ML20217G218
| ML20217G218 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 10/03/1997 |
| From: | NUCLEAR REACTOR LABORATORY |
| To: | |
| Shared Package | |
| ML20217G214 | List: |
| References | |
| NUDOCS 9710090258 | |
| Download: ML20217G218 (90) | |
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+ SAFETY EVALUATION REPORT FOR FISSION CONVERTER FACILITY O OCTOBER 3,1997 ju 18$EEESkjo Q SR#-0-97 14 OCT 031997 y
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- 1. I NTR O D U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 i
1.1 S cope o f Th i s R e po n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2
- 2. FISSION CONVERTER NEUTRONIC DESIGN..... . . . .. ..... .... ..... .. ... .... ... 2 1 2.1 General Description of the Fission Convener ................................. 2 1 2.2 Fu e l De si g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 2.3 Critlea1ity and Power Distribut10n............................................... 2 6 2.3.1 Criticality Calculation for the Fission Convener................... 2 6 2.3.2 Reactivity Worth of the Fission Converter..........................,2 8 I 2.3.3 Fission Convener Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9 2.3.4 Fission Convener Radial and Axial Power Distribution .......... 2 9 2.3.5 Fuel Element Orientation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 1 4
2.4 Fil t e r/hi odera t or . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 1 2.5 Coll i ma tor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 1 3 ' (J
- 3. El SiON CONVERTER TilERhtAL llYDRAULIC DESIGN..................... 31 3.1 General Description of the Thermalliydraulie System....................... 3 1 3.1.1 Primary Coolan t System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 3.1.1.1 Fis sion Con verte r Tank.................................. 3 3 3.1.1.2 Fission Converter Fuelliousing Grid Design........ 3 7 3.1.1,3 Cover G as System . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 3.1.1.4 C1e a n U p ' S y s t e m . . .. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 3 10 3,1,1.5 hiake Up Coolant System.. . . ... . . . .. . . ,,,, . . .. . . . .. . . . . 3 10 3.1.1,6 Coolant and Cover Gas Sampling...................... 3 10 3.1,2 Second ary Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 1 3.1.3 hia teiial Selec tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 1 3.2 Basis for Thermal f lyd raulic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 12 i
SR#-0-97 14 OCT 031997
TABLE OF CONTENTS b v DE 3.3 Computational Method for the Fissior. Converter Themial liydraulic De s i g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . J 1 2 3.3.I hie thodolog y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 13 3.3.2 Major Cortelations Used in h1ULCil FC........................... 313 3.3.2.1 Correlation for Onset of Nucleate Bolling (ONB)....................................................313 3.3,2,2 Conelations for Onset of Signincant Voiding ( O S V) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 4 3.4 Fis slon Con ve rie r Ope ra ting Limit s.......................... ................... 3 15 3.4.1 Fower Deposition Factor (Fp) and Nuclear llot Channel Faetor (FilC)............................................................315 3.4.2 Fueled Rel; ion Coolant Flow Factor (Fr) and Channel Flow Disparity I actor (dt) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 16 3.4.3 En ginee ti n g Hot Chan n ei Fae tors.................................... 3 17 3.5 Themial I lydraulic Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 20 3.5.1 Dennitions of Safet Limits and Limiting Safety System Settings ...........y .....................................................320 l 3.5.2 Deriva tion of the Safety Limits. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 22 3.5.3 Derivation of the Limiting Safety System Settings................. 3 25
- 4. FlSSION CONVERTER SHUTTER AND MEDICALTilERAFY ROOM DESIGN....................................................................................41 4.1 General Description of the Shutter Design .. .. ... .... . . . .... .. ... ... ... ...... .. 4 1 4.1.1 Cad m i um C u rta i n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1 4.1.2 Collimator S hutter Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 4.1.2.1 Wate r S h u t ter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 4.1.2.2 Mechanical S hutter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 4.2 S h u t t e r Co n t rol s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 4.3 M e d ie a1 Th e ra py R oo m D e sig n.... . .. . .. .... .... .. . ... ....... ... .......... ...... 4 7 4.3.1 M edle ai Therapy Room Door......................................... 4 7
- 5. FIS S ION CONV ERTER FU EL 11 ANDLING . . . .. . . . . . . . . . . .. . . . . . . . . . . . .. . . . . . . .. . . . 5 1 p 3.I Fuel Element Security, Storage, and Quality Assuiance ..................... 5.I s
il SR# 0 9714 OCT 031997
TAllLE OF CONTENTS DEC 5.2 Fuel Eleme nt Self. Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 5.2.I U se o f U n i rradi a t ed Fu eI . ... . ... ... .. ... . .. . . ..... . . ... . . ...... . . ... .. 5 2 5.3 Fue1 Element Remoyal............................................................52 5.4 Fuel Ele me n t Ha nd li n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3 5.4.1- Fuel Elemen t Tran sfer Cask . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4
- 6. S A FETY A N A LYS I S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1 6.I M a x im u m H y po t h e t i e a1 A ec id e n t ... . .. .. .... . .. ... . .. ... . .. ...... ... ... ... . .. ... 6- 1 6.2 Insertion of Exce s s R eac tivi ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3 6.3 Loss o f Prima ry Coola n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 6 3 6.4 Loss of Primary Coolan t Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 7 6.5 Loss of One ol Two Primary Pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 8 6.6 Loss of Off S ite Electric Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 12 6.7 Lo s s o f H e a t S i n k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 14 6.3 Mishandling or Mal function of Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,6- 14 6.9 Ex pe rimen t M alfu nction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 14 6.10 Natura1 D i s t u r b a n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 14 6.10.1 Eart h q u ak e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 14 6.10.2 L i g h t n i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 15 6.10.3 a c v e re S t o m1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 15
- 7. INSTR U M ENTATION A ND CONTR OL S YSTEM , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1 7.1 Fission Converter Nuclear Instrumentation . . . . . .. . .. . . . . . . . .. .. . .. . . .. . . . . . . . .,7- 1 7.2 Fission Converter Themial. Hydraulic Instrumentation Essential for Safety...............................................................................72 7.3 Fis slon Co n verte r S h u td own S y ste m..................... .. ................... 7 2 7,3.1 Operability of the Fission Converter Shutdown System.......... 7 4 7.4 Other instrumeniation............................................................73 3
(V iii SR# 0-9714 OCT 031997
i TAllLE OF CONTENTS i l Eage
- 8. PRE OPERATIONAL TESTS AND INITIAL OPERATION ..... ................. 8 1 8.1 Pre Opera t ion al Test s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 1 l
l l 8.1.I Non Nuclear Instrument Calibration .. .... . . . ....................... 8 1 j 8.1.2 N u'cIe a r instru me n t Ca11bration....................................... 8 2 l 8.2 Ope ra tor Tra i ni n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2 8.3 in i ti al Fu el Loadi ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3 8.4 Fuel Region Flow Distribution Measurement.................................. 8 3 8.5 Reactivity Estimation of the Fission Convener................................ 8 3 8.5.1 Estimation of Integral Reactivity Wonh............................. 8 4 8.5.2 Estimation of Differential Reactivity Worth......................... 8-4 8.5.3 Determination of Reactor Period Associated with Fission Con vene r Ope rat ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4 8.6 initial Approach to the Highest Available Fission Convener Operating Power...............................................................................85 0 0 O
\v iv SR# 0 9714 OCT 031997
- 1. [pt roduction h Boron Neutron Capture Therapy (BNCT) is a binary form of cancer therapy that has the potential to kill cancer cells selectively while sparing nonnal tissues. This can be accomplished by localizing 108 in the tumor cells and delivering thennal neutrons to the target (Ref.12]. Compounds such as p boronphenylalanine (BPA) can be used to locabre l 10B in the tumor by taking advantage of the higher metabolic rate of the tumor cells. A 10B nucleus that absorbs a thennal neutron disintegrates into an alpha particle and 7Li nucleus with an accompanying energy release of 2.79 hieV. The alpha particle and the 7Li nucleus are both heavy charged particles that slow down quickly. The distance that they travel while slowing down is about a cell diameter. As a result, cancerous cells are killed while adjacent healthy ones are spared.
Recently, BNCT trials have been conducted at the hiassachusetts Institute of Technology (MIT) using the M67 beam that has an epithennal neutron flux of l 2.1x108n/cm2s. This beam was first used for BNCT Phase I clinical trials of subcutaneous melanoma of the extremities on September 6,1994 [Ref.1-2]. This was the first epithennal neutron irradiation used for BNCT of a human subject anywhere in the () world. Brain cancer trials were subsequently initiated at the Brookhaven National Laboratory in 1994 and at MIT in 1996. Technical Specification No. 6.5 to Reactor Operating License No. R 37 governs the use of this beam for both human trials and therapy. - The M67 beam takes approximately 2.5 hours to deliver a normal tissue tokrance dose of about 1000 RBE cGy and correspondingly more to tumor. Because of the long irradiation time, a new beam capable of treating a patient in a few minutes is necessary for advanced clinical trials and for routine therapy [Ref.1 1]. Also, the M67 beam contains significant background components and a higher purity beam would improve the therapeutic ratio. The new beam design is based on a fission converter plate driven by the neutrons from the Massachusetts Institute of Technology Research Reactor-ll (MITR ll). Neutrons from the reactor are converted to a fission spectrum by the fission converter plate. A filter / moderator is then used to tailor the neutron spectrum to eliminate unwanted fast neutrons and photons without significantly decreasing the epithermal neutrons (1 eV to Og Y l1 S R#-0 14 OCT 031997
l l l 10 kev) [Ref.1 1]. The cooling of the fuel contained in the fission converter will be provided by forced convection of either 110 2 or D 20 enclosed in a tank. The purpose of the design is to deliver a neutron flux of about 5x1010 to ! 10x1010 n/cm:s with specific fast neutron and specific incident photon doses lower than 2x10-11 cGy em2/ epi n to minimize non selective dose components. An epithermal neutron Dux at this intensity would result in an irradiation time of less than ten minutes at a reactor power of 10 MW (Ref.1 l]. Irradiation times in this range are typical of those used with conventional external beam irradiation facilities such as linacs and are imponant to patient comfon and needed for eventual high throughput of patients. The purpose of this report is to provide the necessary information required for the licensing of the MITR Fission Convener Facility by the U.S. Nuclear Regulatory Commission (NRC) and to demonstrate the safe operation of the fission converter. Te.hnical specifications that constitute limitations for the design and operation of the facility are also included. 1.1 Scooe of This Report O The focus of this report is the design of the fission converter. Use of the fission convener beam for medical therapy is governed by existing MITR Technical Specification
# 6.5. Accordingly, materialin that specification that pertains to patient therapy and beam calibration is not repeated here.
Rdernico i [l-1] W.S. Kiger lli, Neutronic Design of a Fission Converter Based Epithennal Beam for Neutron Capture Therapy, Nuclear Engineer Thesis, MIT,1996.
'[12] D.E. Wazer, R.G. Zamenhof, O.K.11arling, and 11. Madoc Jones, " Boron Neutron Capture 'herapy," in Radiation Oncology: Technology and Biology, edited by P.M. R nd J.S. Loeffler, W.D. Saunders Company, Philadelphia, 1994.
O l2 SR# 0 9714 OCT 031997
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- 2. Ehsisn Connrkr Neuttnnic liesign
'O O' The neutronic design of the fission converter is summarized in this chapter. 2.1 General Description of thtFission ComencI Figure 2.1 is a view of the hilt Research Reactor (hilTR) and its existing components. An isometric view of the hilt fission converter beam and the new medical therapy room is given in Figure 2.2. Figures 2.3 and 2.4 show top and side views of the facility. The fission converter tank will be located to the exterior of the reactor's graphite rencetor in the region previously occupied by the thennal column. The fission convener plate will be centered on the 35.6 cm (14 inch) window in the graphite reDector. A cadmium cunnin (Section 4.1.1) located between the fission converter plate and the hilTR core is used to control the neutron Hux from the hilTR and, therefoie, the fission converter power. To reduce radiation dose in the medical therapy room, additional shutters are located in the collimator region (Section 4.1.2). The neutron Dux from the reactor is thennalized in the D2 0 and graphite reDectors. This thennal Oux is converted to a I.ssion spectrum by the fission convener plate. ' A filter / moderator is then used to tailor the fission spectnim to climinate unwanted fast neutrons and photons withoitt signincantly decreasing epithennal neutrons (1 eV to 10 kev). A collinator then directs the epithermal neutron beam onto the patient. Cooling of the fuel is provided by forced convection of either 112 0 or D 20 enclosed in an AM061 tank. The fission converter's primary coolant will be cooled by a heat exchanger for which the secondary side is connected to the hilTR II's cooling tower through the reactor's secondary coolant system. The fission converter's thennal hydraulic design is discussed in Chapter 3 of this report. A new medical therapy room will be located in the space originally occupied by the hilTR's hohlraum A heavily shielded door provides access to this medical therapy room. The medical therapy room will be adequately shielded to maintain a low radiation dose outside the facility.
