 3

.C. REACTOR POOL AND PRIMARY COOLANT SYSTEM 4 1. General. 4 2. Reactor' Pool 4 3 Shielding 4 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 CORE AND CONTROL ELEMENTS 7 1. Principal Core Materials 7 2. Fuel Elements 7 3. Reflector Elements 8 L 4. Control Elements 8 5. Servo Regulating Element 8 6. Control. Element Drive 8 7 Servo Regulating Element Drive 9 8. Neutron Sources-- 9 F. REACTOR' SAFETY SYSTEMS 9 1. Modes-of-Power Operation 9 .a. Power Operation - Natural Circulation (NC) 9

b. _ _ Power Operation - Forced Circulation (FC) 9 2.

__ Design Features 10 a.. The Reactor Control System -10 b. Process Instrumentation 10 c. "Marter Switch 10 d. Power Level Selector Switch 11 e. Control Element dithdrawal 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 Desigt. Features 14 3. Other Radiation Monitoring Equipment 15 4. High Radiation Area 16 g~g LJ 1

TABLE CF CONTENTS (CONT INUED ) 17 H. FUEL STORAGE 17 1. New Fuel Storage 17 2. Irradiated Fuel Storage 17 I. EXPERIMENTAL FACILITIES 18 J. ADMINISTRATIVE AND PROCEDURAL SAFEGUARDS 19 1. Organization 19 2. Qualifications of Pesonncl 19 3. Responsibilities of Personnel 19 a. Director 20 b. Senior Reactor Operators 20 c. Reactor Operators 21 d. Health Physicist 21 4. Written Instructions and Procedures 21 5. Site Emergency Plans 22 K. OPERATING LIMITATICNS 22 1. General 23 2. Experiments 24 3. Operations 24 a. Site 24 b. Containment 25 c. Primary Coolant System 26 d. Secondary Cooling System 26 e. Reactor Core and Control Elements 29 f. Reactor Safety Systems 30 g. Waste Disposal and Reactor Monitoring Systems 31 h. Fuel Storage 31 () 4. Maintenance 0 11

A. 12 1, teratirn The resctor shall be located at the Rhode Island Nuclear Science Center on three acres of a 27-acre f o rce r military reservation, originally called Fort Eearney 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 reservaticn is centrclied by the State cf Rhode Island t hrough the University of Rhede Island. The reservaticn is in -he Tcwn of Narragansett, Rhode Island on the west shore of Narra:pnSett Bay approximately 22 miles south of Providence, Rhode Island and approximately six miles north of the entrance cf the Bay from the Atlantic Ocean. The Phode Island Nuclear Science Center and various buildings used for research, education and training purposes are located on this 27 acre campus. 2. Exclusien Ares Figure A.1 is a drawing of the Narragansett Bay Ca~ pus showing i the three acre Nuclear Science Center site. The ::cunda ry of this area shall be pos+.ed sith ccnspicuous signs to delineate the area. This three acre area shall be the exclusicn area as defined in 10 CFR 100. 4 i 3. Festricted Area Figure A.1 also shows the location of the reactor building cn the three acre area. The reactor building and attached effice 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 operaticn and utilization of the reactor. It shall be permissible to lccate 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. A.end ent 15

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3-B. C^MTAEMENT 1. Eeactor Buildinc The reactor shall be housed in a building capable of meeting the following functional requirements: In the event of an accident which could involve the release of radioactive material, the confinement building air shall be exhausted through a clean-up system and stack creating a flew of air into the building with 3 negative differential pressure between the building and the outside atmosphero. The building shall be gas tight in the sense that a negative differential pressure can be ma i rit a ined dynamically with all gas leaks occurring inward. The cenfinement 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 tne ventilation and air conditioning system intakes and exhaust, other than those which are a part of the clean-up system. No further action shall be required to establish confinement and place the clean-up system in operation. An auxiliary electrial power system shall be provided at the site to insure tne 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 4 exhaust of at least 10000 cfm frem 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 w air temperature for the control room, If it is installed with a penetration through the wall of the reactor building, it shall have a damper at this penetration which closes when an evacuation button is pressed. Upon activation, the clean-up system shall exhaust air from the reactor building through a filter and a 115 foot high stack, creating a pressure less than atmospheric pressure. The clean-up filter shall contain a roughing filter, an absolute particulate filter, a charcoal filter for removing radioiodine, and an absolute filter for removing charcoal dust which may be contaminated with radioicdine. Each absolute filter cartridge shall be individually tested and certified by the manufacturer to have an efficiency of not less than 99.97% when tested with 0.3 micron diameter dioctylphthalate smoke. The minimum removal efficiency of the charcoal filters shall be 99%, based on OBNL data and measurements performed locally, Gases from 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. Change 4 Amendment 16

. 1 C. FEACTOR POCL A'D PFOLARY O2 LA';T SYSTEM 1. C,=

  • i l The primary coolant system shall consist of the reactor pool, delay tank, heat exchanger, coolant pump, and the associated valves, piping, flow channels and sensors.

During forced cenvection 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 suspensien frame. The cociant water shall flow frcm the inlet flow c h a r.ne l downward tnrough the core to a plenum below the grid box. The coolant water -Mll then flow into tha rutlet flow channel en the cpp' a ito side 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, ^ Reaeter Prri The reactor pool shall be constructed of ordinary concrete with 1/4" thick 6061-T6 aluminum liner and shall have a volure of approxir.ately 36,300 gal. 3. ShielM 2 The reactor pool and primary system shielding shall be adequate to meet the applicable personnel radiaticn protection requirements of 10 CFR 20. 4. Eri-=rv enclant system The primary coolant system shall conform to the following: a. Heat Exchance; The heat exchanger shall be designed to remove he n at the rate generated by the reactor at maximum licensed steady state power frcm the primary water and shall be designed to perform under the maximum primary system operating. temperature and p re s s u re. Replacement heat exchanger shell and tube bundles shall be constructed from stainless steel according to the requirements of Section III, Class C of the AS.VI Boiler and Pressure Vessel Code.

b. Prim ry Puq Number of pumps 1

Type Horizontal mounted, Centrifugal, Single Suction O Change 3, 4

. Materials of construction Worthite Rating 1500 gpm Head 59 feet Design Pressure 75 psig minimum Design Temperature 1500F minimum Motor Type Drip proof, induction, 440 v, 3-phase, 60 cycle c. 221Ay Tath Number of tanks 1 Material of constructicn Aluminum Association Alloy 5083 and 5096 Material Thickness Walls 0.25 inch Dished Heads 0.375 inch Capacity 3000 gal., minimum d. Pri-wry Perircu1At en Piring i Material and thickness Sch. 40 A1. type 3003 aluminum size 8 and 10 inch Design temperature 1500F, minimum Design pressure 100 psig, minimum e. Make-uo 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 sourco Potable water from city main ~ Make-up demineralizer type Mixed-bed single shell, regenerative Make-up demineralizer capacity Norma 1 25 gpm Emergency 50 gpm Water sof tener capacity Normal 50 gpm f. cleanuo system for pr4~arv ceolant water Cleanup pump Capacity 40 gpm Head 100 ft Cleanup demineralizer Type Mixed-bed, single shell, regenerative Cleanup demineralizer capacity Normal 40 gpm Emergency 50 gpm Change 1

c_ _ _ _ _a_s _ _e v_ _e _ e_.v. ,w n. _c e r_ m_m_.a.n.y 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 compcsed of the heat exchanger, cooling towers, pumps and associated valves, piping and sensors. In this s '/ s t e m, water tlows frem the heat exchanger through a control valve to the eccling towers. From the ccoling tower casins, the water is then pumped hack to the heat exchanger. Change 4 9 e

E. PEAOTOR CORE AND CONTROL E'.EMSCji V.- The reactor core and control elements shall.have the following characteristics and nominal dimensions: .' l. E;in:: ipsi Core Materials Fu-1 matrix Alloy, UAlx,U309 -U-235 enrichment Approximstely 93t ruel clad 1100 and/or 6061 aluminum Fuel element-side plates 6061 aluminum cnd fittings 356-76 or 6061 aluminum P.oderator dater Reflector AGOT grade (or equivalent) graphite and/or water Mixture of B C and aluminum, Control elements 4 clad with aluminum Mixture of B C and aluminum. Servo Element 4 clad with aluminum. L 2. Fuel ElementM l. 2.8 inches l' Plate width overall l 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 I l Water gap between plates 0.1 inch l Amendmen* 8,11

-g. Number of plates per fuel element IS U-235 per fuel element 124 grams, nominal Overall fuel element dimer.s io n s 3 in x 3 in. x 40 in. 3. Failrur 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. C or.t r n l Elementa width 10.6 in. Thickness 0.38 in. Overall 1 ,th 54.1 in. Active length 52.1 in. 5. Servo Regulatirc Elamant Shape Square boral twLe Width 2.1 in. Overall length 28.9 in. Active 24.9 in. 6. Centrol Flament Drive Type Electromechanical screw Drive to safety element Electromagnet connection Stroke 32 in, maximum 9 Amendment 11

