Text

Revised SER for Fission Converter Facility
	ML20199H237
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	12/31/1998
From:	; NUCLEAR REACTOR LABORATORY
To:	;
Shared Package
ML20199H226	List: ML20199H220; ML20199H237; ML20199H266;
References
	NUDOCS 9901250150
	Download: ML20199H237 (96)
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O SAFETY EVALUATION REPORT FOR FISSION CONVERTER FACILITY O l l

1 DECEMBER 31,1998 I

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TABLE OF CONTENTS O

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1. INTRODUCTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 1.1 Scope of This Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -2

2. FISSION CONVERTER NEUTRONIC DESIGN.................................... 2-1 2.1 General Description of the Fission Convener ................................. 2-1 2.2 Fu el De si gn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-6 2.3 Criticality and Power Distribution............................................... 2-6 '

2.3.1 Criticality Calculation for the Fission Converter................... 2-6 2.3.2 Reactivity Worth of the Fission Converter.......................... 2-8 2.3.3 Fission Converter Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-9 2.3.4 Fission Converter Radial and Axial Power Distribution .......... 2-9 4

2.3.5 Fuci Element Orientation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 1 2.4 Filter / Moderator . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 1 p 2.5 Col 1imat or . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 2

3. FISSION CONVERTER THERMAL HYDRAULIC DESIGN..................... 3-1 3.1 General Description of the Thermal Hydraulic System....................... 3-1 3.1.1 Primary Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3.1.1.1 Fission Converter Tank.................................. 3-3 3.1.1.2 Fission Converter Fuel Housing Grid Design........ 3-7 .

3.1.1.3 Cover-Gas System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-9 3.1.1.4 C1 e a n-Up S y s t e m . .. . . . . . . . . . . . .. . .. . . . . . .. . . . . . . . . . . . . . . . 3- 10 3.1.1.5 Make-Up Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 1 3.1.1.6 - Coolant and Cover-Gas Sampling...................... 3-11 3.1.2 Secondary Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 1 3.1.3 Material Selection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 12 3.2 B asis for Thermal-Hydraulic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 13

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Esige 3.3 Computational Method for the Fission Convener Thermal Hydraulic Design..............................................................................3-13

< 3.3.1 Me thodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 13 l ,

3.3.2 Major Correlations Used in MULCH-FC........................... 3-14

3.3.2.1 Correlation for Onset of Nucleate Boiling l (ONB ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 14 l

3.3.2.2 Correlations for Onset of Significant Voiding (O S V) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 15 ;

l 3.4 Fission Converter Operating Limits............................................. 3-16 3.4.1 Power Deposition Factor (Fp) and Nuclear Hot Channel Factor (FHC)... .3-16 l 3.4.2 Fueled Region Coolant Flow Factor (Fr) and Channel Flow Disparity Factor (d ) . . . t. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 17 3.4.3 Engineering Hot Channel Factors.................................... 3-18 l 3.5 Thermal Hydraulic Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-21 3.5.1 Definitions of Safety Limits and Limiting Safety System

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l tJ S e t t i n g s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 -2 2 g 3.5.2 Derivation of the S afety Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-23 3.5.3 Derivation of the Limiting Safety System Settings................. 3-27

4. F1SSION CONVERTER SHUTTER AND MEDICAL THERAPY ROOM i DESIGN....................................................................................4-1 l 4.1 General Description of the Shutter Design . .. ... ...... .... .... .. ...... ..... .... 4-1 I

i 4.1.1 Converter Control S hutter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 4.1.2 Collimator S hutter Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 4.1.2.1 Water S hutter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 l 4.1.2.2 Mechanical Shutter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.2 S hutter Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 4.3 M e dical Therapy Room De sign................. ............. . .... . .. ...... .. . . .. 4-7 l 4.3.1 Me dieal Therapy Room Door... ............. ........ .... ........... .. 4-7

5. F1S SION CONVERTER FUEL HANDLING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 b,/

t 5.1 Fuel Element Security, Storage, and Quality Assurance ..................... 5-1 ii t

TABLE OF CONTENTS t

Eil8G 5.2 Fuel Element Self-Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 1 5.2.1 Use ' of Unirradiated Fuel.............................................. 5-2 5.3 Fuel Element R e m o y a 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -2 5.4 Fuel Element Handling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.4.1 Fuel Element Transfer Cask . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 5.5 - Samples............................................................................5-4

6. S AFETY ANALYS IS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 1 6.1- Maxim u m Hypothetical Accide nt................................................ 6-1 6.2 Insertion of Excess Reactivity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.3 Los s of Primary Coolant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6.4 Loss of Primary Coolant Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-8 6.5 Loss of One of Two Primary Pumps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 12 6.6 Loss of Off-Site Electric Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 14 6.7 Loss of Heat Sink . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 14 6.8 Mishandling or Malfunction of Fuel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 15 6.9 Experiment Malfunction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 15 6.10 N a t u ral D i s t u rb a n c e s . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . 6- 15 6.10.1 Eanhquake . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 15 6.10.2 L i g h t n i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 16 6.10.3 Se vere Storm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6- 16

7. INSTRUMENTATION AND CONTROL SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7- 1 7.1 Fission Converter Nuclear Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7,1 17.2 Fission Converter Thermal-Hydraulic Instrumentation Essential for Safety...............................................................................7-2 7.3 Fission Converter Shutdown System........................................... 7-3 7.3.1 - Operability of the Fission Convener Shutdown System.......... 7-5

( 7.4 O t h er In s t ru m e n t a t 10 n . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . 7-5 iii

TABLE OF CONTENTS em

8. PRE-OPERATIONAL TESTS AND INITIAL OPERATION ....................... 8-1 8.1- Pre-Operational Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 1 8.1.1 Non Nuclear Instrument Calibration ........................... ... .. 8-2 8.1.2 Nuclear Instrument Calibration....................................... 8-2 8.2 Operat or Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2 8.3 Initial Fuel Loading . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.4 Fuel Region Flow Distribution Measurement.................................. 8-3 8.5 Reactivity Estimation of the Fission Converter................................ 8-3 8.5.1 Estimation ofIntegral Reactivity Worth............................. 8-4 8.5.2 Estimation of Differential Reactivity Worth......................... 8-4 8.6 Initial Approach to the Ilighest Available Fission Converter O P o w e r. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... .. ... .. . .. ... .. ..8.-4. . . . . . . . . . . . . . . . . . . .

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1. Introduction f3 Y

Boron Neutron Capture Therapy (BNCT)is a binary form of cancer therapy that has the potential to kill cancer cells selectively while sparing normal tissues. This can be accomplished by localizing 10B in the tumor cells and delivering thermal neutrons to the target [Ref.1-2]. Compounds such as p-boronphenylalanine (BPA) can be used to localize 10B in the tumor by taking advantage of the higher metabolic rate of the tumor cells. A 10B nucleus that absorbs a thermal neutron disintegrates into an alpha particle and 7Li nucleus with an accompanying energy release of 2.79 MeV. The alpha particle and the7Li 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 Massachusetts Institute of Technology (MIT) using the M67 beam that has an epithermal neutron flux of 2.1x108 n/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 epithermal neutron irradiation used for BNCT of a human subject anywhere in the world. Brain cancer trials were subsequently initiated at the Brookhaven National l]

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 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to deliver a normal tissue tolerance 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.

l The new beam design is based on a fission converter plate driven by the neutrons from the Massachusetts Institute of Technology Research Reactor (MITR). Neutrons from

' the reactor am 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 l l-1 l

10 kev) [Ref.1-1]. The cooling of the fuel contained in the fission converter will be ;

provided by forced convection of either H2O or D 2O enclosed in a tank.

The purpose of the design is to deliver a neutron flux of about 1x1010 n/cm2 s with !

specific fast neutron and specific incident photon doses lower than 2x10-11 cGy cm2/ epi n to minimize non-selective dose components. An epithermal neutron flux at this intensity !

would result in an irradiation time of less than ten minutes at a reactor power of 10 MW L [Ref. 1-1]. Irradiation times in this range are typical of those used with conventional ;

extemal 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 convener.

Technical specifications that constitute limitations for the design and operation of the facility are also included.

1.1 Scone of This Report The focus of this repon 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 penains to patient therapy and beam calibration is not repeated hem.

References q l

[l-1] W.S. Kiger III, Neutronic Design of a Fission Converter-Based Epithennal Beam for Neutron Capture Therapy, Nuclear Engineer Thesis, MIT,1996.

[1-2] D.E. Wazer, R.G. Zamenhof, O.K. Harling, and H. Madoc-Jones, " Boron Neutron Capture Therapy," in Radiation Oncology: Technology and Biology, edited by P.M. Mauch and J.S. LoefDer, W.B. Saunders Company, Philadelphia, l 1994.

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3 2. Fission Converter Neutronic Design b

The neutronic design of the fission converter is summarized in this chapter.

2.1 General Description of the Fission Converter Figure 2.1 is a view of the MIT Research Reactor (MITR) and its existing components. An isometric view of the MIT 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 convener tank will be located to the exterior of the reactor's graphite reflector in the region previously occupied by the thermal column. The fission converter plate will be centered on the 35.6 cm (14 inch) window in the graphite reflector. A 1

converter control shutter (Section 4.1.1) located between the fission converter plate and the MITR core is used to control the neutron flux from the MITR and, therefore, the fission convener power. To reduce radiation dose in the medical therapy room, additional shutters are located in the collimator region (Section 4.1.2). l g The neutron flux from the reactor is thermalized in the D2 O and graphite reflectors.

d This thermal flux is converted to a fission spectrum by the fission converter plate. A filterhnoderator is then used to tailor the fission spectrum to climinate unwanted fast neutrons and photons without significantly decreasing epithermal neutrons (1 eV to 10 kev). A collimator then directs the epithermal neutron beam onto the patient.

Cooling of the fuelis provided by forced convection of either H2O or D 2O enclosed in an Al-5083 tank. The fission converter's primary coolant will be cooled by a heat exchanger for which the secondary side is connected to the MITR cooling towers through the reactor's secondary coolant system. The fission converter's thermal 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 MITR'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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fissd,;, ' g- * ;e 0 398cm E , : (1% Borated .- ,

Polyethylene -

ht r Collimator 4

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90cm # - - a

.4

. -4 D

.9

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, ,% 90cm

v. , 9 . . . ,

5 9 .

280cm I Figure 2.3 Too View of the Fission ConverterFacility t

I

,I

fN /'

V)

V d Water Filled v . Fission Converter 600cm

. Tank

~

.. I "'" Biological ' 't ll . . '.- .- -

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.* 437cm Mo. derat.or ,

l l

Collimator 1

i ' '

[(Water Shutter) . ,. 3 ticm smu" m, 9 <

l l

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/

N Patient Position

" Cadmium k p 1 Shutter ~N,l N g

i jg k h .

swwww Imad

s 4 't

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m 3,. -

kHydraulically Operated 151cm Fast Shutter 191cm 450cm 120cm Graphite 1% Borated Reflector Polyethylene 14*x14"x14* Window in Graphite i

Figure 2.4 Side View of the Fission Canverter Frility 1

t

_ _ _ _ _ . _ - __.__._m_ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ . , . - . . _ , . . - . . . . m -. , - -

l 2.2 Fuel Desien

'V The fission converter plate consists of an array of MITR fuel elements arranged in a fuel grid plate assembly that is located in the fission converter tank. The fission convener plate can contain up to cleven MITR fuel clements.

The fission converter will normally use cleven MITR fuel elements with a small cicarance between them. The gaps are blocked by the grid plate so the amount of the bypass flow is minimized. Fewer elements may be used provided that dummies having the same exterior geometry as an element are ernployed so as to maintain fluid flow behavior and provided that the thermal-hydraulic cdteria established in Chapter 3 of this report am satisfied. Because the fission converter uses the s une fuel element as the MITR, the fuel specifications given in MITR Technical Specifications shall apply for the fission converter.

2.3 Criticality and Power Distribution 2.3.1 Criticality Calculation for the Fission Converter

/h

( ) The criticality of both the fission converter and the coupled core-converter system v

were studied. The Monte Carlo N-Particle (MCNP) code has been used for neutronic studies of the fission converter. The MCNP model of the MIT Research Reactor has been '

l extensively validated in the core region and in the thermal column region where the fission l l

converter will be located [Ref. 2-1,2-2].