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3
, 2.2 Evel Design l The fission converter plate consists of an array of MITR ll fuel elements arranged in a fuel grid plate auembly that is located in the fission converter tank. The fission converter plate can contain up to eleven MITR il fuel elements. l The fission convener will nonnally use eleven MITR Il fuel elements with a small clearance between them. The gaps are blocked by the grid plate so the amount of the bypass flow is rninimited. Fewer elements may be used provided that dummies having the s une exterior geometry as an element are employed so as to maintain Guld flow behavior and provided that the thennal hydraulle criteria established in Chapter 3 of this report are satisfied.13ecause the fission converter uses the same fuel element as the MITR il, the fuel I specifications given in MITR Il Technical Specification # 5.2 shall apply for the fission converter. 2.3 Qilkalltv and Power Distrib ittlen 2.3.1 Criticalltv Calndation for the Fis1[on Converter O The critiedity of both the fission converter and the coupled core convener system were studied. The Monte Carlo N Particle (MCNP) code has ten used for neutronic studies of the fission converter. The MCNP model of the MIT Research Reactor has been extensively validated in the core region and in the thermal coh.mn region where the fission converter will be located [Ref. 21,2 2). To calculate the kerf of the fission converter, criticalite alcution v : fission converter beam and core albedo model were performed usi' < K cede opvia c" >1CNP [Ref. 2 1). The cfIcetive multiplication factors (ken) cah Med v bre M abk: 2 . The kerr values calculated for2a D 0 cooled system are O ; Or pu of v w MrNil fuel and 0.344 for fresh MITR Il fuel. For an 1120 coo' :pm tb w muel slues are 0.514 and o.618 for partially spent and fresh MIT 4 (v.i ivrn A dy. iw a the kert predicted is much smaller than unity, a criticality weeu o o 6 2-u SR# 0 9714 OCT 031997
O Table 2.1: Gilicality Analysis of the Fission Converter 235 Coolant Fuel (g U) kertConverter Alone D02 , 312 0.26810.001 (Spent hilTR Il fuel) D02 510 0.34410.001 (Fresh MITR li fuel) I12 0 312 0.51410.001 ppent MITR Il fuel) l 112 0 510 0.61810.001 (Fresh MITR ll fuel) Nnic: The statistical uncertainty listed with each value represents one standard deviation [Ref. 2 11 O Table 2.2: Reactivity Change Associated with Onening of the Cadmium Cunnin to the Fission Converter Coolant Fuel (g 2350) kerr Reactivity Curtain Closed _ Curtain Open ( W k) DO2 312 1.0045510.00048 1.0049010.00036 0.0003510.00060 (Synt MITR- (45176 mp) 11 Puel) 112 0 510 1.0041710.00051 1.0054310.00050 0.0012510.00071 (Fresh MITR. (159190 m ) 11 Fuel) Nois: The statistical uncenalnty listed with each value represents one standard deviation [Ref. 2 3) 27 SR# 0-97-14 OCT 031997
2.3.2 & activity Wonh of the Fission Convem l The kerr values of the coupled MITR core convener system were calculated similarly using the K code option of McNP lRef. 2 3). Criticality calculations to estimate l the reactivity insertion to the MITR caused by fission converter operation were done to assess the potential for interaction between the reactor and the fission converter. These results are shown in Table 2.2. These calculations show that operation of the fission converter by opening the cadmium cunain will cause a change in reactivity of 0.0003510.00060 Ak/k (45i76 mp) for a D2 0 system using pasily spent fuel and O.0012510.00071 Ak/k (159190 n.p) for an H2O system ming fresh fuel. The cases cited bound the anticipated range of reactivities. (Notc: Yhe reactivity worth of the fission converter was also estimated using the diffusion theory code CITATION. This was done by modifying the CITATION input file for the MITR-il model to include a simplistic model of the fission converter facility in the thermal column region both with and without fuel. This method of calculation is recognized to be less accurate than the Monte Carlo approach for the geometry in question. The results of these runs show the reactivity worth to be 0.0000257 ok/k (3 mp) [Ref. 2 4).) O The actual scactivity worth of the fission convener will also be determined during pre operational testing (Section 8.5), but the above referenced calculations indicate that it is within the existing technical specification limit (TS# 6.1) of the MITR il for moveable experimental facilities (0.002 Ak/k or 254 mp). MITR il Technical Specifications provide several approaches for limiting the reactivity associated with an experimental facility. Specification No. 6.1 irt 'oses limits depending on whether the experiment is classified as moveable, non secured, or secured. Specification No 6.4 imposes a limit on the allowed period for experiments related to reactor control research. The latter approach provides more flexibility because it permits any combination of reactivity and a rate of change of reactivity provided that a certain minimum period is not exceeded. Accordingly, this approach is used for the fission converter. The Tission converter will be exempt from MITR il Technical Specification No. 6.1. Instead, a reactor period limit of 50 seconds shall apply during opening of the cadmium shutter that controls the fission convener power Reactor controls can be used to compensate for the change in reactivity resulting from the presence of the fission converter as long as the reactor period is longer than 50 seconds. O 28 SR# 0 9714 OCT 031997
2.3.3 Eission Conve21er Power 3 (O The hionte Carlo N Particle (51CNP) code has been used to calculate the fission convener power. 'the calculated values are summarized in Table 2.3. As shown, the maximum power generated by the fission converter is 251 kW with hilt reactor power at 10 h1W [Ref. 2-1]. At this power level, each fuel elenient generates an average of 22.8 kW compared to an average poiver of 208 kW per fuel element in the hilTR Il core (5 h1W, 24 fuel elements). Table 2.3: Calculated Fission Converter Power Fission Convener Fission Converter Coolant Fuel (g 235U) Power at 5 hiW Power at 10 h1W Reactor Power Reactor Power (kW) (kW) D0 2 312 81.5 163.0 (Spent hilTR il Fuel) 10.3% i.0.3% D20 510 105.4 210.8 (Fresh hilTR il Fuel) 10.2 % 10.2% 1120 312 83.4 166.8 (Spent hilTR il Fuel) 10.2% _ 10.2 % 112 0 510 125.5 251.0
~
(Fresh hilTR il Fuel) i0.2% 10.2% Noic: Data are from the coupled core criiWality calculations. The statistical uncenainties listed as a percent for each value represents one standard deviation [Ref. 2 1]. 2.3.4 Fission Converter Radial and Axial Power Distribution The power distribution within the fission converter plate was calculated using h1CNP [Ref. 2 5]. These calculations were done aumaing that the reactor operates at 5 h1W and that fresh hilTR fuel elements are used in the fission converter. Table 2.4 lists the fission convener power, fuel plate maximum power, and the corresponding nuclear hot channel factor. The fuel plate maximum power is used in Chapter 3 as the maximum power in the coolant channel for the thermal hydraulic limit O 2-9 SR#-0-97 14 OCT 031997
O Table 2.4: Results of Fission Converter Radial Power Distribution [Ref. 2 5) 12Tsion Fuel Pldte Fuel Plate Noelear Converter Avg. Power Max. Pour llot Channel Power (kW) (kW) (kW) Factor
- ll 20-cooled 125.7 0.76 1,12 1,47 Fresh Fuel l
D20-cooled 104.9 0.64 0.98 1.53 l Fresh Fuel
- Nuclear llot Channel Factor = Fuel Plate Max. Power / Fuel Plate Avg. Power Table 2.5: Axial Power Distribution for the Fission Converter [Ref. 2 5]
Ileight Above Fuel Avg. Linear Power Standard Center Line (cm) Distribution (W/cm) Deviation (%) 27.2 18.4 3.95 24.9 13.2 4.24
- 2 I .3 16.4 3.20 16.6 19.9 2.91 11.8 21.2 2.93 7.1 21.3 2.78 2.4 22.7 2.71 2,4 23.0 2.72 7.I 23.8 2.84 11.8 20.4 2.9J
-16.6 I8.6 3.04
-21.3 17.2 3.20 24.9 I 5.8 4.48
-27.2 17.7 4. I i f) b 2-10 SR# 0 9714 OCT 031997 1
I
calculations. Table 2.5 shows the vertical power profile of the hottest plate. The vertical power pofile is similar to a cosine profile except that higher power is predicted to occur at the ends because of higher moderation (water peaking) at the top and bottom of the fuel elements. 2.3.5 Fuel Element Orientation The preceding calculations were made for fuel that is oriented so that the individual fuel plates are " edge-on" toward the MITR as shown in Fig. 3.4. If the elements were to te rotated by 90'so that one entire plate were facing the MITR calculations show that the peaking in that plate would be signincantly greater [Ref. 2 6]. Administrative procedures will be used to ensure that fuel elements are loaded with proper orientation. (Mok: , Rotation of a fuel element by 90' is physically impossible unless it is first invened.) 2.4 Filter / Moderator Thermal neutrons from the reactor are converted to a fission spectrum by the converter plate. A filter / moderator is then used to tailor the fission spectmm to climinate l unwanted fast neutrons and photons without a significant decrease in the epithennal q neutrons (1 eV to 10 kev). Filtra'.lon is accomplished through the use of resonance V scattering materials with large scattering neutron cross sections in the fast energy range,1/v behavior in the thennal region, and a relatively low, flat cross section at epithermal energies [Ref. 2 l]. To reduce contamination from thermal neutrons, thennal neutron absorbers 6 such as Cd, 'Ul f, or Li can be used. liigh Z material such as bismuth or lead can be used to reduce the contamination from photons. Material selection for the filter will be made to ensure that a phase change because of elevated service temperatures is precluded under both normal and accident conditions. In addition, the selection will minimize decomposition and accumulation of long tem) activities [Ref. 2 11, 2.5 Collimatet The resulting beam from the filter is focused by a collimator that is lined with a layer oflead (~15 cm thick). In addition to positioning the epithermal neutron beam onto the patient, the collimator region serves as the location for shutters that control delivery of _ the beam (Section 4.1.2).
.]
2 11 SR#-0-97 14 OCT 031997
References
!2-i) W.S. Kiger 1lI, Neutronic Design of a Fission Converter Ilased EpithennalIIcam I for Neutron Capture Therapy, Nuclear Engineer Thesis, MIT,1996.
[22) E.L. Redmond II, J.C. Yanch, and 0.K. liarling, " Monte Carlo Simulatl 1 of the Massachusetts Institute of Technology Research Reactor," Nuclear TechnoIngy, i Vol.106, April 1994. [23] File Calculation (Reactivity Change Associated with Cadmium Cunain Opening - MCNP Calculations) [2 4) File Calculation (Reactivity Change Associated with Cadmium Cunain Opening - ;
~
CITATION Calculation) [2-5) File Calculation (Fission Convener Plate Power Distribution Calculations) l [2-6) File Calculation (Fission Convener Plate Fuel Element Orientation)
~
O i O 2 12 SR#-0 9714 OCT 031997 l ._ _ _ . _ . . , - .-.-_ _ - . . ~ . _ _ - _ . _ _ _ - _ . . _ _ _ _ - - - .
- 3. Fission Coinerter Theriltal liydrmtlle Desigt1 V
The thermal hydraulic design of the fission converter is summarized in this chapter. The fission converter has been designed to accommodate either an 2110 or a D;O primary coolant system. The design limits include consideration of the consequences of credible deviations from the operating wnditions as well as allowance for manufacturing tolerances. In addition, the design limits of the thennal hydraulic system are chosen to nr vide a reasonable safety margin beyond the desired operating range. Specifically, the limiting condition for operation is set to prevent incipient boiling in the fueled region of the converter, Such a conservative design assmes a wide margin to the real safety limit set to prevent fuel cladding failure. Accident analysis of the facility is covered in Chapter 6 of this report. l 3.1 Gtneral Description of the Themial livdraulle System The MITR fission converter is designed to be cooled by forced convection using either 112 0 or D 20 as the primary coolant during nonnal operation. The primary coolant is enclosed in an Al.6061 tank and is cooled by a heat exchanger for which the secondary side g is connected to the MITR's cooling tower through the MITR's secondary coolant system. l ( A schematic of the fission converter heat removal system is shown in Figure 3.1. ) Provision is also made to allow the fission convener to operate at low power with natural circulation. Natural circulation is achieved by removing the inlet pipes, which are used for forced convection, from the downcomers. The purpose of enabling natural circulation operation is to facilitate activities such as flux measurements in the fueled region. In order to setup this type of activity, the fission converter's top shield lid ir removed. The activity can be perfonned either with the lid in place or removed. Analysis of the radiation level associated with operation of the fission converter in the natural circulation mode with the lid remos ed is given in Appendix 3.2. The fission converter primary coolant system consists of two pumps, one heat exchanger, a cleanup system, a make up coolant system, a cover gas system, and associated valves and piping. The pumps can be operated singly or in parallel. The reason for using two pumps is to provide redundancy, part of the fission convener primary coolant flow is diverted to the cleanup system, which consists of an ion column and filters, to purify the coolant. The cleanup system could also be used to adjust water chemistry if
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O 31 S R#-0 14 OCT 031997 I
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<3 lon Column Figure 3.1 Schematic Drawine of Fission Converter Heat Removal System m 3-2 S R#-0 14 OCT 031997 I w ] necessary. An auxiliary pump is used to maintain a small flow through the clean up loop. p It may also be used to maintain flow through the heat exchanger after fission converter b shutdown to remove decay heat. At the top of the fission converter tank, a cover-gas system is provided. During normal operation, pressure in the converter tank is maintained at atmospheric by this cover-gas system. Protection against overpressure is afforded by a relief valve on the gasholder. This valve is set at about 5 psig. In addition, there will be a rupture disc set at 7.5 psig, Figure 3.2 shows the cover gas system. The tank will be tested at a pressure of 10 psig. llelium will be used as the cover gas. The secondary coolant is light water. The secondary coolant from the heat exchanger is merged with that from the MITR heat exchangers and then sent to the MITR's cooling towers. The assion converter secondary coolant system is discussed in Section 3.1.2. 3.1.1 Primary Coolant System 3,1,1,1 Fission Converter Tank O V The fission converter plate is enclosed in a tank made of Al 6061. Figures 3.3 and 3A illustrate the fission converter tank design. The tank wall is 1.27 cm (0.5 inch) in thickness. The downcomer wall thickness is 1.27 cm (0.5 inch). The fission converter design pressure and temperature are below the limits of the ASME Boiler & Pressure Vessel Code. Hence, the fission converter tank is exerr3 from the specifications of that code. However, ASME code sections 11 and IX for materials and welding specifications will be used voluntarily to ensure that sound practice is followed in construction of the tank [Ref,3-12]. The larger upper portion of the fission converter tank rests on a concrete upper shield block which serves as the primary tank support. This upper shield block is supported by the reactor concrete biological shield. A removable aluminum block is placed adjacent to the fuel elements on the patient side of the fission converter. The purpose of this aluminum block is to provide flexibility in the amount of the moderation of the neutron beam to the fission converter medical therapy room. For example, if light water is used as the fission converter primary coolant then the presence of the aluminum block would reduce moderation of the neutron beam. The pri. mary inlet and outlet pipes which penetrate the top shield lid can be removed for natural circulation operation. 3-3 SR#-0-97-14 OCT 031997 O . r- - - - - - - -C)<}- - Sampler- - --{><} q To ~MITR-Il l l_ l~ stack l l-A l' l I I Heated I r-- . recombiner-- ---d ! r--i l I I l l- ; IPRV- l _, _ J l < % A- l k- l- Blower-0 1l [ r- ,--- ------- l Rupture [ 4 _ ;
- disc l l l ~l
, I. _l Gas holder .; 4_ _ _ or precision ' pressure j l , reducer valve ' Storage A Slope drainage and-gas &. liquid makeup . line coolant Helium tank gas tank Figure 3.2 Schematic Drawine of Fission Converter Cover Gas System O v 3-4 S R#-0 14 .OCT 031997 COOLANT OUTLET PIPE [ (I.D 1.5") p I-- ' C m - lp C00LANT INLET PIPE (I.D 1") \ WATER LEVEL +7 \s (l g e SURFACE 1 i ! TANK WALL DOWNCOMER W LL% /~(THICKNESS 1/2") (1/2") 7 i e ~ l ! f cALUMINUM BLOCKS f - -11 MITR-II FUEL ELEMENTS l l E01C: Detailed geometry of the fuel elements l l :: is shown in Figure 3.4 / \ Figure 3.3 Isometric View of Fission Converter Tank Design r^N / SR#-0-97 14 OCT 031997 l ilIl!llIl \lll s m d n c e o 4 m ht 3 c o o 6 1 b 1 6 t a 0 se 1 r t u c u r t s t n e r f e f i n d f i o o k t c e c e s o S u a l r a c B e e w o b l a c m u n i L n i r i k t e n m a m u Tr m l a y A e t s r t e o v n n / o i k s n a m s C n o u e i T m m r i s s e c c o F t o 9 2 C e o a 8 1 ht N . r f 0 1 o o 3 3 t 1 c w m R a e i V l e a n o m i t c e s s a s o r C \ s 4 3 t) ns ee t u i F e r u g ma l m c leP 2 E 9 5 l 1 6 e( 1 u F o : mN?9$ a oO e G$ l; : 3.1.1.2 Fission Converter Fuel flousing Grid Design O V The fuel elements are arranged in a closed-packed array in a housing with a fixed fuel grid plate and side walls as shown in Figure 3.5. The fuel elements are held down by gravity. The maximum hydraulic lift created by the coolant is much less than the element weight. The results of the calculations of the hydraulic lift are given in Table 3.1. Table 3.1 Fuel Restraint Calculation Results [Ref. 3-1] Fuel Element Downward Force 34.6 N (7.8 lbf) j Maximum flydraulic Lift Force Generated
- 7.5 N (1.7 lbf)
- based on primary flow rate of 10 kg/s (162 gpm),
50% error band is assumed for both primary flow rate and friction pressure drop calculation. O d 3.1.1.3 Cover-Gas System A cover,-gas system is provided at the top of the fission converter tank. All free surfaces of the coolant are blanketed with helium which is supplied from a gasholder. The helium blanket perfoir . * 'etions:
- 1. Prevents air with entrained moisture from entering the system and coming in contact with the D 2O (if used), and degrading it.
- 2. Inhibits nitrous oxide formation from moist airin the presence of high radiation fields.
- 3. Provides an inert nonradioactive atmosphere that minimizes41Ar production.
- 4. Provides an inert atmosphere for the circulation of the byproducts of radiolytic decomposition through a recombiner.
3-7 SR#-0-97-14 OCT 031997 .A l lI1II{l 1)i1) ~ "5 ~ o 0 4 9 0 Lz i t 2 3 i /, 1 M i ~ 7 7 I 7 l 2 I ~ 7 - 7 l 7 i 2 t "7 7 I 7 2 h t e ~ l a 7 7 1 P - d 7 i 1 r 2 t G l ~ e 7 u 7 I F 7 A r 2 l i - e t A e r '6 v 3 ~ N n 9 7 O o 1 3 7 7 I I T C C n E n E o S i s i s ~ 7 F - 7 1 7 5 l 3 2 i e r I u ~ g - 7 i - 7 1 F ' 7 2 H 5 r4 ~ 7 e 7 f x 1 7 m a5p 2 h 2 y H - C 1 T _ ~ ~ 7 _ 7 1 _ 7 2 H "9 _ - 0 3 . - ~ l _ 7 7 l . 7 r 2 E A y Y* mk&S he C ll
- 5. Permits monitoring of fission product gas concentrations in the cover gas.
Abnormal levels would be indicative of an incipient clad failure of a fuel element. IIelium from a high pressure manifold is supplied through reducing stations to a constant pressure gasholder or a precision pressure reducing valve. From there, helium is supplied to the free surface in the fission converter primary coolant system. If the pressure builds up, helium will be vented to the MITR stack through a relief valve. The production of a flammable concentration of either D2 or 112 from the disassociation of either heavy or light water in the fission converter is a concern because the fission converter will be a closed system. Accordingly, a recombiner will be installed and utilized. The MITR's heavy water reflector is also a closed system and it too operates with a recombiner, MITR Technical Specification #3.3 specifies that the reflector's recombiner be operating whenever the reactor is in use and that, in the event of a recombiner failure, reactor power be reduced to a nominal level (200 kW for D 2 ,200 kW for 112 ) unless sampling shows that the concentration of flammable gas is quite low (27o for D2,1% for 112). Such stringent requirements are not required for the fission convener because: O V
- 1. The maximum thermal neutron flux in the reflector is 5x1013 n/cm2 s whereas that in the fission converter is calculated to be 1.3x103 I n/cm2 s, a factor of 380 less.
- 2. Patient set up and removal time is, under the best of circumstances, an hour or more. Thus, even if in routine use, the fission converter's duty cycle (i.e., time at power) would be at most a few hours per day.