9 j O( j ervs Peguladng rierant Drive 7. c Type Electromechanical screw Drive to eternent connection f r. ; ': --- w (no scre.m) Stroke 25 An. ma ximu:. Position indication accuracy 0.02 in. 9. -N utren scurres Start-up Source Number 2 Type Plutonium-beryllium Unit Source Strength 1: 106 neutrons /sec, minimum Maximum Pcwer Level with Plutonium-beryllium sources installed 10 ra Operational Source Number 1 'f\\ Type Ant, mony-beryllium Scurce Strength 2 x 106 neutrons /sec. minimum F. PEACTOR SAFETY SYeTEMe 1. ti;;de s cf Power Oceration There shall be two modes of power operation: a, Power Ocerari+n - Natural Circulation HJC) Power operation NC shall ce 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 oceratien - rerced circulation frei 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 PC operation. J I l Change 4

~ ~..- , 2. Desien ren tu res a. -The Reacter Centrol Sy m The reactor safety system shall consist of sensing devices and. associated circuits which automatically sound an alarm and/or produce e reactor scram. The systems shall be designed cn the tail-safe principle (de-energizing shall cause' a scram). Table F.l and F.2 describe the arrangement and req 2ireeents of the Jefety system, b. Process In t rurent at ien Process 'instrumentat ion with readout in the contrn roem shall be provided to permit measurement of the flow rate, temperature, and. conductivity of the primary coolar.t and the flow rate of the secondary coolant. In additton, a second primary flow indicating device with readout in the control room shall be located between the reactor (utist plenum and the reactor outlet header. After normal working hours, an independent protection system, separate from the system described in Section K.3.a, shall be used to monitor certain items in the reactor building and alarm in the event of an abnorral condition. The alarm channels provided are: n (1) A fire in the reactor room, u (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. M3?ter Switch A key lock master switch shall be provided with three positionar "off", " test", and "on". These positions shall i: have the following functions: (1) The "off" position shall de-energize the reactor control circuit.

o

. (2) The " test" position shall energize the reactor control circuit exclusive of the centrol blade magnets. (3) The "on" position shall energire the reactor centrol circuit including the control blade magnets. d. Prwer Level Eelector Ewitch A power level selector switch shall be provided witn f ur I positions: "0.1 MW", "1 MW", "2 MW", and "5 MW" These positons shall have the follcuing functions: (1) The "O 1 MW" position shall activate all safety system sen.ars except thosr indicated in Table F.1. (2) The "1 MW" and "2 MW pcsitions shall activate all safety system sensors. (3) The "5 MW" position shall scram the reactor. e. Ccntrol Ela~ ant Withdrawil Int e rlo ck s Interlocks shall prevent control rod withdrawal unless all of the following conditons exist: (1) The master switch is in the "on" position, (2) The safecy system has been reset, (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 mc. a than one control element at a time. f. Servs Svstem Control 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 control system shall be designed so that innediately following a scram the Servo control shall automatically return to the manual mode of ope ra tior. Change 4 O

. TA9tr r.1 PEATTOR EAFETY SYSTEM )- Sensor or Trip Device No. of1 Switches Trip Set Ala rm Set or Sensors Point Point Short Period. 1 3 sec, min. 7 sec. min. ,High Neutron Flux 2 Max, of 130% of 110% mex. full scale with a 2.6.!W max. s High Temperature of Primary 1130F max. Coolant ~ Entering-Core During Forced Convection Cooling

  • High Terperature cf Primary 125cr max.

123CF max. Coclant Leaving Core During Forced. Convection Cocling* Low Flow Rate of Primary 1 1200 gpm, 1350 gym, Coolant

  • min.

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

  • 1 X

X Coolant Gates Open* 1 per gats X X Neutron Detector Iligh 1 per Decrease of _ Voltage Failure in Linear power 50 volts max. Level Safety Channels supply Manual Scram;(Switch.at 2 X X l bridge ~and on console)' High Conductivity of 1 Equivalent Primary Coolant to 2pmho/cm at 250C, max. Safety Blade Disengaged 1 X Log'N - Period Amplifier ~ 1 X X . Failure PegLlating Rod at.Either 1 X Limit of Travel Low Flow Rate of Secondary 1 800 gpm, Coolant

  • min.

Bridge Movement 1 X X ~No Flow Thermal Column

  • 1 X

X f \\-

  • These functions are. bypassed when the Power Level Selector Switch is in the "O.1 MW" position.

Change 3,4,5 i

i { ~ ~,r- ;' TA1312 F _2, REACTOR MUCINAR INSTRUMENTATION Channel Detector. Sensitivity. Range Information Information Information Recorded to to to information . Operator Logic Element Servo System (Scram)' Retractable ' Neutrons-Source-. Neutron Relative gas filled approximately-level Flux power level i l Start-up B-10 filled l 12 counts /nv to full None None on log proportional power scale Neutrons-Source Power level Power level Log N Fixed fission approximately level to Period Period scram None log scale counter 7 cps /nv 3x106 and period watts l l Linear Compensated Neutrons-I watt Power. 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 i 1

14 r' G. WASTE DISPOSAL A_W FACILITY M

  • N I T O RI NC-SYSTEMS 1,

Waste Disresal Syster* Design Features 3 LISuid FadiSactive Waste DM EC331 Sy3 tem All liquid uaste (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. Gaseous Radienrtive W3ste rispes31 System All gaseous radioactive waste from the beam pcrts, thermal

column, pneumatic irradiaion system and all other radioactive gas exhaust points associated with the reacter itself shall be collected in a ma.ifold and discharged to the reactor stack through an absolute filter, blower and damper.

c. Salid Endio*ctive Waste Ster 3?e Solid Radioactive wastes shall either be stored in radioactive waste storage containers located within the reactor building or removed from the site by a ecmmercial licensed organization. 2. Area and Exhaume nas Mrniter nemien reatures s)' a. Three fixed gamma monitors employing suitable detectors shall be e:r. ployed in the reactor building. Each of these shall have the following characteristics: 1) A range consistent with the expected radiation levels in the area to be monitored (0.01 to 10 mr/hr, 0.1 to 100 mr/hr, or 1 to 1,000 mr/hr). 2) A radiation dose rate output indicated in the control room. 3) An adjustable high radiation alarm which shall be a aunicated in the control room. Amendment 12 Change 2,3

. l 4) The three fixed gamma monitors shall be lo ca'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 fer new fuel elements, and at the reactor bridge. b. A gamma monitor 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 .,ys t ems. The characteristics of these monitors shall be as stated in e. 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 consistent with expected radiation levels in the area being monitored (0 to 10 mr/hr or 0 to 50 mr/hr). 2) A radiation dose rate output at the instrument. 3) An adjustable high radiation alarm to alarm at the instrument and create both an audible and visual signal, d. A stack exhaust gas monitor system shall be provided which draws a representative sample of air from the exhaust gas. The monitor with indicators and alarms in the control room, shall have the followi., characterics: 1) A beta parciculate 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. Othar RaMintien Meniterins Ecuirment a. Portable survey instruments for measuring beta-gamma dose rates in the range frem.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/em2 sec shall also be available to the facility, b. Reactor excursion monitors shall be placed in the facility for measuring gamma and neutron doses in the event of an accident. Amendment 5 l ~ c. A radiation monitor shall be provided to monitor all g persons leaving the re sct e r room for tota-gamma W c o nt a. Tin a t ion, 4, }Jigh RndiatimJ A*ei Ouring reactor cperation, the dose rate fr:m the delay tank may be in excess of 100 millirem per hour. On three sides, the tank shall he shielded. On the fourth side, the tank is shielded using a "ma:e" so that a c c a:3 3 to the tank is posssible through a dr ar equipped with a lock 9

...-_______...~m_. . H. FUEL STOFAGE 1. New Fuel St m *;A 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 r gg shall be less than 0.S. e 2, Trtidiated Fue.i_e* m un Two types of irradiated fuel element storage racks shall be provided. One type of rack shall contain spaces for nine fuel assemblies and shall have approximate over-all dimensions of 35.5 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 tacks of 12 it.ches. This 18 fuel asserbly 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 irradi.ited fuel storage racks shall have a maximum Ke g g of 0.8 for all conditons of moderation possible at the site. Storage spaces shall be provided for at least 36 f ui.; assemblies. l 1. FYPERIMENTAL FAEILITIES The permanent expe rimental facilities shall consist of the following: 1. Thermal column. 2. Beam ports; two 8 inch dia, ar.d four 6 dia. 3. A six inch diameter through port. 4. Radiation baskets, i l 5. A two-tube pneumatic tube system. 6. Dry gamma cave. l f I l (.- \\., I

-1g-J, AtMI?!!ETP3t?IYr AL'S r RC T EE L'P11L _ E Ar rTJ Apr s 1,

ALLSLLcn The Rhode Island Atomic Energy Cer.ission (RIAEC) shall have the responsibility f or the safe crbraticn of the reacter.