To calculate the kerr of the fission convener, criticality calculations of the fission converter beam and core albedo model were performed using the K-code option of MCNP

[Ref. 2-1]. The effective multiplication factors (kerr) calculated are listed in Table 2.1. The kerr values calculated for a D 2 O cooled system are 0.268 for partially spent MITR fuel and 0.344 for fresh MITR fuel. For an H O2 cooled system, the kerr calculated values are 0.514 and 0.618 for partially spent and fresh MITR fuel, respectively. Because the kert predicted is much smaller than unity, a criticality accident is not credible.

D \

l:V l

2-6

i O Table 2.1: Criticality Analysis of the Fission E'onverter 235 Coolant Fuel (g U) kerrConverter Alone D02 312 0.268i0.001 (Spent MITR-II fuel)

DO2 510 0.344i0.001 (Fresh MITR-Il fuel)

HO2 312 0.51410.001 (Spent MITR-II fuel)

HO2 510 0.61810.001 (Fresh MITR-II fuel)

NeIn: The statistical uncertainty listed with each value represents one standard deviation

[Ref. 2-1].

'%/

Table 2.2: Reactivity Change Associated with Opening of the Converter Control Shutter to the MITR Coolant Fuel (g 235U) kdf Reactivity CCS Closed CCS Open (Ak/k)

D02 312 1.00455i0.00048 1.0049010.00036 0.00035i0.00060 (Spent MITR- (45176 m )

II Fuel)

II2 0 510 1.0041710.00051 1.0054310.00050 0.00125i0.00071 (Fresh MITR- (159190 m )

II Fuel)

Neig: The statistical uncenainty listed with each value represents one standard deviation

[Ref. 2-3].

[ \

.r 2-7

2.3.2 Reactivity Wonh of the Fission Converter The keft values of the coupled MITR core-convener system wem calculated similarly using the K-code option of MCNP [Ref. 2-3]. Criticality calculations to estimate i the reactivity insenion 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 convener by opening the convener control shutter (CCS) will cause a change in reactivity of 0.00035i0.00060 Ak/k (45176 m ) for a D 02 system using partially spent fuel and 0.00125i0.00071 Ak/k (159190 m ) for an 110 2 system using fresh fuel. The cases cited bound the anticipated range of reactivities. (Note: The reactivity wonh 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-II 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 wonh to be 0.0000257 Ak/k (3 m ) [Ref. 2-4].)

The actual reactivity worth of the fission converter will also be determined during pre-operational testing (Section 8.5) and annually thereafter. The above refemnced 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 m ).

MITR-II Technical Specifications provide several approaches for limiting the ,

reactivity associated with an experimental facility. Specification No. 6.1 imposes limits l depending on whether the experiment is classified as moveable, non-secured, or secured.

Its is anticipated that the reactivity worth of the CCS will be less than the movable limit (0.2%AK/K). Routine reactor control actions can be used to compensate for the change in reactivity resulting from opening of the CCS as long as the reactivity wonh is within the limit.

2.3.3 Fission Converter Power The Monte Carlo N-Panicle (MCNP) code has been used to calculate the fission converter power. The calculated values am summarized in Table 2.3. As shown, the maximum power generated by the fission converter is 251 kW with MIT mactor power at 2-8

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

10 MW [Ref. 2-1]. At this power level, each fuel element generates an average of 22.8 kW O l Q compared to an average power of 208 kW per fuel element in the MITR-II core (5 MW, 24 l

fuel elements). ,

i Table 2.3: Calculated Fission Converter Power Fission Convener Fission Convener Coolant Fuel (g 235U) Pcwer at 5 MW Power at 10 MW l Reactor Power Reactor Power '

(kW) (kW)

DO 2 312 81.5 163.0 (Spent MITR-II Fuel) i0.3% i0.3% l DO 2 510 105.4 210.8 (Fresh MITR-II Fuel) i0.2% i0.2% i 1

HO 2 312 83.4 166.8 (Spent MITR-II Fuel) i0.2% i0.2%

H2O 510 125.5 251.0 (Fresh MITR-II Fuel) 10.2 % i0.2 %

l Ecig: Data am from the coupled core criticality calculations. The statistical uncertainties listed as a percent for each value represents one standard deviation [Ref. 2-1].

2.3.4 Fission Convencr Radial and Axial Power Distribution The power distribution within the fission convener plate was calculated using MCNP [Ref. 2-5]. These calculations were done assuming that the reactor' operates at 5 MW and that fresh MITR fuel elements are used in the fission converter. l l

Table 2.4 lists the fission converter power, fuel plate maximum power, and the corresponding nuclear hot channel factor. The nuclear hot channel factor is the ratio of the I fuel plate maximum to the fuel plate average power level. Hence it is the same for both 5 MW and 10 MW power operations. 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

._ _ . . ._ . - _ _ . _-. . .._-.. . - - . . - _ - . - . = - - . ..-

O Table 2.4: Results of Fission Converter Radial Power Distribution Fission Fuel Plate Fuel Plate Nuclear Converter Avg. Power Max. Power Hot Channel Power (kW) (kW) (kW) Factor

H20-cooled 125.5 0.76 1.12 1.47 Fresh Fuel ,

D20-cooled 105.4 0.64 0.98 1.53 Fresh Fuel

Nuclear Hot Channel Factor = Fuel Plate Max. Power / Fuel Plate Avg. Power Egic: Data is from Ref [2-5] normalized to that from Ref. [2-1].

l Table 2.5: Axial Power Distribution for the Fission Convener [Ref. 2-5]

O Height 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 21.3 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 1

-7.1 23.8 2.84

-11.8 20.4 2.90

-16.6 18.6 3.04

-21.3 17.2 3.20

-24.9 15.8 4.48

-27.2 17.7 4.11 O

2-10

calculations. Table 2.5 shows the vertical power profile of the hottest plate. The venical power profile 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 be rotated by 90 so that one entire plate were facing the MITR, calculations show that the peaking in that plate would be significantly greater [Ref. 2-6]. Administrative procedures will be used to ensure that fuel elements are loaded with proper orientation. (Note:

Rotation of a fuel element by 90* is physically impossible unless it is first invened.)

l 2.4 Filter / Moderator Thermal neutrons from the reactor are converted to a fission spectrum by the convener plate. A filter / moderator is then used to tailor the fission spectrum to climinate unwanted fast neutrons and photons without a significant decrease in the epithermal O

V neutrons (1 eV to 10 kev). Filtration is accomplished through the use of resonance scattering % c,.Mais with large scattering neutron cross sections in the fast energy range,1/v behavioi m the thennal region, and a relatively low, flat cross section at epithermal energies

[Ref. 2-1]. To reduce contamination from thermal neutrons, thermal neutron absorbers such as Cd, *B, or ki can be used. High Z material such as bismuth or lead can be used to reduce the contamination from photons.

Material selection for the filter is made to ensure that a phase change because of elevated service temperatures is precluded under both normal and accident conditions. In addition, the selection minimizes decomposition and accumulation of long term activities

[Ref. 2-1]. Satisfactory beams can be obtained with combinations of aluminum and aluminum oxide, aluminum and aluminum fluoride, aluminum and Teflon @ , and aluminum and graphite. Furthermore, these materials have engineering propenies such as mechanical strength, temperature resistance, radiation stability, and fabricability which make them suitable for this application. We have initially chosen to use aluminum /reflon@

because of its somewhat lower cost, ready availability, and very easy fabricability compared to the other material combinations.

O v

2-11

1 2.5 Collimator l

The resulting beam from the filter is focused by a collimator that is lined with a i 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 i the beam (Section 4.1.2).

i l

l l

References )

[2-1] W.S. Kiger III, Neutronic Design of a Fission Onverter-Based Epithennal Beam )

for Neutron Capture Therapy, Nuclear Engineer Thesis, MIT,1996.

]

.[2-2] E.L. Redmond II, J.C. Yanch, and O.K. Ilarling, " Monte Carlo Simulation of the l Massachusetts Institute of Technology Research Reactor," Nuclear Technology, O Vol.106, April 1994. .

[2-3) File Calculation (Reactivity Change Associated with Cadmium Cunain Opening - '

MCNP Calculations)

[2-4] File Calculation (Reactivity Change Associated with Cadmium Cunnin Opening - l

- CITATION Calculation)

[2-5] File Calculation (Fission Convener Plate Power Distribution Calculations)

[2-6] File Calculation (Fission Convener Plate Fuel Element Orientation) i i

1

'O l

2-12

l l 1

3. Fission Converter Thermal Hydraulic Design l l[m v]

The thermal hydraulic design of the fission converter is summarized in this chapter. I The fission converter has been designed to accommodate either an 2H O or a D2 0 primary l coolant system. The design limits include consideration of the consequences of credible deviations from the operating conditions as well as allowance for manufacturing tolerances.

In addition, the design limits of the thermal hydraulic system are chosen to provide a reasonable safety margin beyond the desired operating range. Specifically, the limiting condition for operation is set to prevent indpient boiling in the fueled region of the convener. Such a conservative design assures 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.

3.1 General Description of the Themial Hydraulic System The MITR fission converter is designed to be cooled by forced convection under I normal operating conditions using either H2 O or D2 O as the primary. The primary coolant l is enclosed in an Al-5083 tank and is cooled by a heat exchanger for which the secondary A

(v) side is connected to the MITR's cooling towers through the MITR's secondary coolant I l

system. A schematic of the fission converter heat removal system is shown in Figure 3.1. I Provision is also made to allow the fission converter to operate at low power with natural circulation. Natural circulation is chieved 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 Hux measurements in the fueled region. In order to set up this type of activity, the fission converter's top shield lid is removed. The activity can be performed either with the lid in place or removed. Analysis l of the radiation level associated with operation of the fission converter in the natural circulation mode with the lid removed 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 Gow is diverted to the cleanup system, which consists of an ion column and filters, n

( )

v 3-1 l

l l

l

- * ~ - ,

. ~i. _

,_[ . , ,

,g. .

I

! ' Fill In # M O Storage & Makeup Tonk n )

N-M- !

Primary 9 l' U Cooto n*, !

Pump <, E '

N U X ;

Sanple x " :

ctean up eung N =

'l es

(

)

C Heat N h y ) Exhonger I

h U U

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.k 4 Sample X -

U ,

o t

MITR-II Secondary Coolant Systen l Fission Converter I Tonk j i

! Filter Filter i Sanple =

X M Inn Column M i k =

[ Figure 3.1 Schectic Drawing of Fission Convener Name Removal System 1

3-2

c to purify the coolant. The cleanup system could also be used to adjust water chemisty if

( necessary. An auxiliary pump is used to maintain a small flow through the clean-up loop.

It may also be used to maintain flow through the heat exchanger after fission convener shutdown to remove decay heat.

At the top of the fission convener tank, a helium 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. This valve is set at about 4 psig or lower. In addition, there will be a rupture disc set at 5 psig.

Figure 3.2 shows the cover-gas system. The tank will be tested at a pressure of 10 psig.

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 Ession converter secondary coolant system is discussed in Section 3.1.2.

3.1.1 Primary Coolant System 3.1.1.1 Fission Converter Tank The fission convener plate is enclosed in a tank made of Al-5083. Figures 3.3 and 3.4 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 t!. /G.E Boiler & Pressure Vessel Code. IIence, the fission converter tank is exempt from the specifications of that code. However, ASME code sections II 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 convener tank rests on a concrete upper shield block which serves as the primary tank support. This upper shield block is supponed 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,iflight water is used as the fission convener primary coolant then the presence of the aluminum block would reduce moderation of the neutron beam.

O o

3-3

I O '

i r- - - - - - - --{><}- - Sampler - -C><} q j To MITR-II l l

stack l .

l A I :

l I l l

I I r-- recombiner - ---y I

r--1 1 I i

I PRV l _ ._ _a ;

I <

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- l l 1------------

l T F- 7 i Rupture ,__ ,_ g l l l !

disc , ; l l

l y A

I I I i Storage i A Slope drainage and makeup i

gas & liquid coolant Helium line tank gas tank Figure 3.2 Schematic Drawing of Fission Converter Cover Gas System 3-4 r j

A U .__

e COOLANT OUTLET PIPE (I.D 1.5")

- COOLANT INLET PIPE lC -

(I.D 1")

s ,/'

\ WATER LEVEL

\s g(~

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+ SURFACE O '

r^sx witt T(THICKNESS 1/2")

DOWNCOMER WALL % _ '

o (1/2") ,

fALUMINUM BLOCKS i,

/

ell MITR-II FUEL ELEMENTS

/ Ncte: Detailed geometry of the fuel elements

- is shown in Figure 3.4 p: -

N g

O 3-5

p; ( ,.