Given the above, it is more appropriate to operate the fission converter recombiner periodically as a maintenance requirement rather than routinely as an item of required equipment. It is therefore proposed to operate the fission converter recombiner for five hours per month under the conditions specified by its manufacturer. (Noir The MITR reflector's recombiner operates continuously, 720 hours / month. There is more than a factor of 300 difference in the flux levels. The volumes of the two systems are essentially equivalent,260 gals for the reflector versus 180 gals for the fission converter tank. lience, the fission converter recombiner would be needed about 2.5 hours / month for a 1007c duty (3 V 3-9 S R#-0 14 OCT 031997 cycle. The specified five hour minimum is conservative by a factor of two and places no o restriction on the duty cycle.) b The fission converter cover gas is also circulated past a radiatiori monitor (pancake GM with stainless steel window or equivalent detector). This monitor would sense any buildup of fission product gas such as would occur should there be an incipient clad defect. 3.1. l A Clean-Uo System A portion of the fission converter coolant from the heat exchanger flows through a pump, an ion column inlet filter, a mixed bed ion exchanger, and an ion exchanger outlet filter. The purified coolant then rejoin the rest of the primary coolant that is being pumped back into the converter tank. The clean-up loop is located downstream from the heat exchanger to maintain a low coolant tempeiature so as to protect the temperature-sensitive resin. Two conductivity cells (one at the inlet and one at the outlet of the clean-up loop) are used to monitor the conductance of the primary coolant flowing through the clean up A loop. U This system can also be used to remove decay heat from the fission converter fuel and to adjust the pH of the coolant should that be necessary. 3.1.1.5 Make-Un Coolant System The coolant levelin the fission converter tank is monitored by a level probe. If the fluid level in the converter tank is below a preset level, additional coolant can be added manually from the make-up water tank. 3.1,1.6 Coolant and Cover-Gas Samnling The purpose of the coolant sampling system is to permit access for sampling the primary loop at any time during operation. Typical analyses might include gross p-y activity as well as pH and chloride concentration. If the coolant in use is D 2 0, a tritium analysis should also be performed. The frequencies for these analyses are monthly for p gross p-y, pH, and chloride and quarterly for tritium and isotopic purity. The latter is V 3-10 S R #-0-97-14 OCT 031997 - required only if D 20 is used as the primary coolantc These areihe frequencies followed for 4 the MITR. The purpose of the cover gas sampling system is similar. It pemiits the sampling of radioactive gases and dissociated gases (i.e.,11 2 and D2 ). i 3.1.2- Secondary Coolant Systent - The secondary coolant system consists of piping and valves. The secondary light water coolant from the MITR Research Reactor is used for the fission converter. Cool secondary coolant is drawn from the basins of the MITR cooling towers. This flow is combined into a single pipe which penetrates the reactor's containment shell leading into the = equipment room. The flow is then separated to provide secondary coolant to the MITR's components and the fission converter. The pump provides flow to the shell sides of the fission converter heat exchanger. The exit flow from the fission converter heat exchanger combines with the exit flow from the MITR heat exchangers to form a single exit pipe . which returns to the cooling towers, p The MITR II's secondary coolant system is designed to remove 6 MW of heat with normal operation at 5 MW The addition of a 125 kW heat load is within the capability of the reactor's secondary system. (Nels: In the event that the MITR's power is upgraded to 10 MW, the upgraded secondary system will also be sized to allow for this additional load, q - 250 kW maximum in this case.) In the event that the fission converter coolant is D2 0, then provision should be 4 made to preclude t' e possibility o.' tritium release in the event of a heat exchanger leak. The philosophy of MITR Technical Specification No. 3.8(4) is followed here as well. Namely. whenever secondary coolant is flowing through (ne fission converter heat exchanger, the - MITR's secondary coolant will be sampled d' aily_ for tritium and the level of the fission converter tank will be monitored either by an alarm in the MITR control room or by hourly readings. -'3;l.3 Material Selection . - Material selection for the fission convener is based on structural requirements, t . nuclear properties, availability, and cost. 3-11 < S R#-0 14 OCT 031997 All materials, including those of the converter tank, in contact with primary coolant, shall be aluminum alloys, stainless steel, or other materials that are chemically compatible with each other and with 110 2 and D 02 coolant, except for small non-corrosive components such as gaskets, filters, and valve diaphragms. 3.2 Basis for Titermal-Hydraulic Design The basis for the fission converter's thennal-hydraulic design is that, under conditions of forced convection,its primary coolant system removes 300 kW (50 kW more than the maximum fission convener power for 10 MW reactor power, fresh fuel ar.d light water coolant) of heat from the fission converter fuel and transfer it to the secondary coolant system without onset of nucleate boiling. Another design feature is that the system can operate at low power with natural circulation witk ? uccedirm the coolant temperature limit. The requirement for natural circulation operation is to facilitate activities such as Oux measurements in the fueled region for which it is desirable to have minimum disturbances from the flow, Sufficient margin for possible deviation of parameters is taken into account in the thermal hydraulic limits calculations. i b The system also fulfills several other functions. The coolant pool above the fuel elements provides shielding for the fission converter top and a reservoir of coolant as a heat sink for emergency conditions. Also, the fissica converter primary coolant system acts as a barrier against the escape of fission products to either the reactor building or to the secondary coolant system. 3.3 Comnutational Method for the Fission Converter Thermal Hydraulic Design A computer program (MULCH FC) was written to model the primary and secondary coolant systems of the fission converter. The fueled region is modeled in detail to calculate the coolant and the fuel clad temperatures as a function of axial position for both average and hot channels. This computer program can be used to calculate steady-state operating conditions, safety limits, and the limiting safety system settings of the fission converter. It has been used in the current analysis for the thermal hydraulic limits calculations. Methodology and key correlations used in this computer program are discussed in the following sections. A brief description of MULCH-FC is given in Appendix 3.1 to this chapter. 3-12 SR#-0-97-14 OCT 031997 u
- n. 3.3.1 Methodoloey V
~~ Both the fission converter's primary and secondary coolant systems are modeled in the MULCH-FC code. Each component of the coolant system is modeled as a control volume. The energy equations for the control volumes are then solved simultaneously to , obtain the temperatures associated with each control volume. The fueled region is further divided into hot channel and average channels. For the calculations described here, each channel consisted of ten axial nodes. Fuct and coolant temperatures are solved for each node using energy conservation equations. For forced convection, the flow rate is taken as e constant. For operation with natural circulation, the pressure drop equation is solved to obtained the flow rate through the fueled region. The temperatures in the average and hot channels are then calculated based on the natural circulation flow rate. I Safety limits (SL) and limiting safety system setdn.a.s (LSSS) can be calculated based on the steady-state operating conditions of the hot channel, it is assumed in the current analysis that the hot channel is the coolant channel that produces the highest power j and has the lowest primary flow rate among all the coolant channels. To calculate the SL Q and LSSS, engineering hot channel factors (Section 3.4.3) are first applied to the hot channel to obtain the maximum coolant and fuel clad temperatures. Appropriate correlations are then used to determine if the limits are exceeded. 3.3.2 Maior Correlations Used in MULCH-FC 3.3.2.1 Correlation for Onset of Nucleate Boiline (ONB) Sudo et al. suggested the Bergles-Rohsenow correlation for the prediction of ONB for narrow rectangular coolant channels (Ref. 3-2]. This suggestion was based on comparisons of correlations with experimental data. Sudo et al. also concluded that the Bergies-Rohsenow correlation predicts the lower limits of the measured ONB temperatures for given heat fluxes and there exists a margin between the predicted and measured ONB temperatures. The Bergies Rohsenow correlation predicts the fuel clad temperature at which the ONB occurs. ' 0(v , 3-13 , SR#-0-97-14 OCT 031997 n ^ 0 .0m OA63 P 9 Iclad. OND = Tsat + 0.556 pi,t56 (3 I) where Tctad, ONn is the fuel clad temperature ('C) at which ONB occurs. Tsai is the saturation temperature ( C), q" is the local heat flux (W/m2 s), and P is the pressure (bar). 3.3.2.2 Correlations for Onset of Significant Voiding (OSV.) 4 Two correlations are used in the MULCH-FC to predict OSV. The first one is the Saha Zuber correlation which has been widely used to predict OSV (Ref. 3-3]. q' D, Cpr Xeosy = - 0.0022 g if Pe < 70000 (3.2) and Xeosv = - 154 HG I *D (3'3) O where Xe osv is the thermal equilibrium quality at which OSV occurs, Xe equals (H-Hg)/Hrg, and H-is the local liquid enthalpy, Other symbols are defined as: GDep Cr Pe is the Peclet number Pe = , q" is the heat flux (W/m2 s), Crp is the liquid heat capacity (J/kg *C), kr is the liquid thermal conductivity (W/m 'C), De is the equivalent diameter (m), Hfs is the difference between saturated gas and liquid enthalpies (J/kg) Hrg = Hg - Hr , and G is the mass flux (kg/m2 s), The second correlation was recently proposed by Kowalski et al. to predict OSV for p coolant channels with finned surfaces (Ref. 3-4]. It is: () 3-14 S R#-0 14 OCT 031997 p o.18 Xc (3A) osv = 004460 H g The equilibrium qualities predicted by the above two relationships are compared and the lesser of the two, which is the one that corresponds to a lower coolant temperature, is used to check for OSV. This yields a conservative prediction of OSV using the MULCil FC code. 3.4 Fission Converter Oncrating Limits The operating limits include the power deposition factor, nuclear hot channel factor, fueled region coolant flow factor, flow disparity factor, and the engineering hot channel factors. These factors are used in the calculation of the thermal hydraulics limits (Section 3.5). The nuclear hot channel factor, fueled region coolant flow factor, and the flow disparity factor will be measured during the initial startup test (Sections 8.4 and 8.6(Sa)). If the measured values are less conservative than the ones used in the existing calculations, l these calculations will be performed again and the operating condition or limits adjusted as p necessary. The engineering hot channel factors are determined by statistically combining \ the uncenainties associated with design, calculation, and measurement. These uncertainties are governed by the MITR quality assurance program as well as by testing and calibration procedures forihe fission converter. 3,4.1 Power Deposition Factor (Eo)_imd_ Nuclear Hot Channel Factor (Fgc) The power deposition factor defines the percentage of the fission power deposited in the fueled region (both fuel and coolant) of the fission converter tank. It is expected that at least a few percent of the fission power will be deposited outside the fission converter's fueled region because of the long mean-free path of gammas and fast neutrons compared to the small size of the fission converter (Ref. 3-5]. However, the power deposition factor for the fission converter is conservatively assumed to be 1007c in this analysis. That is,it is assumed that there is no energy escaping the fission converter tank. The nuclear hot channel factor defines the ratio of the maximum power deposited in the hottest fuel plate to the average power per fuel plate. The nuclear hot channel factor ( 5 used m the current analysis is derived from the radial neutron flux distribution calculation ( 3 15 SR#-0 97-14 OCT 031997 of the fission converter using MCNP (Section 2.3.4). The nuclear hot channel facies used i q in the thermal hydraulic calculations is 1.53, which is obtained for the conditions that D2 0 f (_./ coolant and fresh fuel elements are used in the fissien converter. This combination of coolant and fuel is the limiting case. To ensure that the current analysis covers the possible operating conditions of the fission converter, the following condition needs to be satisfied. FpxFilC s 1.0 x 1.53 = 1.53 (3.5) 3.4.2 Fueled Region Cov. ant Flow Facter (F I )rd Channel Flow Disoarity Factor (df) The coolant flow factor is defined as the ratio of the fission converter primary coolant flow which actually cools the fueled region to the total flow. Ideally, the fueled region should be designed so that 100% of the coolant flows through the fuel elements. liowever, part of this flow bypasses the fueled region because of design / manufacturing tolerances such as clearances between the fuel elements. The MITR ll has a core coolant Gow factor of 0.921 which was determined experimentally during the reactor's initial /~ startup test. The same value is assumed in the current analysis for the fission converter. \ This is a conservative assumption because the MITR-ll has multiple paths which would allow bypass Dow. These include anti siphon valves, natural convection valves, in core sample facilities, and dummy elements. In contrast, in the fission converter tank, the only possible bypass flow paths are the clearances between the fuel elements and the clearances between the walls of the plate housing and the fuel elements. The channel flow disparity factor is defined as the ratio of the minimum flow to the average flow in the coolant channels. It is: dr= W"". g avg (3.6) where Wmin is the minimum flow rate measured in all the coolant channels and Wavg is the average flow rate in the coolant channels. The flow distribution in th< MITR-Il core was measured during the reactor's initial .Q startup test. The minimum flow through a fuel element is 937c of the average core flow k) 3-16 S R#-0 14 OCT 031997 ) rate [Ref 3 6]. The flow distribution within a fuel element has also been measured n experimentally using a dununy el ment. The ratio of the minimum channel flow rate to the d average channel flow rate within a fuel element is 0.929 [Ref. 3 6]. So, the worst case channel flow disparity factor of the MITR-ll can be c.dculated av dr = 0.93 x 0.929 = 0.864 (3.7) It is assumed in the current analysis that the fission converter channel flow disparity factor is the same as that of the MITR il. The minimum flow in a fission converter coolant channel is then calculated in the thermal hydraulic analysis using the following equation: Wmin n W ' ' x Fr x dr (3.8) where W e,iot is the total primary coolant flow rate, and Ne is the number of coolant channels in the fission converter fueled region. O d To ensure that the current analysis covers the possible operating conditions of the fission converter, the following condition needs to be satisfied: Fr X fd 5 0.921 x 0.864 = 0.796 (3,9) 3.4.3 Engineering Hot Channel Factors The engineering hot channel factors account for possible deviations from nominal design specifications that may affect the thermal hydraulic calculation results. Specifically, they are defined for channel enthalpy rise, film temperature difference, and heat flux [Ref. 3-7]. These parameters are divided into sub factors that can be combined either multiplicatively or statistically to obtain the engineering hot channel factors, it has been concluded that it is overly conservative to combine the sub-factors multiplicatively [Ref. 3-7,3-8]. So the statistical approach has been used in the current study. Table 3.2 is a summary of the engineering hot charmel factors. /O () 3-17 S R#-0 14 OCT 031997 4 The engineering hot channel sub factors considered in the current study include p those for reactor power measurement, power density measurement / calculation, fuel density k tolerances, flow channel tolerances, fuel meat eccent ; city, and %: transfer coefficient prediction. Numerical values for these sub-factors were rnostly adopted from the MITR-il S AR and reference should be made to the MITR il SAR for definitions and derivations of these sub-factors [Ref. 3 9). A " vertical" approacl. is used in the current analysis to calculate the maximum fuel clad temperature and the maximum coolam temperature using the engineering hot channel factors. This approach is the standard conventional method noted in Ref. 3 8. To calculate the maximum coolant temperature, use Tc, y = Tin+ F gi AT c (orli,y=Ilinc + F g iAli ifc boiling occurs) (3.10) .O U 3-18 SR#-0-97-14 OCT 031997 Table 3.2 [ (h) Engineering llot Channel Egetors Used in Fission Converter Thermal-Hydraulic Calg.nlations Enthalpy Rise Reactor power measurement 1.05 Power density measurement / calculation 1.10 Fuel density tolerar.ces 1.026 Flow measurement 1.05 Flow channel tolerances 1.089 Eccentricity 1.001 Fn, Statistical 1.154 Film Temperature Rise Reactor power measurement 1.05 Power density measurement / calculation 1.10 (] 'd Fuel density tolerances 1.05 Flow channel tolerances 1.124 Eccentricity 1.003 Heat transfer coef0cient .L20 Frr, Statistical 1.265 IIcat Flux Reactor power measurement 1.05 Power density measurement / calculation 1.10 Fuel density tolerances 1.05 Eccentricity 1.003 Fq Statistical 1.123 b' qts: The engineering hot channel factors are obtained by combining the sub-factors 1(fj - l}3' I/2 , where fj denotes sub-factors statistically using the equation F = 1 + .J [Ref. 3-8]. rm !v) 3-19 S R#-0 14 OCT 031997 L To calculate the maximum fuel clad temperattire, use: Tw, st = Tn + Fg; Arc + Pg AT w (3.11) where Tin _ is the coolant channel inlet temperature, I lijn is the coolant enthalpy at the channel inlet, uTe is the coolant temperature rise, Alle is the coolant enthalpy rise, ATw is the 61m temperature rise (temperature difference between coolant and clad), Fu is the engineering hot channel factor for enthalpy rise, FrN is the engineering hot channel factor for 61m temperature rise, Te,st is the maximum coolant temperature because of design / manufacture deviations, lic.ht -is the maximum coolant enthalpy because of design / manufacture deviations, and Tw,ht is the maximum fuel clad temperature because of design / manufacture deviations. v Notice that ATc (or AHe), and ATw are calculated for every axial node in the hot channel with the operating limits taken into account. Tc.ht (or lic,st) and Tw.ht are then calculated and checked if the safety limits or the limiting safety system settings are exceeded. 3.5 Thermal Hydraulic Limits The fission converter has been designed to operate in a safe manner under all credible conditions. The thermal hydraulic design limits have been chosen to provide a safe margin beyond the desired operating range. Calculations of the design limits also include considerations for deviations from design specifications. 3.5.1 Definitions of Safety Limits and Limiting Safety System Settings The safety . nits for the fission converter are established to maintain the integrity of the fuel clad although aluminum melts at approximately 660 C (1200 F),it begins to o soften signif.cantly at about 450*C (842 F). The softening temperature is therefore used as b 3-20 SR#-0-97-14 OCT 031997 the criterion to guarantee the structural integrity of the fuel elements (Ref. 3-9), Critical p heat Dux (CliF) is normally used as the criterion of fuel overheating. Ilowever, because V the coolant flow path in the core is a multichannel design, there exists the possibility thrt now instabilities could occur before reaching CilF limitations. If How instability does occur first, it would have the effect oflowering the flow rate to the hot channel signi0cartly and thus lowering the burnout limits. Flow instability is a complicated phenomenon and the changes in flow rate are very difficult to predict. A conservative assumption to us: the onset of flow instability (OFI) in the MITR Il SAR produced a safety limit for tb maximum steady state power that is considerably below values based on applicable CilF correlations (Ref. 3 9] Various correlations have been developed for the prediction of OFl.110 wever, the effect of an axial heat flux distribution is not included m these correlations. OSV, on the other hand, can be more accurately predicted for various heat flux distributions. Also it has been observed experimentally that OSV occurs before OFI (Ref. 310), llence, OSV is assumed as the criterion for the safety limits in the current study. The limiting safety system settings (LSSS) are established to allow a sufficient margin between the normal operation conditions and the safety limits. Or set of nucleate g boiling (ONB) is chosen as the criterion for the LSSS derivation. This guarantees that V boiling will not occur anywhere in the fueled region as long as the limits are not exceeded. Specifically, the LSSS is set for: Operation with forced convection:
- 1. The maximum fission converter power (P),
- 2. The maximum steady-state average primary outlet temperature (Tout),
- 3. The minimum primary flow rate (W p), and 4, The minimum coolant level in the fission converter tank (H).
