ML20199H220
| ML20199H220 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 01/14/1999 |
| From: | Bernard J, Hu L NUCLEAR REACTOR LABORATORY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20199H226 | List: |
| References | |
| TAC-M99844, NUDOCS 9901250147 | |
| Download: ML20199H220 (74) | |
Text
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6 h NUCLEAR REACTOR LABORATORY AN INTERDEPARTMENTAL CENTER OF MASSACHUSETTS INSTITUTE OF TECHNOLOGY -
JOHN A. BERNARD 138 Albany Street, Cambridge MA 02139 4296 Actwation Analysss Der:ctor Telefax No. (617) 253 7300 Coolant Chemistry Ovector of Reactor Operations "#**'"*d'"'"'
Pmcipal Research Engineer Tel. No (61353-4202 Reactor Engineering January 14,1999 1
Nuclear Regulatory Commission l Attn: Document Control Desk Washington, D.C. 20555
Subject:
Request for Amendment of Facility Operating License No. R-37 for the Massachusetts :
Institute of Technology Research Reactor (MITR); Docket No. 50-20 ;
Dear Sir or Madam:
i On 3 October 1997, the Massachusetts Institute of Technology submitted an amendment l' request for the operation of a fission converter in conjunction with our research on neutron capture therapy. That submittal included a safety evaluation report, a proposed technical specification (No. 6.6), as well as minor charges to other technical specifications. On 20 July 1998, the Massachusetts Institute of Technology received a request for additional info mation relative to the amendment request. Enclosed is our response to that mquest together with the revised safety evaluation report (SER) and the revised wording to proposed MITR Technical Specification No.
6.6. !
Your earliest attention to this request would be most appreciated. Correspondence concerning this request should be directed to Dr. Bernard. l Sincerely,
' (. . r lam ikk' Lin Wen Hu, Ph.D. ohn A. Bernard, Ph. .
Relicensing Engineer Dimctor !
MIT Nuclear Reactor Laboratory MITNuclear Reactor Laboratory JAB /lwh Enclosure cc: USNRC - Senior Project Manager, ;
NRR/PDND; Mail Stop 0-11-D-19 \ 1 USNRC -
Region I- Project Scientist O. Harling, MIT Principal Investigator BNCT Project P. Busse, BIDMC Climcal Principal Investigator .
R. Zamenhof, BIDMC Project Coordinator and Principal Investigator l MITRSC i
'i50030 .
l 9901250147 990114 PDR ADDCK 05000020 p PDRL 1
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Response to Request for Additional Information !
l Massachusetts Institute of Technology ;
Q Docket No. 50-20; License No. R-37 (TAC No. M99844) !
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- 1. The fission converter based epithermal neutron irradiation facility has been designed so that it can operate with2 H O or 2D 0 or even with a mixture of coolants. The keff using either H2 O or D20 is much less than 1.0, as shown in Table 2.1. Our neutronic calculations have shown that a somewhat better quality beam, at the target position,is obtained with D 2 0. The heat removal system will function adequately with either D 2 0 or H2 0. Our preference is to operate with D20, but should we not be able to obtain D 2 0 coolant in a cost free loan, we will use H20. We do not intend to switch coolant between D2 0 and H2O routinely.
There is ofcourse a major difference in the precautions required for handling D 2 0 versus H2 0. We will use our experiencewith the D2 0 reflector in the MITR to assure that the tritium hazard is adequately managed. However, because of the low power of the fission converter and its low capacity factor, less than 10%, the specific activity of the tritium in the fission converter coolant will be much lower than that in the D20 of the MITR reflector.
- 2. As discussed in question 13 below, the fission converter tank will be built to ASME Code Sections H and IX. Other aspects of the Fission Converter Facility, includingpiping, shielding, shutters, and the medicalroom, either have or will be designed with the involvement of licensed professional engineers (nuclear and O mechanical) and reviewed by the reactor engineering staff and the Reactor SafeguardsCommittee.
- 3. The grid plate is designed to allow minimal clearance in order to reduce bypass flow. The amount of bypass flow is implicitly limited by TS# 6.6.2.l(2), which was used as the basis of the thermal hydraulic limit (SL and LSSS) calculations.
The calculated SL and LSSS are valid as long as Fr x dr > 0.8, which takes into account both the effects of bypass flow and flow disparity among fuel elements as well as within the fuel element. The primary coolant flow through each fuel element will be measured during the initial startup test.
- 4. A step in the existing reactor startup procedure ensures that the cadmium curtain is fully inserted before and during reactor startup. This procedure will also be applied to the converter control shutter. In addition, there will be an interlock between CCS position and the reactor's withdraw permit circuit. This interlock will necessitate the CCS beingin the fully closed position in order to commence reactor startup (TS# 6.6.4.7(e)). The interlock will be equipped with a key operated bypass to allow testing of the interlock and to allow bypassing of the interlock once the reactor is critical. Use of this bypass will be in accordance with existing MITR procedures. It should be noted that there is no safety significancein terms of either reactivity or radiation if a reactor startup is O conducted with the CCS open. The concern is for the safety of the fission 1
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converter and in particular that the CCS not be opened with the reactor at power l
[ and flow not established within the fission converter. Note: An additional safety feature that addresses the above concern is the interlock associated with the l fission converter operation panel ON/OFF key switch. This interlock ensures j that the CCS is closed whenever that key switch is in the OFF position. Also l whenever the keyswitch is in the ON position, the low coolant level reactor scram l l is activated and either the fission converter primary coolant flow scram is also I activated (forced convection) or a reactor scram at 15 kW fission converter power is operable (natural convection). Hence, it is not possible to open the CCS without adequate heat removal.
- 5. We have revised the proposed TS to utilize the reactivity worth for movable l experiments as specified in existing TS# 6.1. We believe that the reactivity worth !
of the converter control shutter (CCS) will be substantially less than the existmg j limit (0.2% AK/K). i 1
- 6. These results are from two sets of MCNP calculations,the first set was done by l l W. Kiger and the second was by S. Sakamoto, both of whom were graduate
! students at MIT. Both were supervised closely by MIT faculty. Discrepancies are due to MCNP's random particle transport process and are less than 0.5%
between these calculations. The tables have been revised to be consistent.
'O O 7. The purpose of the filter / moderator (F/M) is to moderate fission neutrons into the epithermal range (1 eV- 10 kev) and to reduce the intensity of fast neutrons. A large number of transport calculations have been carried out to examinea variety of F/M materials and configurations, see 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 properties such as mechanical I strength, temperature resistance, radiation stability, and fabricability which make them suitable for this application. We chose to use aluminum / Teflon @because of its somewhat lower cost, ready availability, and very easy fabricability compared to the other material combinations. This additional information has been incorporated in the revised SER.
- 8. Nucleate boiling includes subcooled boiling and saturated boiling. The former refers to the condition where small bubbles are formed on the heated surface but l collapse when they enter into the bulk coolant. The latter refers to the condition j where the small bubbles coalesce with each other upon leaving the heated surfaces and form largerbubbles. The bubbles then collapse and form a vapor core with liquid film on the heated surface with higher heat fluxes. Onset of nucleate boiling (ONB) or incipient boiling defines the condition where small bubbles first start to 2
l 1
1 p form on the heated surface. The definitions are included in the SER as well as in lQ i
the basis of TS 6.6.1. !
- 9. In the event ofloss of fission converter primary flow, there is no provision for natural circulation flow because a natural circulation loop can not be established l with the inlet pipes in the downcomers. A safety analysis was performed for this transient (see SER section 6.4). The results showed the maximum fuel temperature during this transient is well below the fuel softening point (450 C). With the l reactor automatic scram, the maximumfuel temperature is 100 C and no boiling will occur. If the reactor automatic scram failed and the fission converter shuts down with automatic converter control shutter closure, the maximum fuel temperature is 139 C. Note that for the second scenario cold water will flow down from the top to wet the fuel surface with the vapor flowing upward when boiling occurs. This will provide adequate cooling for the fuel plates. (Ref: " The i
Effect of Flow Direction and Magnitude on CHF for Low Pressure Water in Thin Rectangular Channels", K. Mishima and H. Nishihara, Nuclear Engineering and Design 86(1985)). Iflow temperatures are desired, the cleanup loop could be used to provide cooling. Please see response to question #10.
- 10. For any givenpatient irradiation,the fraction of time spent in patient setup will lp likely exceed the time spent under the beam. According, for the anticipated duty
- ( cycle of the fission converter, the decay heat load will be small. There may, however, be other sources of heat that are deposited in the fission converter. This willinclude both gammaheating and fissions caused by leakageneutrons through the CCS. The temperature of the fission converter coolant is monitored by an alarmed instrument. If the alarm is received,the corrective actions might include use of the cleanup pump to lower the temperature. The SER has been modified to include thisinformation.
- 11. There will be a burst disk or equivalent present on the cover gas system that will be set for 5 psig. The converter tank was hydrostatically tested at 10 psig.
Hence, any prolonged operation at or below this value will have no consequences.
In addition, the cover gas will have a feed and bleed control to compensate for pressure changes expected during normal startup and shutdown transients. This system will be set to bleed at 4 psig or lower.
- 12. Under normal operation, the fission converter primay system will operate at a pressure that is about 10 psi lower than that of the secondary system. Should an l overpressure event occur, the burst disk will open before the primay pressure exceedsthat of the secondary. The MITR-Il secondary water radiation monitor will be operating and the secondary water will be sampled daily for tritium content. Any leakageto the fission converter primary system will be seen as a l
3
- level or a conductivity increase. The tritium concentration of the fission converter primary system using D2 0 will be at least two orders of magnitude lower than that of the reactor D2 0 system. llence, the consequences of a leak into the secondary system will be much less than that from the reactor D2 0 system.
Similarly, fission product activity will be much lower in the fission converter so the consequences of a leak will be much less than the reactor primary system.
Leak detection will be present on and around the fission converter piping and will alarm in the reactor control room if a leak is detected. This will be in place regardless of which coolant is used. It is planned that the secondary flow will continue even if the fission converter is in a shutdown condition. Therefore, the secondary pressure will remain higher than the primary pressure. If this is not the l case (maintenance activity to the secondary system, for example), the fission l
converter heat exchanger will be isolated and vented.
- 13. We are committing to the use of ASME Code sections 11 and IX and will l document materials and welding specifications. No seismic loading considerations 1 were made per se, although an analysis was made in SER Section 6.3 which showed that an instantaneous LOCA would not result in a fuel cladding failure.