The RIAEC shall appoint a Director of Cporations and a Femetet i't ilization Cerrittee cerisisting of a mir.irrun o f five corte.3, as !alicws: (1) The Director of Operatiens (0) T!.e Reactor Escility health I hysicist (3) A qualified representative frem the faculty of Dr:wn University (4) A qualified representative from the faculty of Providence College (5) A qualified representative from the faculty of the University of Rhode Island. 1 A qualified alternate may serve in lieu of one of the abcve. The Director and Health Physicist are not eligible for chairmanship of the C o rami t t c e. The Reactor Utilitation Committee shall have the following functions: a. Review proposals for the use of the reactor considering the suitability of the reactor for the propoced use and the safety factors involved. 5

.__ -.. _.m b. Approve or disapprove preposed use of the teactor. 1 c. Review at least annually the operating and emergency procedures and the overall radiation safety aspects of the facility. The Reactor Utilization Committee shall maintain a titten record Of its findings regarding the above. t 2. Oud ificatiens e t ro rnrmel a. The Director of Cperations shall have at least a bachelers degree in one of the physical sciences or engineering, and be shall be trained in reactor technology and be 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. Tha reactor operators and senior operators shall be licensed in accordance with the provisions of 10 CTR 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 alternated appointed by the RIAEC. 3. Pescensibilities of Persennel a. Director (1) The Director shall have responsibility for all activities in the reactor facility which may affect reactor operations or involve radiation hazards, including controlling the admission of personnel to the building. This responsibility shall encompass administrative control of all experiments being performed in the facility including those of outside agencies. (2) It shall be the responsibility of_ design the Director to insure that all proposed experiments, 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

. b. geMer Rearter Geraters (1) A licensed senior reactor cperator shall be assigned each shift and be responsible for all activ3cies during his shift which may affect reactor cperatxon or involve radiation hazards. The reactor operators on duty shall be responsible directly to the senior operator. (1) The reactor operations which affect core reactivity shall not be performed without the senior eperator on duty et.eadily available on call. The senior cperator ahall be present at the facility during initial startup and approach to

power, recovery from an unpaanned or unscheduled shutdown or significant reduction in power, and refueling.

The name of the person serving as senior eperator as well as the tire he assumes the duty shall be entered in the reactor log. Wnen the senior operator is relieved, he shall turn the operation dutie, over to another licensed senior operator. In such instances, the change of duty shall te 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 crerators (1) The responsible senior operator shall designate for his shift a licensed operator (hereafter called " operator") who shall have priraary responsibility under the senior operator for the operation of the reactor and all associated control and safety

devices, the proper functioning;of which is essential to the safety of the reactor or personnel in the facility.

The operator shall be responsible directly to the senior operator. (2) Only one operator shall have the above duty at any given time. Each operator shall enter in the reactor log the date and time he assumed duty. (3) When op6 rations are performed which may affect core reactivity a licensed operator shall be stationeu in the Control room. When it is nicessary 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 bt a licensed operator. In such instances, the chaage 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 cperator en duty, shall be responsible for the operatien of 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 teactor without reference to higher authority, when in his judgement such action appears advisable er necessary for the safety of the reacter, related equipment, or persennel. Any person working on the reactor bridge shall Le similarly authorized to scram the reacter by pressing a scram butten located on the bridge, i. Hesit h Physicist ~ The Health Physicist shall be responsible for assuring that adequate radiation montoring and control are in effect to prevent undue exposure of individuals to radiation. 4. Written Lut r u ct i e s as PreeMures 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 menber of the staff O shall be familiar with those procedures and instructions for which he has responsibility. 5. Eite ru rgenev Pinns The Rhode Island Nuclear Science Center shall have available the services of other state agencies for dealing with certain types of emergencies. The RIAEC shall enter into an 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 at.nosphere. The emergency vehicle shall contain equipment such as portable radiation monitors, respirators, and a particulate air sampler. Communications using the statewide emergency network shall be available. Personnel of the Civil Defense Agency and of local fire departments shall have received training frem 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

o 4 K. grigTMr; L!M!nTICMS 1. Seneral O The foll, wing administrative controls shall be employed t3 assure the safe creration of the facility: a. The reme.or shall not be cretated whenever there are any significar.- defects in fuel elements, c;ntrol rods, or control circuitry. b. The reactor centrol and safety system must t' e turned en and functioning preperly and ar apprcpriate neutren Scurce must be in the core during any change which can affect core reactivity. c. During cperations which could affect core reactivity, a licensed cporator shall be stationed in the control re:m. Communications between the control roce and the senicz reactor operator directing the eperation 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 cf tne scram or power dec; ease 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 u s i ri g 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. 9 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 made 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 cores whose nuclear properties have been previously determined from critical experiments shall be as follows: l i All core loading changes shall be performed under the j supervision of a person having a senior operator's license. During the operation there shall be present in addition to the desig7ated senior reactor operator at least one other technically qualified person who shall act as an observer. Change 6 l l l

l -:3-The exact procedure to be followed for a l particular reloading eperaticn will be detetmined by the cbserver and the senior resctor operator in charge of the operation before the cperatien begins. Each step in the procedures shall be considered by both persons, and no step shall te taken withaut the concurrence of both. 2. M erim nts a. " Experiments" as used in this section shall t:0 censtrued as any apparatus or device installed in the corc regicn which is not a compenent of the core. b. The Reactor Utilization C e rr.'ni t t e e shall review and approve all experiments tefore initial performance at the faci)1ty. 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 Peactor Utilization Ccmmittee. The docurnentation shall include at least: (1) The purpose of the experiment, (2) A description of the experiment, and (3) An analysis of the possible hazards associated with the performance of the experiment, c. All use of experimental facilities shall be approved Ly the Director of Operations, d. Tha 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 ta the core simultaneously, their combined reactivity shall not exceed 0.006, e. The calculated reactivity worth of any single independent experiment not rigidly fixed in place shall not exceed 0.0008. If such experiments are connected or otherwise related so that their ccmbined reactivity could be added tu the core simultaneously, their combiaed reactivity worth shall not exceed 0.0008. f. No experiment shall be installed in the reactor in such a manner that it could shadow the nuclear instrumentation system monitors and thereby give erroneous or unreliable information to the control system safety circuits. g. No experiment shall be installed in the reactor in such a manner that it could fail so as to interfere with the insertion of a reactor control element.

0 S -:4-h. No experiment shall be performed involv'ng materials used in such a way that they might credibly result in an explosin. i. No experiment shall be performed involving materials which could credibly contaminate the reactor pool causing corrosivo 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 cf the core. 3. Quanti:ns a. Site Control of access to the reactor facility shall be the responsibility of the Dirnetor of Operations, b. Centafn-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 opened during reactor operation. If a door is to remain open, an indivi-tual from the reactor operations staff is continuously in attendance at the door. b. The building clean-up system is operable. (2) Require-ents for Petest of Ernfirement (a) Method of Retggt 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 conditoning and normal ventilation has turned off. 3. Dampers on all ventilating ducts leading to the outside have closed. 4. Building cleanup system-air scrubber and fresh air blower come on. Change 4

4. , b 5. The negative differential pressure U between the inside a t.d outside of the building is at least 0.5 inches of water. This shall be determined by reading the differential manometer located in the contcol room. (b) Fremiency of Retest The building cleanup 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 from the reactor building and passing through the charcoal scrubber. The remaining - air will be provided by a separate blower from an uncontaminated source. This shall create a pressure in the building which is equivalent to at least 0.5 inch of water below atmospheric pressure. c. Primarv ceeinnt system (1) The minimum depth of water above the top of the active core shall be 23 feet. (2) No piping shall be placed in the pool which could cause or fail so as to cause a siphon of the pool water to below the level of the ten inch coolant lino penetrations. (3) MAkeur; Syst em The effluent water of the primary coolant water makeup system shall be of a quality to insure compliance with K 3.c.($) and (6) below. (4) cleanuo 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) celow. (5) The primary coolant shall be sampled at a minimum frequency of once per week and the samples snalyzed for gross radioactivity, pH, and conductivity in accordance with written procedures. Corrective action shall be taken to avoid exceeding the limits listed below Amendment 10 O Change 4 ~.