Oneration with natural circulation:
- 1. The maximum fission converter power (P),
- 2. The maximum fission converter tank mixing temperature (Tmix), and
- 3. The minimum coolant level in the Ossion converter tank (H).
Thus, the fission converter can operate with all four (for forced convection) or three (for natural circulation) parameters simultaneously approaching their limits without there being ,3 boiling in the fueled region. D 3-21 S R#-0 14 OCT 031997 3.5.2 Derivation of the Safety Limits ("'N O The safety limits for the fission converter are calculated based on the conservative assumptions that:
- 1. Fp x F lic = 1.53 for the hot channel (Section 3.4.1),
- 2. The hot channel has the minimum flo.v rate, that is. Fr x dr = 0.796 (Section 3.4.2), and
- 3. The engineering hot channel factors summarized in Table 3.2 are appropriate.
The axial power distribution used in these calculations is the same as that listed in Section 2.3.4. The calculations were done using the MULCil FC code which is described in Section 3.3. Figure 3.6 shows the resuhs of the fission converter safety limits calculatien for forced convection. The safety limits were determined for three different coolant levels (normal coolant height is 2.6 m above the top of the fuel). Iterations were made of th I p fission converter power and the average primary outlet temperature for given primary flow
- d rates to find the points corresponding to onset of significant voiding (OSV). The curves show the combinations of fission converter power (P), primary flow rate (Wp ), and average primary outlet temperature (Tout) for a certain coolant height (H) at which OSV may occur in the hot channel. Points to the left of the curve represent operating conditions at which OSV will not occur as long as the operating limits are satisfied.
Figure 3.7 shows the results of the fission converter safety limits calculation for natural circulation. The safety limits were determined for a coolant level of 2.4 m, which is the top of the downcomers. Iterations were made of the fission converter power and the fission converter tank mixing temperature to find the points corresponding to OSV. The natural circulation flow rate is determined using the pressure drop equation based on the fission converter power and mixing temperature. The curve shows the combinations of fission converter power (P), and fission converter tank mixing temperature (Tmix) for a coolant height of 2.4 m at which OSV may occur in the hot channel. Points to the left of the curve represent operating conditions at which OSV will not occur as long as the operating limits are satisfied. n 3-22 S r.#-0 14 OCT 031997 i 9 ,,,, ,,,,, ,,, ,,,.,,,,, 8 - H =2.6 m -- --+ -- - ; H=2,I m 3 l 7 -- --+-- + -- P (kW) : : . Wp (gpm) . H=1.6 m i > . 6 -- -- = l l 5 -- +- = l . 4 ' ' ' ' ' ' 60 65 70- 75 80 85 Tout ( C) Figure 3.6 Fission Converter Safety Limits for Forced Convection Q\. V 3-23 S R#-0-97-14 OCT 031997 ,m V) i 40 .... , ,.... ........ .........i.... .... i i ; - i I { !, J I i ! ! - 35 --- a -t- -- *- II = 2.4 r ~ ~- . t . ! i , , l ; i - }Q . . . 9... ...p... , . . . . . . 4 y- !-- --- 4 i: , . i - g . i 3 1 a 25 ..........+. .... c ...L. . ;.~......-.,.. .- a . i i . ! s 20 -- - - - - - '-- - O l ! ;V : i l -
- 1 -
i + 15 -- +- r -t- -- i i ! i~ - i . ! ; i 1 . i 10''''''''''''''''''''''''''''''''''- 40 45 50 55 60 65 70 73 80 Tmix( C) Figure 3.7 Fission Converter Safety Limits for Natural Convection (O V i 3-24 SR#-0-97-14 OCT 031997 __J 3.5.3 Derivation of the Limiting Safety System Setlings O V The limiting safety system settings (LSSS) calculations for the fission converter are based on the c.ame assumptions that were made to obtain the safety limits (Section 3.5.2). The calculations were also performed using the MULCH-FC code. Figure 3.8 shows the results of the calculation of the fission converter LSSS for forced convection. The limiting safety system settings were determined for a primary coolant floiv rate of 45 gpm and a constant fission converter coolant level of 2.1 m with iterations being made for the fission converter power and the steady state average primary outlet temperature to find the conditions corresponding to the onset of nucleate boiling (ONB). The curve shows the combinations of fission converter power (P) and steady state primary outlet temperature (Tout) for which ONB mry occur in the hot channel. Points to the left of the curve represent the operating conditions at which ONB will not occur as long as the operating limits are satisfied. 1 l Because the fission converter's maximum predicted power is 250 kW, the l following limiting safety system settings are chosen for the fission converter with forced o convection: U Variable Limitine Safety System Setting P 300 kW (max) Wp 45 gpm (min) Tout 60 'C (max) H 2,1 m above top of fuel (min) Notice that the LSSS temperature calculated for 300 kW and 45 gpm is 63 C (see Figure 3.8). The calculated LSSS for natural circulation operation coincides with the result of the safety limits (Figure 3.7). The reason for this is as follows. For natural circulation, both the heat flux and Dow rate are low. Hence the axial coolant temperature rise exceeds the film temperature rise (radial direction). This difference is exacerbated when the engineering ,/'\ 3-25 SR#-0-97-14 OCT 031997 ) _ o l i l t i 450 ,,,, ,,,, ,,,, ,,,, ,,,,, ,,,, 400 - - - . -. H = 2.1 m ..I Wp = 45 gpm - 350 - - - - - - - - - - - - = l . iit . _ d . . c., . l- . 300 --... . _ . . J. . -250 - - - t-~~ 3 i, 200 ' ' ' ' ' ' ' ' *I>' 45 50- 55 60 - 65 70 75 Tout ( C) . Figure 3.8- Fission Converter Limiting Safety System Settings for Oneration with Forced Convection s 3-26 S R#-0 14 - OC1' 031997 factors are applied. The net effect of the dominance of the axial coolant temperature rise is r] that the maximum coolant temperature and the n.aximum fuel clad temperature V simultaneously approach the saturation temperature. As a result, both ONU and OSV coincide and the safety limit curve was used to determine LSSS for natural circulation. The results were: Variable Limitine Safety System Settine P 25 kW (max) Tnux 60 C(max) H 2.4 m above top of fuel (min) Notice that the SL temperature calculated for 25 kW is 63 C (see Figure 3.7). It is also worth noting that ONB does always occur before OSV and,if were not for the application of the engineering factors, this difference would be apparent for the above calculation. O V 3-27 S R#-0 14 OCT 031997 l
- Annendix 3.1 - Description of the MULCil FC Code s A multi Cilannel analysis (MULCil ll) code was developed at the MIT Nuclear Reactor Laboratory for the safety analysis of the MITR. The code models the primary and secondary coolant systems with special emphas:s on analyzing detailed thermal hydraulic
. conditions in the fueled region. The hot channel is modeled in parallel with the average channels in order to predict the flow distribution among them during transients. A point- -kinetics subroutine is included in the code. Therefore, coupled neutronic thermal hydraulic effects can be modeled. The MULCH Il code has been benchmarked using steady state and transient experimental data for the MITR il (Ref. 3 11].- The steady-state experimental data were taken from the hourly operation log. The operation conditions cover a wide range of cooling tower outlet temperatures and_ heat exchanger fouling factors. The transient' experimental data were obtained from pump coastdown experiments that were performed during the MITR-It's initial startup.- Calculations of onset of flow instability compared satisfactorily with correlations derived from experimental data. The MULCHAllcode was modified for the fission converter to evaluate possible O-thermal hydraulic design options. The new code is designated as the MULCH-FC code. L Special features of the MULCH FC code include:
- 1. Both primary and secondary coolant systems are modeled, 2, The fueled region is modeled as average and hot channels with the hot channel representing the most limiting condition,
- 3. The axial direction of the fueled region is divided into small nodes so that the power distribution is modeled in more detail.
- 4. The " worst case" results (in which the uncertainties associated with the design parameter deviations are considered) are given as well as the "best estimate" results, O
3-28 S R#-0 14 OCT 031997
- 5. - A two-phase flow model is included that covers the thermal-hydraulic f] conditions from onset of nucleate boiling to bulk boiling. and
(/
- 6. Assorted benchmarked correlations for rectangular coolant channels under low pressure conditions are used in the code.
The MULCH.FC code was used in the current study for the following purposes:
- 1. To determine system design parameters such as flow rates, temperatures, and heat exchanger capacities, etc.
- 2. To establish safety limits and limiting safety system settings for both forced convection and natural circulation.
- 3. To analyze the transient resulting from the loss of one of two fission converter primary pumps.
( O 3-29 SR#-0 97-14 OCf 031997 Arrudix 3 2 - Dose Rate calculatimuvithout Top Shield.1jd Calculations were performed using MICROSillELD to estimate the dose rate cortesponding to a fission converter power of 25 kW without the top shield lid. The i MICROSillELD code is user friendly software for dose rate calculations. A gamma yield and energy library is supplied which covers a wide range of materials that are used for nuclear applications. A photon s.ource can also be supplied by the user. In the dose rate calculations, the fission converter fueled region is modeled as a 6.5 l ' cm x 72 cm x 66 cm rectangular aluminum block. Coolant above the fueled region is 2.4 m, which is the proposed LSSS for coolant level with natural circulation. Prompt gammas are assumed to be the primary source of radiation. The effect of 16N is assumed negligible. An eighteen energy group prompt gamma model is used in this calculation. The energy groups and their as'ociated yields are derived from the fission prompt b.mma spectrum. The calculated result is 560 mR/h a: the coolant surface. This dose rate h not in excess of those occasionally encountered during certain maintenance operations, and it has been demonstrated that administrative actions can provide controls under such conditions that are in accordance with ALARA. O 4 O 3 30 SR# 0-97-14 OCT 031997 \\ Rtinentc1 [31) File Calculation (Fuel Restraint) [32) Y. Sudo et al., Experimental Study of incipient Nucleate lloiting in Narrow Vertical Rectangular Channel Simulating Subchannel of Upgraded JRR 3, J. of Nuclear Science and Technology,23[1], Jan.1986. [33) P. Saha and N. Zuber, Point of Net Vapor Generation and Vapor Void Fraction in Subcooled Boiling, Proc. of Fifth International lleat Transfer Conference, Vol. 4 (1974). ( [3 4) J.E. Kowalski, P.J. Mills, and S.Y. Shim, Onset of Nucleate Boiling and Significant Void on Finned Surfaces, ASME FED:Vol. 99,1990. [35) File Calculation (Power Deposition in Fission Converter Tank - MCNP I Calculation) [36) G.C. Allen, The Reactor Engineering of the MITR il Construction and Startup PhD Thesis, Nuclear Engineering Department, MIT,1976. [37) J.M. Rust, Nuclear Power Plant Engineering, liaralson Publishing Co.,1979. [38) N.E. Todreas and M.S. Kazimi, Nuclear Systems ll lilements of Thermal Hydraulle Design,11emisphere Publishing Corp.,1990. [39] MITR Safety Analysis Report, MITR Staff,1970. [310] T. Dougherty et al., Flow Instability in Vertical Channels, llTD Vol.159, Phase Change lleat Transfer, ASME 1991. [311] L. W Itu and J.A. Bernard, Development and Benchmarking of a Thermal-liydraulics code for the MIT Nuclear Research Reactor, Jomt International Conference on Mathematical Methods and Supercomputing for Nuclear Applications, 6 10 Oct.1997, Saratoga, NY. [3-12] File Memos (ASME Code Requirements for the Fission Converter Tank). (313) File Calculations (Fission Converter Safety Limits and Limiting Safety System Settings) 3 31 SR#-0-9i 14 OCT 031997
- 4. Elssion Convrrier Shutter amLAlnlital Therapy IhmnLihmigli
'~'T I 4.1 General Descriotion of the Shutter Onign 1 The shutters for the fission converter beam consist of a cadmium curtain, a water shutter in the collimator, and a mechanical shutter at the end of the collimator. These shutters fulfill several functions. The cadmium curtain, which is located between the reactor core and the fhsinn convener plate, controls the neutron Dux to the converter and, therefore, the fission converter power. The shutters in the collimator reduce the radiation dose in the medical therapy room by a factor of about one thousand. These shutters are also used to provide the specified irradiation dose to the patient. The McNP (Monte Carlo N Particle) code was used for the shutter design. The controls and indicator lights for the shutters, cadmium curtain, and fission converter medical therapy room door are conformed to those in use for the existing medical room. 4.1.1 Cadmium Cunnin The cadmium curtain, which is located between the reactor core and the fission convener plate, reduces thennal neutrons incident on the fksion converter plate to less than Q 1% [Ref. 41]. It is composed of a 0.0508 cm layer of cadmium sandwiched between two 0.635 cm layers of aluminum. Figures 4.1 and 4.2 show the fission converter power, as well as the epithermal neutron Dux and the fast neutron dose rate at the patient position as a function of cadmium cunain height above the fuel centerline. Two cables, attached at the top corners of the cadmium cunain, connect the cunain to an electric motor. A frame welded on the fission convener tank will limit the movement of the cadmium cunain to the venical direction only, Drive-in and drive out limit switches are used to stop the motor and to prevent the cadmium cunain from being driven beyond its physical limitations. Limit switches are also used to indicate if the cunain is either fully closed or fully open, in the event of a failure of the cadmium curtain to close fully, the reactor operator can lower reactor pow:r or scram the reactor to shut down the fission converter. The maximum speed for opening the cadmium curtain is limited by the reactor period. Specifically,it is required that the change in reactivity associated with the operation of the fission converter not generate a reactor period shorter than 50 seconds (Section A i 2.3.2). L w/ 41 SR#-0 9714 OCT 031997 N 100- ,,,,i,,,,,,,,,,,,.,,,,,,,,,,,,,,,,,,,,, (G , 1.4a.+10 .1 Power ~ O Flux + , - - 1.2e+10 - 80 -- b ; d - w o { - 1.0e+10 $ 60 - - m l $u _ - - 8,0e+9 "g m ~ o $o ~
- 'B C 40 -- - .-. 6.0e+9 l
o 9 - U - . -e-e - ,o . - 4.0e+9 m m - - iE 20 -- - f'J 1 w - - 2.0e+9 0 - @Q ; (*-l -l -l -l -l O.0c+0 -100 -75 -50 -25 0 25 50 75 100 Shutter Height above Fission Converter Fuel Centerline [cm) Figure 4.1 Fission Converter Power and Eoithermal Neutron Flux at the Patient Position as a Function of Cadmium Curtain Height. (Ref,411 O N 42 SR#-0-97 14 OCT 031997 I 1 100 ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,, 12 ^ Power : ^ ; Fast Neutron Dose Rate ;_ 10 e 80 - - u u - y " -8 3 60 - o - - .E l a. . E u s v - t . - s o s -
- 1 C 40 - E o -
Q U ~ -4 e - .2 - 0? . ! (w E 20 -- . (a) . - -2 0 =l 6; ; 'l -l -l -l -l 0 -100 -75 -50 -25 0 25 50 75 100 Shutter Height above Fission Converter Fuel Centerline [cm] Figure 4.2 Fission Converter Power and Fast Neutron Dose at the Patient Position as a Function of Cadmium Cunnin Hr.ight (Ref. 41) .fh V 43 SR#JL97-14 OCT 031997 The time for the cadmium cudain to go from its full open to full-closed position will be measured as part of the pic-operational testing (Chapter 8), and verified at least annually O thereafter. The maximum allowed value is 60 seconds based on assumptions made in the analysis ofloss of fis.; ion converter prima.y flow. Specifically,it was shown that upon a loss of primary flow and in the concurrent absence of a reactor scram, no fuel damage because of overheating would occur " the cadmium curtain took 60 seconds to close (Section 6.4). This 60 seconds requir tnent is only for closure of the cadmium curtain. Opening the curtain is governed by :' $0 second period requirement (Section 2.3.2) and hence may be longer than 60 seconds. This difference in opening and closing times could l be achieved through the use of two motors of different speeds or a single variable speed luotor. 4.1.2 Collimator Shutter Design Shutters in the collimator are needed to ensure the safety of personnel working in the medical therapy room and to control the radiation dose to pa lents. A two pan shutter design will be used: a water shutter and a fast acting mechanical shutter. The latter is
- located at the end of the collimator.