Any seismic damage would be bounded by this analysis. I
- 14. The aluminum block placed between the fuel and the wall of the fuel tank, serves j'~'N to minimize the coolant / moderator near the fuel. This removable aluminum block V is planned to be used with either H 2 O or D2 0 coolant. It has been shown that if H2O coolant is allowed to fill the space occupied by the block, the beam will be overmoderated and a poor epithermal beam will result. TS# 6.6.4.5 was added to specify the associated requirements on this aluminum block. Calculations of K,y for the fission converter have been made for fission converter operation both with and without the aluminum block. The maximum Ken with the block removed using H2 O and fresh fuel is 0.67010.0012, compared to 0.589i0.0012 with the block installed. The calculated results for the hot channel factors and fission converter powers are summarized below (file memos attached in Appendix A):
Configuration Hot Channel Factor Total Power (kW)
(MITR at 5 MW)
With Al Block Fresh Fuel. H2O cooline 1.47 i 0.013 125.7 i 0.1 Fresh Fuel, D2O cooline 1.53 i 0.014 104.9 i 0.1 Without Al Block Fresh Fuel, H2O cooling 1.55 1 0.007 158.2 i 0.4 Fresh Fuel, D2 O cooline 1.57 i 0.007 141.5 t 0.4 m 15. The low-power / natural-circulation operational mode of the fission converter is
) intended for the initial startup test when measurements require the lid to be off.
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Because the tritium concentration is relatively low in fresh D 2 0, tritium release to the building atmosphere should be extremely low. If the tank needs to be open for maintenance / refueling, an analysis will be performed prior to such operation to predict the associated stack release and address the procedures required for personnel radiation protection. This is the same approach currently followed for the MITR's D 2O reflector tank.
- 16. Closure of the fission converter tank is provided by a bolted hatch with a rubber 0-ring seal. It is both gas and water tight.
- 17. No. The minimum operational requirement of the recombinerwas derived based on the assumption that the fission converter operates on a 100% duty cycle.
Therefore, the recombiner can be operated any time within a month.
I 8. The following statements are added to section 3.5 of the SER:
" Onset of nucleate boiling (ONB, also called incipient boiling) defines the condition where bubbles first ctart 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 stan to detach regularly. "
- 19. No. The calculations will be redone only if the parameters used are less conservative than the measurements. For example,if the measured hot channel factor (Fu) is higher than that used in the thermal hydraulic limit calculations, the calculations will be redone based on the measured value. TS# 6.6.5 has been added to address the initial startup reporting requirements for the fission convener.
- 20. The question is correct in that the fission converter SLs and LSSSs are based on 10 MW operation of the MITR. The question is also correct in that the nuclear hot channel factor obtained from Table 2-4 was determined for 5 MW operation.
However, this factor is valid at any power level, including 10 MW, because it is the ratio of the fuel plate maximumpower to the fuel plate averagepower. This ratio is independent of the actual power level provided that no power dependent feedback mechanism such as boiling has altered the thermal hydraulic behavior of the converter. This is in fact the case for the fission converter for a reactor power level up to at least 10 MW. Hence, the entire discussion on pages 3-22 to 3-25 is relevant to 10 MW operation. (File memo attached in Appendix F)
- 21. " top of fuel" means " top of fuel element". This has been corrected in the SER and TS.
s 22. The reviewer has raised a valid point that a margin should exist between LSSS and s
SL for natural convection cooling. Therefore, a 5 C margin has been added to 5
establish the LSSScurve. The LSSS setting for the fission converter power is thus lowered to 20 kW from 25 kW. A marginalready existed between SL and LSSS for forced convection. Analyses for both forced convection and natural convection have been perfonned which show that the margins are adequate.
Namely, the margins are sufficient so that automatic protective actions will correct an abnormal situation before a SL is reached. (File memo attached in Appendix G)
- 23. TS# 6.6.2.2 has been modified to utilize the reactivity worth for movable experiments. Please see the response to question #5. The opening speed of the converter control shutter (CCS) is therefore no longer an issue. The total reactivity effect of the CCS drop is the negative value of the reactivity worth of the CCS, which depends on the fuel / coolant combination of the fission converter system. Measurements will be made to confirm its reactivity worth during the initial startup testing.
- 24. These two figures were made based on the MITR power of 5 MW, using D2 0 coolant and spent fuel elements. The 251 kW figure was calculated for 10 MW with H 2O coolant and fresh fuel elements. The captions for these two figures have been changed to indicate the conditions of these calculations.
A 25. We will not use a variable speed motor except possibly in the startup testing.
U During routine operation a fixed speed motor will be used. Provisions to protect against inadvertent movement of the converter control shutter (CCS) are listed below:
- a. An interlock ensures that the fission converter primary flow must be established in order to open the CCS (forced convection operation). (TS#
6.6.4.7 (a))
- b. An interlock ensures that the fission converter coolant level scram must be enabled in order to open the CCS. (TS# 6.6.4.7 (b))
- c. An interlock ensures that the water shutter and mechanicalshutter will close automatically when the medical shield door is opened. The interlock will also prevent opening of the shutters if the shield door is open. (TS#
6.5.5 (a),(b))
- d. An interlock ensures that the CCS will close automatically when the CCS control panel key switch is in the OFF position. (TS# 6.6.4.7 (d))
For additional discussion and an additionalinterlock for the CCS, please refer to our response to question #4.
- 26. Calculations of the dose rate in the medical room with all shutters closed, including the water shutter, indicate that water without dissolved B-10 will be adequate to dg reduce dose ratesin the direct beam at the patient position to ~1 mrem /hr. B-10 6
c may be added if a further reduction in neutron or hydrogen capture gamma dose
()
rate is desired. The reactivity effect of B-10 is negligible. (file memo attached in Appendix B) i l
- 27. The remotely operated valve in the water shutter system will be either solenoid or air-operated. This valve will be normally open so that in the event of a loss of ,
power, the shutter will fill. This valve is located outside the medical room. I Hence, if manual operation were needed, no entry into the medical room would be required. In order to empty the water shutter, the valve will shut and the main pump, located outside the medicalroom, will start pumping water to the supply tank. Because the power failure mode of the valve is to keep the water shutter closed, there are currently no provisions to manipulate the valve manually.
- 28. The followinglist summarizes our planned shutter controls, their locations, and manual / automatic override if applicable.
A. Fission Converter Operation Control Panel (Located in Control Room)
A1. Fission Converter ON/OFF Key Switch This key switch enables flow and level scrams as well as interlocks B1
( through B3. It also energizes the fission converter operation control (3,,/ A2.
panel.
Converter Control Shutter OPEN button A3. Converter Control Shutter CLOSE button A4. CCS Remote Open Permission Key Switch The above controls come with indication lights.
B. Interlocks The following interlocks are required by technical specifications.
Bl. Fission Convener Primary Flow Scram (forced convection) / Converter Control Shutter Open Interlock (TS# 6.6.4.7 (a))
B2. Fission Converter Coolant Level Scram / Converter Control Shutter Open Interlock (TS# 6.6.4.7 (b))
B3. Medical Room Shield Door / Water Shutter and Mechanical Shutter Open Interlock (TS# 6.5.5 (a),(b))
B4. Loss of Electrical Power / Water Shutter and Mechanical Shutter Close Interlock (TS# 6.5.5 (c))
- B5. Medical Room Control Panel Key Switch Off / Water Shutter and
( Mechanical Shutter Close Interlock (TS# 6.6.4.7 (c))
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B6. Reactor Startup / CCS Fully Closed Interlock (TS# 6.6.4.7 (e))
(" ')
C. Fission Converter Medical Room Control Panel (MRCP)
C1. ON/OFF Key Switch C2. Minor Scram button C3. Converter Control Shutter (CCS) Close Button C4. CCS OPEN/ CLOSED Display C5. Mechanical Shutter OPEN/CLOSE buttons and display C6. Water Shutter OPEN/CLOSE buttons and display C7. Med Room door OPEN/ CLOSED display C8. CCS OPEN Button (operable only with remote permission key switch ON, see A4)
D. Medical Room Emergency Control Panel (Inside the medical room)
(TS 6.5.5 (e))
DI. Converter Control Shutter CLOSE button D2. Mechanical Shutter CLOSE Button D3. Water Shutter CLOSE button
,O Position Displays for the above.
E. Emergency Manual Shutter Controls (TS 6.5.5 (d))
El. Mechanical Shutter Our answers to other sub-questions are listed as follows:
- a. The CCS, water shutter, and mechanicalshutter can be closed from inside the medical room. No significant dose rates exist in the medical room when all shutters are closed (~ l mrem /hr). Therefore, no scram button is required in the medicalroom. Also, the water and mechanicalshutters are interlocked to close when the medical room door is opened. A fixed radiation monitor is provided so that the dose rate in the room is known prior to entry.
- b. The table below shows the dose rates at the patient position in Rem / hour for various shutter conditions (adopted from Appendix H). In all cases, the dose rate exceeds the criteria for a high radiation area if one of the shutters fails to close with the reactor at power. The corrective action for f]
a such a situation is to lower the reactor power. This is the same approach 8
,- 3 that is used for the current MITR medicalroom shttiers. Also, it should
(
) be noted there is a minor scram button on the medici wm control panel.
1 Condition of Shutters (MITR-Il at 5 MW) Dose Rate at Patient Position (Rem /hr)
CCS Open, Water and Mechanical Shutters 0.1 Closed Mechanical Shutter Open, Water Shutter 15 and CCS Closed Water Shutter Open, Mechanical Shutter 0.2 and CCS Closed
- c. The medicalroom control panel key switch is interlocked with all three shutters. The shutters will close automatically when the key switch is in the OFF position.
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- d. Opening of the CCS is under the control of a licensed reactor operator. l Please see #36 for detailed information.
rm ,
- 29. The medical room shielding is designed to reduce the radiation levels, from the medicalbeam,on the outside of the room to < 1 mrem /hr for all possible fission converter facility operating scenarios including MITR at 10 MW. Patients will be observed visually duringirradiations 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 in the medicalroom. This approach is used in the existing MITR medical room. A permanent radiation monitoring system with readouts at the medical control panel outside the irradiation room will provide information concerningambient radiation levels in the medical room. A similar system is currently used in the MITR medical room in the reactor basement. I Note: This information has been added to SER section 4.3. I i
- 30. The MITR administrative procedures apply to the fission converter facility. PM 1.9.1 states that "If not part of an approved procedure, bypasses must be l individually approved before implementation by the Duty-Shift-Supervisor or j Reactor Superintendent. The bypass, authorizer's initials (or number of approved procedure) must be recorded on the bypass log sheet at the time of l implementation." The medical room door bypasses will normally only be used for i tests and experiments.