-:i-pH 5.5 to 7.5 conductivity 2 nmho/cm (6) The radioactive materials contained in the pool water and in the prima ry 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 eperation, the primary coolant flow rate shall not be less than 1200 gpm. During determinations et reactor power by coolant heat balances, the coolant flow rate may be reduced to 600 gpm providing all other aspects of these Technical Specifications are met. d. Accendary Csolin? System (in TLe 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 exceedin'; the pH limit given below: pH 5.5 to 9 (2) The concentration of radionuclidos 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 eperated using forced convection cooling. (3) If the radioactive materials contained in the secondary coolant exceed a radionuclide concentration in excess of the values in 10 CFR 20, Appendix B, Table I, Column II, above background, the reactor shall be Jhutdown and the condition corrected before operation using the secondt;y cooling system resumes. (4) The secondary coolant system shall be placed in operation as required during power operation utilizing forced convection in order to maintain a primary coolant core outles temperature of 1250F or below. e. Reactor Corn and Control Elements (1) The reactor shall not contain in excess of 35 fuel elements. There shall be a minimum of four operable control elements. Amerdment 6,14 Change 3,4,7

i (2) The limiting theraal and hydraulic core characteristics based on a 28 element, graphite reflected core are specified below: (a) Maximum Heat Flux 47,000 BTU /hr ft2 (b) Maximum Core Specific rewer 1,120 watt /gm U235 (c) Maximum Tuel Surface 1970F Temperature (d) Coolant Velocity during 2.65 ft/sec, min. Forced Ccnvection Cooling (e) Coolant Inlet Temperature 1150r max. (f) Average Coolant Temperature 1000 max. Rise (g) Primary System Bulk outlet 125Cr max. Coolant Temperature (h) Temperature Margin in ?rimary 430F Coolant (Tsat-Tsurft (1) Number of Coolant Passes 1 Through Core (3) Principal __Wclear Characteristics of th _.C0re C (a) Corn and Centrol svatem pe3ctlyity gg 1. The reactor shall be subcritical by at least in Ak/k from the cold, Xe-free, critical condition with the most reactive control element and the { Porvo egulating element fully withdrawo. 2.

  • he maximum worth of the servo regulating element shall be 0.7%

Ak/k. (b) Maximum Reactivity Addition Rato - Ak/k/sec 1. By servo regulating element maximum of 0.0002 2. Manual by control element maximum of 0.0002 O Amendment 13 Change 4

. (c) Reactivity Ecefficients 1. Temperature coefficient approximately -0.5 x 10-4/oC (calculated) 2. Void coefficient approximately (core average) -1.9 x 10-3 4 void / (calculated) (4) Princiral cere ^reratine Li mi tathns (a) M a x i-m Prol *e~rerature L4-!tatima The pool water temperature shall not exceed 1300r. (b) Reactivitv t!mitatiena 1. Excesa Reactivity The cold, clean excess reactivity for any core used in the reactor shall not exceed 0.047 2. Mint-um shutdewn Margii_ All reactor cores used shall be such that they would be suberitical if any single control element and the servo regulating element were withdrawn. (c) Reactivity coefficient L!~!tation The reactor power coefficent (as inferred by the control rod movements required to compensate for changes in power) shall be negative. (d) centrol Element Drive Performance Re ma i ro-o nt a All control element drives shall meet the following specifications: 1. The control drive withdrawal rate shall not be more than 3.6 inches per minute. 2. For the electronic scram system, the time from initiation of a scram condition until control element release shall not exceed 100 milliseconds. 3. The time from initiation of a scram condition until the control element is fully inserted shall not exceed 900 milliseconds. 4. It shall be demonstrated at least every 3 months that the above specifications are met. Change 4,7 w

.;9 (e) Ferve Rem 213 ting Element Srive Performance Pe111 rom nts If in use during operation, the servo regulating O element drive shall meet the following specificatiens: 1. The drive withdrawal rate shall not be more than 78 inches per minute. 2. It shall be demonstrated at least once per :* n '. h that the above specification is met. (f) rissirn egnalty timit The fission density limit for

alloy, uranium aluminide, and uranium oxide fuel shall meet the following speci'ications:

1. The fission density limit shall be 0.5 x 1021 fissions /cc. 2. The fission density of all fuel elements wh2:h have burnup shall be calculated at least quarterly. f. nearter safety E w.m (1) The reactor safety system shall be operable during all reactor operation. The safety system shall be checked cut before each start-up and functionally tested for calibraticn at least monthly. (2) It shall be 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 second with a signal to noise ratio at least 3 t o 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) rollowing 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 Providence, Rhode Island. The central station shall be O provided with written instructions on the steps to be taken following an alarm. Amendment 8t' Change 4

. g. Waste Sisrcsai nM Penet,rr M niteri~2 systam l (1) The liquid waste retention tant discharge shall flow to a monitor staticn in the reactor building where the effluent shall be batch sart. pled and the gross activity per unit vo l utte determined before release. All off-site releases shall be directly into Narragansett Dsy. (2) Gaseous radioactive waste shall be disposed of sing the reactor stack. Disposal limits shall Ocnfctm to the f ollowing tatsle. In this table, the MPC stated is for individual isotopes and mixtures contained in Column 1, Table II, Appendix B of 10 CFR 20, 1 2 Type of Activity Maximum Curies Curies per second to per second to t be released averaged released over one year Particulate Matter and Halogens with half-lives 140 X MPC (ue/cc) 14 X MPC (ue/cc) longer than 9 days All other Radioactive 105 X MPC (uc/cc) 104 X MPC (uc/cc) Isotopes 0 (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 organization. (4) The exhaust gas monitor shull be calibrated to alarm at an instantaneous release rate which instantaneously exceeds the limits stated in Colamn 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 de'.ection equipment. Adopate spare parts shall be on Sand to allow necessary repairs to be made during the maintenance or calibration outages of the monitors. O Amendment 12

. 0 (6) The area and the primary and secondary coolant system monitors shall be adjusted to alarm at a maximum l reading of 2 mr/hr or 200% of the normal radiation levels in their area, whichever is larger. ( "F ) The door which controls entrance to the " mare" leading to the delay tank shall be locked with the key in the possession of the Director or a licensed senior operator. Entrance to the delay tank high radiation area shall require the presence of the Hemith l direct reading portable radiation monitoring Physicist or a licensed senior operator and the use of e quipment. h. Egel Stegggg (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 the criticality ufe storage racks described in Section H. Only one fuel assembly may be i inserted or moved from a storage rack at a time. O' (3) Safety against inadvertent criticality shall be I 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 tack with a sheet of aluminum covered cadmium, 4, Maintenana -(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 start-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. l L l LOL

,Part C Technical S}'ecifications Peview and Modif i at.len 111T RO DUC TI O11 i There are numerous Technical Specification chanaes required as a result of the use of the LEU fuel in t he Rhode Island !?uclear Science Center :eactor. Parts A and B of the Safety Analysin Report tuuch on many of them. As a result of the Rhode Island I;uclear Science Center review process, additicnal changes which reflect current conditicns or clarifications cf sone Technical Specificaticn sections are alao included in the final Technical Specification versien. Appendix A is a copy of the Rhode Island !;u c le a r Science Center current Technical Specifications. Appendix B is a copy of the Technical Specifications wit h the changes included as a ter.sult of the SAR and review process. The double vert ical lines adjacent to a section designates the nection which has the proposed changes. Implementation of the final approved Safety Analysia Peport will be a difficult task for the Rhode Island 13uclear Science Center. Conditions outside the control of the licensee, such as key staff retirements, budget cuta, small operating staff etc., increase the difficulty and will curtail the operation of the facility during the conversion process. The Rhode Island I;u c le a r Science Center acknowledges the assistance of Argonne t;ational Laboratory in the prepa ration of the Saf ety Analysis Report. O

I l l i c. APPENDIX B i i ? o PROPOSED RHODE ISLA!!D NUCLEAR SCIE!1CE CENTER REACTOR TECHNICAL SPECII'ICATIO!iS l y s l t b ? r O

TABLE CF CONTENT 3 PAGE _O A. SITE 1 1. Location 1 2. Exclusion Area 1 3. Pestricted Area 1 4. Principal Activities 1 Figure A.1 2,2a B. CCNTAINMENT 3 1.Reacter Building 3 C. REACTCR POOL'AND PRIMARY COOLANT SYSTEM 4' 1. Gencral 4 2. Reactor Pool 4 3. Shielding 4 4. Primary Coolant System 4 a. Kent Exchanger 4 b. Primary Pump 4 c. Delay Tank 5 d., Prietry Recirculation Piping 5 5 e. Make-up System f. Clean-up System for Ptimary Coolant System 5 D. SECONDARY COOLANT CYSTEM 6 E. REACTOR CCP,E AND CONTROL ELEMENTS 7 1. Principal Core Materials 7 2. Fuel Elements 7 3. Reflector Elements S 0 4. Control Clements 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 (TC) 9 2. Design 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 O. 1 L

TABLE OF CONTENTS (CONTINUED) H. FUEL STORAGE 17 1. New Fuel Storage 17 2. Irradiated Fuel Storage 17 I. EXPERIMENTAL FACILITIES 17 J. ADMINISTRATIVE AND PROCEDURAL SAFEGUARDS 18 1. Organization 18 2. Cualifications of Personnel 19 3. Responsibilities of Personnel 19 a. Director 19 b. Senior Reactor Operators 20 c. Reactor Operators 20 d. Health Physicist 21 4. Written Instructions and Procedures 21 5. Site Emergency Plans 21 K.- CPERATING-LIM 1TATIONS 22 .1. General 22 2. Experiments 23 3. Operations 24 a. Site 24 b. Containment 24 c. Primary Coolant System 25 d. Secondary Cooling Systcm 26 'e e. Reactor Core and Control Elements 26 f. Reactor Safety Systems 29 g. Waste Disposal and Reactor Monitoring Systerns 10 h. _ Fuel Storage 31 4. Maintenance 31 O I O 11 bi i i