4.1.2.1 Water Shuttet The water shutter located in the collimator provides neutron and ganuna attenuation. Figure 4.3 illustrates the water shutter design. To increase neutron attenuation and decrease hydrogen capture gammas, the water may contain dissolved 10B (~1% by weight). At the start of an irradiation, the water shutter is full and the water tank locatec' above the shutter contains sufficient empty volume to accommodate the contents of the water shutter. The solenoid operated valve in the line between the water tank and the shutter is deenergized and the valve is open. To open the water shutter, the solenoid operated valve is closed and the contents of the shutter are pumped to the water tank. To close the shutter, the solenoid operated valve is opened and water Dows by gravity from the tank. (Nnte: It will be possible to operate the solenoid control valve manually if necessary.) Conductivity, or other suitable probes, will be used to indicate the water shutter position (full closed or full open) at the medical control panel. When the water 44 S R#-0-97 14 OCT 031997 n To MITR-Il From stack U Makeup O x x* tert #x lon column Supply Tank (590L) ) Solenoid J valve g >C Gravity fill line O (2" pipe) Pumped drain 135 line inches (l" pipe)- Fast shutter region Water \ shutter -- (470L) - Beam-center line Figure 4.3 Fission Convener Water Shutter System O 45 SR#-0-97 14 - OCT 031997 , shutter is emptied, the tank will be Olled with air. Calculation has shown that the 41Ar , production is insignificant [Ref. 4 2]. It is desired that the nonnal opening and closing time be less than 120 seconds. Ilowever, there is no safety signincance to this number. The 120 second figure is chosen solely for reasons of efficient operation. I 4.1.2.2 Mechanical Shutter A fast acting mechanical shutter composed of lead and a hydrogenous material such as polyethylene will be located at the end of the collimator closest to the irradiation position to provide shielding from both gamma radiation and fast neutrons. The mechanical shutter will be operated by a pneumatic or hydraulle cylinder. The shutter will close automatically by gravity when pressure is lost and in the event of electrical power failure. Manual control is also provided. Limit switches are used to indicate the mechanical shutter position (full closed or full opened) at the medical control panel, it is desired that the opening and closing time for this shutter be less than four seconds. This time is desired so that patient exposure to the beam will start and stop in a step-like manner. However, a ramp shaped start and stop is also acceptable. The rapid (4 second) shutter cycling time is a matter of convenience, it is not necessary for either personnel or patient safety. 4.2 Shutter Controls Controls for the fission converter shutters will be located outside the medical therapy room at the medical control panel, inside the medical therapy room, and in the reactor control room. The controls at the fission converter medical control panel will consist of open and close buttons with appropriate position indicators for the cadmium , water, and mechanical shutters. In addition, there will be a reactor minor scram button. The control panel itself will be activated by means of a key switch. When the key is removed, these controls cannot be used. it will also be possible to close the water and the mechanical shutters manually from a location adjacent to the fission converter medical . - control panel. The controls inside the fission converter medical therapy room will consist of close buttons for the cadmium, water, and mechanical shutters. It will not be possible to open these shutters from inside the medical therapy room. Lights that indicate the status of each shutter will be located at the fission converter medical control panel. The cadmium curtain can be closed from the reactor control room, The reactor control room will be O 46 S R#-0-97 14 OCT 031997 i supplied with " fully closed" indicator lights or the equivalent such as a scam panel alann for the cadmium, water, and mechanical shutters. 4.3 hiedical Therapv Room Design The new medical therapy room for the fission converter based epithennal neutron beam will be located in the MITR il's hohlraum area (Figure 2.1). Curved pipe ducts to prevent radiation streaming will be provided for the necessary connections and cable runs from the medical therapy room to the control centers. The interior of the fission convener medical room will be accessible to both medical and reactor personnel who are assisting patients. These individuals will be registered as radiation workers. The shutters have been designed so that radiation levels in the fission converter medical therapy room will not exceed 5 mrem /h with all shutters closed with the MIT Research Reactor at 10 MW. This figure is based on a guideline exposure of no more than 100 mrem per week and a 20 hours per week occupancy of the fission converter medical room. The latter is a conservative figure. Au personnel,except the patient, will be clear of the room before any shutters are opened. 4.3.1 Medical Therany Room Door Access to the therapy room will be provided by a motor driven sliding door. During irradiation, this shielded door at the entrance to the therapy room is closed. This door will normally be controlled by push buttons located both inside and outside the door, as well as at the medical control panel. In an emergency, the drive mechanism can be disconnected and the door can then be operated manually. In the event that the bearings which support the door were to be damaged, the door could be opened manually by supporting it venically with the reactor's overhead crane. Lights on the medical control panel and in the reactor control room will indicate the position of the shielded door. An interlock shall prevent opening of any of the three shutters that control beam delivery unless the medical therapy room's shielded door is closed. Funhermore, the shutteis shall also close automatically when the door opens. If required by a particular experiment, this interlock may be bypassed for runs at low fission converter power or when adequate shielding can be provided to allow personnel to work in the area without radiation hazard, in all cases, sufficient warning will be afforded by a radiation warning lamp outside the door. . O 47 SR# 0 9714 OCT 031997 indications of high radiation levels and the position of the door will be displayed at O the medical control panel. The same signals will be available in the reactor control rcom. Communications will exist between both of these centess and the medical therapy room. The shielded door will be locked when the facility is not in use. 1 References (41) W.S. Kiger 111 Neutronic Design of a Fission Converter Based Epithennal Beam for Neutron Capture Therapy, Nuclear Engineer Thesis, MIT,1996. (42) File Calculation (41Ar Calculation for the Water Shutter System). i O 48 SR#-0-97-14 OCT 031997
- 5. Elshluli ConXctler FutLilamtling b V The fission converter will use the same type of fuel, either inadiated or new, as does the MIT Research Reactor. Many of the provisions that pertain to the storage and hndling of the MITR fuel can therefore be applied without modification to the fission converter's fuel. An issue that is potentially unique to the fission converter is that of self-protection because the fission converter operates at a maximum of 250 kW. This and other related issues me discussed in this chapter. In addition, the impact of tritium on fission converter component handling is discussed. Tritium production will be a concern only if D20 is used as the fission converter's coolant.
5.1 Eucl Element Securitv. Storage. and Onlity Assurancs The provisions of the MITR security plan apply to the fuel used for the fission converter. Storage and handling of fission converter fuel shall be in accordance with MITR Technical Specification # 3.10,except as noted in Section 5.3 of this teport. Fuel burnup will be in accordance with MITR Technical Specification # 311(2e). Also, applicable are the requirements of the MITR quality assurance program for fuel. O d 5.2 Fuel Element Self-Protection The fission converter will be fueled with elements that have been irradiated in the MIT Research Reactor except as noted in Section 5.2.1 below. Calculations and measurements have established that discharged MITR fuel (average overall depletion ~10 %) will remain self protecting for at least a decade [Ref. 5-1]. lience, the self-protection criterion will be met for the fission converter if MITR fuel with a significant burnup history is used, llowever, circumstances might a ise where it would be desirable to utilize relatively fresh fuel in the fission converter. Such fuel can be maintained self-protecting if a strategy of the type given below is followed: (a) Fresh fuel elements will be pre-irradiated in the MITR core if they arr planned to be used in the fission converter. Calculations show that a fresh fuel element will remain self protecting for 300 days ifit is irradiated in the MITR core at 5 MW (25 elemNs in core,200 kW average power output per element) for 150 hours. This is equivalent to a burnup of 3 x 104 kWh per ekment. !O V 51 S R #-0 14 OCT 031997 (b) The pre irradiated element will then be placed in the fission converter. An element { with a burnup of 3 x 104 kWh will remain self protecting if the fission convener is operated at 80 kW for at least 60 hours (minimum burnup of 436 kWh for each fuel element) cvery month. The actual sequence used to maintain a fission converter fuel element in a self protecting state would depend on that element's unique power history. The point is that it is possible to achieve self protection. A protocol to assure self protection of fission converter fuel will be prepared and approved prior to each refueling. l 5.2.1 Use of Unirradiated Fud l Fresh unirradiated fuel may be utilized in the fission converter provided that the )- total number of non self protecting elements on site conforms to the MITR security plan. 5.3 Ettel Elemen' Removal l MITR Technical Specification # 3.10(4) specifies that prior to transferring an irradiated element, that element shall not have been operated in the reactor core at a power O' level above 100 kW for at least four days. This requirement can not be translated directly to the fission converter because of the different numbers of elements in the reactor core and in the fission converM. In addition,it might be desirable to operate the fission converter in excess of 100 kW and refuel within less than four days Such operation will be possible provided that the power history is acceptable. The following conditions are equivalent to the limit given in the MITR Technical Specification # 3.10(4): (a) Continuous operation at or below 50 kW for the four days prior to refueling. (b) A maximum operating time of 4.8 hours per day at or below 250 kW during the four days prior to refueling. (c) A maximum bumup of 436 kWh per fuel element during the four days prior to refueling. The first of these criteria is equivalent to the MITR core 100 kW limuation except that eleven fuel cl:ments are involved as opposed to twenty-two or more. The second and third are based on an equivalent power history . 52 SR#-0-97 14 OCT 031997 s I A study was conducted to calculate the fuel plate temlerature during fuel element \ removal. it was assumed that the fission converter was operated continuously at its maximum power of 250 kW until four days prie, ;c removal of the fuel element. During those four days, operation was as specified in the preceding paragraph. It was also assumed that all heat transfer was by radiation alone. The maximum clad temperature was calculated to be 313'C which is well below the Al 6061 softening temperature of 450 'C. 5.4 Fuel Element Handling During nonnal refueling, depleted or partially depleted fuel elcuents may be removed from the fission convener and stored in one of the approved MITR fuel storage areas. Also, panly depleted fuel elements may be moved within the converter from one position to another and new or spent fuel elements may be inserted into the converter. The main problem associated with .; pent fuel handling is the shielding of fission-product decay gamma rays. A transfer cask provides the nenssary shielding during movements. l Any transfer ofirradiated fuel out of the convener tank will be made by lifting the fuel element into the specially adapted fuel transfer cask (Section 5.4.1). The fucJ transfer cask will be positioned on the access hole of the shield block above the convener tank. Prior to positioning the transfer cask over the access hole, ad quate shielding will be provided between the transfer cask and the water in the convener tank. After the fuel element is fully lifted into the transfer cask, the shutter in the cask bottom is closed. The cask is then transferred by crane to the spent fuel storage pool in the basement where it is positioned over the discharge funnel. The bottom shutter is then opened and the fuel element is lowered into the pool and placed in one of the cadmium lined boxes in the storage pool. Refueling operations for the fission convener will be scheduled based on fuel burnup and planned operation. Records of fuel element transfers will be entered in the reactor console log book. Refueling preparation and fuel element transfer procedures equivalent to those specified for the MITR will be utilized. O 53 S R#-0 14 OCT 031997 5.4.1 Fuel Element Transfer Cas1 . O The cask for transferring spent fuel is a steel weldment, filled with lead for shielding and equipped with a bottom shutter. The cavity has a diameter of 16.5 cm (6.5 in.) and a length of 102 cm (40 in.) above the shutter, its capacity is therefore that of a single fuel element. It is nonnally used only in the containment building. The same fuel element transfer cask that is used for the MITR will be used for the fission converter, 1 l l References [51] File Calculation (Fission Converter Fuel Dose Rate Calculation) [52] File Calculation (Fission Converter Refueling Requirement) [53] File Calculation (Maximum Fuel Temperature During Complete Loss of Coolant) 5-4 SR#-0 9714 OCT 031997 _=
- 6. Salill.Alinivsis O
V 6.1 Maximum llypothetial Accident The maximum hypothetical accident (MilA) for the fission converter is that a maximum of fisc coolant channels are blocked by a foteign object, which causes four fuel plates to melt. This is the same scenario used in the MITR il SAR. A review of the fission converter primary coolant system indicates that Dow blockage in a fuel element is very unlikely. The coolant flows from the converter tank, through small diameter tu'ocs in the heat exchanger, and then through two pumps before entering the downcomers and nowing up through the fuel elements. Any foreign material that might statt 'o ,:irculate in the system would have to be small enough to pass through the heat exchanger tubes in order to reach the fuel elements. Therefore, the object would be too small to cause significant flow blockage. One scenario whereby blockage could occur is that a foreign object falls to the bottom of the tank during a refueling. When the pumps are started, the flow could pick up this object and cause it to block the entrance of several fuel element coolant channels. In (v9 order for this to happen, the object would base to fall through the fuel housing matrix when a fuel element was removed. The size of the opening in the fuel housing matrix would restrict the dimensions of the object to those of a fuel element nozzle, if tbc material were small enough to enter the triangular entrance in the element nozzle,it might possibly reduce the flow rate in a maximum of five coolant channels (six plates). Because the two fuel plates on the outer regions of the blocked area will be cooled from one side, the only melting that might occur would involve the inner four fuel plates. Experience with fuel plate melting both at the Material Testing Reactor (MTR) and at the Oak Ridge Research Reactor have shown that fuel plate melting because of now blockage does not propagate beyond the affected now channcis. Although the nearby plates were discolored, cooling by the unaffected channels was sufficient to prevent propagation of the melting [Ref,6-1,6 2), A study has been conducted to calculate the maximum radiation dose to an e individual located at the exclusion area boundary of the reactor during the first two hours of MITR-ll's MilA [Ref. 6-3), The MilA for MITR il is postulated to be a coolant flow 61 S R#-0-97 14 OCT 031997 m blockage in the hottest channels of the center fuel element. This will lead to an oserheating (~ of a maximum of four fuel plates. The study conservatively assumed that the entire active portions of all four plates melt completely and release their inventory of fission products to the primary coolant. Ihcape of fission products to the containment because of fuel melting is restricted by the pool of water above the core. The following approaches were used to evaluate the I major release paths to the exclusion mea during the design basis accident: ! 1. An analysis of the reactor's physical systems was made to determine the fission product release from the containment shell. The dose from leakage was calculated using a standard Gaussian diffusion model and local meteorological data.
- 2. Gamma radiation reaching the beundary area by direct penetration of the containment shell was detennined using standard shielding calculations. A Compton scattering modei was developed and applied to photon scattering from air (skyshine) and from the steel containment roof.