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l 7 s, 31. The maximum dose for a fuel plate melting accident is covered in section 6.1, )
() " Maximum liypothetical Accident"(four fuel plates melted). The following l infomiation has been added to the SER for clarification: ,
l if the fission converter 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 elements /11 elements) x 595 mrad = 65 mrad at 21 m (front fence), or 41 mrem at 8 m (back fence). The thyroid dose from containment leakageis (250 kW/5000 kW)(24 elements /11 elements) x 118 mrad =13 mrad.
l The calculated whole-body doses are significantly lower than the annual dose limit to a member of the general public (500 mR).
The 595 mrad and 118 mrad are the estimated doses, whole-body and thyroid respectively, at the nearest point of public occupancy duringthe first two hours of the MITR-Il maximum hypothetical accident.
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- 32. Situations of this type would be coveed by tne existingMITR emergency plan. l In response to the sub-questions:
- a. 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 l products should the clad be scratched or otherwise damaged. It should be O noted that all MITR fuel handling tools have a safety lock feature that prevent this type of accident.
- b. A large heavy load could not impact the converter because it would be deflected by surrounding shielding. Also, administrative procedures, which are followed for the reactor core, would preclude situations that could lead to this type of accident. Should such an accident somehow occur, the consequences would be limited because the fission converter operation cycle, even at a full patient load, will be a small fraction of a day.
This is because most of the time spent on each patient will be for setup.
Hence the fission product inventory will be low. Our calculation estimates that the fission product inventory is 31660 curies per fuel element 10 minutes after shutdown for a refueling. In the event that a fuel element was damaged in the converter, calculated results show that the total whole body dose would be 2.8 mrem at 8 meters (ground release)and 3.9 mrem (plume release) at 21 meters from the containment. The thyroid dose would be 0.5 mrem during the first two hours. The above figures are less than 0.5% than those calculated for the MITR (5 MW) design basis j accident. (file memo attached in Appendix C) l In the event that all eleven fuel elements in the fission converter were damaged,the total whole body dose would be 30.8 mrom at 8 meters and n 42.9 mrem at 21 meters from the containment, the thyroid dose would be V 5.5 mrem during the first two hours.
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- 33. The calculation has been redone based on a more realistic power function. A slight increasein the maximumfuel temperature from 132 C to 139 C was calculated )
using the power function shown in Figure 4.1 in the fission converter SER. (file memo attached in Appendix D)
- 34. The proposed fission converter overpower setpoint is 275 kW, which is 110% of l the maximum anticipated design power, as shown in Tables 2.3 and 7.1. The l automatic response that occurs at the overpower setpoint is CCS closure. An l l alarm will be provided at 110% or less of the fission converter's nominal operating l power (TS# 6.6.4.8 and 6.6.2.5(3)). This alarm shall not exceed 275 kW. The fission converter's nominaloperating power was calculated for combinations of l reactor power, primary coolant, and fresh / spent fuel element using MCNP. The results will be verified during the initial startup test. The setpoints for fission converter protective actions are summarized in Tables 1 and 2 in our response to ;
question #37. As stated above, one of the factors that determines the alarm j setpoint is the reactor's power level (5 or 10 MW) which in turn affects the fission converter's nominal operating power. The setpoints for the other j protective actions do not depend on the fission converter power level because the i fission converter flow is always established for operation of the fission converter y at 250 kW, which corresponds to a reactor power of 10 MW.
- 35. The function of the " design power" safety channelis to shut down the fission converter in case of an overpower condition. The design power is based on the LSSS (300 kW) regardless of the actual configuration (type of coolant and fuel) of the system. This safety channel is specified as " power"in Table 6.6.2.5-1 in TS#
6.6.2.5. The " nominal operating power" channel will provide an alarm in the reactor control room if the fission converter power exceeds 110% of its nominal operating power as specified in TS# 6.6.2.5 (3). The magnitude of the nominal operating power, which depends on the configuration of the system + was determined using MCNP and will be confirmed during the initial startup tests. !
The latter provides an extra margin to the design power. It is currently planned to use one neutron detector that will provide settings for both " design power" and
" nominal operating power".
The following statement has been added to section 7.4 of the SER:
" The fission converter medical room control panel (MRCP) is a dedicated coinrol panel for instruments, such as the shutters, that are related to the beam control and use of the fission converter medical room. The fission converter control panel l (FCCP) houses the control and display of the process system instruments of the fission converter."
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- 36. The converter control shutter (CCS) OPEN/CLOSE buttons will be located on a panelin the reactor control room. The panel is energized by an ON/OFF key 11
, switch. Administrative controls will ensure proper possession of the key as with
( other keys associated with reactor operation. The CCS shall be opened from the LJ 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, pennission 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-i console operator's permission before operating the CCS. This is the current practice of using the D2 0 and H2 O shutters for the M67 beam currently used for the BNCT clinical trial. In answer to specific sub-questions:
- a. Pre-operational checks of the fission convetter will be performed by l licensed personnel (or a trainee under direct supervision) in the same way that reactor system pre-operational checks are conducted. Upon completion of these checks, all systems are operational and the converter can be placed on line by opening the CCS. This would be done by a l licensed person (or a trained non-licensed person) upon direct receipt of I
authorization from the reactor console operator.
- b. If non-licensed individuals wish to qualify on operation of the CCS, a written qualification program will be prepared. We have such a program for the existingMITR medicalfacility. It is structured alongthe lines of the existing MITR training programs for the reactor operators. Operation of the CCS by qualified non-licensed personnel meets the requirements of 10 CFR 50.54 because the licensed console operator retains direct control.
Specifically,the person operating the CCS must request permission from the console operator to open the CCS. Moreover, this permission must be requested immediately before opening the CCS. Please refer to existing MITR TS# 6.5.16.
- c. No. There are no fission converter operations that require the presence of a senior reactor operator.
l d. The console operator's responsibilities for the fission converter will not be any different than they are now for experiments that have the potential to affect reactivity. The fission converter is not a distraction, in the sense that it is a separate activity distinct from the operation of the reactor. The two are coupled and the operator cannot properly monitor reactor l operation unless he/she is also given the authority to direct fission j converter operation. The sub-question implies that it might be desirable to l m split the responsibility. To do so would make it very difficult for the 12 l
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l l
l
-m console-operator. One person, the console operator, should be in overall
{) charge of anything that can affect the recctor,
- e. In case of an emergency,the console operator's responsibility is either to close the CCS or scram the reactor as appropriate. Subsequent actions, which would take place outside the control room, would be the l responsibilities of other licensed operators.
l
- 37. The interface of the fission converter safety system with the reactor control and safety system is shown schematically in Figure 1. Tables 1 and 2 outline the protective actions and proposed setpoints for corresponding major transients for i forced convection and natural convection operation. A jumper (or equivalent) will be used to bypass the flow scram and CCS automatic closure for lower power (natural convection) operation. For forced convection cooling, protection against a j fission converter overpower condition is provided by an alarm at 110% of nominal operating power and an automatic CCS closure at the over-power setpoint 275 kW. A reactor scram on fission converter overpower is not needed because the reactor itself will have already scrammed on high power. For natural convection cooling, protection against a fission converter overpower condition is provided by 1 a reactor scram at the reactor power which corresponds to fission converter power 15 kW. This difTerent approach is necessary because an overpowcr condition can l occur on the fission converter during natural convection cooling even though the reactor itselfis operating within its licensed operating power.
l l
l Table 1 Protective Actions for the Fission Converter Transients Related to Safety (Forced Convection)
Automatic Proposed Transient Automatic ConverterControl Sctpoints HeactorScram Shutter Closure (LSSS)
Overpower X 275 kW (300 kW)
Overtemperature X 55 *C (60 *C)
Low Coolant Level X X 2.4 m (2.1m)
Low Primary Flow X X 50 gpm (45 cpm)
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T- 5 m 5 5<T-SMF c r aC C; - 'iI an = mua . l i o r e Cm 4 >L 2< e L 2e v 4ev 4e ml ml s 4 + - g p 5l >F p5l<F o m0ow m0 w 4 4 OPF NaC n e l 1 l ~ _ "able 2 Protective Aetions for the Fission Conserter Transients Related to Safety (NaturalConvection) Automatic Proposed Transient Automatic Cons erterControl Setpoints ReactorScram Shutter Closure (LSSS) Overpower X X 15 kW (20 kW) Overtemperature X $$ C (60 *C) Low Coolant Level X X 2.4 m (2.4m)
- 38. The MITR Quality Assurance Program has been modified to include the fission converter. This means that all test procedures associated with the verification of the system design and its subsequent operation must be prepared in writing and be subject to a formal safety review. This review will examine the proposed test for (1) detemiination of the presence of an unreviewed safety question, (2) reactor safety issues, (3) industrial safety issues, and (4) ALARA considerations. In addition, the purpose of the procedure, any prerequisites, and required equipment are documented. Test results are filed formally and are available for review by cognizant regulatory authorities. The fission converter is treated under the MITR y, Quality Assurance Program in the same way that any major reactor component or system is addressed.
- 39. We anticipate that a series of calorimetrics (i.e. Power = Z rhc p AT) will be performed to calibrate the nuclearinstrumentation. The fission converter's heat removal system will be placed on line. The reactor's power will be adjusted to i some low level,500 kW for example. A calorimetricwill then be performed and the result compared against the signals from the fission converter's nuclear instruments. The reactor's power will then be raised to some higherlevel,1 MW for example,and the procedure repeated. The result will be a curve of nuclear instrument signalversus thermal power. This process is iterative. The " initial conservative operating condition" will be documented in the Q/A package that is prepared in conjunction with the startup testing of the fission converter.
- 40. This question is no longer relevant because the TS limit on the CCS operation has been changed to specify the reactivity worth.
- 41. After startup testing, a constant speed, AC motor driven at the 60 Hz line frequency will be used. The reactivity associated with opening the converter control shutter will be determined during the initial startup testing. Proposed TS#
14 l l 6.6.2.2 has nowbeen modified so that the opening speed of the CCS is no longer O an issue.