A. fitg 1-L11AL12u The reactor shall be 1ccated at the Phede Island Nuclear S e ie r.c e Center on three acres of a 27-acre forror military reservaticn, origin, illy called Fort Fearney and new calle.d the Narragansett Pay Carpus of the University of Phode Taland. Tae "niversity of Phede Island is a state agency. The 27-acre reservaticn is centrolled by th+t State of Rhode I s l a r.d thr.qh the "niversity of Rhode Island. The reservation is in the Tcan Of Narragansett, Phode Island on the west shore of Narragsnsett Day appteximately 22 niles south of Frevidence, Ph:de 131and and approximately six miles north of the entrance of the Bay from the Atlantic Ocean. The Phede 1mland Nuclear !. c i e nc e Center and various buildings used for research, education t r. d training purposes are located on this 27-acre carpas. 2. EyciusE n Area Figure A.1 is a draw.tng of the Narragansett Bay Cattpus showing the three scre NucJear Science Center site. The bcundary of this area shall be pot]ted with conspicaous signs to delineate the area. This three acre area shall be the exclusicn area as defined 4.n 10 CFR 100. 3. Fm tricted Arga Figure A.1 also shows the lo:ation cf the reactor butiding on the three acre aro&. The reactor bailding and attached ef fice laboratory wing t.h e ll be considered the restricted area as 4 defined in 10 CFR 20. 4. E.1r.c i n i A;;.1vitleg The princips1 ac*.tvities carried on within the Instricted and exclusion area shall be those associated wit h operation and b 3 utilization of the reactoc. It shall be permissAtle to locate additional E lear Science Cente.r or University of Rhode I31and 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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1 .. B. cetTMnMENT r 1. peneter Building ? The reactor shall be housed in a building cepable of meeting the following f unctional requi re rents: i 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 building with a negative differential pressure between the building and the outside atmosphere. The building i shall be gas tight in the sense that a negative i differential pressure can be maintained dynamically with all F?s leaks occurring inward. The continement and clean-up systems shall become operative when a building evacuation button is pressed. This action shall: (1). turn off all ventilation fanc and the air conditioner system and (2) close the dampers on the ventilation and air conditioning system intakes and exhaust, other than those which are a part of the clean-up-system. No further action shall be required to establist afinement 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 leact 4000 cfm, operates in conjunction with additional exhaust blowc r (3 ) which provide an additional exhaust of at leaar 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 1: at least 14000 cfm. During normal operation, the building is at a pressure somewhat belong i atmospheric. The control room air conditioner shall be a self-contailed unit, thermestatically controlled, providing constant air temperature for the control room. If it is. installed with a penetration through the wall of the reactor tuilding, it =shall have:a damper at this penetration which closes when an evacuation button is pressed. Upon activation, the clean-up system shall oxhaust air from the . reactor building through la.fi'.ter'and a 115 foot high stack, l creating a preesure less than atmospheric pressure. The clean-l up-filter shall ; contain' a roughing filter, an absolute l-particulate filter,.a charcoal filter for removing radiciodine, ( srd an absolute filter for removing charcoal dust which may be - ntam.iqaud with radioiodin3 Each absolute filter cartridge l n oil te individually twsted and certified by the manuf acturet ,.have an efficiency of-not less than 99.97% when tested with .3 micron. diameter diocty1phthalate smoke. The minimum cemoval of ficiency of the charcoal filters shall be 994, based on ORNL data and measorcments performed locally. Gaues.-f rom the beam ports, the rmal - column, pneumatic system, and all other radioactive gas exhaust points shall be exhausted to-the stack through a roughing and absolute filter system. k .a, ......_._.,b. .m.. ,_-+_,_-...m_.:.wm ,.S i _,. 4,A. . m.~r ,. c

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GranGal The primaty ccolant system snall corsist of the reactJt p.w l, delay

-k, heat exchanger, ccolant purp, an3 t f.e aS8ociattd va.vos, piping, flow channels and sensors.

Curirg fcceed convectien ecoling, ccolant w a t re t shall Lo sqrlied to the are L' y an aluminum lino connected to the inlet fl;w :- h a n n e l wh::h is en one side of the suspension frame. The - clant uter shall flow from the inlet f';w channel dcwnward through +he core to a pienum below the grid box. The

c'. ant water shall then flow into the outlet ficw channel en the cpposite side ;f the suspension frame and then through a discharge.ine to
  • Le delay tank, coolant pu p, heat exchanger and then return t:

ne coolant inlet line. 2. D m 'a r Peel The reactor pool shall be constructed of ordinary con: rete with 1/4" thick 6061-T6 aluminum liner and shall have a /clume ,f approximately 36,300 gal. 3. Etielding The reactor pool and primary system shielding shall be adequate to

meet, the applicable personnel radiacion p r o i. -

. ion requirements of 10 CFR 20. l' r 1:.a ry ce e 1 a nt Ryft.try., The primary coulant system shall conform to the following: a. Heat rygg The hear exchanger shall be designed to remove heat at the rate generated by the tractor at maximum licensed steady state power from the primary water and shall be designed to perform under the maxirnum primary system operating temperature and pressure. Replac ernent heat exchanger shall and tube bundles shall be constructed from stainless steel according to the requirements of Section III, Class C of the ASME Doller and Pressuro Vessel Code,

b. Prim rv Pq Nusber of pumps 1

Type Horizontal mounted, Centrifugal, Single Suction O

.-___.._____.____m_..___....._..._. 5-i O r.aterials of co nst ruct. ion Wortnite Rating 1500 gpro Head 59 feet Design Pressure 75 psig minimum Design Temperature 1500F minimun Motor Type Drip proof, induction, 440 v, 3-phase, 00 cycle i c. Delay _lanh i Number of tanks 1 i Material of construction Aluminum Association Alloy 5083 and 5086 Material Thickness Walls 0.25 inch Dished Heads 0.375 inch capacity 3000 gal., minimum d. Sr N rv Regirculatien Pipin2 Material and thickness Sch. 40 A1. typo 3003 aluminum Size 6 and 10 inch o Design temperature 150er, minimum Design pressure 100 psig, minimum 1 e. g3ke-un system A check valvo shall be installed in the line between the j pot abl-2 water supply and the make-up and cleanup domineralizer to prevent entry of potentially contaminated water into the potable water supply. P -Water source Potable water from city main Make-up domineralizer type Mixed-bed single shell, regonexative Make-up deminerallger capacity Normal 25 gpm Emergency 50 gpm Water softener capacity Normal 50 gpm f. Cleanun System f or P rL?a ry coolant Water l-Cleanup pump Capacity 40 gpm Head 100 ft Cleanup demineralizer Type Mixed-bed, single shell, regenerative cleanup demineralizer capacity Normal 40 gpm Emergency 50 gpm -"et'-'Mr' 9r e t -' f df* C' M mx.*H+-- eb--'t"T-s-T-Mw*d-r7N'4-+8^etv--W .e Pf fN1 r*-f N

  • wet W ry' vv wemre vp Y'T'w-verM wpq"--ww y"eFww*vu w'FTU9Ws'==

Y- '*w4 'V->'-C~ WT V'

-(. D. LLC.:lC12Y C22 M T.EXETDi The secondary coolant system shall carry the heat rejected from t rar prirna ry coolant at the heat exchanger to the atmosphere at coaling towers. It shall be composed of the heat oxchanger, cooling tcwers, pu rps and associat ed valves, piping ana senscrs. In this system, water ficws frcm the neat ex:nar.ger through a :onttol.alve to t he COolin7 towers. From the cooling tower basiM, the watt 13 t f en p;rped ta 9 to the heat exchtnjer, t_. _r v r 2, r.n v. r p.v r_ _m s_ +,.. _. cve-ro _-c- -m An emergency core cooling system shall 1.e in place to provide a minimum cf 4 GPM directly to the core grid box fat a mi n ite duration of 6 hours. e o O f 9 1 i

+ 7 E.- ptACTOL CEPE A*!D C0!HDOL EtrMf?fta The reactor core and control elements shall have the following characteristics and nominal dimensionst 1, Princical Core Materigla 0 S 2-A2 diepersion ruel matrix 31 U-235 enrichnent -Approximately 201' l ruel clad 6061 aluminam Fuol element side pitoos 6061 aluminum End fittings 356-76 or 6061 aluminum ) Moderator' Water Reflector-Graphite AGOT grado (or equivalent graphite and/or water l Reflector-Beryllium Beryllium-aluminum clad Mixture of D C and aluminum, Control elements 4 clad with aluminum Servo Element Stainless steel 314 ll 2. Puel Eloronts r; Plate wjdth overall 2.81 inches Active plate width 2.4 inchos maximum Plate length overall 25 inches Active plate length 23.5 inches l-l Plate thickness 0.06 inch l Clad thickness 0.02 inch Puel matrix thickness 0.02 inch l Water gap between plates 0.1 inch I - s. G r-r ry., e. ,,,-eci w-- ,,p.v.y.e-w yw,,y, , w nw yv -w y...w-<,,,e,,r#n,w,,mwv--,-,,,---w,wr,%wo ,, ren nw m*w-w m yewmww w y y gr my= -we j re.