,,3 O d U -

Eng: Tank is not symmetrical because of different structures at both ends.

Fuel Element Aluminum Block (15 Plates) !

=

= 130.89cm 16.34cm

wmmmaswmmma

= 01.12cm

=

= 103.35cm 10.61cm 16.92cm .

1f Reactor Core Figure 3.4 Cross-sectional View of the Fission ConverterTank -Lower Section i

i Associated requirements on this aluminum block include compliance with the operating (m) limits for Dow distribution and hot channel power deposition. The removable aluminum block shall be installed in the Hssion converter tank unless calculations have been performed to show compliance of the above limits for another configuration. Other configurations could include but are not limited to a block of a different material, the absence of the block, or a combination of a solid material and coolant. Calculations of K, for the Dssion converter have been made for fission converter operation both with and without the aluminum block. The maximum K, with the block removed using H2 O and fresh fuel is 0.670 i 0.0012, compared to 0.589 i 0.0012 with the block installed. The K, of the fission converter with the aluminum block removed is higher than that with the aluminum in place, but still significantly lower than the criticality limit.

The primary inlet and outlet pipes which penetrate the top shield lid can be removed for natural circulation operation.

3.1.1.2 Fission Converter Fuel Housing Grid Desien The fuel elements are arranged in a closed-packed array in a housing with a Dxed ,

I 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. ,

l l

Table 3.1 Fuel Restraint Calculation Results [Ref. 3-1]

l Fuel Element Downward Force 34.6 N (7.8 lbf)

Maximum Hydraulic 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.

l

(.

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31.1.3 Cover-Gas System n

A cover-gas system is provided at the top of the fission converter tank. All free surfaces of the coolant are blanketed with helium. The helium blanket performs five functions:

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 air in the presence of high radiation fields.

3. Provides an inen non-radioactive atmosphere that minimizes 41 Ar production.

4. Provides an inert atmosphere for the circulation of the byproducts of radiolytic decomposition through a recombiner.

5. Permits monitoring of fission product gas concentrations in the cover gas.

ly Elevated levels would be indicative of an incipient clad failure of a fuel element.

Helium is supplied from gas tank (s) through a pressure reducing valve. From there, helium is supplied to the free surface in the fission convener primary coolant system.

If the pressure builds up, helium will be vented to the MITR stack through a relief valve set at 4 psig or lower. In addition, there will be a burst disc set at 5 psig to prevent over-pressure.

The production of a flammable concentration of either D2 or H2 from the disassociation of either heavy or light-water in the fission converter is a concem 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 Specincation #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 D2, 200 kW for H 2) unless sampling shows that the concentration of flammable gas is quite low (2%

for D 2,1% for H ).2 Such stringent requirements are not required for the fission convener

() because:

3-9

i l

l 1

i l

( / l. The maximum thermal neutron Dux in the reDector is 5x1013 n/cm2 s whereas that x i in the fission converter is calculated to be 1.3x1011 n/cm2 s, a factor of 380 less. l i

2. Patient set up and removal time is, under the best of circumstances, an hour or more. Thus, even ifin routine use, the fission converter's duty cycle (i.e., time at power) would be at most a few hours per day.

1 I

l Given the above, it is more appropriate to operate the Dssion converter recombmer periodically as a maintenance requirement rather than routinely as an item of required equipment. It is therefore proposed to operate the fission convener recombiner for five j hours per month under the conditions specified by its manufacturer. (Nnte: The MITR reflector's recombiner operates continuously, 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months / month. Them is more than a factor of 300 difference in the Dux levels. The volumes of the two systems are essentially equivalent,260 gals for the reDector versus 180 gals for the fission converter tank. Hence, the fission convener recombiner would be needed about 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months / month for a 100% duty cycle. The specified five hour minimum is conservative by a factor of two and places no l restriction on the duty cycle.)

A V The MITR's technical specifications for D2 or H2 concentrations in helium am adopted for the fission converter system.

The fission converter cover gas is also circulated past a radiation 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.

(Note: This gas monitor does not fulfill a safety function because the fission convener is a closed system. Hence the fission convener may be operated without this detector.

Monitoring of a fuel element failure is achieved from a -y analysis of the fission converter's primary coolant.)

3.1.1.4 Clean-Un System A portion of the fission convener coolant from the heat exchanger Dows through a pump, an ion column inlet filter, a mixed bed ion exchanger, and an ion exchanger outlet filter. The purified coolant then rejoins the rest of the primary coolant that is being pumped OG back into the converter tank. The clean-up loop is located downstream from the heat 3-10 i

i I

exchanger to maintain a low coolant temperature so as to protect the temperature-sensitive resin.

Two conductivity cells (one at the inlet and one at the outlet of the cican-up loop) are used to monitor the conductance of the primary coolant flowing through the clean-up loop.

This system can also be used to remove decay heat from the fission convener fuel if a high coolant temperature alarm is accived. It is not anticipated that the system will need to be used for shut down heat removal. The cleanup system can also be used to adjust the pH of the coolant should that be necessary.

3.1.1.5 Make-Up Coolant System The coolant level in 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 Sampling 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. If the coolant in use is D20, a tritium analysis shall also be performed. The frequencies for these analyses are monthly for gross -y, pH, and quarterly for tritium. The latter is required only if D 2O is used as the primary coolant.

These are the frequencies followed for the MITR.

The purpose of the cover gas sampling system is similar. It permits the sampling of radioactive gases and dissociated gases (i.e., H2 and D2).

3.1.2 Secondary Coolant System The secondary coolant system consists of piping and valves. The secondasy 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 3-11 i

l l

l equipment room. The flow is then separated to provide secondary coolant to the MITR's O

components and the fission converter. 'Ihe 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.

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. (Note: 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, 250 kW maximum in this case.)

In the event that the fission convener coolant is D2 0, provision has been made to preclude the possibility of tritium release in the event of a heat exchanger leak. Under normal operation, the fission converter primary system will operate at a pressure that is about 10 psi lower than that of the secondary system. Should an overpressure event occur, the burst disk will open before the primary pressure exceeds that of the secondary. The MITR-II secondary water radiation monitor will be operating and the secondary water will l be sampled daily for tritium content. Any leakage to the fission converter primary system will be seen as a level or a conductivity increase. Irak detection will be present on and around the fission convener piping and will alarm in the reactor control room if a leak is l 1

j detected. This will be in place regardless of which coolant is used. It is planned that the I secondary flow will continue even if the fission convener is in a shutdown condition. l Therefore, the secondary pressure will remain higher than the primary pressure. If this is not the case (maintenance activity to the secondary system, for example), the fission converter beat exchanger will be isolated and vented.

3.1.3 hiaterial Selection Material selection for the fission converter is based on structural requirements, nuclear properties, availability, and cost.

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 H 2O and D O 2 coolant, except for small non-corrosive components such as gaskets, filters, and valve diaphragms.

4 3-12

/N

) 3.2 Basis for Thermal-llydraulic Desien The basis for the fission convener's thermal-hydraulic design is that, under conditions of forced convection,its primary coolant system removes 300 kW (50 kW more than the maximum fission converter power for 10 MW reactor power, fresh fuel and light water coolant) of heat from the fission convener fuel and transfers it to the secondary coolant system without the onset of nucleate boiling. Another design feature is that the system can operate at low power with natural circulation without exceeding the coolant temperature limit. The requirement for natural circulation operation is to facilitate activities such as Dux 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.

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 fission converter primary coolant system acts as a barrier against the escape of fission products to either the reactor building or to the

/%

Q secondary coolant system.

3.3 Computational Method for the Fission Converter Thermal Hydraulic Desien A computer program (MULCII-FC) was written to model the primary and secondary coolant systems of the Hssion 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 fimiting safety system settings of the fission convener. 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.

3.3.1 Methodolocv o Both the fission converter's primary and secondary coolant systems are modeled in ly the MULCll-FC code. Each component of the coolant system is modeled as a control 3-13

1 volume. The energy equations for the control volumes are then solved simultaneously to Q obtain the temperatures associated with each control volume. The fueled region is fuuher divided into a hot channel and average channels. For the calculations described here, each channel consisted of ten axial nodes. Fuel and coolant temperatures are solved for each node using energy conservation equations. For forced convection, the flow rate is taken as a constant. For operation with natural circulation, the pressure drop equation is solved to obtain 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.

Safety limits (SL) and limiting safety system settings (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 channelis the coolant channel that produces the highest power and has the lowest primary flow rate among all the coolant channels. To calculate the SL 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 temperatums. Appropriate i

correlations are then used to determine if the limits are exceeded.

3.3.2 Major Correlations Use d in MULCH-FC N)

Onset of nucleate boiling (ONB), also called incipient boiling, defines the condition where bubbles first start to form on the heated surface. Because most of the liquid is still subcooled, the bubbles do not detach but grow and collapse while attached to the wall.

Onset of significant voiding (OSV) describes the condition where the bubbles grow larger on the heated surface and detach regularly.

3.3.2.1 Correlation for Oncet of Nucleate Boiling (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.

3-14 I

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

i I

I:

l' l

<- - " 0.463 P0.0234 Q Tclad, ONB = Tsat + 0.556 1082 P l.156 (3.1) ;

4 l where Tclad, ONB is the fuel clad temperature ( Cut which ONB occurs,

! - Tsat is the saturation temperature ( C),

q" is the local heat flux (W/m2 s), and P is the pressure (bar). I

, 1 l

3.3.2.2 Correlations for Onset of Sienificant Voidine (OSV)

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 Xeosv = - 0.0022 if Pe < 70000 (3.2)

Hrg kr _ {

and Xeosy = - 154 HG if Pe 2 70000 (3.3) where Xe osy is the thermal equilibrium quality at which OSV occurs, Xe equals (H-Hf)/Hrg, and H is the local liquid enthalpy. Other symbols are defm' ed as:

Pe GDep Cr is the Peclet number Pe = '

kr 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),

Db is the equivalent diameter (m),

Hrg is the difference between saturated gas and liquid enthalpies (J/kg)

Hrg = Hg - Hr , and i G is the mass flux (kg/m2 s),

l O

]

3-15 i

The second correlation was recently proposed by Kowalski et al. to predict OSV for coolant channels with finned surfaces [Ref. 3-4]. It is q" Pe o.18 Xc (3.4) osv = 0.0446 G H fs .

The equilibrium qualities predicted by the above two relationships are compared and the l

lesser of the two, which is the one that corresponds to a lower coolant temperature, is used i to check for OSV. This yields a conservative prediction of OSV using the MULCH-FC l code.

l 3.4 Fission Convener Onerating 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 i disparity factor will be measured during the initial startup test (Sections 8.4 and 8.6(Sa)). l If the measured values are less conservative than the ones used in the existing calculations, these calculations will be performed again and the operating condition or limits adjusted as necessary. The engineering hot channel factors are determined by statistically combining the uncenainties associated with design, calculation, and measurement. These uncenainties are govemed by the MITR quality assurance program as well as by testing and calibration procedures for the fission converter.

'.4.1 Power Deposition Factor (Fy and Nuclear Hot Channel Factor (F3 c)

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 st 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 100% in this analysis. That is, it is assumed that there is no energy escaping the fission converter tank.

3-16

1 i

1 1

The nuclear hot channel factor defines the ratio of the maximum power deposited in l Q the hottest fuel plate to the average power per fuel plate. The nuclear hot channel factor I used in the current analysis is derived from the radial neutron Dux distribution calculation of the fission converter using MCNP (Section 2.3.4). The nuclear hot channel factor used !

in the thermal hydraulic calculations is 1.53, which is obtained for the conditions that D2O coolant and fresh fuel elements are used in the fission 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.

Fpx F li c 51.0 x 1.53 = 1.53 (3.5) 3.4.2 Fueled Region Coolant Flow Factor (Fp and Channel Flow Disparity Factor (d3 The coolant flow factor is defined as the ratio of the fission convener primary coolant flow which actually cools the fueled region to the total flow. Ideally, the fueled g3 region should be designed so that 100% of the coolant flows through the fuel elements.