- 42. "in-core temperature distribution" is changed to " temperature distribution in the fission converter plate"
- 43. The thermal hydraulic limits of the fission converter are set conservatively for a maximum powerof 250 kW. This power level can only be achieved with use of light water, fresh fuel elements and reactor power of 10 MW. The fission converter can operate safely regardless of the reactor power as long as the thermal hydraube limits are met.- Therefore,it is not necessary to specify limits for 5 MW and 10 MW separately. However, there will be an alarm for the nominal operating power. Please see response to question # 34.
- 44. The reference point for water height should be the top of fuel element. This has been corrected in the SER and TS. j i
- 45. Please see response to question #22.
- 46. All calculations previously discussed in the SER that was provided to the NRC l were done for eleven fuel elements. Current calculations show that the SL and :
< LSSS are the same for ten fuel elements for forced convection but the differences ! are quite significant for natural convection (see Appendix E). TS# 6.6.2.1(4) has been modified to state the requirements for each operating condition. >
- 47. Yes. TS# 6.6.2.l(8)is now changed to "...in place and scaled..". i
- 48. "or its equivalent" was used with " cadmium curtain" in TS# 6.6.2.2 as a provision )
of possible material change of the curtain, such as using Boral instead of (or in ! conjunction with) cadmium. The term " cadmium curtain" is now changed to " converter control shutter" or CCS in the SER and associated TS to reflect the function of the shutter without referring to a specific material.
- 49. The technical specification for the CCS closing time has been deleted because a malfunction of the automatic reactor scram on low primary flow is not considered a credible event. Therefore, a surveillance requirement on the closure time for the CCS is not necessary.
- 50. TS# 6.6.2.4 specifies the H2 /D 2 concentration in helium. The MITR is an open system and has an air blanket over the primary coolant that is contained in the core tank. Therefore, TS# 3.4 was written for H2 in air which has a higher 02 concentration.
15
- 51. TS# 6.6.2.6 (4) has been added to address actions that should be taken in the event of out-of-specification water chemistry.
- 52. According to TS# 6.6.3, functional tests of the instrumentation shall be made prior to operation of the fission converter if the converter is not scheduled to be used and use becomes necessary. The functional tests will require bypass of the reactor scrams for fission converter primary Dow and coolant level to prevent a reactor scram during these tests. " neutron Dux level"in TS# 6.6.3 is now changed to " power level"to match TS# 6.6.2.5. Chloride content has been deleted from the primay coolant sampling requirement TS# 6.6.3 (4), the reason being that the conductivity will reDect an increase in the chloride level.
We do not think that a TS for initial startup of the fission converter is necessay. We will develop a startup plan and associated procedures. These documents will be available for review before the initial startup testing.
- 53. The sample assemblies that might be introduced into the fueled region of the converter are subject to TS# 6.6.4.4, which refer to TS# 6.6.2.l(4) and TS# 5.2.
The former states that the requirements for sample assemblies and the latter specifies the criteria for MITR's in-core sample assembly design.
- 54. Wordingin TS# 6.5 has been changedand now it applies to the neutron beams from both the basement medicalroom and the fission converter medical room.
" Cadmium shutter"is the same as the " cadmium curtain". It has been renamed as " converter control shutter"(CCS), please see response to question #48. This has been changed in both the SER and TS for consistency. Please refer to response to question #36 for the CCS control.
- 55. Please refer to reply #36 for the CCS control. The CCS can be operated by any licensed operator, by any operator candidate who is in training and who is under the direct supervision of a licensed SRO, or by a non-licensed person who has completed a written qualification program on the fission converter. In all cases, the reactor console operator has overall authority and his/her permission will be required prior to any CCS operation. TS# 6.6.5 has been added to address the initial startup reporting requirements for the fission converter.
O V 16 .w__ O 1 l APPENDIX A , O O p MIT NRL Fisalen Conv:rt:r Group ~ i Memorandum l To: FCB Group i ! cc: File , From: K. Riley l . Date: 08/12/98 l Re: MCNP criticality calculations of the FCB fuel with the aluminum block iemoved from the fuel tank. l l Several calculations to evaluate the effective multiplication constant (keff) of the fission converter with and without the aluminum block inserted in the fuel tank, and with various coolant and fuel loading conditions have been completed. The MCNP model of the fuel tank matches the most recent design that was l submitted to Artisan Engineering Inc. in January 1998. It includes the removable l aluminum block that is placed between the fuel and the tank wall, on the patient side of the fuel. When the block is removed, it is replaced with either H2 O or D2 O , coolant, as appropriate for the situation being modeled. Spent fuel contains 312 g 285 U per element, and fresh fuel contains 510 g assU per element. The table below summarizes the results of the calculations. It is clear from the results that even with the block removed, using fresh fuel and H2O coolant, the system is still very subcritical and poses no credible risk from a criticality accident. Configuration ken Spent fuel, D2 0 cooling, block in place 0.2636 1 0.0008 l; Spent fuel, D2 0 cooling, block removed 0.3910 1 0.0010 i Fresh fuel, H2 O cooling, block in place 0.5893 1 0.0012 4 Fresh fuel, H2O cooling, block removed 0.6698 1 0.0012 lO i , , ,-n , GJ = . , I l It should be noted that the two reference cases shown in the table above (spent fuel, D 2O cooling, block in place and fresh fuel, H2 O cooling, block in place) differ slightly from the same calculations that were performed by W. S. Kiger [1] that are listed in the SER. Due to differences in the tank and coolant geometry, different k.n values have been calculated. The SER can be updated to reflect these new values, if so desired, but the differences are small and all values are much less than unity. The configurations shown in the above table represent the lowest and highest expected k.n values when the aluminum block is removed. Calculations for other configurations can be performed if necessary.
- 1. W. S. Kiger 111, Neutronic Design of a Fission Converter Based Epithermal Beam for Neutron Capture Therapy, Nuclear Engineers Thesis, MIT,1996.
2 N . i , MIT NRL Finion Ccnv:rt:r Graup l lVl Memorandum l To: FCB Group c cc: File ~ i From: Kent Riley l l Date: 08/31/98 Re: Hot channel factor and power calculations with the aluminum block removed from the FCB fuel tank MCNP calculations have been completed to determine the power distribution for the case of fresh fuel (512 g '85U per element) using light or heavy water coolant, with the aluminum block removed from the FCB fuel tank. The power in each fuel plate ! of each element was tallied, and a hot channel factor was calculated for each case. The MCNP model of the fuel tank matches the most recent design that was submitted to Artisan Engineering Inc. in January 1998. The removable aluminum L(] block that is placed between the fuel and the tank wall is removed for this (/ calculation and is replaced with light water coolant. The output of the MCNP calcu!ation is scaled to correspond to an MITR 11 reactor power of 5 MW. The MCNP filenames to reference for this calculation carry the prefix kjr505 (light water) and kjr506 (heavy water). The table below summarizes the results from this calculation, as well as those calculated by S. Sakamoto (memo 1/13/97) under the same conditions, but with the aluminum block in place. The hot channel factors with the block removed are found to be slightly higher than those with the aluminum block in place. In all cases the plate with the highest power is plate 91, which is the first plate of element number 7. Configuration Hot Channel Total Power Factor (kW) Fresh Fuel, H2O cooling 1.47 i 0.013 125.7 i 0.1 Fresh Fuel, D2O cooling 1.53 1 0.014 104.9 i 0.1 Fresh Fuel, H2 O cooling, block removed 1.55 i 0.007 158.2
- 0.4
- Fresh Fuel, D2 O cooling, block removed 1.57 i 0.007 141.510.4 L
i f () 1 - Augurt 31,1998 lt is also important to notice in the above table that in both cases with the block removed, the total power of the fission converter is significantly higher. An earlier memo (K. Riley 8/13/98) reported a total power of 145.1 kW for the case of fresh fuel, light water cooling and the block removed. The value reported above is correct, as a mistake in the post processing of the MCNP output was discovered for the value reported earlier. For either light or heavy water cooling, fresh fuel, aluminum block removed, and a reactor power of 10 MW, it appears that the nominal anticipated power of 250 kW will be exceeded (283 kW for D2O and 316 kW for H2O). Also attached to this memorandum are plots of the transverse power profile (power per plate) and a summary of the power distribution in each fuel element for the two cases described above. O 2 == _ _ _ _ - _ . . . _ , _ ~ . . . - ... -.~ w - - - _ . - Transverse Power Profile in the Fission Converter Fuel Al block removed, H2O Cooling, Fresh Fuel - 1.400 - 1.200 - N i lg 1.000 - o <, i_ I E 0.800 - ; E I. 5 a. 0.6 0 0 - . i im ' !o3 0.400 O i 0- 0.200 - i ! ! l 0.000 i I I 0 15 30 45 60 75 90 105 120 135 150 165l Fuel Plate Number i i' o O O Transverse Power Profile in the Fission Converter Fuel Al block removed, D2O Cooling, Fresh Fuel g 1.400 5 1.200 - g 1.000 - E 0.800 u. { 0.600 fo 0.400
- n. 0.200 0.000 !