_.. _ _. _, _ _ _ _.. _ _ _ _ _ _. _. _ _ _ _ _. _. ~ _ _ _ _. _ _ l .. p. !!urt e r of plates ter fuel i eierrent 22 U-235 per fue; element 2~' grams, n c t. i n a l Overall fuel element dirensions 3 .n x 3 in. x 4: i 3-huil c L^. L k :'A n L --.CLA5 hkLn a:LDuiy11La Cverall reflecter elerent 3 in v. 3 in. > 4 2 d ime n s i r-n a, nominal Nominal c:la d thickness

in Noninal graph te dimer 2ns 2.4 in. x

'.9 an 2t in. !Jcminal Bety11ium air.cnslens 2.94 in. x 2.94 In. x 29 in 4. Control Elen c a Wi d t. h 10.6 in. Thickness 0.39 in. Overt 11 length 54.1 in Active length S2.1 in. b Zcuz.i..;;uLALLG2 L1ccnL Shape Square stainles ste 1 ll Width 2.1 in. Overall length 28.8 in. Activo 24.9 in, f. Cont rC1 file'*ent Driyg Type Electromechanical screw Drive to safety element Electrcmagnet connection Stroke 32 in. maximum 9 I

.. i i 7 gerve Pegulatine riemr.t Drive Type Electr: mechanical screw Drive to element connection took screw (no scram) Jtr ke 2 6 i n. ru x imum l Fosi':icn indi:4tien accuracy !, 0.02 in, j 8. Meutren scur m Start-up Sourte !!umbe r 2 Type Flutonium-bary111um Unit Source Strength 1 x 106 neutrons /sec. minimum Maximum Power Level with Plutonium-beryllium sources installed 10 Mw operational Source 11ambe r 1 Type Antimony-beryllium Source Strength 2 x 106 neutrons /sec. minimum F. PEACTOR EArPTY MYE?rMS 1. Medes of Power Oceration There shall be two modes of power operation: i a. P ower.. C;g rat ien - Natural Circulatien (NCl Power operation fiC shall be any reactor operation performed with the reactor cooling provided by natural circulation. The reactor power shall not exceed 0.1 MW during PC operation. b. Pcwer Operatien - rerced circuintien tre) h FC shall by any reactor. operation Power operation performed with reactor cooling provided by forend circulation. The reactor power shall not exceed 2 MW during FC operation, t


,,w--

-a .w-. -n-w ,.,w,.-, w r

. 2, Design reatures a. The Reacter Centrol Eystem 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-energiring shall cause a scram). Table F.1 and F0 describe the arrangement and requirements of the safety system. b. Proces* In=tru-entation Process instrumentation with r63 out in the centrol room shall be provided to permit measurement of the flow rate, temperature, and ccnductivity of the primary coolant and the flow rate of the secondary coolant. In additien, a second primary flow indicating device with readout in the control rooin shall be located between *he reactor outlet plenum and the reactor outlet header. After normal working hours, a r. independent protection system, separate frem the system described in Secticn 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) in alarm condition from any other selected feature. c. Master E wit h. A key lock V: er switch snall be provided with three positions; ">n , " test", and "on". These positions shall a have the following functions: (1) The "off" po sit ion.shall de wnergize the reactor control circui* O

A . [f (2) The " test + position ~shall energize the reactor control circuit exclusive of the control blade magnets. (3) Tne "on" position shall energize the reactor-control circuit including the control blade =j magnets. d. Emr Leysi selecter sjetit.ch A power level selector switch shall be provided with four positions: "O 1 MW ", "1 MW ", "2 MW", and "5 Mw" These p sitions shall have the following functions: (1) The "O,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 activato all safety system sensors. (3)' The "5 E posit, ion shall scram the reactor. e. centrel rioront withdrawal Interiorks Interlocks shall prevent control rod withdrawal unless all. of the following conditions exist-(1) The-master switch is in the "ca" position, .f( (2) The safety system has been reset, (3) The Log N amplifier switch is in the " operate" t l

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. serve Svetem Cent rol Interlock 1 ( Interlocks shall-preveni. 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-automa icslly return to the manual mode of I l ope ra tion.

. TAsir F.1 - orACT P_EAFETY SYETIM Sensor or Trip Device No. of Switches Trip Set AlIrm Set or Sensors Point Point Short Teriod 1 3 sec. min. 7 sec. min. High Neutron Flux 2 Max. of 120% of 110% max. ll full scale with a 2.4 MW rax. High Terperature of Primary 1130F max. Coolant Entering Core During Forced Convection C90 ling

  • High Terperature of Primary 1250F max.

1:30F max. Coolant Leaving Core During Forced Convection Cooling

  • Low Flow Rate of Primcry 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 2cale max. High Pool Temp 1 1250F 120CF ll Bridge Misalignment

  • 1 X

X Coolant Gates Open* 1 per gate X X Neutron Detector High 1 per Decrease of Voltage Failure in Linear power 50 volts max. Level Safety Channels supply Manual Scram (Switch at 2 X X bridge and on console) High Conductivity of 1 Equivalent Primary Coolant to 24mho/cm at 2500, max. Safety Blade Disengaged 1 X Log N - Period Amplifier 1 X X Failure Regulating Rod at Either 1 X Limit of Travel Low Flow Rate of Secondary 1 000 gpm, Coolant *

min, Bridge Movement 1

X X 9 No Flow The rmal Column

  • 1 X

X

  • These f unctions are bypassed when the Power fevel Selector Switch is in the "0.1 MW" position.

h f% (V v v TABIE F_2 PMCtL11UCIIAR INSTRUMENTATICl{ Channel l Detector Sensitivity Range " formation Information Information Pecorded to to to Information i Operator Logic Element Servo System (Scram) Retractable tieutrons-Source Neutron Relative gas filled approximately level Flux power level Start-up B-10 filled 12 counts /nv to full None None on log scale proportional power Neutrons-Source Power level Power level 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 Power level. level ion chamber approximately to Power level Level scram Power level linear scale safetr 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 1 n l ._- en,

. i G, WA9TE DISPOSAt AND FACILITY MS*II?ORING EYETEME 1. Wasit_21Apual System Dcaign Featurn a. Liquid Radicactive Waatt Disp nal System Ali 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. G nen n Radicactive Waste Diagonal _Sy;1cm All gaaeous radioactive waste from the beam ports, thermal

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

C. Eclid Radioactiva Waste Sterage Solid Radioactive wastes shall either be stored in radioactive wtste storage containers 1ccated within the reactor building or removed from the site by a commercial licensed organization. 2, h..and Exhaust G is Maniter Desien Features O s a. Tnree fixed gamma monitors employing suitable detectors shall be employed in the reactor building. Each of these shall have the following characteristics: 1) A range consistent with the expected radiation levels in the area to be monitored (0.01 to 10 mr/hr, 0.1 to 100 mr/hr, or 1 to 1,000 mr/hr) 21 A raaiation dose rate output indicated in the control room. 3) An adjastable high radiation alarm which shall be annunciated in the control room. O

. 4) The three fixed gam a monitors shall be located to de t. e c t radiation as follows: At the pcol biological shield between a beam port and tne thermal colu.n, above the storage container for new fuel elements, and at the reactor bridge, b. A gamma monitor shall be provided near the primary coolant system, and an additional one shall be provided near the secencary coolant system for use in determining the presence of a b n o rra a ll y high concentrati-ns of radioa ctivit y in these systems. The characteristics of these monitors shall be as stated in a. above. c. Six additional direct reading area monitors.mploying Gaiger tube detectors shall be provided to monitor the pneumatic system receiver stations, the t.eam port areas, and other areas as required. Each of these shall have the fcilowing characteristics: 1) A range consistent with expected radiation levels in the area ceing monitored (0 to 10 mr/hr er 0 to 50 mr/hr). 2) A radiation dose rat.e output at the instrument. 3) An adjustable high radiation alarm to alarm at the instrument and create both an audible and visual signal. O d. A stack exhaust gas monitor system shall be provided which draws a representative sample of air from the exhaust gas. The monitor with indicators and alarms in the control roce, 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 n eint ion Monitorin, E m2iment a. Portable survey instruments for measuring beta-ga:ma 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 1 n/cm2 sec 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 gamma and neutron doses in the event of an accident. O J

..r.._.-____._._. M' c. A radiation monitor shall be provided to monitor all i persons leaving the reactor room for beta-gamma .'W' contamination. i' .l -4. Mich Radiatien Area During reactor operat; ion, the dose rate from the delay tank may be in excess of 100 millirem per hour. On three sides, the tank _shall be shielded. On the fourth side, the tank is ] shielded using a " maze" so that access to the tank is possible 1 threvgh a door equipped with a lock. i: ll h i f y._,. b a 9 1 4 4. t n j :. 1 .. - -.. _..... _ _ _.. _,.. _ _ _. _..., _ - _ - _.~

s 'm..r.