V However, pan of this flow bypasses the fueled region because of design / manufacturing tolerances such as clearances between the fuel elements. The MITR-II has a core coolant now factor of 0.921 which was determined experimentally during the reactor's initial startup test. The s. ne value is assumed in the current analysis for the fission converter.

This is a conservative assumption because the MITR-II has multiple paths which would allow bypass flow. 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 = (3.6) avg p where Wmin is the minimum flow rate measured in all the coolant channels and Wavs is the average flow rate in the coolant channels.

3-17

1 The flow distribution in the MITR-II core was measured during the reactor's initial startup test. The minimum flow throt.gh a fuel element is 93% of the average core flow rate [Ref. 3-6). 'Ihe flow distribution'within a fuel element has also been measured ,

experimentally using a dummy element. The ratio of the minimum channel flow rate to the ,

average channel flow rate within a fuel element is 0.929 [Ref. 3-6]. So, the worst-case l

channel flow disparity factor of the MITR-II can be calculated as:

dr = 0.93 x 0.929 = 0.864 (3.7) ,

j It is assumed in the current analysis that the fission converter channel flow disparity factor

1. is the same as that of the MITR II.'

The minimum flow in a fission converter coolant channel is then calculated in the thermal hydraulic analysis using the following equation: j l

Wmin = ', x Fr x dr (3.8) l O where WP, tot is the total primary coolant flow rate, and Ne is the number of coolant channels in the fission converter fueled region.

To ensure that the current analysis covers the possible operating conditions of the fission convener, the following condition needs to be satisfied:

Fr x dr 2 0.921 x 0.864 = 0.7% (3.9) 3.4.3 Engineering Hot Channel Factors l

l l

" The engineering hot channel factors account for possible deviations from nominal design specifications that may affect the thermal hydraulic calculation results. Specifically, i 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. 3-3 18 l

7,3 8]. . So, the statistical approach has been used in the current study. Table 3.2 is a summary of the engineering hot channel factors.

The engineering hot channel sub-factors considered in the current study include those for reactor power measurement, power density measurement / calculation, fuel density tolerances, flow channel tolerances, fuel meat eccentricity, and heat transfer coefficient prediction. Numerical values for these sub-factors were mostly adopted from the MITR-II SAR and reference should be made to the MITR-Il SAR for definitions and derivations of these sub-factors [Ref. 3-9].

A " vertical" approach is used in the current analysis to calculate the maximum fuel clad temperature and the maximum coolant 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+ Fu AT c (or H e,y = H;n + FHAH cif boiling occurs) (3.10)

O O

3-19 t

- Table 3.2 V )

Engineering Hot Channel Factors Used in Fission Converter -

Thermal-Hydraulic Calculations Enthalpy Rise )

Reactor power measurement 1.05 Power density measurement / calculation 1.10 Fuel density tolerances 1.026 Flow measurement 1.05 j Flow channel tolerances 1.089 l Eccentricity 1.001 Fu, Statistical 1.154 j Film Temperature Rise Reactor power measurement 1.05 m Power density measurement / calculation 1.10 i,

m, )- Fuel density tolerances 1.05 Flow channel tolerances 1.124 Eccentricity 1.003 Heat transfer coefGcient 120 !

FNr. 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 h[ein: The engineering hot channel factors are obtained by combining the sub-factors statistically using the equation F = 1 + (fj - 1)2" , where fj denotes sub-factors

[Ref. 3-8].

(~

(_>

3-20

1 i

n To calculate the maximum fuel clad temperature, use:

Tw, y = Tin+ Th AT c+ FATAT w (3.11) where Tin is the coolant channel inlet temperature, Hin is the coolant enthalpy at the channel inlet, ATc is the coolant temperature rise, AHe is the coolant enthalpy rise, ATw is the film temperature rise (temperature difference between coolant and clad),

FH is the engineering hot channel factor for enthalpy rise, FAT is the engineering hot channel factor for film temperature rise, Tc,M is the maximum coolant temperature because of design / manufacture deviations, He,M is the maximum coolant enthalpy because of design / manufacture deviations, and (qj Tw,M is the maximum fuel clad temperature because of design / manufacture deviations.

Notice that ATc (or AHe), and ATw me calculated for every axial node in the hot channel with the operating limits taken into account. Tc,g (or He,u) and Tw,u are then calculated and checked if the safety limits or the limiting safety system settings are exceeded.

3.5 Thermal Hydraulie Limits The fission convener 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. j i

Onset of nucleate boiling (ONB) is chosen as the criterion for the derivation of limiting safety system settings (LSSS). Onset of significant voiding (OSV) is chosen as l

the criterion for the derivation of safety limits (SL). ONB, also called incipient boiling, l l defines the condition where bubbles first start to form on the heated surface. Because most l

O j 3-21 1

l

l l

of the liquid is still subcooled, the bubbles do not detach but grow and collapse while attached to the wall. OSV describes the condition where the bubbles grow larger on the heated surface and detach regularly.

) 3.5.1 Definitions of Safety Limits and LimitiDgjafety System Settines The safety limits for the fission converter are established to maintain the integrity of l the fuel clad. Although aluminum melts at approximately 660 C (1200 F), it begins to '

soften significantly at about 450*C (842 F). The softening temperature is therefore used as ;

the criterion to guarantee the structural integrity of the fuel elements [Ref. 3-9]. Critical l heat Dux (CIIF) is normally used as the criterion of fuel overheating. However, because i the coolant flow path in the core is a multichannel design, there exists the possibility that now instabilities could occur before reaching CIIF limitations. If flow instability does occur first, it would have the effect oflowering the flow rate to the hot channel significantly I

and thus lowering the bumout limits. Flow instability is a complicated phenomenon and the changes in Dow rate are very difficult to predict. A conservative assumption to use the onset of flow instability (OFI) in the MITR-II SAR produced a safety limit for the l maximum steady-state power that is considerably below values based on applicable CIIF i correlations [Ref. 3-9]. Various correlations have been developed for the prediction of !

OFI, However, the effect of an axial heat flux distribution is not included in these l correlations. OSV, on the other hand, can be more accurately predicted for various heat Oux distributions. Also it has been observed experimentally that OSV occurs before OFI

[Ref. 3-10]. Hence, OSV is assumed as the criterion for the safety limits in the current j study. i The limiting safety system settings (LSSS) are established to allow a sufficient margin between the normal operating conditions ar.d the safety limits. Onset of nucleate ;

boiling (ONB)is chosen as the criterion for the LSSS derivation. This guarantees that i boiling will not occur anywhere in the fueled region as long as the limits are not exceeded. }

Specifically, the LSSS is set for: l l

Oneration with forced convection:

1. The maximum fission converter power (P),

2. The maximum steady-state average primary outlet temperature (Tout),

l 3. The minimum primary flow rate (Wp), and

4. The minimum coolant level in the fission converter tank (H).

3-22

l 1

Oneration with natural circulation:

1. The maximum Dssion converter power (P),

2. The maximum Gssion converter tank mixing temperature (Tmix), and

3. The minimum coolant level in the Dssion converter tank (H).

Thus, the nssion converter can operate with all four (for forced convection) or three (for natural circulation) parameters simultaneously approaching their limits without there being boiling in the fueled region.

1 l

3.5.2 Derivation of the Safety Limits The safety limits for the Hssion converter are calculated based on the conservative assumptions that:

1. FpxFilC = 1.53 for the hot channel (Section 3.4.1),

1

2. The hot channel has the minimum flow rate, that is, Ffx dr = 0.796 l f

m (Section 3.4.2), and f, '

3. Tie 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 MULCH-FC code which is described in Section 3.3.

The safety limits are calculated for the Ussion convener with either ten or eleven fuel elements. It was found that the difference in the safety limits for the cases with ten and 11 fuel elements is very small for forced convection operation. Therefore, the proposed safety limits can be used for either ten or eleven fuel elements for forced convection operation. For natural convection, the difference is significant and therefore the proposed safety limits can only be used for eleven fuel elements.

Figure 3.6 shows the results of the fission converter safety limits calculation 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 the (O fission converter powerand the average primary outlet temperature for given primary flow 3-23

l rates to find the points corresponding to onset of significant voiding (OSV). The curves show the combinations of fission convener power (P), primary flow rate (Wp ), and average primary outlet temperature (Tout) for a cenain coolant height (H) at which OSV l

may occur in the hot channel. Points to the left of the curve represent operating conditions l at which OSV will not occur as long as the operating limits are satisfied. I Figure 3.7 shows the results of the fission convener safety limits calculation for natural circulation. The safety limits were determined for a coolant level of 2.4 m, which is l

the top of the downcomers. Iterations were made of the fission converter power and the j fission convener 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 convener 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. ;

O l I

)

l l

O i 3-24

i O

9 .... .... .... .... ....

8 ' H=2.6 m

. H=2.1 m P (kW) : H=1.6 m ~

Wp (gpm) . > .

6 - -

l

'\ .

i l -

, 1 4 .... ....

60 65 70 75 80 85 Tout ( C)

Figure 3.6 Fission Convener Safety I.imits for Forced Convection l' (for either ten or eleven fuel elements) i I

l d

1 iO 3-25 4

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

1 i

l l

l l l

1 i

t./ ;

I l

40 ,,, ,,,, ,,,, ,,,,, ,,,, ,,,, ,,,, ,,,, )

l

. l

' 1 35 H = 2.4 m '-

30 r

F a 25

, v .

A .

f~ 20 l.( : :

15 t

10 ' '

( 40 45 50 55 60 65 70 75 .80 Tmix( C) l l

I l

- Figure 3.7 Fission Converter Safety Limits for Natural Convection l

(for eleven fuel elements only) i i

I i

)-

f l 3-26

3.5.3 Derivation of the Limiting Safety System Settings O The limiting safety system settings (LSSS) calculations for the fission converter am based on the same assumptions that were made to obtain the safety limits (Section 3.5.2).

The calculations were also performed using the MULCH-FC code.

The LSSS are calculated for the fission converter with either ten or eleven fuel elements. It was found that the difference in the LSSS for the cases with ten and cleven fuel elements is very small for forced convection operation. Therefore, the proposed LSSS can be used for either ten or eleven fuel elements for forced convection operation. For natural convection, the difference is significant and therefore the proposed LSSS can only be used for eleven fuel elements.

Figure 3.8 shows the results of the calculation of the fission converter LSSS for forced convection for either ten or eleven fuel elements. The limiting safety system settings were determined for a primary coolant flow rate of 45 gpm and a constant fission convener 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 convener power (P) and steady-state primary outlet temperature (Tout) for which ONB may 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.

Because the fission converter's maximum predicted power is 250 kW, the i following limiting safety system settings are chosen for the fission convener with forced convection: ,

Variable Limiting Safety System Setting P 300 kW (max)

Wp 45 gpm (min)

Tout 60 C(max)

H 2.1 m above top of fuel elements (min)

O 3-27

1 Notice that the LSSS temperature calculated for 300 kW and 45 gpm is 63 C (see Figure

[U ,) 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 flow rate are low. Hence, the axial coolant temperature rise exceeds the film temperature rise (radial direction). This difference is exacerbated when the engineering factors are applied. The net effect of the dominance of the axial coolant temperature rise is that the maximum coolant temperature and the maximum fuel clad temperature simultaneously approach the saturation temperature. It is worth noting that ONB does I l

always occur before OSV and, if were not for the application of the engineering factors, this difference would be apparent for the above calculation. Therefore, a 5 C margin is added to establish the LSSS curve to allow adequate response time for appropriate actions.

The resulting LSSS cmve is shown in Figure 3.9. This added margin corresponds to about six minutes of heat up in the mixing area with the fission converter power at 20 kW.

The following limiting safety system settings are chosen for the fission converter with natural circulation:

(

QJ Vnriable Limitine Safety System Setting P 20 kW (max)

Tmix 60 C(max)

H 2.4 m above top of fuel elements (min)

Notice that the LSSS temperature corresponding to 20 kW at natural circulation is 63 C (see Figure 3.9).

Analyses for both forced convection and natural convection have been performed which show that the margins are adequate. Namely, the margins are sufficient so that automatic protective actions will ccrrect an abnormal situation before a SL is reached.