0 15 30 45 60 75 90 105 120 135 150 165 Fuel Plate Number S . e /' Al block removed, D2 O cooling, fresh fuel Element 1 Element 2 Element 3 Element 4 Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) 1 0.975 16 0.602 31 0.730 46 0.951 2 0.793 17 0.594 32 0.727 47 0.950 3 0.695 18 0.594 33 0.739 48 0.959 4 0.627 19 0.589 34 0.769 49 0.969 5 0.620 20 0.597 35 0.759 50 0.964 6 0.588 21 0.593 36 0.779 51 0.981 7 0.570 22 0.600 37 0.823 52 0.965 8 0.569 23 0.605 38 0.804 53 1.001 9 0.558 24 0.620 39 0.820 54 1.006 10 0.556 25 0.637 40 0.799 55 1.028 11 0.559 26 0.632 41 0.827 56 1.025 12 0.559 27 0.629 42 0.855 57 1.017 13 0.565 28 0.650 43 0.895 58 1.046 14 0.587 29 0.693 44 0.912 59 1.053 15- 0.597 30 0.715 45 0.935 60 1.098 TOTAL 9.42 9.35 12.17 15.01 AVERAGE 0.628 0.623 0.812 1.001 PF 1.55 1.15 1.15 1.10 Element 5 Element 6 Element 7 Element 8 Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) d 61 1.131 76 1.181 91 1.346 106 1.087 62 1.091 77 1.134 92 1.214 107 1.063 63 1.088 78 1.140 93 1.167 108 1.041 64 1.082 79 1.128 94 1.145 109 1.004 65 1.068 80 1.084 95 1.123 110 1.027 66 1.057 81 1.105 96 1.127 111 1.018 67 1.067 82 1.118 97 1.126 112 1.007 68 1.078 83 1.106 98 1.098 113 0.977 69 1.093 84 1.145 99 1.078 114 0.963 70 1.080 85 1.152 100 1.101 115 0.948 71 1.057 86 1.158 101 1.052 116 0.945 72 1.104 87 1.138 102 1.065 117 0.926 73 1.092 88 1.163 103 1.077 118 0.914 74 1.118 89 1.170 104 1.065 119 0.927 75 1.132 90 1.258 105 1.081 120 0.940 TOTAL 16.34 17.18 16.86 14.79 AVERAGE 1.089 1.145 1.124 0.986 PF 1.04 1.10 1.20 1.10 O At block removed, D2 0 Cooling, Fresh Fuel Page 1 of 2 5, [ . O l Q Plate # Element 9 Power (kW) Element 10 Plate # Power (kW) Element 11 Plate # Power (kW) 121 0.929 136 0.702 151 0.618 122 0.882 137 0.655 152 0.595 123 0.866 138 0.658 153 0.588 124 0.829 139 0.646 154 0.560 e 125 0.830 140 0.642 155 0.553
- 126 0.809 141 0.623 156 0.559 127 0.772 142 0.622 157 0.554 128 0.769 143 0.619 158 0.571 129 0.739 144 0.600 159 0.574 130 0.748 145 0.594 160 0.587 l
131 0.733 146 0.601 161 0.592 132 0.719 147 0.580 162 0.641 133 0.698 148 0.576 163 0.683 134 0.711 149 0.596 164 0.763 135 0.707 150 0.615 165 0.904 l TOTAL 11.74 9.33 9.34 l AVERAGE 0.783 0.622 0.623 l PF 1.19 1.13 1.45 t O ; I i ! l i i O Al block removed, D 2O Cooling, Fresh Fuel Page 2 of 2 s . e' Al block removed, H2 O cooling, fresh fuel Ci g Element 1 Element 2 Element 3 Element 4 Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) 1 0.932 16 0.669 31 0.878 46 1.078 2 0.722 17 0.666 32 0.852 47 1.069 3 0.652 18 0.672 33 0.854 48 1.056 4 0.598 19 0.671 34 0.880 49 1.091 5 0.543 20 0.688 35 0.872 50 1.106 6 0.536 21 0.710 36 0.905 51 1.096 7 0.546 22 0.719 37 0.908 52 1.105 8 0.575 23 0.735 38 0.903 53 1.093 9 0.577 24 0.748 39 0.915 54 1.128 10 0.585 25 0.732 40 0.936 55 1.107 11 0.596 26 0.766 41 0.944 56 1.146 12 0.603 27 0.762 42 0.966 57 1.182 13 0.627 28 0.770 43 1.001 58 1.183 14 0.637 29 0.820 44 1.032 59 1.202 15 0.671 30 0.858 45 1.082 60 1.263 TOTAL 9.40 10.99 13.93 16.91 AVERAGE 0.627 0.732 0.928 1.127 PF 1.49 1.17 1.17 1.12 Element 5 Element 6 Element 7 Element 8 Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) 61 1.238 76 1.328 91 1.486 106 1.238 62 1.217 77 1.313 92 1.353 107 1.171 63 1.195 78 1.302 93 1.285 108 1.154 64 1.199 79 1.259 94 1.302 109 1.141 65 1.204 80 1.232 95 1.239 110 1.116 66 1.200 81 1.233 96 1.245 111 1.102 67 1.216 82 1.218 97 1.212 112 1.089 68 1.183 83 1.224 98 1.211 113 1.072 69 1.183 84 1.227 99 1.207 114 1.088 70 1.172 85 1.216 100 1.196 115 1.090 71 1.179 86 1.244 101 1.184 116 1.083 72 1.244 87 1.300 102 1.170 117 1.085 73 1.254 88 1.310 103 1.184 118 1.078 74 1.272 89 1.352 104 1.199 119 1.068 75 1.331 90 1.487 105 1.263 120 1.088 TOTAL 18.29 19.24 18.74 16.66 AVERAGE 1.219 1.283 1.249 1.111 PF 1.09 1.16 1.19 1.11 O Al block removed, H2 O cooling Fresh Fuel Page 1 of 2 p~ Element 9 Element 10 Element 11 (.' Plate # Power (kW) Plate # Power (kW) Plate # Power (kW) 121 1.081 136 0.868 151 0.662 122 1.007 137 0.809 152 0.638 123 0.960 138 0.792 153 0.612 124 0.974 139 0.756 154 0.590 125 0.956 140 0.759 155 0.563 126 0.954 141 0.739 156 0.565 127 0.931 142 0.733 157 0.571 128 0.913 143 0.706 158 0.570 120 0.888 144 0.688 159 0.592 130 0.872 145 0.680 160 0.582 131 0.894 146 0.657 161 0.578 132 0.870 147 0.652 162 0.596 133 0.869 148 0.657 163 0.651 134 0.842 149 0.661 164 0.717 135 0.860 150 0.655 165 0.899 TOTAL 13.87 10.81 9.38 AVERAGE 0.925 0.721 0.626 PF 1.17 1.20 1.44 r^s gI .G Al block removed, H2 O cooling, Fresh Fuel Page 2 of 2 4 . .+-m - uaa - h-w= A a 2 A-__u ___ _-am.m._A_, e~. _ m._h -*C -hp _e4@ w s,h-.- a.h-n *e4.4. - 4e+J m& -4.. N O APPENDIX B O O MIT NRL Fisti:rt Ccnv:st:r Grcup (vD Memardm To: FCB Group cc: File From: Kent Riley Date: 08/31/98 Re: Reactivity effect of' B in water shutter system. Regarding Ouestion 26 of the NRC response to the SER for the fission converter facility, the reactivity effect of adding ' B to the water shutter system will be negligible. The water shutter is located approximately 1 meter from the fuel in the fission converter facility. Also, a 0.020"!ayer of cadmium is placed in the beamline, after the filter / moderator, a location which is between the fuel in the fission converter tank and the water shutter in the collimator region. This layer of cadmium will ^ prevent any neutrons that might be thermalized and/or backscattered in the water shutter from reaching the fuelin the fission converter facility. The addition of ' B to ~ the water shutter will have no impact on the flux of thermal neutrons that reach the fuel after scattering in the water shutter and will therefore have no effect on the reactivity of the fission converter fuel. l i l [) % 1 ,u ,_, a- --s ..-on- m - m-, +-~ a-m 4 A-- a-K- ~ ^- " A-"---O-- -A-4--- v-"^- - ' -'-=a+= 0 APPENDIX C O l O , Oi;;y I NUCLEAR REACTOR LABORATORY b AN INTERDEPARTMENTAL CENTER OF kg l MASSACHUSETTS INSTITUTE OF TECHNOLOGY j JOHN A. BERNARD 138 Albany Street, Cambrdge, MA 02139-4296 Achvation Analysis Director Telefax No. (617) 253-7300 Coolant Chemistry Director of Reactor Operatons "#"'"*""' Tel. No. (617) 258-5860 Reactor Engineenng Pnncipal Research Engmeer September 10,1998 MEMORANDUM From: Lin-Wen Hu W To: Fission Converter Files
Subject:
Radiation Doses after a Material Handling Accident in the Converter
- 1. A material handling accident in the fission converter is assumed to occur under the following scenario:
- a. The fission converter operated infinitely at 250 kW followed by 4 days of operation at 50 kW to prepare for refueling (TS# 6.6.2.4 (3a)).
- b. During refueling, a fuel element dropped in the converter tank and the whole element (15 plates) was damaged.
( ! 2. The fission product activity was estimated using DKPOWR. The total decay activity for the fission converter fueled region 10 minutes after shutdown is 3.483x10 5
j Curies. Assume that the fueled region consists of 11 fuel elements, therefore the 4
fission product activity for each element is 3.166 x10 Curies.
- 3. A recent study (Q. Li, Estimate of Radiation Release for MIT Research Reactor During Design Basis Accident, MS Thesis, Nuclear Engineering Dept., MIT,1998) analyzed the off-site radiation release from the MITR during design basis accident i (DBA) using an updated source term. The calculated total fissien product activity in 6
4 melted fuel plates is 7.16 x10, activity for each fission product nuclei is listed in the attached table.
- 4. 'Ihe radiation doses resulted from the above-described scenario are then obtained by using the doses calculated for 5 MW MITR and the ratio between these activities.
Note that for the MITR DBA it was assumed that 4 fuel plated melted while for the fission converter it was assumed that a whole fuel element was damaged. The following results are estimated for two hours after the accident.
MITR (5 MW) Fission Converter (250 kW)
Total whole-body,8 m 644 mR 2.8 mR Total whole-body,21 m 887 mR 3.9 mR
[ l Thyroid 112 mR 0.5 mR V
4 O 5. It is thus concluded that radiation doses resulted from a fission convener material mishandling accident are insignificant (< 0.5%) compared to those for the MITR design basis accident.
l l
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1 Estimate of Radiation Release for MIT Research Reactor During Design Basis Accident i by Qing Li Submitted to the Department of Nuclear Engineering in partial fulfillment of the requirements for the degree of Master of Science in Nuclear Engineering at the MASSACHUSETTS INSTITUTE OF TECHNOLOGY May 1998
@ Massachusetts Institute of Technology 1998. All rights reserved.
Author............................................................
. Department of Nuclear Engineering i
May 8,1998 Ce r t ifi ed by . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
John A. Bernard Director, MIT Nuclear Reactor Laboratory Thesis Supervisor
. C e r t i fi ed by . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Jacquelyn C. Yanch Professor, Nuclear Engineering Department Thesis Supervisor l A c cep t ed by . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Lawrence M. Lidsky Chairman, Department Committee on Graduate Students
Table A.1: Total Core Fission Product Inventory Isotope Half-life A(secd) Yi (%) Qi (x106 Ci) 5MW 6MW 7MW 8MW 9MW 10MW Kr 85m 4.36h 4.41 & 5 1.5 0.6490 0.7788 0.9086 1.0384 1.1682 1.3000 87 78m 1.48 & 4 2.7 1.1700 1.4040 1.6380 1.8720 2.1060 2.3400 !