+-: me y*

r osm 1. Scw Fucl Etenn New fuel shall be stored in a security container in " egg c: T te " boxes. Sheet cadmium at least 3.020 inches thick shal1 1,. fastened around the outside cf the bcxes in the regien which centains the fuel. The nurter of fuel elenents which can N placed in each box shall not exceed three. F:r all c e n.i : t.. s cf moderation possible at the site Feff shall te less thar : a 2. 1rrnstatzi Fuel S tmp Tw: types of irradiated fuel element st o ra ge racks chall r.: previded. One type of rack shall centain spaces for 2ne fee; aseemblies and shall have approximate cver-all dimensions :f 35.5 in, wide by 26 in. high by 6.25 in, thick, and shall N fixed to the pool wall. At least two of these racks shall te previded. The second type of rack shall consist of two cf the nine fuel assembly racks described above attached together witt a minimum space between the center lines of fuel asser nies 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 neutren absorbing material. At least one 18 fuel assertly :ack shall be provided, and the rack may be moved within the pool. The fuel stcrage racks may also be used to store core ecmponents other than fuel assemblies. The irradiated fuel storage racks shall have a maxinum Ke f f cf 0.8 for all conditions of moderation possible at the sit e. Storage spaces shall be provided for at least 36 fuel assentlies. I. FJP E RIMF'!T A L FACILITTPS The permanent experimental facilities shall c:nsist of the following: 1. Thermal column. 2. Beam ports; two 8 inch dia, and four 6 dia. 3. A six inch diameter through port. 4. Radiation baskets. 5. A two-tube pneumatic tube system. 6. Dry gamma cave. O

.~.-.. - ~ - ~. -.. - - - ~.. ~.. . -.~..-.

  • J.

ADMINISTPATIVE AND PROCEDUP"L IAFE00APf3 1. 01;;a niz atien 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 from the faculty of Brown University (4) A qualified representative from the faculty of -Providence College (5) A qualified representative from the faculty of the -University of Rhode Island. A qualified alternate may serve in lieu of one of the above. The Director and Health Physicist are not eligible for chairmanship of the Committee. The Reactor Utilization Committee shall have the following functions: L a, Review proposals for the use of the reactor I considering the suitability of the reactor for the proposed use and the safety factors involved. i l l l l l l

9 b.

App:ove or disapprove proposed use of the reactor. c. Rev ew at least annually the operating and emergency pro :edures and the overall radiation safety aspects of.ne facility. The Reteter Utilization Committee shall maintain a written record of its findings regarding the above. 2. Oualificati+ns ~f Per? nnel a. The Director of Operations shall have at least a tachelors degree in one of the physical sciences or engineering, and he shall be trained in reactor technolcgy and be a licensed senior operator. b. The staff Health Physicist shall be professionally trained and shall have at least a bacbeJors degree in one of the physical or biological sciences or engineering. He shall have experience such as may heve been gained through employment in a responsible technical position in the field of health physics. c. The reactor operators and senior operators shall be licensed in accordance with the provisions of 10 CFR 55. d. In the event of temporary vacancy in the position of Director of Operations or the Health Physicist, the functions of that position shall be assumed by qualified alternates appointed by the RIAEC. 3. Re s nc n m i b i lit ie s of Personnel a. Director (1) The Director shall have responsibility for all activities in the reactor facility which may affect reactor operations or involve radiation bazards, including control 1.ing 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 Directvc to insure that all proposed experiments, design modifications, or changes in operating and emergoney procedures are performed in accordance with the license. Where uncertainty exists, the Director shall refer the decision to the Reactor Utilization Committee. 6

. b. Senior Reactsr Op rators .v (1) A licensed senior reactor operator shall be assigned each shift and be responsible for all activities during his shift which may affect reactor operation or involve radiation hazards. The reactor operators on duty shall be responsible directly to the senior operator. (2) The reactor operations which affect core reactivity shall not be performed without the senior operator on duty or readily avuilable 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 duties over to another licensed senior operator. In such instances, the change of duty shall be logged and shall be definite,

clear, and explicit.

The senior operator being relieved of his duty shall insure that all pertinent information ' is logged. The senior operator assuming duty shall check the log for [ information or instructions. c. Re met e r Ocerators. (1) The responsible senior operator shall designate for his shift a licensed operator (hereaf ter called " operator") who shall have primary responsibility under the senior operator for the operation of the reactor and all associated control and safety

devices, the proper functioning of which is essential to the safety of the reactor or personnel in the facility.

The operator shall be responsible directly to the senior operator. (2) Only one operator shall have the above duty at any given time. Each operator shall enter in the reactor log *he date and time he assumed duty. (3) When operations-are perf ormed 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 dut; by proper notation in the reactor log.

. (4) The operator, under the senior operator on duty, shall be responsible for the operation of 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 similarly authorized to scram the reactor by pressing a scram button located on the bridge. d, Health Phvnici c b The Health Physicist shall be responsible for assuring that adequate radiation monitoring and control are in effect to prevent undue exposure of individuals to radiation. 4. Written Instructions and Precedures Detailed written operating instructions and procedure.s shall be prepared for all normal operations and maintenance and for emergencies. These procedures shall be reviewed and approved by qualified personnel before ase. E ach member of the staff shall be familiar with those procedures and instructions for which he has responsibility, u 5. Site Emercencv Plans The Rhode Island Nuclear Science Center shall have available the services of other state agencies for dealing with certain types of eme rgencies. 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 co=nunications 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. Tne emergency vehicle shall contain equipment such as portable radiation monitors, respirators, and a particulate air sampler. Communications using the statewide emergency network shall be available. Personnel of the Civil Defense Agency and of local fire 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 l 1 1

. K NroATIM" tiMI T AT Mm s l 1. C;lneral The following administrative controls shall be employed te assure the safe operation of the facility: a. The reactor shall not be operated when'ver 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 scurce must be in the core during any change which can affect core reactivity. c. During cperations which could affect core reactivity, a licensed operator shall be stationed in the control rocm. Communications between the control rocm and the senior reactor operator directing the operation shall be maintained, d. The operator shall not attempt to start up the reactor following an automatic scram or unexplained power decrease until the senior operator has determined the cause of the scram or power decrease and has authorized a start-up. e. The reactivity of all core loadings to be utilized in operating the reactor shall be determined using unirradiated fuel elements or elements containing fission products in which the effect of xenon poisoning en total O core reactivity has decayed to 0.05% delta k/k or less, f Critical experiments shall be performed under the supervision of the Director or other competent supervisory scientist licensed as a senior reactor operator. During tne experiment there shall be present, in addition to this licenaed supervisor, at least one other technically qualified person who shall act as an independent observe.. 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 made at the time of each fuel element addition or other core change which could significantly affect core reactivity. g. The basic operating principles for the assembly end reloading of cores whose nuclear properties have been previously determined from critical experiments shal. 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. O l

3

  • T' e exact procedure to be followed for a a

particular reloading operation will be determined by the observer and the senior reactor operator in charge of the operation before the operation begins. Each step in the procedures shall be considered by both persons, and no step shall be taken without the concurrence of both. 2. eveeriments a. " Experiments" as used in this section shall be construed as any apparatus or device installed in the core region which is not a component of the core. b. The React.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 documer.ted for the study of the Reactor Utilization Committee. The docunentation shall include at least: (1) The purpose of the experiment, (2) A description of the experiment, and (3) An analysis of the possible hazards associated with the performance of the experiment, c. All use of experimental facilities shall be approved by the Director of Operations. d. The absolute value of the reactivity worth of any single independent experiment shall not exceed 0.006. If.such experiments are connected or otherwise related so that their combined reactivity could be added to the core simultaneously, their combined reactivity shall not exceed 0.006. e. The calculated reactivity worth of any single independent experiment not rigidly fixed in place shall not exceed 0.0008. If such experiments are connected or otherwise related so that their combir.ed reactivity could be added to the core simultaneously, their combined reactivity worth shh11 not exceed 0.0008. f. No experim4nt shall be installed in the reactor in such a menner that it could shadow the nuclear instrumentation system monitors and thereby give erroneous or unreliable information to the control system safety circuits. g. No experiment shall be installed in the reactor in such a manner that it could fail so as to interfere with the insertion of a reactor control element.