3, An analysis for radiation streaming was performed for the truck airlock which is the O largest containment penetration. For the hilTR il accident analysis, the fission products in the fuel at the time of the accident are assumed to be in equilibrium for a steady state reactor power of 5 htW. This assumption is conservative for the hilTR because it does not operate continuously. For the fission converter, such an assumption would be even more conservative because patient set up time requires an hour or more under the best of circumstances while patient treatment time will be ten minutes or less. 11ence, even if the Ossion convener facility were in continuous use, the fission converterfucI region would be at power for at most a few hours per day, Given this short operating cycle, most of the important fission products, including 1131, will not reach saturation. Another important difference to note between the fission converter and the hilTR il is that the fission converter operates at a maximum power of 125 kW. At this power level, each fuel element generates 11.4 kW, compared to 208 kW per fuel element for the h11TR il core. (Nele: For h11TR at 10 htW, the figures are 250 kW,22.7 kW, and 416 kW respectively.) O 62 SR# 0 97-14 OCT 031997 l l The calculated results for the hilTR il hillA are given in Figure 6.1 and a summary of the estimated doses are listed in Table 6.1. Even with the conservative assumption about O Ossion product equilibilum, the estimated maximum external dose to an individual located at the nearest point of public occupancy during the first two hours of the h11TR ll hillA is 595 mrad to the whole body and the intermal dose is 118 mrad to the thyroid, llecause the l fission converter operates at about 5.5% of the h11TR ll's power density ((125 kW / $000 kW) (24 elements /1i elements)) and has a much lower capacity factor, the dose from its Mil A would be expected to be considerably less. ' 6.2 Insertion of Execss Reactivity insertion of excess reactivity is not considered a credible accident because the fission converter is a highly suberitical system. Analysis has shown that the key for the fission converter is 0.62 or lower, depending on the type of coolant used and the amount of U 235 in the fuel. 6.3 Loss of Primary Coolant The re are two initiating events that could result in a loss of pdmary coolant accident O for the fission convener - primary pipe breakage and fission coavener tank failure. The fission converter's safety system provides cenain protections against this type of accident. In the case oflow coolant level, the cadmium curtain will automatically close, in the case of low primary coolant flow, a cadmium curtain closure is automatically initiated, if the cadmium curtain is closed when the coolant level remains above the fuel elements, calculation has shown that the residual heat can be removed initially by coolant evaporation (Ref. 6 4]. A calculation was made to analyze the effect on fuel temperature if the primary coolant were completely lost from the fission convener tank. No credible scenario could be found which would result in rapid and complete coolant lost. Therefore the following should be considered a bounding analysis. The following conservative assumptions were made in this analysis:
- 1. The fission converter has been operated continuously for 5 years at 250 kW before shutdown. It is assumed that 100% of the decay heat is deposited in the fuel p region, d
63 SR#-0-97-14 OCT 031997 j! IIll e. o D . e - y : - a 0 d o 5 o D b r. e n e l i s o r o i e e D M t s t o a l a D m ] a c n m 3-S a t o o C 6 t i T c t e a 0 f. e a g e t r a 4 n R S t k o [ e a i e I n e sc I un e L / P ll a R l ct xs T I e T1 M l e t ,- 1 n i S ia me ) A ir M H , 0xA ( 3 a M t Mr _ n o _ e f e m s s n t o i l u D a s t e g e n R n s o e i o t C s r D S o e m D t d fy o t e l , 0n r y a s e 2a f d c o i b l e o S D h A c b-r S n e i a a l o t w t A e o s h n d a i D W e r g h u n o a o k _ ie H a sc o e 0 l a un w L SA g l 1 Ei ct xs m r ia me ir D T l T 1 t a o i nA 6 M e r u g i F - ~ 6 5 4 3 2 1 0 O o c 0 0 d 5 $ ll n c1 ,,k eSi# O w ~eS L Table 6.1 Estimated Doses from all Modes of Radiation Release Duringa hilTR Il Maximum flynothetical Accident (Ref. 6 3] Component of the Dose Dose (mrad)(c) S m (a) l 23 m (l) Whole body: Containment leakage 27 27 Steel Dome Penetration 3 27 Shadow Shield Penetration 43 21 Air Scattering i14 147 Steel Scattering 192 373 Total 379 595 Thyroid: Containment Leakage 118 118 (a) Boundary of restricted area (b) Nearest point of public occupancy (c) Calculation assumes that duration of release is two hours.
- 2. The primary coolant in the converter tank is completely lost at the time of fission O
V converter shutdown.
- 3. Convection and conduction to air and surrounding materials are neglected. The only heat transfer paths are conduction within a fuel plate and radiation transfer from the fuel plates to the reactor biological shield (graphite / concrete) region. The graphite / concrete temperature is assumed to be 150*C.
This calculation was made for the hottest fuel plate (hot channel factor is 1.53). The initial temperature of the fuel plate was assumed to be 106'C. The maximum temperature in the fuel plate occurs at the centerline (x=0) because of radial conduction. Figure 6.2 shows the heat transfer path of a fuel plate, qx" is the heat flux because of radiation heat transfer to the graphite area and qy~ is assumed negligible because the heat transfer resistance to other fuel plates is much higher than that in the x-direction. Figure 6.3 shows the maximum temperature in the hottest fuel plate in the fission converter. The temperature rises rapidly during the first hour to 383*C, which is 67 C lower than the fuel softening temperature (450*C). It then decreases because of decreasmg decay power (Ref. 6-4), 6-5 SR#-0-97 14 OCT 031997 J O - 4Y "O i I End Plate sk l C adding i i y sk Y \ l Fuel Meat 9Y "O AX Figure 6.2 lleat Transfer Paths from a Fuel Plate During a Comniete Loss sf Coolant l Accident 400 350 O Q e 300 3 250 2 - 200 150 { m 3 - 100 Fission Converter Power 250 kW Ilot Channel Factor 1.53 "** I 50 0 0 2 4 6 8 10 Time (hr) Figure 6.3 Maximum Fue! Temnerature During a Complete loss of Coolant Accident O 6-6 S R#-0 14 OCT 031997 1 6.4 Loss of Primary Coolant Flow The fission conveiter is designed so that low primary flow ' ss than 50 gpm) will automatically initiate both a reactor shutdown and a cadmium c" .losure. Either one of these actions will shut the fissloa converter down.110 wever, ths idmium cunain will take 60 seconds (or less) to close while the automatic reactor shutdow will take less than one second. If the reactor shuts down automatically upon loss of fission converter primary coolant, the fission converter will shut down rapidly because of the absence of the incoming neutron flux from the reactor. Temperature elevations in the fuel and the coolant l will be small and no boiling will occur in the fuel region (Ref. 6 5). An analysis was made of the fuel and coolant temperatures if only the cadmium curtain responds to the low fission converter primary flow signal (Ref. 6 6), initial conditions and assumptions made in this analysis were as follows:
- 1. The initial fission converter power is 250 kW.
- 2. Steady state average prirmuy coolant outlet temperature is 55'C.
- 3. Primary coolant flow rate undergoes a step change to zero.
- 4. Cafmium curtain starts to close after a one second instrument delay time.
Fission converter power then decreases linearly over 60 seconds.
- 5. Only conduction heat transfer to the coolant from the fuel plates is taken into account in this analysis. Convection heat transfer and heat transfer to fuel housing materiais are neglected.
The calculation was made for the hot channel where both the coolant and fuel temperatutes are the highest in the core region. To simplify the calculation, the coolant and the fuel plate were lumped as single nodes (no axial dependence). A homogenous equilibrium mixture model was used in the coolant region to model the coolant phase change. Figure 6.4 shows the hot channel fuel plate temperature es a function of time. It is assumed that there is a prompt jump of fuel temperature from about 75 'C to 95 *C 6-7 SR#-0 9714 OCT 031997 (p) v 140 . . . ..., , , , . . . - , , .g . . , ,. , . .. ...... . . . . . . ,.i.4 . , . . . . . . . . - . . . . . , - , , . - , . . . . . . . , . . + ..ro- - - - ....y .- . . . .- ...-4,. . . . E.j . L.. . . . . .. , . a # .+..,...,.,. , , .+... 4..+.. . . . ' ~ ~ ' ~ $ 120 - ~ ......,.. j.. .... .. . . . . , . ... .. . . p. o # . _ , .- . . + . . , .. c . , .- . . .. ;. . . i _ . .. . g .....; . ; 4 4.. , . i...; .. ; . . . 4 . . .. . . 4. . . . . . . ;. . . . . .. .-p... . . 4 .. ..; .. i . i .. t 4... i _ i .. i.. _i. 4 44.. . .. ; . 4 . . .., ....;.. . . ;.. .. : t..J .. . i . . . .. . ; .. . .........i.J.. ..i 1..:. ' ..+. ~ ' .. t'..'i .. . . . . .. .; _ . c. 110 .. ; : .. t , ...{ .p . ..p..' ..,{. .{.1 . i . . . ! .4. . ...p ., 7. q. . 4 . f, . . 7. . . .;... g .. ...9.-, , ,~ .. ..y.,........ .
- 4. . . . .......q _.
.,.q.7 ..p...;..q.. . . 7. 7. . . . .,).4.. . . + ..g . 7 . . t. . . . j. . . . p. .,..3.+.- ..+.-,-...e..4,.. .#.+.+.+ . . , ..#. .+ .j .. ..+.....#..+. . . 4 . ..,..-..i. +,-4... .....{... . i . ..& .. . ( . 1. . .. , ...,...i... 4.. L ..i 1.. .. ..[ . + ..; .4 . ..; -.; . g. ; . .. ..i.144 ~. ...;. 4..;!....i....l. i.. 4.. .J ...;.. i...&. .. ...i...,...+.L. . 4 4. ...i 1. . , ..j ..J ..; .1 . . .. i . 4.-. e . . . . 90 l ..jt ...j. ..i'.4...i' '.. 'j ...I ...*... ... l .4. ...i. 4..l. -i.. ...i t 4..; 4. .; ... .. ... ; . . L- . .' i l C 'O 20 30 40 50 60 l i Time (s) Figure 6.4 Calculated Hot Channel Fuel Temocrature During a Lonpf Primary Flow Transient (Cd cunain closure. no scram) because of the ~sudden decrease in forced convective heat transfer. The maximum fuel plate temperature during this transient is 132 'C, which is significantly lower than the fuel softening temperature of 450*C. Coolant temperature, coolant mass, coolant enthalpy, and static quality of the hot channel are shown in Figures 6.5,6.6,6,7, and 6.8, respectively. The coolant temperature holds at 106*C after 8 seconds into the transient because vapor starts to form. The coolant mass in the coolant channel then stans to decrease because the channel volume is constant (assume no flooding). 6,5 Loss of One of Two Pdmary Pumos The fission converter primary coolant system is designed to use two primary pumps operating in parallel during normal operation, it is intended that either pump be sufficient to deliver a primary flow rate higher than the scram set point (50 gpm) and that the patient treatment continue if one of the pumps fails during an irradiation. Accordingly, the following analysis was performed to show that this transient would not result in an O o 6-8 SR#-0-97 14 OCT 031997 l excessive coolant temperatuie. The main issue involved in this analysis was the time delay associated with the mixing of the coolant in bulk in the convener tank. The primary outlet O4 temperature sensor will measure a lower temperature than the average coolant outlet temperature during this transient because of this mixing effect. i 1 The accident analysis is made under the following assumptions.
- 1. The fission converter power is 250 kW.
- 2. The fission converter primary flow rate undergoes a step change from 100 gpm to 50 gpm (scram setpoint). (Noic: The design flow rate for one pump operation is 60 gpm or higher.)
- 3. The initial steady. state primary coolant outlet temperature is 50*C.
- 4. Instantaneous mixing is assumed in the fission converter tank region. This was shown to be a conservative assumption [Ref. 6 7].
O V 69 SR#-0-97 14 OCT 031997 .,, i , I i y 1 , , ~ , i 1 , s, i. . - - 1 1, . ,. . i . T. T I e . T. I T. .' 1 'w', . , . , , ,. . [ . i , , , . , . , , . . ,. s . . .- . . .- , i 4 . . ... . . . w -e e.. . . . . . . . . + = 110 ...-.. - . . . , . 4. . - . . , .. - 4 e e t , ', .t i +
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. - . . . . . . . ., i - . .-.t-+-. . . O - ** r . .. , . .+ m . . . , . . . . , . g 4. . . ... 4, . - , -, .. 4 , 4 .,;.. , . . . 4 e 4 . ..4 ...e. --,4" t- 'e-*-8 . t - 4
- 4- 4- t 6 "t"*+.4*
.e,.. ,,-. . e. . . , .g..v. ...'+..9.4 . . . ,.4~,..... +. 90 - },.g. ..4.4.., ..,.. ,.. ,m 9 ..,j.. j...,. . . . j. ..;...... ; q...,p .i-, .;..... . .e -q . .p...-.-. ............... ..j. a..,, q~...=. . . . , . . . . . .y... . . .e. . ..e.. . . ..9 q ..p., .s...a..... 4. , , 4 , p. ... 4 4,4 4 . . . . . . . 4 . H 80 ...,..e ...+..4.~,4... ..+.4..e.. .e.+.,, ., ...e....,.go.+.. ..+..,p.6..4... 4..44 s...,.... . . . a . .e . . -.i...+...i...,,q... o=* , ... f ..y .. 4 .. ..i........+.. ._p.4o ,..4 [.. mem 4 .. 4 4.. 4.. ..+...j.... . ......4.. 4 4 . . i 4. . < . ..,.9..4.,,,., .4 ..<4, ..4...,.4.. ..J..4..9..... . 8 70 e,....., + , ......e. . . . . ... -- y , e .. y... r. .. ; 4... ; . . e..p ..q .. .ee...... ,. .{p. . +. ...e,4.. 4 ,..4..-e. ..e..,,, , ..q.., ,4 . ts ......e.q....p. ... q..e. .3,.e.,._.. q . y .. . . .. .4... p . 9. . . , . . e. . 4 .. . 4,.-+.. 4.4.9 -.4 ...,e.. , , . , a . 4 .. . . . . ... q . 4 m . 4 . . . . . . , . ..e...... ~.,..j-,.....,.4.. . ..d.....4.. . . _ - . .e . 4 . ..f - j ..g , ,y . ..j .- f . .j.-,. .4 .. . .....a... ..6.. .. m ._ p . 4..j _ q... 4...e L.. .q.., ; . . . . . i. 4 . . a .1. ; .a .. 4..6.-.; , . . .. L . 7., p.q q .. .j j ; . g= . 3 .. ; . 7. 7.. . 7 7 g ..q.. .. p. p ..j .j . p .g.4. 7 0 10 20 30 40 50 60 Time (s) l Figure 6.5 Calculated Hot Channel Coolant Temperature Durine a Loss of Primary Flow Transient (Cd Curtain Closure. No Scram) ep\ s- , 0.05 , , , , , , , , , , ,, , , , , , , , , , , , i ..,.44....... e , , . ........4...e. .......;... . . . . ...;........e. - - :,,..p. . . . . l . : : > - * ..j...q, .4...q. . . p. 4 9. ,4~ . . . . p .. y ..t..6 (. . .4i- i .9+. d, ...:[ ..j : ...: ;. ..+....+.. ..d. .6.., , .I . . . [. . . + . . ..h , , .. d . 4. [ .. ., , . .q....4..-... .-...,......... . .. . . . . .-......q..m,.. .+ . ' e . . ,. . . . . . . ... 0.Q4 %p 4- e. 4 . .(...,. .. ... ... ...J.,.(... 4 9a , 6 .+,.e ,'4.. .. . .. . . , . . . , . . . , . , ,. . .;..... ..4.+ , . , . - . ... ..e....... A l . . . ' ; i j ! ...g....,,,.4.,. ..t--4-t- * -'t.'"* +-4 *- ti == f + ( ' ,- , + ** -*,*'t*"i" 4- * ,;g - 4..... ;. .. , ..6.,...,...e.. . . . . . ..._....;...,... .J...... y . ...,.... .6.... 0.0 m v ......+....... ..e.,.. . .4......4....... ...e.. . 4 , , 4.. . . + . . . ....e.4... .d : . . .: . . > < , i 1.. ..ei ..e.. . . . ...a... .. 4. .. e. ....!....e..... ....,..4..i..,.. ..4.....p...... ... ,,e, . . : : .. e.. e . .e . . .......,.......4 ...e. . . . , .... .......e........ . 4. .. . e . . e . , ,
- s .
. ....+..q....... .....-+.. i ..e. . ' . .'. . . . . . ... ..4.... . . . . . . ; ..4.......+.. . . . ' ' i e== 0.0a. , , . . , } _....m....e. 8 _ e. . e. . , . ...e... . . ... e. . . , . . . ,.. . . . . . . .. . . + . . + m. gi ..g.... i .i. h - k - 1 .f. j..i,(. ...$. . .-+p . m- ...,.,( $ $ s g. _ . . . ,g .e..e..... - ...+... , .. . . . ~ . ..~.,.-._ . . . . . . , ..e....... e . , . . . . , . .. . ...e. ,...7,., . . . . . . 4..-4... 0.01 .,..p......,... . . , . . , s._ . _ . . ._, ... ,..+4.. , . . , , . . . . , . 4 .3 .. +.r . , , . . , . . , . . , . . . , , . . . . . .+ .+ .- 0 i .'. '. .' ' ' ' ' ' ' 1 ' ' ' ' + ,. . .e. 0 10 20 30 40 50 60 Time (s) Figure 6.6 Calculed liot Channel Coolant Mass in liot Channel Durine a Loss of Primary Flow Transient (Cd Curtain Closure. b o Scram) /ms \v/ 6 10 SR#-0 14 OCT 031997 40 iie i i. v v i i v i i i i A 9 . .pi 4qe .4.+. . . .j - f ..v e u. 94 .q . 4..:( u 4 ..i.24 ,i qm .ig e ,i. 4- ,....,g..(.ug- ,. .. ,j.j .. p . } .. . .. i ., j i . 4 4,. j .,. 4 . { . ,4m , . , , . . .4m , ,. .-.-,.3. . . ._- . .j .. j . 4.q.. . ., .. q... , 4 ..{. 7 e . .. p ;. p . _q . . .p. 9 .,..p..o.. .. p. .p. q.. p. ,99.eq- e.9.q . .. g i . p. p . 35 . , q..je s .. . g.a.} q . 7.. s.p ., .l f.. ..f..9m . . . .. p . g .. j _ 9 4.e -t* . . . p. % .p- .. p. . p . ,4. , .9 p., , p ,q ... p .4 , . .3 q , -, . p s .q.4 a.. m . m e. . a .+ ; -p.4... m. . 4.q . 4. p.o .e ,.+.. ..a p..n . . .4..j L. . ] - 30 m .6 % .4.q. .. p . . , . . . p.4
- 4. 3 +
mp+ 4. 4.. . := . g,, e r- , pv o,e p+p . %4, 9 . . a ,..,-,q.,. 9 p.j_94 ., p 4 9.o p. ~..m q. .. . m .p. i ...A~ 44......{4d- ,4.-4. 4.. . - .,4. 4 4.. p 4.a. 4. ~ ..e.4 4 @4.. l . p..pd .p. d. ., ' ., ,, .f 4.. . ..p.. e . 4. .. 94a. .. . c 5 . . p ..j . . 1... { .. p .4 . . 7...{ ...h ..