88 2.77h 6.95 & 5 3.7 1.6000 1.9200 2.2400 2.5600 2.8800 3.2000 l Xe 131m 12.0d 6.68 & 7 0.03 0.0130 0.0156 0.0182 0.0208 0.0234 0.0260 133m 2.3d 3.49 & 6 0.16 0.0692 0.0830 0.0969 0.1107 0.1246 0.1380 133 5.27d 1.52 & 6 6.5 2.8100 3.3720 3.9340 4.4960 5.0580 5.6200 135m 15.6m 7.40 & 4 1.8 0.7780 0.9336 1.0892 1.2448 1.4004 1.5600 {
135 9.13h 2.11 & 5 6.2 0.4130 0.4956 0.5782 0.6608 0.7434 0.8260 '
138 17m 6.79 & 4 5.5 2.3800 2.8560 3.3320 3.8080 4.2840 4.7600 1 131 8.05d 9.96 & 7 2.9 1.2500 1.5000 1.7500 2.0000 2.2500 2.5100 132 2.4h 8.02 & 5 4.4 1.9000 2.2800 2.6600 3.0400 3.4200 3.8100 133 20.8h 9.25 & 6 6.5 2.8100 3.3720 3.9340 4.4960 5.0580 5.6200 134 52.5m 2.20 & 5 7.6 3.2900 3.9480 4.6060 5.2640 5.9220 6.5700 135 6.68h 2.89 & 5 5.9 2.5500 3.0600 3.5700 4.0800 4.5900 5.1000 Br 83 2.4h 8.02 & 5 0.48 0.2080 0.2496 0.2912 0.3328 0.3744 0.4150 84 30m 3.85 & 4 1.1 '
0.4760 0.5712 0.6664 0.7616 0.8568 0.9510 Cs 134 2.0y 1.10 & 8 0.0* 2.8600 3.4320 4.0040 4.5760 5.1480 5.7200 136 13d 6.17 & 7 0.006* 0.4140 0.4968 0.5796 0.6624 0.7452 0.8280 137 26.6y 8.27 & l0 5.9 2.3100 2.7720 3.2340 3.6960 4.1580 4.0200 Rb 86 19.5d 4.11 & 7 2.8&5* 0.6120 0.7344 0.8568 0.9792 1.1016 1.2200 Te 127m 90d 8.82 & 8 0.056 0.0242 0.0290 0.0339 0.0387 0.0436 0.0484 127 9.3h 2.07 & 5 0.25 0.1080 0.1296 0.1512 0.1728 0.1944 0.2160 129m 33d 2.43 & 7 0.34 0.1470 0.1764 0.2058 0.2352 0.2646 0.2940 129 72m 1.60 & 4 1.0 0.4320 0.5184 0.6048 0.6912 0.7776 0.8650 l 131m 30h 6.42 & 5 0.44 0.1900 0.2280 0.2660 0.3040 0.3420 0.3810 131 24.8m 4.66 & 4 2.9 1.2500 1.5000 1.7500 2.0000 2.2500 2.5100 132 77h 2.50 & 6 4.4 1.9000 2.2800 2.6600 3.0400 3.4200 3.8100
! 133m 63m 1.83 & 4 4.6 1.9900 2.3880 2.7860 3.1840 3.5820 3.9800 i 134 44m 2.63 & 4 6.7 2.9000 3.4800 4.0600 4.6400 5.2200 5.8000 89
4 I
Table A.1: Total Core Fission Product Inventory Isotope Half-life Ai (sec-2) Yi (%) Q! (x105 Ci) 5MW 6MW 7MW 8MW 9MW 10MW Sr 91 97h 1.99e-5 5.9 2.5500 3.0600 3.5700 4.0800 4.5900 5.1000 Ba 140 12.8d 6.27 & 7 6.3 2.7200 3.2640 3.8080 4.3520 4.8960 5.4500 Ru 103 41d 1.96 & 7 2.9 1.2500 1.5000 1.7500 2.0000 2.2500 2.5100 105 4.5h 4.28 & 5 0.9 0.3890 0.4668 0.5446 0.6224 0.7002 0.7790 106 1.0y 2.20 & 8 0.38 0.1640 0.1968 0.2296 0.2624 0.2952 0.3290 Rh 103 36.5h 5.27 & 6 0.9 0.3890 0.4668 0.5446 0.6224 0.7002 0.7790 Tc 99tn 6.04h 3.19 & 5 0.6 0.2590 0.3108 0.3626 0.4144 0.4662 0.5190 Mo 99 67h 2.88 & 6 6.1 2.6400 3.1680 3.6960 4.2240 4.7520 5.2800 Sb 127 93h 2.07 & 6 0.25 0.1080 0.1296 0.1512 0.1728 0.1944 0.2160 129 4.6h 4.32 & 5 1.0 4.3200 5.1840 6.0480 6.9120 7.7760 8.6500 Nd 147 11.3d 7.10 & 7 2.6 1.1200 1.3440 1.5680 1.7920 2.0160 2.2500 La 140 40.2h 4.79 & 6 6.3 2.7200 3.2640 3.8080 4.3520 4.8960 5.4500 Ce 141 32d 2.51 & 7 6.0 2.5900 3.1080 3.6260 4.1440 4.6620 5.1900 9 143 144 32h 290d 6.01 & 6 2.76 & 8 6.2 6.1 2.6800 3.2160 3.7520 4.2880 4.8240 5.3600 2.6400 3.1680 3.6960 4.2240 4.7520 5.2800 Zr 95 63d 1.27 & 7 6.4 2.7700 3.3240 3.8780 4.4320 4.9860 5.5400 97 17h 1.13 & 5 6.2 2.6800 3.2160 3.7520 4.2880 4.8240 5.3600 Nb 95 35d 2.29 & 7 6.4 2.7700 3.3240 3.8780 4.4320 4.9860 5.5400 ptoJ2. ( . 04- N t o Table A.2: Values of N'/Ny33 for Neutron-Capture Influenced Isotopes at 4r = 4 x10 a Isotope N!/Ny33 Xe 135 1.05 x10-6 Cs 134 1.4 x10-1 Cs 136 3.6 x10-4 Cs 137 1.5 x10 Rb 86 8.0 x10-d
()
i 1
90
Table 5.1: Total Dose aL5 MW Component of the Dose Dose at 8m (Rem) Dose at 21m (Rem)
Whole-body :
Containment Leakage 1.38 & O2 1.38 & O2 Steel Dome Penetration 6.60 & O3 5.09 & O2 Shadow Shield Penetration 4.81 & O2 2.31 & O2 Air Scattering 2.21 & 01 2.67 & 01 Steel Scattering 3.54E-01 5.32501 Total 0.644 0.887 Thyroid:
Containment Leakage 1.12 & 01 1.12 & 01 The whole body dose which includes gamma and beta dose and the thyroid doses from all sources at the front and back fences are listed below. In the whole body dose, the scattering gamma doses contribute the highest portions, one or two orders of magnitude greater than those from other sources. The results are listed in Tables 5.1 through 5.6. The exclusion area doses as a function of reactor power are plotted in Fig. 5-1. The whole-body dose at 21 meters is greater than that at 8 meters. The thyroid doses at both distances are almost equal.
The regulation gives a limitation of 300 rem for thyroid dose and 25 rem for gc whole-body dose. Our results show that the doses released in a postulated design W"' basis accident of the MIT Research Reactor at a power level of 5 MW up to 10 MW
- f. s. .
are well below the limitation.
i.
l G , l l
80
e t
(x /-
)
T' b V k./ ~~
,' 1 decay power calculation, epri code dkpowr FISSION CONVERTER 250 KW OPERATION nnue = 1 nu eer of nuclides described with fission histories (pos. value = fission rate input for each nuclidel (neg. value = fission density input for each nuclide) ntpow = -2 number of at-power time periods (pos. value = power periods input in seconds)
(neg. value = power periods input in hours) ntcool= 1 number of elapsed cooling times at which decay power desired (pos. value = cooling times input in seconds)
(neg. value = cooling times input in hours)
(sero value - use built-in default cooling times) jtunit= 1 output time units 1 = seconds 2 = hours 3 = days jpunit. 2 output decay power units 1 =mev/s 2 = watts jeunit= 2 cutput decay energy units 1 =mev 2 = joules 3 = watt-h default fractional uncertainties used:
u235 fission recoverable ene;gy .010 pu239 fission recoverable energy .010 u238 fission recoverable energy .010 reactor power level .020 1 s table 1 ln f ins' E fission history
, gg n (cu 2 G C W. u.)
8/s, for use with 1979 ans 5.1 standard 9/s. for use with fits to cinder.10 sumation calculatiens total
+
u238 th232 u233 u235 u238 pu239 pu241 fission delta u235 pu239 u238 tim, thermal thermal fast neutron fast thermal thermal fast thermal thermal rate, timk fission fission fission capture fission fission fission fission fission fission 8/s step \ , hours W
1 1.000E+10 7.903E+15 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+C0 7.803E*15 0.000E*00 0.000E+00 0.000E+00 7.803E*15 2 9.600E*01 1.561E+15 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 1.561E*15 0.000E+00 0.000E+00 0.000E+00 1.561E*15 1
table 11
- q. s @ fo N decay power calculated with the 1979 ans 5.1 decay power standard cooling decay power, watts
+ -
fission products limited heavy elements f.p. decay power uncertainty, watts time, S total gmax u239 np239 total u235 pu239 u238 total **
seconds u235 pu239 u238
- ..-- - - - . . -. .}
Page 1 of 4 TAFE14 9-9-98 4:52p e e
- - _ - _ _ _ - - _ _ _ _ _ _ _ - - _ _ _ - - . _ . . _ - ,n .
(g gg
, f% $
N ' TAFE12 Srf-98 _4:32p..- L- _.____ _ . __
-_. _Jage l_of 4 _ .__ -._. _\ /_ .