D 3-28

l i

O 1 i

)

450 .... .... .... .... .. , ,.,,

1 400 H = 2.1 m .l Wp = 45 gpm .

350 -h '

g -

e- : -

A .

~

3M O -

V -

250 %

. 1 200 ' '- - - .... ....'

45 50 55 60 65 70 75 Tout ( C)

Figure 3.8 Fission Converter Limiting Safety System Settings for Oneration with Forced Convection (for either ten or eleven fuel elements)

O o

3-29 l

l l

i

\

s i

i 1

40 ,,, ,,,,, ,,,,, ,,,, , , , , ,,,, ,,,, ,,,,

i 1

35 +-

[ H = 2.4 m -

30 p

.w 25 0, .

(~'

20 4

N .

15 l

i 10 35 40 45 50 55 60 65 70 75 i l

Tmix( C) ;

i Figure 3.9 Fission Converter LSSS for Operation with Natural Convection (for eleven fuel elements only)

O 3-30

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

Appendix 3.1 - Description of the MULCH-FC Code O A multi-Channel analysis (MULCH-II) 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 emphasis 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-II code has been benchmarked using steady-state and transient experimental data for the MITR-II [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-II's initial startup. Calculations of onset of Dow instability compared satisfactorily with correlations derived from experimental data.

The MULCH-II code was modified for the fission converter to evaluate possible thermal-hydraulic design options. The new code is designated as the MULCH-FC code.

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, l

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 3-31

..-. . .. - - - - - - . = _ _ -. - . .. -.-.. --_ - . - -

l

5. A two-phase flow model is included that covers the thermal-hydraulic l t

(]s conditions from onset of nucleate boiling to bulk boiling, and

6. Assoned benchmarked correlations for rectangular coolant channels under low pressure conditions are used in the code.

i 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. l t

l l

2. To establish safety limits and limiting safety system settings for both forced ;

l l convection and natural circulation.

i l

l

3. To analyze the transient resulting from the loss of one of two fission converter

[

primary pumps. i O l l

l I

i l l l l I- 3-32 i .4 , , - _ - . , _ ,

Appendix 3.2 - Dose Rate Calculation without Top Shield Lid O Calculations were performed using MICROSIIIELD to estimate the dose rate for fission converter operation without the top shield lid. The MICROSIIIELD code is user-frien'dly 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 source can also be supplied by the user.

In the dose rate calculations, the fission convener fueled region is modeled as a 6.5 cm x 72 cm x 66 cm rectangular aluminum block. Coolant above the fueled region is 2.4 m, which is the propose.d 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 associated yields are derived from the fission prompt gamma spectmm.

The calculated result is 560 mR/h at the coolant surfsee for fission convener power of 25 kW (or 450 mR/h for 20 kW). This dose rate is 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 0 -ith =

0 3-33

o T References (G

[3-1] File Calculation (Fuel Restraint)

[3-2] Y. Sudo et al., Experimental Study of Incipient Nucleate Boiling in Narrow Venical Rectangular Channel Simulating Subchannel of Upgraded JRR-3, J. of Nuclear Science and Technology,23[1], Jan.1986.

[3-3] P. Saha and N. Zuber, Point of Net Vapor Generation and Vapor Void Fraction in Subcooled Boiling, Proc. of Fifth International Heat 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.

[3-5] File Calculation (Power Deposition in Fission Convener Tank - MCNP Calculation)

[3-6] G.C. Allen, The Reactor Engineering of the MITR-il Construction and Startup, ,

PhD Thesis, Nuclear Engineering Department, MIT,1976.

[3-7) J.M. Rust, Nuclear Power Plant Engineering, Haralson Publishing Co.,1979.

[3-8] N.E. Todreas and MS Kazimi, Nuclear Systems II Elements of 7'.cmml O Hydraulic Design, Hemisphere Publishing Corp.,1990.

\)

~

[3-9] MITR Safety Analysis Report, MITR Staff,1970.

[3-10] T. Dougherty et al., Flow Instability in Venical Channels, HTD-Vol.159, Phase Change Heat Transfer, ASME 1991.

[3-11] L.-W. Hu and J.A. Bemard, Development and Benchmarking of a Thermal-Hydraulics code for the MIT Nuclear Research Reactor, Joint Intemational Conference on Mathematical Methods and Supercomputing for Nuclear Applications, 6-10 Oct.1997, Saratoga, NY.

[3-12] File Memo (ASME Code Requirements for the Fission Converter Tank).

[3-13] File Calculation (Fission Converter Safety Limits and Limiting Safety System Settings)

[3-14] File Calculation (Thermal Hycraulic Limits for Ten Fuel Elements Operations) i

\

O l 3-34 l

l l

l

, 4. Fission Converter Shutter and Medical Therany Room Design 4.1 General Description of Shutter Design The shutters for the fission converter beam consist of a converter control shutter I (CCS), a water shutter in the collimator, and a mechanical shutter at the end of the collimator. These shutters fulfill several functions. De CCS, which is located between the reactor core and the fission convener plate, controls the neutron flux 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, CCS, and fission converter medical therapy room door conform to 4 the criteria for those in use for the existing medical room. l 4.1.1 Converter Control Shutter The converter control shutter (CCS), which is located between the reactor core and

(" the fission converter plate, reduces thermal neutrons incident on the fission converter plate to less than 1% [Ref. 4-1]. The CCS consists of 1/16" aluminum followed by 20 mils of 2

cadmium and 1/4" of Boralyn@ or 1/2" of Boral (either is equivalent to 50 mg/cm og toB).

Figures 4.1 and 4.2 show the fission convener power, as well as the epithermal neutron flux and the fast neutron dose rate at the patient position as a function of CCS height above the fuel centerline.

Two cables, attached at the top corners of the CCS, connect the shutter to an electric l motor. A frame welded on the fission converter tank limits the movement of the CCS to the venical direction only. Drive-in and drive-out limit switches are used to stop the motor and to prevent the CCS from being driven beyond its physical limitations. Limit switches are also used to indicate if the CCS is either fully closed or fully open. In the event of a failure of the CCS to close fully, the reactor operator can lower reactor power or scram the reactor to shut down the fission convener.

l The reactivity wonh of the CCS will be determined during the initial startup testing l of the fission converter and annually thereafter. The reactivity worth of the CCS is lp expected to be less than that of a movable experiment (0.2%AK/K).

(

4-1

. x x. . . a= ,. =_ x . ,, : : . .. . a. .- .

- -. : .- l i

100 1.4e+10

d. Power

--9-- Flux l- 1.2e+10 r- 80 --

h a

2- 1.0e+10 l u - -

l

'o -

l h 60 -- l T

A .

-- 8.0e+9 u - E o -

. O Y - -

b o ~

~

- .- 6.0e+9 , .g I o 40 -- . o u . - c -

.9 -

-- 4.0e+9

\

O.Q m 20 --

i l

l

- 2.0e+9 0 *' -'

=-l -l -l -l -l O.De+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 EnithermalNeutron Flux at the Patient Position as a Function of CCS Height. (MITR power is at 5 MW. heavy water and i

spent fuel elements are used in the fission convener.) [Ref. 4-1] i i

l O

4-2

t l

l 100 .... ....,,,,,,,...,....i.. ,,...,i.... 12

^

-- Power

--4>-- Fast Neutron Dose Rate -

-- 10 r- 80 -- -

)w .

w -

g

-8 y.

a 60 -- -

.5 4 -

5 u - -

o -

-6 %,

t .

O y .

?6 .

c 40 -- -

o o O ~

-4 c -

o .

()

V 0 -

E 20 --

-2 o ..a . .i....i....i....i....i.... O I I I i i I 4

-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 CCS Height. MilTR nower is at 5 MW. heavy water and spent fuel elements are used in the fission convener 3 [Ref. 4-1]

, b' O s

l 4-3

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

iq Q . lt is desired that the time for the CCS to go from its full-open to full-closed j position be less than 60 seconds. This figure has no safety significance. However, it was !

- shown in Section 6.4 that upon a loss of primary flow and in'the concurrent absence of a l l reactor scram, no fuel damage because of overheating would occur if the CCS took 60 l seconds to close. !

l- i 4.1.2 Collimator Shutter Desmn l

i Shutters in the collimator are needed to ensure the safety of personnel workmg m 1

,. the medical therapy room and to control the radiation dose to patients. A two-part shutter l design will be used: a water shutter and a fast-acting mechanical shutter. The latter is located at the end of the collimator. ,

l 4.1.2.1 Water Shutter l- t l

The water shutter located in the collimator provides neutron and gamma 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 l weight). There is no reactivity effect to the fission convener system if the water contains ,

dissolved 10B. !

u

~At the stan of an irradiation, the water shutter is full and the water tank located above the shutter contains sufficient empty volume to accommodate the contents of the l

water shutter. Initially, the remotely 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 remotely L operated valve is closed and the contents of the shutter are pumped to the water tank. To l close the shutter, the remotely operated valve is opened and water flows by gravity from the tanki Conductivity, or other suitable probes, will be used to indicate the water shutter L position (full closed or full' open) at the medical control panel. When the water i

!O l r

4-4 l

7.~g Fill / Vent =

X Burst Disk

(

D#

) Cover Gos (outo fill / vent system)

X

, X I H Low-Level Sensor rion Switch l l Ve it Line V nt un s Level Couca Clean-up System Supply Tonk o 3 Flow Me er N0i" u -c><a : :

C><3-f Auto Volve (N.O.)

.4 Shutter Closed Sensor Medical Room f')

V Water Shutter tz N

-c><a

-O Shutter Open/ Pump Off Sensor X Droin V

Main Pump X

Cleon-up Pump Figure 4.3 Fission Converter Water Shutter System

.ID v

4-5

shutter is emptied, the tank will be filled with air or helium. Calculation has shc,wn that the p)

( 41Ar production is insignificant [Ref. 4-2]. It is desired that the normal opening and closing time be less than 120 seconds. However, there is no safety significance to this number. The 120 second figure is chosen solely for reasons of efficient operation.

Interlocks are provided to ensure that the water shutter closes automatically when the medical room door is open or the medical room control panel keyswitch is turned to the OFF position.

4.1.2.2 Mechanical Shutter A fast-acting mechanical shutter composed oflead 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 an electric motor connected to a ball screw. This motor will be supplied by an un-interruptable power supply, and will close the shutter automatically in the event of normal electrical power failure. Manual operation is also provided. Limit switches are used to indicate the mechanical shutter position (full closed or full open) at the medical Q control panel. Interlocks are provided to ensure that the mechanical shutter closes automatically when the medical room door is open or the medical room control panel keyswitch is turned to' the OFF position.

It is desired that the opening and closing time for this shutter be less than a few seconds. This 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 shutter cycling time is a matter of convenience. It is not necessary for either personnel or patient safety.

4.2 Shutter Controls Control stations are located outside the medical therapy room at the medical control panel, inside the medical therapy room, and in the reactor control room. Each station is equipped with those centrol functions that are commensurate with safe operation of that station. The controls at the fission converter medical room control panel will consist of open and close buttons with appropriate position indicators for the CCS, water, and p mechanical shutters. In addition, there will be a reactor minor scram button. The control V panel itself will be activated by means of a key switch. When the key is removed, these 4-6

l controls cannot be used. It will also be possible to close the mechanical shutter manually.

/O V The controls inside the fission converter medical therapy room will consist of close buttons for the CCS, 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 room control panel. The CCS can be opened only from the reactor control room by a licensed operator. In the event that the reactivity worth of the CCS is small and can be compensated using reactor automatic control, permission may be given to trained non-licensed personnel to open the CCS from the medical room control panel. This is done by tuming a key switch ON in the control room to energize the CCS OPEN button at the medical room control panel. The CCS can be closed from the reactor control room, the medical room control panel, and inside the medical room. The reactor control room will be supplied with indicator lights or the equivalent such as a scam panel alarm for the CCS, water, and mechanical shutters.

4.3 Medical Therany Room Design The new medical therapy room for the fission convener based epithermal neutron g beam will be located in the MITR-II'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 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months per week occupancy of the fission convener medical room. The latter is a conservative figure. All personnel, except the patient, will be clear of the room before any shutters are opened.