- p. 4 u}... ...{. 4~ . 4..p ...p . q{ .. y ..
y 4.q.. F 4 4 {. 4 . ...j ..+.4 ..j .p.4, ,p ... .. j .. p. ,q. .. .j 4 4. .q.. 4 m ' .. f ,4. 4.. 4 4 e4.. 4.m q... g , q .m Q m.4.. .. p.p.4 .h.. .m , 9 .. g ., .p.. . 9. .p q. . ..}o. 4 4.ye ..q q ,.y.. }.g .. . .. ....p..p.{.. .....e ...g y . 9,..l.q .... .. .. ..,9 . .. p .. p., ,. .q ., p . .. m.p.4.q. .% p .m. 4.. q.g.p..p., 9. 9..
- 9. .
20 {.p.. .. q . . , ' . q. . ., Q ..g.g .{p . . q.l...h.p. m 4 ... ... , ,.. 4 , .p .. .{ ...j a {. e{m . . , . . . q. o.p +. + . .q ..-- , ... p p.. ...e.. p. ...{....p j .. y .. .9. . g{,q.. . 4.y. 4 . .. . 4..p . p. p .p .. ~p.5.... m.p.. 4. 4 .e., .e.9.. j .. .p. m4.. {...hq7 .. .. p . 4... ,. 4.. . 4 .. ...p4 a. j....i ., . ....{ .q-p m,4 p 4.. 15 .; _...p . y _p554 . .. _' - _p .p.p.p4_ _p +.+ , . .p ..... .+.4_ _ J ... .4 .. q.4.p a- a4.. i q.p.g .p4. . o. 3,..>...4 ... 4 4,.4...L. ...q ~..j .1 - ,p. 4... 4...b. pp 4... ...p... . 4.o?.- e.4..p.+ . ..4., ,....4.... 4... . 4 b.l 4... 10 4_p... . ...p.. .M4 ' '
- o. ' '
' ' ' ' ...p.j...p,4 ' ' ..b4 ' ' 4 0 10 20 30 40_ _ 50 _60 1 Time (s) Figure 6.7 Calculated Hot Channel Coolant Enthainy During n Loss of Primary Flow - Transient (Cd Cunnin Closure. No Scram) ! , . , i i , , , i, , , , i i , , , ,, . 4...e , 7. .+. . ...t. . 4. 4. ;... . . a ....- 5._ 3....; . ....'..t.1.-, .. '... j , ! i ...,5. ...i q.., .. . t .t. 4.. . .p.} . ! ... ...i.,. t..., f I i 4.j l . 9. 4... .. .. . f ., ...p.. ; . .q.. . i .. ! . .
- 4. . ,9 .. . .. {. . p. 9 .,. ..q . .. .. . .g 4... ..
.p.}..+ . .. o.9. 4..p 4.... p..p .. .p . 0.8 ~- ' ' ' I 6 l e..h' l .. 4.._ ! i . I
- m. . f....h. . .4 4. . ..,p.4" . , . . . p. ..- f .q[ .. 9..d'. ., . . .q . . 4. .
. j i a j , i { l : 1
- 12 ' I y I
, . .+ .g . ..e . 9. .. ..........4.. - . .g...p .p n.j. . ..g. 9 n,p4 .. h., . . .jl . . . . 4... h.j... h i. 4 .. ...h . ...h .h... .,d ., .. e ., ......d..,..p., Mr. :.. ..i x 0.6 : M = h. 4..q 1 f . -.h.e.h. .,4 i I ; 4.h.. ...{ o . { .. h ...h. ; 2 ..4 t i . ... ..h,,..... .{, .h.. t i 6 .h. j - 9 . e. I m4. ..pi 4... . 9...p.,e..p. . .q ...p.o. ... p. . pop..p.9.. ....p. ..+e . .pp, 9.. .p...p. .9...y.+ d. .. i
- e i : : i . i i : i 9..
a 0'4 ...' ' . I p' I-p.. I +~ 4. ...I...I. 4...p . 4' 4. . .. 9.. o ' ' ' ' ' ' ' ' ' ' j 14.q... ' ...}...j...p.a .. 44 p.p.p. 4..p..p.g .. gw.p . [' % ...f.,...... ..,p.. o. . . . , + ...p,..,,,.9... . o+ . + . . ... .p. .g. .. I f l . ...t,..,Iu.g. 4.. i . i .. 4 9. 4' .. 9 % . 9...,f .. po ,: I e . 4.o +t 4m 4m ..'.. .. .W,i ...p., b . i s t 0.2 4 . ( ;! ; i : . . +..e e + . .-r .. e u e .. . - 4 i ....p..}.....i.,+i. . ,.e) .+E . 6' , i i .
- ! ! ! ! i = ! i ! i i I i _ .!...9.i.q. . .
a !.... .! ...eoe . ' .. ....t., . . i ..i .e , ..!. ... i ..e..e. . ...e.... i I i . 4. p..e. . . + . . ..i i .. ' . , . . . ... ...p..p.9 .[ . .4 ..i...!..' i 1 ) i .q.. 9i ... .;. . .. i . 9 .o..e. . ...p , 0 , . , $ ; . . : 4 i , # : . - i i . ..,.% 9... p. ...j. . .. j. ...p. .y ~.m p. p..p.. ...p...p. 4 ...p 1. p.y.. . ... ,..,t,. p . p. ,.... e i i a i i i e e i a e i i e a e i a i e a a e 0 10 20 30 40 50 60 Time (s) Figure 6.8 Calculated Hot Channel Coolant Static Ouality During a Loss of Primary Flow Transient (Cd Curtain Closure. No Scram) %4 6 1l' SR# 0 9714-- OCT 031997 Figme 6.9 shows the calculated average primary coolant outlet temperature and the mixing area (upper fission converter tank) temperature. Notice that the maximum difference between u" average primary coolant outlet temperature and the mixing a:ea temperature is about 9.5'C. So the average primary coolant outlet temperature will not exceed the LSSS temperature limit (65'C) under all conditions during this transient. Also, the calculation predicts a decrease in the average primary coolant outlet temperature after about 30 seconds. This is the result of a decreasing primary coolant inlet temperature as shown in Figure 6.10. The fission converter primary coolant system approaches a new steady state operating condition about 300 seconds after initiation of the transient. 6.6 Loss of Off Site Electric Power The MITR il scrams automatically upon loss of electric power. This in turn would cause a shutdown of the fission convener. llence, temperature elevations in the fuel and coolant would be small and no boiling would occur in the fuel region [Ref. 6 5). The wt.ter and mechanical shutters of the fission converter are designed to close automatically in the event of a power failure. The medical room door can be operated manually for patient removal. 'O V Emergency power is supplied by the MITR ll emergency power system for selected equipment and instmments such as :
- 1. Fission convener medical therapy room radiation monitor.
- 2. Intercom between the fission convener medical therapy room and its associated medical control panel area.
- 3. Intercom between the fission convener medical control panel area and the reactor control room.
- 4. Emergency lighting of the fission converter medical therapy room and its associated medical control panel area.
- 5. Safety channels listed as follows:
. Fission converter power
- Primary coolant outlet temperature
- Primary flow rate
. Coolant level + Manual reactor minor scram from the fission converter medical control panel 6 12 SR# 0-9714 OCT 031997 -65 ,,, , , , , ,, ,,,, ,,,,; ,,,, .1,.1.. ; ..L.; ..L ..+ ;. ..,p.4 ,. . 4...a.. .e ..+._ .. .; _;. ..: a. .. .a .a . .i.. . . . i ,...,-.+..4.e . 6. .. .j , . . -4 .. i .. 4! ., .! . . +1 . +i ...f. . .,i . . i c.l .- ..+..a...t+i.. ~=.....-e.. M 1..;.. Average Prima y Coolant Outlet Temperature .-. .1...} ..1 n-M.4 .. 444.. . .,i'i + . . +i . . . . . . . .- .44.
- 4. o, . .
60 A~}. 1 1 . i i 4 i i ; i> g- - . .L.......j.-q. .p..7.. y' yl .. j... . ...j....q... p . ., .4.. .....p..... q.. ..} . . . , ,...g,..... ..... ..+..- .. . . d.. ir; ..h...b. 4 ... 1. .. ...l...h .1 .d ! ' i ! Ii --- X..1 h,. i ! i ii 1 g 55 ; -;,, ...l.;., 3 1 .4.,; -.u, .. ,. . . ......+.. ,...+.. _ ...l...a . . 4..p p a -. 4-.. i -.. 1.a .... ..p. .... yK .. .. ..+7.{ ...!....q. .[ . h '{ .
- l. .
4 1 1 j t i 5 .l .- ....~ 2 i .}. . . 9y .* 3t Miking Area Coolant TemperaturelTF 50 ~~ ..L. 4.L. .a..I ! t j 1 i.1.. 1 _4.. ..l l 1.. i i i .L ..i.i I : i i i.. 'l 4 l . . 4. .+4 . . . .. ..p..'.. i -4..+...p..... 4' ...iv . ..., ......., . . . ~+.i..,:..,... - 5 ) i 1 ..+.. . . . . .~ , . .. . ... .. . . ._1- J v. d. ..1..1. ... 45 i A . .l'_ ' I I I I O 50 100 150 200 250 300 Time (s) Figure 6.9 Calculated Average Fuel Outlet Temnerature and Mixing Area Temocrature During the Loss of One Pumn Transient 44 ] I ! I I I I I I l ' I I' 1 Y ' ! i , 4 ! I f f 1 I I ; !It I 4.. ...b..3... - j ..4..~ ..p.,.g .. ..+..4....{. . q. .p.4. 1. . . . .ip q. t g s ! - .r j f f 1 i i... . 4. . f -.4..d. l ,. -.l . i .. .. ' .. 4. . 4. .i. . 4 . .......i... ] r ) : ....-4..+.1 1 . . , . . .. . #. 1. . i.
- i. i
......e..9..., ,1 { i . ...,.... l..,,.-9... ...q. . i , ....,...q.+. .{ . , 3 i! ,,,1..9......y... - ,i . .m..,.......q... 4 $. !I i i !i r
- l. ..i 9. ll1 = . ! i i
. !. l -. i i ,i - 4L 1 ei i . I . 2 4.. i , Ii:
- ....o li il
!.! lt 4 s..s.. . .,,,,,,, i I. I 1 1 i .;; - e - ! q 40 y\p' I'I' ! . Primary Coolant !! k' ' ' ' Inlet ' 'Temperature ' ' ' ' 4 +' . j >!i. .! 7. l +1_ ...!;. . . . L : ., _... ..4.. A. ., , ..#. 4 -;. : l %r N i i....e....... ..q . .,.!.. . ... . _t.. ..i ..i, _ . f kk l H i g h h .+ . .l.._ g...;... ...j. . p. ...g q. .1. 3... j... ...,... ,.j .q... i i i > > 3g , ; i ! i < ii i . . . ,t... ..,t ..i . . i .. ..; ..i ._i _N.J ! ! < i 4,. - i , , > * : .-..i. 1 1
- . - 1 : i4 i 3 i i. . .
1 4 > ; . .+.. .y .4.. - . ..p..9.. .p. . 9 . . q . 9. .. a.. . 2p .. , . ..g .+: . i ....f., . ,+. .+ m.{ 4 ~ i j ! . . .!,.., ..t ... ' ! .,! .. . ... .' . - , , ! I i I ..a....... . . t. ...4,..i... : .. 4.,q .. . . . + ' '< 'f t ' ' i I ! 360 .' ' 50 100 150 200 250 300 Time (s) Figure 6.10 Calculated Core inlet Temt vure During the Loss of One Pumn Transient 6 13 SR#-0-97 14 OCT 031997 =_ r 6.7 Loss oflisaLSink t The MITR il uses two secondary pumps. The reactor will shut down automatically upon low secondary now (450 gpm for either pump). If the cooling tower heat transfer capacity is lost and the fission converter and the reactor continue to operate, elevated reactor primary coolant outlet temperatures will cause the reactor to scram automatically. liigh temperature at the fission converter primary coolant outlet will cause the cadmium curtain to close. Because the temperatures will increase slowly due to the nature of this transient, the reactor operator also can manually shut down the reactor if deemed necessary. 6.8 Mishandjing or Malfunction of Fuel The fission converter uses the same fuel as the MITR The fuel handling tools and procedural considerations that are in use for the MITR will also be used for the fission converter. S 6.9 Exneriment Malfunction The fission converter is not designed to accommodate experiments within the fission converter tank. Thus. experiment malfunction is not considered a credible accident. 6.10 Natural Disturbances Safety analyses for natural disturbances for the MITR Il apply for the fission converter M t of the following is from the MITR II SAR. 6.10.1 Earthw ss A s-ismic study of the Cambridge area is described in the Section 2.3 of the MITR-11 SAR The Cambridge area lies in the Boston Basin which has been relatively free of earthquakes in recorded times. In view of the past seismology records and the conservative design of the fission converter, it is unlikely that earthquake damage poses any hazard. Funbermore, reactor shutdown is expected to occur in the event of a significant earthquake o and therefore, would also shut down the fission converter. U 6 14 SR#-0-97 14 OCT 031997 6.10.2 Lightning The reactor containment shell is grounded to a heavy copper conductor buried below the natural water table. Lightning arrestors are attached to the ventilation exhaust stack and are groundeu to the buried copper conductor. Consequently, lightning is not expected to affect the facility directly. Ilowever, an electrical power outage may occur. ! During an electrical power outage, the reactor is automatically scrammed and this would shut down the fission converter. l l 6.10.3 Severe St +m The reactor building conservatively confomis to wind load criteria of the Massachusetts building codes. The reactor building is also protected from excessive pressure variations by the vacuum breakers and the pressure relief system described in the MITR-Il S AR. Loss of off site electrical power is certainly possible during a severe storm. In that event, the reactor automatically shuts down and hence so does the fission converter, The ,G reactor core tank and the fission converter tank remain filled with coolant which would V> provide the required shutdown cooling. The co,ntrol room is equipped with instruments that indicate wind speed and direction as well as barometric pressure, If the storm appears to be of a nature that might cause a power failure, the reactor is shut down until the storm has passed. ( 6-15 SR#-0-97-14 OCT 031997 t 4 a __ m ' Reference 5 ~ -[61) Dykes, J.W., et al.,"A Summary of the 1962 Fuel Element Fission Break in the
- MTE",1D017064, February 1965.
[6 2] Tabor, W.H.," Fuel Plate Melting at the Oak Ridge Research Reactor", Ab'S Transactions 8 Supplement,36 (July 1965). (6-3) ' R. Mull Erclusiori Area Radiation Release During the bilT Reactor Design Basis . -- Accident, MS Thesis, Mll,1983. _[6-4) File Calculatien (Maximum Fuel Temperature During Complete Loss of Coolant) [6 5] File Calculation (Fuel and Coolant Temperature Calculdtion Before Reactor Automatically Shut Down) .[66] File Calculation (Fuel and Coolant Temperature Calculation During Loss of Fission L Converter Primary Flow with Cadmium Curtain Closed in 60 seconds) (6 7} L-W. Hu, Thennal Hydraulic blixing Transients in the bilT Research Reactor Core Tank, Ph.D. Thesis, Nuclear Engineering Departmec,t, MIT, Feb.1996. , O O O 6-16 SR#-0 97-14 OCT 031997
- 7. -Instrumentation and Control- System O The instrumentation and control system for the fission cometter is summarized in this chapter. Signals that are'important for th safe operation of the fission converter are displayed in the control room for the MIT Research Reactor. This chapter is divided into four sections: nuclear instrumentation, thermal-hydraulic instmmentation essential for safety, fission convener shutdown mechanisms, and other instmmentation that is not essential for safety.