6.0000E*02 1.7552+03 0.E00E+00 0.000E+00 1.7 5E+03 1.030 0.000E+00 0.000E+00 0.000E+00 3.356E.01 0.000E+00 0.000E+00 5.388E+01 the ans 5.1 standard limite use of the pulse functions to 10**3 seconds cooling after any power history. however, all of the standard pulse functions result from precise fits to decay power calculated with endf/b-iv data for cooling times to 10**13 seconds, calculated decay powers do not include the absorption correction g.
- gmam upper-bound absorption correction multiplier values are taken from table 10. ans 5.1 decay power standard, which lists gman values at six points per decade (1. 1.5.2. 4. 6. 8.) for cooling times from 1 to 10**9 seconds. values resulting from linear interpolation or estension of ans 5.1 table 10 tabulated values are marked with an asterisk t*) .
- the total uncertainty is calculated from algorithms included in the standard, incorporating the uncertainty in each pulse function. the uncertainty is the energy recoverable from each fissionable nuclide. and the uncertainty in the reactor power.
1 table 111 decay energy calculated with the 1979 ans 5.1 decay power standard cooling integrated decay energy from shutdown. joules
+
time. fission products limited heavy elements
+
seconds u235 pu239 u230 total u239 np239 total 4
6.0000E+02 1.239E*06 0.000E+00 0.000E+00 1.239E+06 0.000E+00 0.000E+00 0.000E+00 calculated decay energies do not include the effects of neutron absorption.
1 table iv decay power calculated with pulse function fits to cinder-10 sumation calculations using processed endf/b-v data, fission product decay power. vatts
+
cooling th232 u233 u235 u238 pu239 pu241 tim. fast thermal thermal fast thermal thernal seconds fission fission fission fission fission fission total
+
6.0000E+02 0.000E+00 0.000E+00 1.806E+03 0.000E+00 0.000E+00 0.000E+00 1.8063E+03 calculated decay powers do not include the effects of neutron absorption.
1 table v decay energy calculated with pulse function fits to cinder-10 summation calculations using processed endf/b-v data, fission product decay energy. joules 4
cooling th232 u233 u235 u238 pu239 pu241 time. fast thermal thermal fast thermal thermal seconds fission fission fission fission fission fission total
+
6.0000E.02 0.000E+00 0.000E+00 1.232E+06 0.000E+00 0.000E+00 0.000E+00 1.2322E+06 calculated decay energies do not include the effects of neutron absorption.
I table vi decay activity calculated with pulse function fits to cinder-10 sumation calculations 3 using processed endf/b-v data.
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g 19 6.60 7.50 0.000E+00 0.000E*00 1.376E-03 0.000E*00 0.000E+00 0.000E+00 1.tT61E-03 total 00 7.50 0.000E+00 0.000E+00 9.343E+02 0.000E+00 0.000E*00 0.000E+00 9.3426E+02 *
- the swa of beta and gamma group spectra may dif fer slightly f rom total >
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i
.i.
- m. . . , -em m-- _ e, we .- -.m.e w
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---m--- -,w---- - - - - - - - -- - --
O APPENDIX D O
O
,G<]e?;p l NUCLEAR REACTOR LABORATORY AN INTERDEPARTMENTAL CENTEA OF MASSACHUSETTS INSTITUTE OF TECHNOLOGY kg#/ -
JOHN A. BERNARD 138 Albany Street, Cambnoge. MA 02139-4296 Activation Analysts Duector Tehetax No (617) 253-7300 Coolant Chemistry Director of Reactor operations Nuclear Medicine '
Pnncipal Research Enginee, Tel. No. (617) 58-5860 Reactor Engineenng August 17,1998 l
MEMORANDUM l From: Lin-WenHu W To: Fission Converter Files i
Subject:
Rev. Complete Loss of the Fission Converter Primary Coolant flow with !
Automatic Cadmium Curtain Closure l
- 1. This calculation was updated using the calculated fission converter power as a function of the cadmium shutter height. A linear power profile was used previously.
- 2. The calculated was performed using MathCAD. All other assumptions remain the same as stated in the memo dated February 29,1997.
'Ihe calculated maximum fuel temperature is 139.2 C at 20 seconds after initiation of
(]/
\'
3.
the transient. The fuel temperature then decreases because of the decreasing fission converter power. Bulk boiling is predicted to occur about 10 seconds after the initiation of the transient. The equilibrium quality increases to about 0.39 at 60 seconds.
1 l
l v l
g/fi / c y 1
1 Fission Converter Complete Loss of Primary Coolant Flow Analysis assumptions: l
- 1. Step change of Primary Coolant Flow to zero !
- 2. Reactor fails to scram automatically, Cd curtain starts to drop after 1 s instrument delay time.
- 3. Initial power is 250 kW, primary flow is 100 gpm,llot channel factor = 1.53 l
- 4. No convection cooling I
- 5. Ilent capabilities of fuel, clad, and coolant in the core region are the only heat sink.
- 6. Neglect the heat transfer to other structure materials 1
thermal conductivity (W/m C) l kfuel = 42.1 kclad :186 kerud .: 2.08 kl : 0.68 ky : 16.1 10
thickness (m)
-2 dfuel = 0.0381 10 delad .= 0.050810 derud : 2.5410 dgap = 0.11210
mass (kg)
Mfuel .= 0.083 Mclad = om6 2 2 coolant channel volume (m^3)
Vchannel := '
2 heat transfer area on one side of fuel plate (assume no fin effect)(m^2)
Area .: 0235 2
density (kg/m^3) pl(T) : 1000.1 + 0.0026863 T- 0.0054424 Y + 1.232410 T' pv ; 0.7 heat capability (W/kg C) (Ref. McGuire's Thesis) ll20.9 T+ 2.2210' cpclad(T) : " '
T) -
3675 2712.6 enthalpy (J/kg) hg = 268210' hf : 444.4610' hfg : hg - hf 1
effective heat transfer coefficient between fuel and coolant 1 1 U(Xe) = if Xe<0.0, I dgapi,. dgap dfuel delad derud 2 dfuel delad derud 2 ikfuel kclad kerud kl f , kfuel kclad kerud (Xe kv + (1 - Xe) k]),
t_ shut c 60.0 Cd curtain closes in 1 minute U(0.0) = 1.18 10' U(1.0) = 28.73 The power changes as a function of shutter position is obtained from Kiger's thesis 164 closing rate Rate =
t_ shut tl.= 12 = 11 = 18.293 t2 = 41.707 Rate Rate
~
P_ shut (t) .= if 15t1,f78 - t- 10',if tst2, 74 - (t - t1)- 10',2 10'
\ 11 , (t2 - 11),
tg :k k := 0. 60 100 , , ,
g i O
P.sbut Q so - -
1000 I '
0 I b I 0 10 ;* 30 40 50 60 I
k 3
P(1,t_ shut) = if t $t_ shut + 1,if 151.0, 5 0 1.53, 1.53 P_ shut (t - 1) ,2.5 11152 (111527810' l.
j :1. 6100 dt : = 0.01 tj .= (j - 1 ) dt l
Initial conditions Two .: 64 H w ,::4193 Tw, Xe, := 0.0 l (this is the hot channel outlet temeperature corresponding to the steady-state average core outlet temeperature of 55 C)
Tf, - Tw, + 3 0.0,t_ shut) Tf, = 94.94 fuel temp increases upon loss of flow y/ U(0.0) Area
i (x
(_ ') Results 140 g g g g g l
120 Tf.
L iOO i
l i i i i i i 0 10 20 30 40 50 60 70 ($) !
'i l
Figure 1 Fuel Temperature l l
V\
'u./ 120 i i i i i i i l
iOO Tm)
~
80 - !
I ' '
60 0 10 20 30 40 50 60 70
($ )
'i Figure 2 Coolant Temperature 1
I l
l
\x.J
I l
e i
rn .
( mass j = {pl(Twp-(1 - Xe j ) r pv Xe] Vehannel
. \/
l 0 045 , ,
l 0.04 -
1 0.035 -
1 maar
~j 0.03 -
)
0.025 -
]
l I I I I I I 0.02 O 10 20 30 40 50 60 70
- t. (
J Figure 3 Coolant Mass in the channel
(
A) 85 5 i i i i i i I
1000 -
u$ ;
l 10
- 500 -
' I I I I I 0
O 10 20 30 40 50 60 70
'i (>)
Figure 4 Coolant Specific Enthalpy l
l O
f v)
~-
e (3
(
w/
49 --
30 - -
Hw.
J mast j 20 - -
10 -
1 I I I I I O 10 20 30 40 50 60 70 8.
J Figure 5 Coolant Enthalpy
./N' t, 1 i i i i i i w/
i i
1 i
l xe _ _
j
__j o.3 f i i f f 0
0 10 20 30 40 30 00 70 1 1 Figure 6 Coolant Equilibrium Quality i
6 V
i i
l
(
(
\
. Check for total energy balance Total energy deposition
- t_ shut i
Qd : P(t,t_ shut) dt Qd = 3.55510'
.O I-Total energy absorbed in the system l
- Tf,,oo ofuel = (Mfuel cpfuel(T) + Mclad cpciad(T)) dr L
Tf, 6100 Qcoolant = ((Hwj - Hw)_3)-(Xef pv, (1 - Xej ) pl(Tw,))) Vchannel l j=1 i Qa :: Qfuel + Qcoolant j' Qfuel = 877.443 Qcoolant = 3.471 10' Qa = 3.55910" l
Dev : 1 - -Qd Qa j
(.
Dev = 8.932* 10 i 1
Write results to a data file ;
i :0. 610 A,,, :: t, 3, A3,3 :Tf,. ,o Ai ,2 : Tw3 ,n A 3,3 : mass j.in A g,, = Hw, jo mass, ,o 1
l A i ,5 = Hw33, A g,,': X e .3 g ,
WRITEPRN(FC) : A i
v 0
L Solve for energy equations dt Tfj _ , + ( P(tj ,t_ shut ) - UIXe
' ' j , , i Area-(Tf j _ , - Tw. 3
_ , ,l --
1 Mfuel cpfuel! Tf, _, , ; 4 Mclad cpcladi Tf, _ 3 , .
Tf2j
- -~-
dt liw -
Hw) .
j~ ' + Vchannel-(Xe j - _3 pv+ (1 - Xe _3) pl(Tw) ,)] U(Xel' ' J Area-(Tfj ~ 3 -
j Tw) 11w )
i liw.J ' ' 5hf, j_,106) ,
[ Xej 4193 Ilw;~'- hf i 11wj _ ,5hf,0.0,i 51.0, liw.J ' ' - hf 3i .
hfg hfg ,3,o/ , .l I
O n
b
O .