The medical room shielding is designed to reduce the radiation levels, from the medical beam, on the outside of the room to q l mrem /hr for all possible fission converter facility operating scenarios including MITR at 10 MW. Patients will be observed during irradiations by visual observation using TV camera (s) and a shielded viewing window.

This is the same as the observation of patients in the current medical room. A boron or lithium containing paint, plaster, or layer, e g., boral will be used to reduce wall activation p)

L.

in the medical room. This approach is used in the existing MITR medical room. A 4-7

permanent radiation monitoring system with readouts at the medical control panel outside

{) the irradiation room will provide information conceming ambient radiation levels in the medical room. A similar system is currently used in the MITR medical room in the reactor basement.

4.3.1 Medical Therapy Room Door Access to the therapy room will be provided by a shielded door. During irradiation, this shielded door at the entrance to the therapy room is closed. This door will either be manually operated or controlled by push buttons located both inside and outside the door.

If the shielded door is driven by a motor, the drive mechanism can be disconnected and the door can then be operated manually in an emergency. In the event that the bearings which support the door were to be damaged, the door could be opened manually by supporting it vertically 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 the water and mechanical shutters that control beam delivery unless the

~~3 medical therapy room's shielded door is closed. Furthermore, the shutters shall also close (V automatically when the door opens. If required by a particular experiment, this interlock may be bypassed for runs at low fission convener power or when adequate temporary shielding can be provided to allow personnel to work in the area without radiation hazard.

In all cases, sufficient waming will be afforded by a radiation warning lamp outside the .

door. l Indications of radiation levels and the position of the door will be displayed at the medical control panel. The same signals will be available in the reactor control room.

Communications will exist between both of these centers and the fission converter medical room. j 1

Beferences

[4-1] W.S. Kiger Ill, Neutronic Design of a Fission Converter-Based Epithermal Beam for Neutron Capture Therapy, Nuclear Engineer Thesis, MIT,1996.

gm [4-2] File Calculation (41Ar Calculation for the Water Shutter System).

d 4-8

, 5. Fission Converter Fuel Handline (v)

The fission convener will use the same type of fuel, either irradiated or new, as does the MIT Research Reactor. Many of the provisions that penain to the storage and handling of the MITR fuel can the.refore be applied without modification to the fission converter's fuel. An issue that is potentially unique to the fission convener is that of self-protection because the fission converter operates at a maximum of 250 kW and a low capacity factor. This and other related issues are 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 D 02 is used as the fission converter's coolant.

5.1 Fuel Element Security. Storace. and Ouality Assurance The provisions of the MITR security plan apply to the fuel used for the fission convener. 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 report. Fuel burnup l will be in accordance with MITR Technical Specification # 3.11(2e). Also applicable am o the requirements of the MITR quality assurance program for fuel.

5.2 Fuel Element Self-Protection ,

i The fission convener 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

~40 %) will remain self-protecting for at least a decade [Ref. 5-1]. Hence, the self-protection criterion will be met for the fission converter if MITR fuel with a significant burnup history is used. However, circumstances might arise 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: l l

(a) Fresh fuel elements will be pre-irradiated in the MITR core if they are 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 l clements in core,200 kW average power output per element) for 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months . This l . -w,

, is equivalent to a burnup of 3 x 104 kWh per element.

5-1 l

(b) The pre-irradiated element will then be placed in the fission convener. An element with a burnup of 3 x 104 kWh will remain self-protecting if the fission converter is operated at 80 kW for at Icast 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months (minimum bumup of 436 kWh for each fuel element) every month.

The actual sequence used to maintain a fission convener 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.

5.2.1 Use of Unirradiated Fue!

(

Fresh uninadiated 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 Fuel Element Removal 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 level above 100 kW for at least four days. This requirement can not be translated directly to the fission convener because of the different numbers of elements in the reactor core and in the fission converter. 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 belovi 50 kW for the four days prior to refueling.

(b) A maximum operating time of 4.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 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 limitation except that eleven fuel elements are involved as opposed to twenty-two or more. The second and third are based on an equivalent power history .

5-2 i

L

A study was conducted to calculate the fuel plate temperature during fuel element removal. It was assumed that the fission converter was operated continuously at its ,

maximum power of 250 kW until four days prior to 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 weH below the Al-6061 softening temperature of 450 C [Ref. 5-2].

i 5.4 Fuel Element Handling During normal refueling, depleted or panially depleted fuel elements may be removed from the fission convener and stored in one of the approved MITR fuel storage areas. Also, partly 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 spent fuel handling is the shielding of fission-product decay gamma rays. A transfer cask provides the necessary shielding during movements.

Any transfer ofirradiated fuel out of the converter tank will be made by lifting the fuel element into the specially adapted fuel transfer cask (Section 5.4.1). The fuel 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, adequate shielding will be provided between the transfer cask and the water in the converter 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.

\

5-3

5.4.1 Fuel Element Transfer Cask O 'Ihe 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 j single fuel element. It is normally 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.

5.5 Samples The fission converter is designed to be operated under forced convection with either 1 ten or eleven fuel elements. If ten elements are used, then one position will be available for sample irradiation. Sample assemblies that might be introduced into the fueled region of the converter are subject to TS# 6.6.4.4, which refers to TS# 6.6.2.l(4) and TS# 5.2. The former states the requirements for sample assemblies and the latter specifies the criteria for MITR in-core sample assembly design. The samples themselves will be subject to review in accordance with TS# 6.1.

O References

[5-1] File Calculation (Fission Converter Fuel Dose Rate Calculation)

[5-2] File Calculation (Fission Converter Refueling Requirement)

[5-3] Fi:e Calculation (Maximum Fuel Temperature During Complete Loss of Coolant) i!

O 5-4

3

6. Safety Analysis g)

(

U 6.1 Maximum Hvoothetical Accident The maximum hypothetical accident (MHA) for the assion converter is that a maximum of five coolant channels are blocked by a foreign object, which causes four fuel plates to melt. This is the same scenario used in the MITR-II SAR.

A review of the fission converter primary coolant system indicates that Gow blocky in a fuel element is very unlikely. The coolant flows from the converter tank, through small diameter tubes in th

heat exchanger, and then through two pumps before entering the downcomers and flowing up through the fuel elements. Any foreign material that might start to circulate in the system would have to be small enough to pass through the heat exchangr 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 Dow could pick up V this object and cause it to block the entrance of several fuel element coolant channels. In order for this to happen, the object would have to fall through the lower grid plate when a fuel element was removed. The size of the opening in the lower grid plate would restrict the dimensions of the object to those of a fuel element nozzle. If the 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 Dow channels. 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

, individual located at the exclusion area boundary of the reactor during the first two hours of MITR-II's MHA [Ref. 6-3]. The MHA for MITR-II is postulated to be a coolant flow 6-1 t

blockage in the hottest channels of the center fuel element. This will lead to an overheating o i f

w) of a maximum of four fuel plates. The study consenatively assumed that the entire active ponions of all four plates melt completely and release their inventory of fission products to the primary coolant.

Escape of fission products to the containment because of fuel melting is mstricted by the pool of water above the core. The following approaches were used to evaluate the major release paths to the exclusion area 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. Tl dose from leakage was calculated using a standard Gaussian diffusion model and local meteorological data.

2. Gamma radiation reaching the boundary area by direct penetration of the containment shell was determined using standard shielding calculations. A Compton scattering model 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 tmck airlock which is the largest containment penetration.

For the MITR-II accident analysis, the fission products in the fuel at the time of the accident are assumed to be in equilibrium for a steady-state mactor power of 5 MW. This assumption is conservative for the MITR 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. Hence, even if the fission convener facility were in continuous use, the fission convenerfuel region would be at power for at most a few hours per day. Given this short operating cycle, most of the imponant fission products, including I-131, will not reach saturation. Another important difference to note between the fission convener and the MITR-II is that the fissien converter operates at a maximum power of 125 kW. At this power level, each fuel element generates 11.4 kW, compared to l 208 kW per fuel element for the MITR-ll core. (Note: For MITR at 10 MW, the figures are 250 kW,22.7 kW, and 416 kW respectively.)

i n 5 V

6-2 E

The calculated results for the MITR-II MHA 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 fission product equilibrium, the estimated maximum extemal doses to an individual located at the nearest point of public occupancy during the first two hours of the MITR-II MHA are 379 mrad at 8 m (back fence) and 595 mrad at 21 m (front fence) to the whole body, and the intemal dose is 118 mrad to the thyroid (see Table 6.1). Assuming the fission convener

. plate has been operating at 250 kW continuously, the whole body dose during the first two hours from the fission converter is (250 kW/5000 kW)(24/11 elements) x 595 mrad = 65 ,

mrad at 21 m and 41 mrad at 8 m. The thyroid dose from containment leakage is (250 l kW/5000 kW)(24/11 elements) x 118 mrad = 13 mrad. l Table 6.1 Estimated Doses from all Modes of Radiation Release During a l MITR-II Maximum Hvoothetical Accident [Ref. 6-3] j

- Component of the Dose Dose (mrad)(c) 8 m (a) 21 m @)

Whole body:

Containment 12akage 27 27 )

Steel Dome Penetration 3 27 Shadow Shield Penetration 43 21 ;

Air Scattering 114 147 l.

i Stee1 Scattering 192 373 l Total 379 595 Thyroid:

Containment leakage 118 118 j

[

f (a) Boundary of restricted area ,

l @) Nearest point of.public occupancy (c) Calculation assumes that duration of release is two hours. '

l P

6.2 Insertion of Excess Reactivity

\ .

! Two scenarios have been considered. These were insertion of excess reactivity in L the fission converter itself and a step reactivity addition to the MITR for which the consequences propagated to the fission convener. Neither scenario was judged to have

! safety consequences.

l 6-3 L

i r

l

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

m O O U 0.6 -

0.5 -

Total Whole-body nose ,

0. 4 -

n S

O.3-Steel Scattering Dose i S l m t E D. 2 -

Air Scattering Dose 0.1 _

Leakage Iletn Dose Steel Penetration Dose l g Shadow Shield Dose _

y Lenkage Gamma Dose s

O i i e i ;

l 10 20 30 l 40 50 Minimum Exclusion Albany St. Maximum Exclusion

Area Distance Arca Distance L Distance f rom Containment (M)

Figure 6.1 Total Two Hour Whole-body Dose Results for MHA in MITR-H [Ref. 6-3] !

t

{

, Insertion of excess mactivity into the fission convener is not considered to be a

) credible accident because the fission convener is a highly subcritical system. Analysis has l shown that the keg for the fission convener is 0.67 or lower, depending on the type of

coolant used and the amount of U-235 in the fuel.

An analysis was performed to estimate the maximum fuel temperature of the fission l

converter if a step mactivity transient occurred in the MITR. It was assumed that the mactivity inserted in the MITR in this accident resulted in a maximum MITR fuel temperature of 450 C (softenir.g temperature of aluminum). The analysis showed that the maximum fuel temperature in the fission converter would be 179 C during this transient.

This is because that the fission convener power density is only about 6% of that of the MITR [6-8]. The most limiting initial conditions for a step reactivity insertion transient are low power and low flow rate. Therefore, the initial conditions used for this analysis were low power and low flow and thus the coolant conditions in the MITR and the fission convener (flow rate and temperature) had very little effect on the calculated fuel temperatures. This analysis showed that if the MITR was not damaged by a reactivity tranisent then neither was the fission converter. It is believed that this is a general C>')

6.3 Loss of Priman Coolant There are two initiating events that could result in a loss of primary coolant accident for the fission converter - primary pipe breakage and fission converter tank failure. The fission converter's safety system provides cenain protections against this type of accident.

In the case of low coolant level, the converter control shutter (CCS) will automatically close. In the case of low primary coolant flow, a converter control shutter closure is automatically initiated. If the CCS 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 converter tank. No credible scenario could be found which would result in rapid and complete coolant loss. Therefore the following l

should be considered a bounding analysis. The following conservative assumptions were

! made in this analysis:

p b

6-5

l

1. The Dssion convener 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 region.

2. The primary coolant in the convener tank is completely lost at the time of fission converter shutdown.

3. Convection and conduction to air and surrounding materials am neglected. The i 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.

l l

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. q" is the heat flux because of radiation heat transfer to the graphite area and q" 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 Drst hour to 383 C, which is 67 C O lower than the fuel softening temperature (450 C). It then decreases because of decreasing decay power [Ref. 6-4].