. h. No experiment shall be performed involving ma te t a a '.s r used in such a way that they might credibly result in an exp2 sion. 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 ons vacant fuel element position within the periphery of the active section of the core. 3. Operstic u a. LLte Control of access to the reactor facility shall be the responsibility of the Director of Operations. b. Centainment (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 butten shall be sealed, except that doors other than the truck door may be opened during reactor operation. If a door is to remain open, an individual from the reactor operations staff is continuously in attendance at the door, b. The building clean-up system is operable. (2) Re m irement? for Rete =t of Confin-ent (a) Method of Retest The building cleanup synter ch?ll be retested by prussing an evacuation button i and observirq that the following functions occur automatically: 1. Evacuation horn blows. 2. air conditioning 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 basement chem lab blower come on. l 'I

. 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 magna helic gauge located in the control room. (b) Erc Gcncy :2L ?17.c2. The building cleanup 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 w it h not more than 1500 cfm caming from the reactor building and passing through the charcoal scrubber. The remaining air will be provided by a separate blower from an uncontaminated source. This shall create a pressure in tne building which is equivalent to at least 0.5 inch of water below atmospheric pressure. c. M m ry Coelant System (1) The minimum depth of water above the top of the active core shall be 23 feet. (2) No piping shall be placed in the pool which could cause or fail so as to cause a ciphon of the pool water to below the level of the ten inch coolant line penetrations. (3) Makeup Svet e 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) Clcuup 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 cnce 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: O

t.(~. Q pH 5.5 to 7.5 conductivity 2 mho/cm (6) The radioactive materials contained in the pool water and in the primary coolant water shall be such that the radiation level one meter above the surf ace of the pool shall be less than 10 mrem /hr. (7) During the forced circulation mode of operation, the primary coolant flow rate shall not be less than 1580 gpm. d. Secrnd uy csmiins Waf.mn (1) The secondary coolart 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 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 f determined at least once per week when not being ~ eperated using forced convection cooling. (3) If the radioactive materials contained in the secondary coolant exceed a radionuclide concentration in excess of the-values in 10 CFR 20, -Appendix B,- Table I,. Column II, above background, the reactor shall be shutdown and the condition corrected.before operation using the secondary cool...g system resumes. (4) The secondary coolant system shall be placed in operation as required during power operation utilizing forced convection in order to maintain a primary coolant core outlet temperature of 1250F or below, e. Reactor core and central Elements (1) The reactor shall not contain in excess of 35 fuel elements. There shall be a minimum of four operable control elenents. O 1 i

u (2) The limiting thermal and hydraulic core characteristics based on a 14 element, graphite h and beryllium reflected core are specified below: (a) Maximam Fear Flux .424 MW/M2 (b) Maximum Core Specific Power S19.48 W/gu235 (c) Maximum Fuel surf ace 1100C Temperature (d) Coolant Velocity during 1.49 M/see Forced Convection Cooling (e) Coolant Inlet Temperature 1150F max. (f) Aterage Coolant Teeperature 100F max. Rise (g) Primary System Bulk Outlet 1250F max. Coolant Temperature (h) Temperature Margin in Primary S.BOC ll Coolant (Teat *Tsurfi (i) Number of Coolant Passes 1 1hrough Core (3) Erincipal Nucinar characteristics of the cera p (a) corn and cantrsi syften p**ctivitY Worth 1. The reactor shall be suberitical by at least 1% Ak/k. f rom the cold, Xe-free, cricical condition with the most teactive cont rol element and the servo regulating element fully withdra wn. 2. The maximum worth of the cervo regulating eloment shall be 0.7% ok/k. (b) Mn4 mm R"act ivitY N14!t ic" Rat" - ^*#k#^" 1. By servo regulating element maximum of 0.0002 2. Manual by control element maximum of 0.0002 1l - ^ ^ - m-----mmmm-___m_____

e. -2 8 - ? (c) Fa ntivity Joefficients 1. Temperature coefficient approximately .82 x 10-4 oc / (es icula t ed) density only 2. Void coefficient approximately teore average) -2.7 x 10-3/% void (calculated) (4) Erincipni core.cparatina 1Jniani;na j (a) M ulmum rool Te:@.cIsture Li-ftations The pool water temperature chall not exceed ~ 12 'aoF, The pool water temp shall be monitored with readout in the Control Room. A trip and alarm shall be incledec' in the system. (b) PeActivity L!*lu*ian.1 l. Imagss Reactivit,y The cold, clean excess reactivity for any core used in the reactor shall not exceed 0.047. 2. Minim m shutdquajiaggin 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) pa er hlty_Ceefficient L M tatica The reactor power coefficie (as inferred by the control rod movements required to compensate fcr changes in power) shall be negat.ive. (d) C2ntrol riament Drive Perfo m nce Requiramants All control element drives shall meet the following specifications: 1. The contrcl drive w Lthdrawal rate shall not be more than 3.6 inches per minute. 2. For the electronic scram system, the time from initiation of a scram condition until control element release shall not exceed 100 milliseconds. 3. The time from initiation of a scram condition until the control element is fully inserted G shall not exceed 900 milliseconds. 4. It shall be demonstrated at least every 3 months that the above specifications are met. i l a

l .29_ (e1 131.X:_.fmMiA2 E19'*a ni_._R.~i'in ___l'.211Ar 3nT.a .=nu i r c"u mL If in use during operation, the servo regulatirg element drive shall meet the follow)ag specifications: I 1. The d;ive withdraral rate shall not be r.o t ? than 70 inches per minute. 2. It shall be demonstrated at leaat once per month that the above specification is met. f. km - c a rn,;; y > r.m (1) The reactor safety system shall be operable during all reactor operaticn. The safety system snall Le enecked out before each start-up and functionally tested for calibration ) at least monthly. (2) It shall be pemissible to continue operations with one or more of the safety system functions that prodace only an alarm temporarily disabled providing that additional procedural co nt rols are instituted to replace the Icst wa fety system olarm f unction (s). (3) The control element withdrawal interlocks ato the serve system control interlocks shall be functionally tested at least cnce per month. 1 (4) During reactor startup or during rauchanica 3 cnanges that could affect e c. re reactivity, the startup range neutron monitoring channel chall be cperable and shall provide a neutron count xate or at least 3 counts per second with a signal to noise ratio at least 3 to 1. (5) The linear level safety channels shall. o t. cead less than 15% of full scale when the reactor is operating at pcwer levels above watt. (6) Follo. ting a redrction in power le"el, the operator chall adjust the servo power schedule to the new power levcl before switching to autcmatic operation. (7) An alarm condition from any eno of the items listed in Section F.2,b. after working hours shall transmit coded information to a continuously manned coatral station in i Providence, Rhoda 1sland. The central statica shall be i provided with written instructions on the steps tc be taken following an alarm. 1 9

> g. ~ ,o-i* g . o (;. .j 'E u l g. yg.n o Disresal and peneter Mnnitsrine syng:n [.%i j ' The liquid waste retention tank discharge shall flow to _(1) a monitor station -in the reactor building where the effluent shall be batch sampled and the gross activity 4 per unit volume - determined bef ore 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 ecnform to the following table. In this table, the MPC stated is for individual isotopes and mixtures contained in Column 1, Table II, Appendix S cf 10 CFR 20 i Type of Activity Maximum Curies Curies per second to per second to be be released averaged i released over one year Particulate Matter and Halogens with half-lices 140 X MP0 gue/cc) 14 X MPC p.e f ec ) longer than 8 days All other Radioactive : 108 X MPC (uc/cc) 104 X F.r0 (uc cc) j Isotopes .k L (3) All radioactive liquid and solid wastes disposed of I. of f-site shell be within the limits established by 10 CFR 20 or shall be removed from the site by a commercial. licensed organization. (4) The exhaust gas monitor shall be calibrated to alarm at an instantaneous release rato which instantaneously exceeds the limits' stated in Colemn 2 for the annual average, release-rate. If tne maximum - pe rmissible stack release raue stated.in Column 1 is exceeded, the reactor shall immediately be placed in the shutdown _ mode of cperation and the situation investigated, -(5).The area, primary and seccnoary coolant system and the lf exhaust gas monitors shall be in operation at-e.11 (j times when control elements or the servo regulating elements are withdrawn; however, individual area coolant system monitors may be taken out of service for maintenance and repair if replaced with portable l radiation detection equipment. Adequate spare-parts shall be on hand to allow necessary repairs to be made during the maintenance or calibration outages of tho j monitors. l j a i

.a - (6) The area and the prura ey and secondary coolant system monitors shall be ad;usted to alarm at a maximum reading of 2 mr/hr or 200% of the no rma l radiation levels in their area, whichever is larger. (7) The cecr which controls entrance to the " maze" leading to the delay tank shall be locked with the Key in the possessian of the Director or a licensed senior opert. cr. Entrance tc the Jelay tank high radiation area shall require the presence of the Health Physicist er a licensed senior operator and the use of direct reading pertable radiation monitoring equipment. h. Enel Storage (1) New fuel shall be stcred in egg crate boxes loca'.ed in a security contaiaer. 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 b( St0 red in the criti:ality 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 fusl assemblies per rack to nine and then positively securing such racks at least 30 cm. apart, or by limiting the number of fuel assechlies to 18 per rack and then covering the two large faces of each rack with a sheet of aluminum covered cadmium. 1. MM erennnee (a) The electronic control and the process control system shall be checked for proper operation and calibration before each reacter start-up. If maintenance or recalibration is required, it shall be performed before reactor start-up proceeds. (b) Maintenance shall be performed with the approval of the Director. Equipment and system maintenance records shall be <cpt to facilitate scheduling and completion of all necessary maintenance. (c) Routine maintenance on al' control and process system components shall be performed in accordance with written schedules and with written procedures. O}}