- 7.1 Fission Converter Nuclear Instmmentation l I lt 5 important to understand that the role of the nuclear instrumentation in a facility-that is capable orachieving criticality such as a research reactor is very different from that in a suberitical facility such as a fission converter. In the former, the power level can be varied arbitrarily through use of the control mechanisms. Also changes in operating conditions may affect 'a reactor's response through reactivity feedbacks. Automatic shutdown on both high power and short period is, therefore, essential in a reactor, in contrast, it is not possible to vary the power level in the fission converter. The power level is fixed by design of the converter and its coupling (i.e., cadmium shutter open or closed) to an associated neutron source. A fission converter's power can be measured viti a calorimetric during conditions of thermal equilibrium and thereafter should remain unchanged until fuel depletion becomes appreciable. The role of the nuclear instrumentation is, therefore, to confirm this power -level when conditions of thermal equilibrium are not possible. -Period measurements are not necessary. Studies of the fission. converter's power (see Table 2.3) show that fission converter power varies linearly with reactor power. The fission converter will be equipped with one or more neutron sensitive nuclear instruments. One such instmment should be operable before the fission converter is-brought to power and utilized. Indication of fission converter neutronic power will be provided in.the reactor control room. In addition,' this signal will cause closure of the fission converter's cadmium curtain in an event of an overpower condition (110% of converter design power or 275 kW). It should be noted that the principal protection against - fission converter overpower is the reactor safety system because an overpower condition on the converter could occur only as a result of a transient on the MIT Research Reactor. 7-1 ' SR#-0-97-14 OCT 031997 in that event, the reactor safety system, which is independent of the fission converter, O would shut the reactor down and hence also shut the fission converter down. V in addition to the cadmium curtain closure signal that is generated at i10% of converter design power, an alamt will be provided at 110% of fission convener operating power. This is provided because the 2501 W design power will only be achieved if the reactor is at 10 MW and if the fission converter uses fresh fuel elements and light water coolant. 7.2 Fission Converter Thermal Hydraulic Instrumentation Essential for Safety The parameters of concern are the outlet temperature of the coolant, the coolant Oow rate, and the coolant level in the fission converter tank. One or more thermocouples or equivalent devices will be used to measure the hot leg outlet temperature of the fission converter. Indication of the fission converter hot leg outlet temperature will be provided in T the reactor control room. In addition, this signal will cause automatic closure of the fission converter's cadmium curtain in the event of an over-temperature condition (Setpoint: 55' C). O A conductance tyr- level probe or an equivalent device will be used to monitor coolant level within the fission converter tank. In addition, this signal will cause both closure of the fission converter's cadmium curtain and an automatic reactor scram in the event of a loss of primary coolant (Setpoint: 2.4 m above top of converter fuel). Fission converter primary flow will be measured u3ing an orifice plate or an equivalent device, Indication will be provided in the reactor control room. In addition, this signal will cause both closure of the fission converter's cadmium curtain and an automatic reactor scram in the event of a loss of flow (Setpoint: 50 gpm). 7.3 Fission Converter Shutdown Svstem The primary means for controlling fission converter power is the cadmium curtain. Controls for opening this curtain are provided at the fission converter medical control panel and controls for closing it are provided both there and in the reactor control room. The backup to cadmium curtain closure is a minor scram of the reactor. This could occur either automatically or manually: , C-7-2 S R#-0 14 OCT 031997 (3 i) Aptomatic Reactor Scraru - The reactor will be automatically scrammed in 'U# the event of fission convener loss of primary flow or low primary coolant level. Such action is not necessary in the case of over-power or high temperature. An overpower condition could only occur as a result of a reactor transient. In that event, the reactor safety system would be automatically actuated. A high temperature condition could occur as a result of either a reactor transient or a fission converter flow problem. In the event of the former, the reactor safety system will be automatically actuated. In the event of the latter, the fission converter low flow signal will be actuated, thereby closing the cadmium curtain and scramming the reactor. ii) Manual Reactor Scram - A manual reactor minor scram button will be available at the fission converter medical control panel and will be operable whenever the fission converter is in use. Table 7.1 summarizes the protective actions for the fission converter transients related to safety, n C Table 7.1 Protective Actions for the Fission Converter Transients Related to Safety. Transient Automatic Automatic Reactor Scram Cadmium Shutter Setpoints Closure Overpower X 275 kW Ovenemperature X 55'C Low Coolant Level X X 2.4 m Low Primary Flow X X 50 gpm V('% 7-3 S R#-0 14 OCT 031997 mc /) 7.3.1 Doerability of Fission Converter Shutdown Svstem V The fission converter is defined as being shut down when the cadmium curtain is fully inserted or when the reactor is in a shutdown condition. The fission converter is defined as being secured when there is no fuel in the fission convener or all of the following conditions are satisfied: (i) The fission converter is shut down, , (ii) The fission convener medical control panel key switch is in the off position and key is in proper custody, and (iii) There is no work in progress within the converter tank involving fuel. Whenever the fission converter is in either a shutdown or a secured condition, the automatic reactor scrams that originate on loss of fission converter primary flow and level are not required. Neither is the manual reactor scram located at the fission converter medical control panel. (~N 7.4 Other Instrumentation V Table 7.2 lists the instrumentation associated with the fission converter. The first four entries are required for reasons of safety as described in Section 7.1 and 7.2. The other instruments listed are not safety-related. O v 7-4 SR#-0 97-14 OCT 031997 c Table 7.2. Fission Converter Ins'mmentation Parameter Instrument Readout Alarm Location Safety Related Design Power Neutron Detector CR Yea Outlet Temperature Thennocouple or Equivalent CR Yes - Coolant Level Conductance Level Probe or CR Yes (Trip Point) Equivalent Primary Flowrate l Orifice Plate or Equivalent CR Yes Not Safety Related ~ Operating Power Neutron Detector CR Yes Hx Secondary Flow Rate Orifice Plate or Equivalent CR Yes l Primary inlet Temperature Thennocouple or Equivalent CR No Secondary Outlet Thermocouple or Local or FCCP No l Temperature Equivalent Secondary inlet Thermocouple or Local or FCCP No - Temperature Equivalent Cleanup System Themloccuple or Equivalent Local or FCCP Yes Temperatere Coolant Conductivity - lon Conductivity Probe Local or FCCP Yes Column Inlet Coolant Conductivity - lon Conductivity Probe Local or FCCP No I3 V) Column Outlet Cleanup System Flowrate Rotometer or Equivalent Local or FCCP No Coolant Level (Indication) Coolant Level Sensor Local or FCCP No Leak Detection Leak Tape or Equivalent Local or FCCP Yes He Gas Holder. Level Manufacturer's Design Local or FCCP Yes Recombiner Flowrate Manufacturer's Design Local or FCCP No Recombiner Temperature Thermocouple Local or FCCP No or Equivalent Primary Coolant Pressure Gauge Local or FCCP No Pressure (@HX) Secondary Coolant Pressure Gauge Local or FCCP No Pressure (@HX) Med Room Gamma Gamma Detector CR and MRCP No Monitor Cd Shutter Position Limit Switch CR and MRCP Yes Mechanical Shutter Limit Switch CR and MRCP Yes Position Water Shutter Tank Level Level Probe CR and MRCP Yes Water Shutter Upper Tank Level Probe MRCP Yes Level Med Room Door Position Limit Switch CR and MRCP Yes Storace Tank Level Gaune Local or FCCP No MRCP: Fission Converter Medical Room Control Panel CR: Control Room FCCP: Fission Converter Control Panel O V 7-5 S R#-0 14 OCT 031997 i
- 8. Pre-operationa! Tests and Initial Operation n
i) This chapter summarizes the p.coperational tests and initial operation of the fission converter. The preoperational tests are designed to ensure that the Gssion converter has been constructed in accordance with the infonnation presented in this report. Specifically, the tests are to prove the satisfactory operation of essential fission converter components. The procedure for the initial fuelloading of the fission converter is described. Calculation has shown that the maximum kerf of the fission converter is 0.62. Hence, a criticality condition is not credible. However, sub-critical multiplication should be monitored for safety during the initial fuelloading. The major steps for the fi'st approach for the fission converter's operation at power are summarized. These include measurement of the flow distribution in the fission converter core region and estimation of the fission converter reactivity effect on the MITR. The initial startup of the fission convener to its highest available operating power will be achieved by a stepwise rise of the reactor's power. 8.1 Pre-Onerational Tests ( Pre-operational tests will be performed to ensure that the facility will operate as O designed. Dummy fuel elements that replicate flow conditions are avaMable for the pre-operational tests. These tests will be used to establish initial compliance with the approved technical specifications. The tests will include component inspection, veriGcation that perfomiance objectives are met, instrument calibrations, and the operability of interlocks. All tests will be conducted in accordance with the existing MITR quality assurance program. 8.1.1 Non-Nuclear Instrument Cahdon Instruments for measuring system pressure, temperature, and flow can be calibrated prior to the initial startup of the fission converter. The techniques used for these calibrations will be those currently employed to calibrate similar instruments on the MIT Research Reactor. Accordingly, these instruments will be calibrated pnor to the initial stanup of the fission converter and technical specifications pertaining to the signals from these instruments shall be in effect during the initial startup, p V 8-1 S R#-0 14 OCT 031997 8.1.2 Nuclear instrument Calibration r ( The nuclear instrumentation associated with the fission converter can not bc
- alibrated in advance of the initial startup. Accordingly, the technical specification requirements associated with signals from these instruments shall not be in effect during the initial startup. The normal method for calibrating nuclear instmmentation at the MIT Research Reactor is to perfonn a calorimetric in which the output of the nuclear instrument is correlated with themial power. In the absence of an operating facility, there can be no thermal power and hence no calibration. This is why the nuclear instrumentation will not be available for use during the initial startup. An attemate to a thermal power calibration would be to perfomi a flux determination using activated foils. The anticipated foil activity could be calculated from theoretical predictions of the flux and those numbers compared to measured data. This comparison would allow a calibration of the nuclear detectors.
However, this technique also requires that the fission converter be operating. l The followir.g procedure will be used to calibrate the fission converter nuclear instrumentation. An operating condition that is known from theoretical analysis to be very conservative and hence to not represent a challenge to the safety limits will be identified. Either a thermal power calibration or a foil activation measuremert or both will be performed under this conservative condition. The information obtained will be used to calibrate the nuclear instrumentation, it should be noted that the maximum fission converter power,250 kW, will be achieved only if the MIT Reseudi Reactor is operating at 10 MW andTresh fuel is used with light water coolant. At present, the MIT Research Reactor's authorized maximum power is 5 MW. Accordingly, the maximum possible operating condition is presently a factor of two below the design operating condition. This affords considerable conservatism in itself. 8.2 Onerator Training Individuals who hold either operator or senior operator licenses at the time of , commissioning of the fission convertet will be given specialized instmction on its design and operation. This will be followed by a written examination equivalent to a requalification examination. Future reactor operator candidates will receive training on the fission converter as part of their initial qualification. O 8-2 SR#-0 14 OCT 031997 8.3 Initial Fuel Loading The initial fuel loading will be made after the pre-operational testing has been -satisfactorily completed. The reactor will be shutdown and the cadmium curtain fully closed before the fuelloading of the fission converter begins. Calculation has shown that the maximal reactivity of the fission converter is 0.62. So, a criticality condition is not credible (Section 2.3.1). Nevertheless, sub-critical multiplicMion will be monitored during the initial fuel loading. The procedure for the initial fuel loading will be the standard technique involving plots of the inverse count rate as the fuel elements are loaded. A neutron source (Pu Be or equivalent) will be located in the center position of the fission convener core. The pattern for loading the fuel will be kept symmetric. The source will be removed prior to insertion of the final fuel element. This element will be loaded subject to the restriction that the 1/M plot shows that the keff for the convener will be less than 0.90, the maximum allowed value for a fuel storage location at the MIT Research Reactor. 8,4 Fuel Region Flow Distribution Measurement ' O V The flow distribution among the fuel elements will be measured with the cadmium curtain closed. A Pitot tube or an equivalent device will be used to measure the flow rates through each fuel element. The minimum flow rate through any of the fuel elements along with the flow distribution within a fuel element (this is known from the MITR-Il stanup test data (Ref. 81]) will be used to determine if the operating limits on the fueled region coolant flow factor aM the channel flow disparity factor are satisfied (Section 3.4.2). 8.5 Reactivity Estimation of the Fission Converter Estimation of the fission converter reactivity is divided into two parts. First, the integral reactivity associated with fully opening the cadmium curtain will be measured. Second, the differential reactivity worth associated with partial opening of the cadmium curtain will be measurea. This latter information is needed to estimate the reactor period corresponding to the rate at which the cadmium curtain is opened. O G 8-3 SR#-0 97-14 OCT 031997 l l 8.5.1 Estimation of Integral Reactivity Wonh (h
- Pre-conditions
- 1. Fission converter fuel loading complete.
- 2. Reactor shutdown with minimal Xenon.
- 3. Fission converter heat removal system operable (not required when cadmium curtain closed).
- Procedure
- 1. Take reactor critical. Level reactor power at a designated low power (e.g.,500 W)
- 2. Record critical data.
- 3. Shutdown reactor.
- 4. Open cadmium curtain fully.
- 5. Repeat steps 1,2, and 3.
! 6. Calculate the reactivity worth of the fission converter by comparison of the critical positions obt iined in steps 2 and 5. fG 8.5.2 Estimation of Differential Reactivity Worth [V A differential curve of the reactivity associated with opening the cadmium curtain will be obtained with the cadmium curtain in various positions (e.g., 25 %, 50 %, 75 % open). This will be done by repeating the procedure described in section 8.5.1. 8.5.3 Determination of Reactor Period Associated with Fission Converter Ooeration The minimum anticipated reactor period will be calculated from the integral and differential reactivity information using the dynamic period equation [Ref. 8-2] (Nnig: The in-hour equation is a speciMized case of the dynamic period equation). The rate of opening the cadmium curtain which results in a 50 second period will then be determined and this information will be used to specify the maximum allowed opening speed of the cadmium curtain. Confirmation of the above calculation will then be made by observing the reactor period associated with opening the cadmium curtain. This will be done with the reactor initially critical at a power level of 1 MW or less. 3 (G 8-4 SR#-0 97-14 OCT 031997 8.6 Initial Apnroach to theHighest Available Fission Converter Operating Power In this section, a procedure for a stepwise increase to the maximum available fission converter power is summarized. The neutron flux from the MITR determines the fission converter power. The design power of 250 kW corcesponds to a reactor power of 10 MW, fresh fuel and light water coolant are used in the fission converter. At present (1997), the maximum operating power for the MITR is 5 MW which will yield a maximum fission converter power of 125 kW. At such time as the MITR's licensed power level is increased to 10 MW, the procedure outlined below will be repeated.
- 1. The interior of the fission converter facility will be checked to ensure that no foreign objects are present.
- 2. All process and radiation monitoring systems will be placed in their normal operating condition with non nuclear instruments calibrated.
- 3. The fission converter's top shield lid and associated shielding will be removed in order to allow the in-core temperature distribution measurement to be made during the initial portion of the stepwise power increase. (Caution: The top lid and shielding shall be installed before raising the fission converter power above 25 kW. The estimated dose rate on the coolant surface for this power level is 560 mR/h as described in Appendix 3.2)
- 4. The converter power will be increased by raising the reactor power in a stepwise manner. Radiation levels and system temperatures will be monitored during each power increment. This procedure will be repeated until the maximum available operating power is attained.
- 5. The following measurements will be made:
a) In-Core Temperature Distribution The in-core temperature distribution will be measured during the initial power ascension. The result of this temperature distribution measurement will be used to identify the hot channel and to determine that both the operating limit for power deposition and the nuclear hot channel factor are satisfied (Section 3.4.1 ). 8-5 SR#-0 14 OCT 031997 ~ b) Process Parameters Fission converter primary inlet and outlet temperatures as well as flow rate will be measured. This information will be used to perfonn a calorimetric. c) Radiation Surveys Radiation measurements will be made outside the fission converter facility and inside the fission converter medical therapy room via remote monitoring, d) NuclearInstrument Calibration The fission converter power will be calculated via a calorimetric. Energy losses because of gamma radiation etc. will be taken into account. The equilibrium neutron count rate associated with the nuclear instrumentation will be measured. Correlation e'these count rates with the calorimetric will be used to calibrate the nuclear instruments. The above procedure for a stepwise increase of the fission converter operating power will be repeated if any one of the following design changes is made: O V 1 The maximum available operating power is increased,
- 2. The fission converter p .iary coolant is changed from H2 O to D2 0 (the hot f
channel factor increases - see Table 2.4 ), or .
- 3. Fresh fuel is used to replace bumed fuel.
References {3-1] MITR Staff, "MITR-Il Startup Report", MITNE 198, Feb.1977. [82) Bernard, J.A., Henry, A.F., and D.D. Lanning, " Application of the ' Reactivity Constraint Approach" to Automatic Reactor Control," Nuclear Science and Engineering, Vol. 98, No. 2, Feb.1988, pp 87-95. . ,r . ( 8-6 SR#-0-97-14 OCT 031997 l}}