- l 9Y =0 1 End Plate aL l C adding "i 'i y Ak Y \

, Fuel Meat 9Y =0 yy Figure 6.2 Heat Transfer Paths from a Fuel Plate During a Complete Loss of Coolant Accident O

6-6

1 1

0 l

i P

^

400 - i 350 -

G 300 l

1 2 !

2 250 2

E 200 o

O I m

150 Fission Converter Power 250 kW i

2 100 -

Hot Channel Factor 1.53 i Emissivity 0.4 l l 50 '

0

_0

_ 2 4 6 8 10 Time (hr) l i

Figure 6.3 Maximum Fuel Temoerature During a Complete Loss of Coolant Accident

O I

6-7

.- , -,-m - _ -w..a

l 6.4 Loss of Primary Coolant Flow i

q

V The fission converter is designed so that low primary flow (less than 50 gpm) will automatically initiate both a reactor shutdown and a converter control shutter (CCS) closure. Either one of these actions will shut the fission converter down. However, the former is the primary mechanism whereby an immediate action is ensured in the event of a fission converter loss of flow.

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 will be small and no boiling will occur in the fuel region [Ref. 6-5].

The following analysis is included as an item of information. Specifically it is shown that closure of the CCS in the absence of a reactor scram will provide adequate I

protection.

l The analysis was made of the fuel and coolant temperatures if only the CCS V 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 powc 4 250 kW.

2. The average primary coolant outle. temperature is 55 C.

3. Steady-state primary flow rate is 100 gpm and it undergoes a step change to zero.

4. CCS starts to close after a one-second instrument delay time. Fission converter power then decreases according to the profile shown in Figure 4.1. It is assumed l that the CCS closes in 60 seconds at a constant speed.

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 materials are neglected.

The calculation was made for the hot channel where both the coolant and fuel temperatures are the highest in the fuel 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 as a function of time. It i is assumed that there is a prompt jump of fuel temperature from 69 C, which corresponds t

6-8

to the hot channel coolant outlet temperature 64 *C, to 95 *C because of the sudden decrease in forced convective heat transfer. The maximum fuel plate temperature during this '

transient is 139 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 t temperature holds at saturation temperature 106 C after 8 seconds into the transient because vapor starts to form. The coolant mass in the coolant channel then starts to decrease because the channel volume is constant (assuming no counter-current flow).

140

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~ ~'~1

~

1 lE 100 i ~I~~I I~I

I~l ..

~

}- ~f a' -

90 0 10 20 30 40 50 60 Time (s)

Figure 6.4 Calculated Hot Channel Fuel Temnerature During a Loss of Primary Elow Transient (CCS closure. no scram)

O 6-9

r 120

(

j j

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~1 g} }_.._7_. .}~j::j-:}~j: .~..j~} ._

3-j~}. 7 ... }~~:} .

0 10 20 30 40 50 60 Time (s)

Figure 6.5 Calculated Hot Channel Coolant Temnerature During a Loss of Primary Flow Transient (CCS Closure. No Scram)

- 0.05

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0 ' ' ' ' ' ' '

O 10 20 30 40 50 60 Time (s)

Figure 6.6 Calculated Hot Channel Coolant Mass in Hot Channel During a Loss of Primary Flow Transient (CCS Closure. No Scram) 6-10

!. l l '

l l

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0 10 20 30 40 50 60 !

l l Time (s)

\

l l

! Figure 6.7 D1culated Hot Channel Coolant Enthainy During a Loss of Primary Flow Transient (CCS Closure. No Scram) 1 ' '

1 -11 i 1 i

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0 10 20 30 40 50 60 l

Time (s) 1 Figure 6.8 Calculated Hot Channel Coolant Static Ouality During a Loss of Primarv e

\ Flow Transient (CCS Closure. No Scram) 6 11

I 6.5 Loss of One of Two Primary Pumps

/#

(

The fission converter primary coolant system is designed to u3e two primary pumps 1

operating in parallel during normal operation. It is intended that either pump be sufficient l

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 excessive coolant temperature. The main issue involved in this analysis was the time delay associated with the mixing of the coolant in bulk in the converter tank. The primary outlet temperature sensor will measure a lower temperature than the average coolant outlet temperature during this transient because of this mixing effect. l l

The accident analysis is made under the following assumptions.

1. The fission convener power is 250 kW.

2. The fission converter primary flow rate undergoes a step change from 100 gpm to 50 gpm (scram setpoint). (Nots: The design flow rate for one pump operation is f 5 gpm or higher.)

3. The initial steady-state primary coolant outlet temperature is 50 C.

)

4. Instantaneous mixing is assumed in the fission convener tank region. This was shown to be a conservative assumption [Ref. 6-7].

Figure 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 the average primary coolant outlet temperature and the mixing area 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 convener primary coolant system approaches a new ;

steady-state operating condition about 300 seconds after initiation of the transient.

O 6-12 3

. _._ m. _ _ _ _ . _ _. _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _

65 ,, . 4 . , , ,

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q. [ yp ._[ly j y._ p' ._

"l~ Average Prim $rh Cooiant Outlet Tdmperatt!re 2 h_. 3 p q t ._.

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50 V Mixing Area Coolant Temperature ~T~1 ~

_.. _.!. ! !. ! .l J ! !! I i 1 J J J _L . .

I i j

, 7 3 45 ' ' ' '

0 50 100 150 200 250 300 Time (s)

Figure 6.9 Calculated Average Fuel Outlet Temnerature and Mixing Area Temnerature During the Loss of One _ Pump Transient O

44 ' ' '

l i i1' 1

~~

_q.T_. _.g

~

qT.1_ i t r 42 . .

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I I I II I I 36 ' ' ' ' '

0 50 100 150 200 250 300 Time (s)

O Figure 6.10 Calculated Fuel Inlet Temperature During the Loss of One Pump Transient 6-13

6.6 Loss of Off-Site Electric Power p)

The MITR-Il scrams automatically upon loss of electric power. This in turn would cause a shutdown of the fission converter. Hence, temperature elevations in the fuel and coolant would be small and no boiling would occur in the fuel region [Ref. 6-5]. The water 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.

Emergency power is supplied by the MITR-II emergency power system for selected equipment and instruments such as :

1. Fission convener medical therapy room radiation monitor.

2. Intercom between the fission converter 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.

v ;

5. Safety channels listed as follows .

Primary coolant outlet temperatum !

= Coolantlevel l

l 1

6.7 Loss of Heat Sink The MITR-II uses two secondary pumps. The reactor will shut down automatically upon low secondary flow (450 gpm for either pump). If the cooling tower heat transfer l capacity is lost and the fission convener and the reactor continue to operate, elevated reactor primary coolant outlet temperatures will cause the reactor to scram automatically. High temperature at the fission convener primary coolant outlet will cause the convener control shutter (CCS) 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.

,x

( )

v 6-14

1 6.8 Mishandling or Malfunction of Fuel O 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 als'o be used for the fission converter. Also, the dropping of a fuel element should not result in a radiation release because

1. the element would fall through water, which would cushion any impact and,

2. the fuel is a cermet which would limit the release of fission products should the clad be scratched or otherwise damaged.

It should be noted that all MITR fuel handling tools have a safety lock feature that prevent this type of accident.

6.9 Exneriment Malfunction The fission convener is not designed to accommodate experiments within the fission converter tank. Thus, experiment malfunction is not considered a credible accident.

O 6.10 Natural Disturbances Safety analyses for natural disturbances for the MITR-II apply for the fission converter. Most of the following is from the MITR-II SAR.

6.10.1 Earthquake A scismic study of the Cambridge area is described in the Section 2.5 of the MITR-II 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 canhquake damage poses any hazard.

Funbermore, reactor shutdown is expected to occur in the event of a significant canhquake and therefore, would also shut down the fission converter.

O 6-15

i 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 grounded to the buried copper conductor. Consequently, lightning is not expected to affect the facility directly. However, an electrical power outage may occur.

During an electrical power outage, the reactor is automatically scrammed and this would I shut down the fission converter.

6.10.3 Severe Storm The reactor building conservatively conforms 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 alief system described in the MITR-II SAR.

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 reactor core tank and the fission convener tank mmain filled with coolant which would provide the required shutdown cooling.

t The control 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 l cause a power failure, the reactor is shut down until the storm has passed. ;

I O

6-16

i References )

l

[6-1] Dykes, J.W., et al., "A Summary of the 1962 Fuel Element Fission Break in the MTR", IDO-17064, February 1965.

[6-2] Tabor, W.H., " Fuel Plate Melting at the Oak Ridge Research Reactor", AN_S Transactions 8 Supplement,36 (July 1965).

[6-3] R. Mull, Exclusion Area Radiation Release During the MIT Reactor Design Basis Accident, MS Thesis, MIT,1983.

[6-4] File Calculation (Maximum Fuel Temperature During Complete Loss of Coolant)

[6-5] File Calculation (Fuel and Coolant Temperatum Calculation Before Reactor Automatically Shut Down)

[6-6] File Calculation (Fuel and Coolant Temperature Calculation During Loss of Fission Converter Priman Flow with Cadmium Curtain Closed in 60 seconds)

[6-7] ' L-W. Hu, Thermal Hydraulic Mixing Transients in the MIT Research Reactor Core Tank, Ph.D. Thesis, Nuclear Engineering Department, MIT, Feb.1996.

[6-8] File Calculation (Maximum Step Reactivity Insenion Accident)

O v

i e 1 l

1 o

"V 1

6-17 l

7. Instrumentation and Control System I'_ s V

The instrumentation and control system for the fission converter is summarized in this chapter. Controls and displays that are imponant for the safe operation of the fission convener am located 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 instrumentation that is not essential for safety.

7.1 Fission Converter Nuclear Instrumentation It is imponant to understand that the role of the nuclear instrumentation in a facility that is capable of achieving 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

()

7 is fixed by design of the converter and its coupling (i.e., convener control shutter open or closed) to an associated neutron source.

A fission convener's power can be measured via a calorimetric during conditions of thermal equilibrium and themafter should remain unchanged until fuel depletion becomes appreciable. The role of the nuclear instmmentation is, therefore, to confirm this power level when conditions of thermal equilibrium are not possible. Period measurements am not necessary. Studies of the fission converter's power show that fission convener power varies linearly with reactor power.

The fission converter will be equipped with one or more neutron-sensitive nuclear instruments. Atleast one such instrument should be operable before the fission converter is brought to power and utilized. Indication of fission convener neutronic power will be j provided in the reactor control room. In addition, this signal will cause closure of the I fission convener's converter control shutter (CCS) in an event of an overpower condition:

110% of converter design power which is 275 kW. It should be noted that the principal protection against fission convener 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 7-1 l

l I

l Research Reactor. In that event, the reactor safety system, which is independent of the (Q

g Ession converter, would shut the reactor down and hence also shut the fission converter

\

down.

In addition to the CCS closure signal that is generated at 110% of converter design power, an alarm will be provided at 110% of the fission converter's nominal operating power. This is provided because the 250 kW 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. The fission converter's nominal operating power depends on the given combination of MITR licensed power, fission converter coolant (H 2 O or D2 0), and U-235 content of the fission converter fuel. The nominal operating power of the fission converter 1 I

shall be estimated prior to initial use and confirmed during initial use.

7.2 Fission Converter Thermal-Hvdraulic Instrumentation Essential for Safety l

The parameters of concern are the outlet temperature of the coolant, the coolant flow l 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

) convener. Indication of the fission converter hot leg outlet temperature will be provided in the reactor control room. In addition, this signal will cause automatic closure of the fission converter's CCS in the event of an over-temperature condition (Setpoint: 55' C).

A conductance-type level probe or an equivalent device will be used to monitor coolant level within the fission convener tank. In addition, this signal will cause both closure of the fission convener's CCS 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 using 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 CCS and an automatic reactor scram in the event of a loss of flow if the fission convener is operated at a forced convection condition (Setpoint: 50 gpm).

O 7-2 i

7.3 Fission Converter Shutdown System n

The primary means for controlling Ession converter power is the converter control shutter (CCS). The convener control shutter (CCS) OPEN/CLOSE buttons will be located on the CCS control panel in the reactor control room. This panel is energized by an l

)

ON/OFF key switch. This key switch also enables the Ossion converter low coolant level I scram as well as the fission converter primary flow scram (for forced convection only).

Administrative controls will ensure proper possession of the key as with other keys associated with reactor operation. The CCS shall be opened from the control room by a l licensed operator. In the event that the reactivity worth of the CCS is small and can be l compensated using reactor automatic control, permission may be given to trained non-licensed personnel to open the CCS from the medical room control panel. This is enabled by turning a key switch on the same panel in the control room to the ON position in order to energize the CCS OPEN button on the medical room control panel. The authorized !

personnel will ask for the on-console operator's permission before operating the CCS. 1 CCS CLOSE buttons will be located also on the medical room control panel as well as inside the medical room. The backup to automatic CCS closure is a minor scram of the reactor. This could occur either automatically or manually:

n b i) Automatic Reactor Scram

a. Forced convection operation: The reactor will be automatically scrammed in 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 (> 275 kW) 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 CCS and scramming the reactor.

b. Natural convection operation: The reactor will be automatically l scrammed in the event of fission convener over-power (> 15 RW) and low primary coolant level. The need for a reactor scram for fission converter over-power during natural convection operation arises because the reactor O

Q safety system is normally set for its licensed power. This power 7-3 1

1

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

i corresponds to a higher fission convener power than is acceptable for natural convection operation. Therefore, the reactor scram corresponding to a fission converter power of 15 kW will provide a timely response to prevent over-heating of the fission converter fuel elements.

ii) Manual Reactor Scram A manual reactor minor scram button will be available at the fission converter medical room control panel and will be operable whenever the fission converter is in use.

Tables 7.1 and 7.2 summarize the protective actions for the fission convener l transients related to safety under forced convection and natural convection, respectively. l Table 7.1 Protective Actions for the Fission ConverterTransisnts Related to Safety (Forced Convection)

Automatic Suggested Automatic Converter Setpoints Transient Reactor Scram Control Shutter (LSSS)

O Closure b Overpower X 275 kW (300 kW)

Ovenemperature X 55 C (60 C)

Low Coolant level X X 2.4 m (2.lm) 4 Low Primary Flow X X 50 gpm l (45 gpm) .

1 Table 7.2 Protective Actions for the Fission ConvenerTransients Related to Safety (Natural Convection)

Automatic Suggested Automatic Converter Setpoints Transient Reactor Scram Control Shutter (LSSS)

Closure Overpower X X 15 kW (20 kW)

Overtemperatum X 55 C (60 *C)

Low Coolant level X X 2.4 m N

(2.4m)

(O 7-4

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

1 I

l 7.3.1 Operability of Fission Converter Shutdown Svstem

)

The fission convener is defined as being shut down when the CCS is fully inserted '

1 or when the reactoris in a shutdown condition. The fission converter is defined as being j secured when there is no fuelin the fission converter or all of the following conditions am satisfied:

l (i) The fission converteris shut down, J

(ii) The converter control shutter (CCS) control panel key switch is in the off l position and key is in proper custody, and j (iii) There is no work in progress within the converter tank involving fuel.

Whenever the fission convener is in either a shutdown or a secured condition, the automatic reactor scrams that originate on lower fission convener coolant level, loss of fission converter primary flow (for forced convection), and high power (natural convection) are not required. Neither is the manual reactor scram located at.the fission ,

converter medical control panel. l 7.4 OtherInstrumentation Table 7.3 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 1 other instmments listed are not safety-related. The fission converter medical room control panel (MRCP) is a dedicated control panel for instruments, such as the shutters, that am I related to the beam control and used of the fission converter medical room. The fission converter control panel (FCCP) houses the control and display of the process system j instmments of the fission convener.

i l

O 7-5

l e Table 7.3 Fission Converter Instrumentation Parameter Instrument Readout Alarm Location Safety Related Design Power Neutron Detector CR Yes Outlet Temperature Thermocouple or Equivalent CR Yes Coolant I2 vel Conductance Level Probe or CR Yes (Trip Point) Equivalent Primary Flowrate Orifice Plate or Equivalent CR Yes Not Safety Related Nominal Operating Power Neutron Detector CR Yes Hx Secondary Flow Rate Flow Switch CR Yes Primary Inlet Temperature Thermocouple or Equivalent CR No Secondary Outlet Themlocouple or Local or FCCP _Ne Temperature Equivalent Secondary Inlet Thermocouple or Local or FCCP No Temperature Equivalent Cleanup System Thermocouple or Equivalent Local or FCCP Yes Temperature Coolant Conductivity -Ion Conductivity Probe Local or FCCP Yes

n. Column Inlet Q Coolant Conductivity -Ion Conductivity Probe Column Outlet Local or FCCP No Cleanup System Flowrate Rotometer or Equivalent Local or FCCP No Coolant 12 vel (Indication) Coolant Level Sensor Local or FCCP N o' '--

Izak Detection Leak Tape or Equivalent Local or FCCP Yes 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 Converter Control Shutter Limit Switches CR and MRCP Yes Position Mechanical Shutter Limit Switches CR and MRCP No Position Water Shutter Tank Level Level Probe CR and MRCP No Water Shutter Upper Tank Level Probe Local Yes 12 vel Med Room Door Position Limit Switches CR and MRCP Yes Storage Tank 12 vel Gauge Local or FCCP No MRCP: Fission Converter Medical Room Control Panel CR: Control Room FCCP: Fission Converter Control Panel O

7-6 I

8. Pre-operational Tests and Initial Operation This chapter summarizes tha pre-operational tests and initial operation of the fission converter. The pre-operational tests are designed to ensure that the fission converter has been constmeted in accordance with the information presented in this repon. 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 keff of the fission converter is 0.62. Hence, a criticality l condition is not credible. However, sub-critical multiplication should be monitored for safety during the initial fuelloading. The major steps for the first approach for the fission converter's operation at power are summarized. These include measurement of the flow distribution in the fission convener fuel region and estimation of the fission convener reactivity effect on the MITR. The initial startup of the fission converter to its highest available operating power will be achieved by a stepwise rise of the reactor's power.

8.1 Pre-Oncrational Tests Pre-operational tests will be performed to ensure that the facility will operate as O designed. Dummy fuel elements that replicate flow conditions am available for the pm-operational tests. These tests will be used to establish initial compliance with the approved l technical specifications. The tests will include component inspection, verification that

! performance objectives are met, instrument calibrations, and the operability of interlocks.

All tests will be conducted in accordance with the existing MITR quality assurance l program.

! 8.1.1 Non-Nuclear Instrument Calibration Instruments for measuring system pressure, temperature, and flow can be calibrated prior to the initial stanup of the fission convener. 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 prior to the initial stanup of the fission converter and technical specifications penaining to the signals from these instruments shall be in effect during the initial startup.

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8.1.2 Nuclear Instrument Calibration

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!V The nuclear instmmentation associated with the fission convener can not be calibrated 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 perform a calorimetric in which the output of the nuclear instrument is correlated with thermal 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 I

be available for use during the initial startup. An altemate to a thermal power calibration would be to perform 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.

The following 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.

/G Either a thermal power calibration or a foil activation measurement or both will be 5

performed under this conservative condition. The information obtained will be used to calibrate the nuclear instrumentation. It should be noted that the maximum fission convener power,250 kW, will be achieved only if the MIT Research Reactor is operating at 10 MW and fresh 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 convener will be given specialized instruction 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 l fission converter as pan of their initial qualification. If non-licensed individuals wish to qualify on the fission convener, a written qualification program will be prepared.

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8.3 Initial Fuel Loading (3

V The initial fuel loading will be made after the pre-operational testing has been satisfactorily completed. The reactor will be shutdown and the convener control shutter (CCS) fully closed before the fuel loading of the fission converter begins. Calculation has shown that the maximal reactivity of the fission convener is 0.67. So, a criticality condition is not credible. Nevenheless, sub<ritical multiplication will be monitored during the initial fuelloading.

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 locate.d 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 converter will be less than 0.90, the maximum allowed value for a fuel storage location at the MIT Research Reactor.

8.4 Fuel Recion Flow Distribution Measurement

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The Dow distribution among the fuel elements will be measured with the CCS 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-II stanup test data

[Ref. 8-1]) will be used to determine if the operating limits on the fueled region coolant flow factor and 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 pans. First, the integral reactivity associated with fully opening the CCS will be measured. Second, the differential reactivity wonh associated with partial opening of the CCS will be measured.

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8.5.1 Estimation ofIntegral Reactivity Worth V

Pre-conditions

1. Fission converter fuel loading complete.

2. Reactor shutdown with minimal Xenon.

3. Fission converter heat removal system operable (not required when CCS closed).

Procedure

1. Take reactor critical. Level reactor power at a designated low power l l

(e.g.,500 W)

2. Record critical data.

3. Shutdown reactor.

4. ' Open CCS fully.

5. Repeat steps 1,2, and 3.

6. Calculate the reactivity worth of the fission converter by comparison of the critical positions obtained in steps 2 and 5. I i

8.5.2 Estimation of Differential Reactivity Worth i

i A differential curve of the reactivity associated with opening the CCS will be l

. i obtained with the CCS in vanous positions (e.g., 25%, 50%, 75% open). This will be done by repeating the procedure described in section 8.5.1.

8.6 Initial Approach to the Highest Available Fission Converter Operating Power l 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 corresponds 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.

n 1. The interior of the fission converter facility will be checked to ensure that no foreign b objects are present.

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wJ 2. All process and radiation monitoring systems will be placed in their normal operating condition with non-nuclear instruments calibrated. I l

3. The fission converter's top shield lid and associated shielding will be removed in order j

to allow the temperature distribution in the fuel region to be measured 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 20 kW. The estimated dose rate on the coolant surface for this power level is 450 mR/h as described in Appendix l 3.2) l l

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4. The converter power will be increased by raising the reactor power in a stepwise i manner. Radiation levels and system temperatures will be monitored during each power increment. This procedure will be repeated until the maximum available l operating power is attained.

5. The following measurements will be made:

a) 1emnerature Distribution The temperature distribution in the fission converter plate 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).

b) Process Parameters Fission convener primary inlet and outlet temperatures as well as flow rate will be measured. This information will be used to perform 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 calorimaric. Energy losses

()

v because of gamma radiation etc. will be taken into account. The equilibrium 8-5

neutron count rate associated with the nuclear instrumentation will be measured.

Correlation of these count rates with the calorimetric will be used to calibrate the nuclear instruments.

The above procedure for a stepwise increase of the fission convener operating ,

power will be repeated if any one of the following design changes is made: l l

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1. The maximum available operating power is increased,

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2. ' The fission convener primary coolant is changed from H2 O to D29 (the hot - !

channel factor increases - see Table 2.4 ), or -

3. Fresh fuel is used to replace burned fuel.

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4. The aluminum block located between the fuel region and the tank wall is l removed or replaced by another approved unit. i l

l References )

[8-1] MITR Staff, "MITR-II Stanup Report", MITNE-198, Feb.1977.

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l TECHNICAL SPECIFICATIONS I

FOR i

O FISSION CONVERTER FACILITY l

o DECEMBER 31,1998 i

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TABLE OF CONTENTS :

EM 6.6 Design and Operation of the Fission Converter Facility............................... 6-34 t

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l 6.6.1 Safety Limits and Limiting Safety System Settings........................... 6-36 l t :

I 6.6.1.1 S afety Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6 j i 1 i 6.6.1.2 Limiting Safety System Settings (LSSS) ......................... 6-40 l 6.6.2 Limiting Conditions for the Fission Converter Operation.................... 6-44 .

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< 6.6.2.1 Limiting Operating Conditions for the Fueled Region .......... 6-44 i

6 6.2.2 Maximum Allowed Reactivity Addition from the Converter l l Control S h u t t e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 7 l 6.6.2.3 Fission Converter Fuel Element Security, Storage, and H a n d l i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 8 6.6.2.4 D2/H2Concentration and Recombiner Operation................ 6-50 9 6.6.2.5 Fission Convener S afety System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-52 l 6.6.2.6 Fission Convener Primary Coolant Quality Requirements ..... 6-56 l 6.6.3 Fission Convener Surveillance Requirements................................. 6-58 i i

6.6.4 Fission Converter Design Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-60 6.6.5 Reporting R e q u i r e m e n t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-63 1

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