4 l
l APPENDIX E O ,
6 O ,
(?%j NUCLEAR REACTOR LABORATORY AN INTERDEPARTMENTAL CENTER OF s
f)
MASSACHUSETTS INSTITUTE OF TECHNOLOGY 4,,' ,,, s'/
JOHN A. BE RNARD 136 Albany $1reet, CambnoDe, MA 02139 4296 Activetion Analysis Director Telefan No. (617) 253-7300 Coola'-A Chemstry Drector of Reactor Operatons Nucker Medicine Tel. No. (617) 258-5860 Pnrupal Rer,earch Engineer Reactor Engineenng September 28,1998 MEMORANDUM From:
To:
Lin-Wen Huh Fission Converter Files
Subject:
Thermal Hydraulic Limits with Ten fuel Elements Operations
- 1. The fission converter operation with one sample assembly in the fueled region (or ten fuel elements) has been analyzed based on the thennal hydraulic limit calculations. It is assumed that no excessive bypass flow is caused by the sample assembly. The LCO for flow disparity (Fr x dr 20.80) should be satisfied. The LCO for the hot channel power generation (Fac x Fp 5;1.53 ) also applies.
- 2. For forced convection, there is not significant change in SL and LSSS for ten fuel elements. For 300 kW, the AT is less than 0.2*C for both SL and LSSS. This is
('"') because both the flow rate and the power increase in the hot channel. The results are attached with the memo. It is thus concluded that the thennal hydraulic limits for forced convection remain the same for 10 or 11 fuel elements.
l
- 3. For natural convection, the AT is about 5'C for LSSS. This is because the natural convection flow rate is driven by the inlet / outlet temperature difference. With 1 reduced total flow area, a higher buoyancy force is needed to offset the higher friction pressure drop. The current SL and LSSS for natural convection thus cannot be used for 10 fuel elements.
p)
(
w/
[~N /] (N
, .- _ kv -
k
..;w b Ce L R &O hk I# a 47 [. O rv\ [ d :- 2 . I \w/
. FC LIMITS CALCULATIC;1 (new flux dist, new coo 1&nt H, 45 gpm) 7/19/97
( [g ,_ , )N )
g .}
50.0e3,25.2.2.83,3.77.2.10 ' LSSS coolant height is 2.10m 1
0.1.0.1 0.
- 0. 0.
1.87.0.41,2.95 0.68.0.5136,1.492 pump coast down 2.03e-3,9.3e-3,0.0508. 1.02,0.0.1 HL primary 3.8870e-5,1.6792e-4.7,04e-3,0.0,7.30,500 RX primary 2.03e-3,1.8e.2.0.0508.1.02,0.0.2 CL primary 0.134,0.135,0.363 1.007,0.0.2 DC 1 0.014,0.037,0.119. 2.65.0.0.2 DC 2 0.039,0.027.0.162,-0.379,0.0.2 DC 3 0.00258,0.00684,0.04064,2.65,0.5.2 Bypass channel 1.2490e-4,0.2434e-5,2.1864e-3,0.66,2.05,150 Core 0.114,0.226,0.187,1.99,0.0.1 FG 0.357,0.359,0.484,1.007.0.0.1 Mixing Area 0.032,0.427.0.203. 7.08,4.58,1 CL secondary g 9.003e-5.3.8895e.4.3.010e.3.0.0,7.3.500 HX secondary 0.032.0.468,0.203,6.97.2.17,1 RL secondary 11.7.62e.4,5.08e-4,2.54e-5,0.05588,0.5683,1.9 0.920$ t 0.9205 2.0E-4 1.0.0.0 1.53 I for D20 0.852,0.892,0.989,1.146,1.174.1.137,1.08.1.039,0.882,0.809 Flux distribution 0.852,0.892,0.989,1,146,1.174.1.137,1.08,1.039.0.882,0.809 g
1.107,1.05,1.063,1.042.1.027,1.019.1.019,1.032,1.123,1.313 Local Peaking '
1.107.1.05.1.063,1.042,1.027,1.019,1.019,1.032,1.123,1.313 0.864 i MITR.11 data 1.154,1.265.1.123 I new e-factors 1/31/97 1.5 i
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1 APPENDIX F i
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4 9
MIT NRL Fizzian Ccnvart:r Group i.
l v.
- 3 Ma11aaidt:11
- l. I L To: FCB Group .
l l
L cc: File j i
From: Kent Riley l
r I oste: 12/02/98 Re: Calculation of Nuclear Hot Channel Factors i
l, l
K' A Monte Carlo model of the FCB fuel tank and beamline, developed by W.S. Kiger (1), has been extensively used to carry out the neutronic design for the MIT Fission Converter Based Epithermal Neutron Beam. This model(developed using MCNP i
! 4a) contains accurate and detailed geometry for all of the major FCB components, including the fuel elements, coolant, fuel tank, as well as beamline components l such as the filter / moderator, collimator and photon shield. The MCNP code makes ,
use of continuous energy cross section libraries; energy group approximations are not required. A portion of the MITR-il reactor is also included in the model to O
! accurately account for any albedo effect from the graphite a.nd D2 O reflector region.
The source term that drives the FCB model is generated from a similar Monte Carlo model of the 5 MW MilR-Il reactor. Developed by E. L. Redmond 11 g), this model includes the full MITR-ll reactor core (with detail in each fuel element), the light
- . water coolant, the heavy water and graphite reflector regions, as well as some of ,
the MITR-Il experimental facilities. The MITR-ll model has been benchmarked via l
=
thermal neutron flux measurements in several experimental facilities of the MITR-II, including the region in which the FCB will be installed. W.S. Kiger made use of the MITR-il model to determine the neutron source in this region. This source, which is I from the SMW MITR-il, has been used for many design calculations, including the calculation of hot channel factors reported in earlier memoranda (K. Riley 8/31/98, S. Sakamoto 1/13/97).
Since an MCNP model of the proposed 10 MW MITR-Il core is not readily available, the reported results for the FCB driven by a 10 MW core have simply been scaled from the 5 MW calculations, where applicable. For example, the magnitude of the
- power and the flux levels in the FCB are expected to increase in proportion with the power of the MITR reactor core. Although the intensity of the source used in the l
FCB model will change when the reactor is changed to 10 MW, the energy and i spatial distribution of the neutrons is not expected to appreciably change. The shape of the power profile in the fuel plates for the FCB is therefore not expected to change with an increase in the power of the reactor. The hot channel factors for a o
C 1
10 MW MITR core are therefore identical to those that have been calculated for the O 5 MW MITR core.
l
- 1. W. S. Kiger lll, Neutronic Design of a Fission Converter Based Epithermal Beam for Neutron Capture Therapy, Nuclear Engineers Thesis, MIT,1996. l
- 2. E. L. Redmond 11, J. C. Yanch, O. K. Harling, ' Monte Carlo Simulation of the Massachusetts Institute of Technology Research Reactor, Nuclear Technology, : j 106, 1-14, (April 1994).
i i
l l
O 4
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t APPENDIX G
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hl N
NUCLEAR REACTOR LABORATORY '
AN INTERDEPARTMENTAL CENTER OF MASSACHUSETTS INSTITUTE OF TECHNOLOGY k hl ~
Actwation Analysis 138 Albany street, Cambndge, MA 02139 4296 Coolant Chemistry JOHN A. BERNARD Teletas No. (617) 253-7300 *' "'
Director Director of Reactor Operations Tel No. (617J258 5860 Reactor Engineermg Pnnctpel Aer,earch Engineer November 24,1998 l
l i
MEMORANDUM l From: Lin-Wen Hu hy a To: Fission Converter Files
Subject:
Determination of SL and LSSS 1.
The fission converter safety limits (SL) and limiting safety system settings (LSSS) were determined based on onset of significant voiding (OSV) and onset of nucleate boiling (ONB), respectively. Note that even the actual SL is defined Ittoisprevent fuel overheating, OSV is chosen conservatively because it is easier to calculate.
believed that adequate margin exists between LSSS and SL so that automatic protective actions will terminate the abnormal situation before a safety limit is l reached. The following analyses were performed to support that argument. i V 2. For forced convection operation, the most limiting transient is loss of primary coolant flow. It is assumed that the reactor scrams in 2 seconds after the LSSS l (300 kW,60 C, and no forced flow) is reached. One second is due to instrument l
delay, and one second due to shim blade drop. Calculated results show that the j maximum fuel temperature is about 112 C, significantly below the cladding softening temperature 450 C. The coolant temperature in the hot channel will rema below the saturation temperature. !
3.
For natural convection operation, the most limiting transient is an overpower i
condition. It is assumed that the operator raises the reactor power at 100 s period inadvertently and cause the CCS to close automatically because of high power. The initial condition is assumed to be 20The kWfission and convener 60 C (LSSS) power isand a one-s instrument delay time is taken into account. )
calculated based on the effects of the 100 s period and CCS closure. The maximum calculated fission converter power is 23 kW and it occurs at about 19 seconds. This l l
is lower than the SL power of 27 kW at 60 C. Note that the change in coolant temperature (Tms) during this transient is insignificant because of the large cool volume in the fission converter tank. It is therefore assumed constant in this analysis. l
- 4. The calculations are attached with this memo.
3 (a 1
l Fission Converter Complete Loss of Primary Coolant Flow Analysis assumptions:
- 1. Step change of Primary Coolant Flow to zero
- 2. Reactor scrams automatically in 2 seconds (Is instrument delay, is blade drop)
- 3. Initial power is 300 kW, hot channel outlet temperature is 69 C (this corresponds to average core outlet temperature 60 C)
- 4. Ilot channel factor == 1.53 l
- 5. No convection cooling '
, 6.11 cat capabilities of fuel, clad, and coolant in the core region are the only heat sink.
- 7. Neglect the heat transfer to other structure materials thermal conductivity (W/m C) kfuel . : 42.1 kclad :186 kerud = 2.08 kl := 0.68 kv = 16.1 10
thickness (m)
-2 -2 dfuel : 0.0381 10 delad = 0.050810 derud : 2.5410 dgap : 0.11210 mass (kg)
Mfuel - S- Mclad g6 2 2 coolant channel volume (m^3)
V d , 8.24310
2 heat transfer area on one side of fuel plate (assume no fin effect)(m^2) 0.0635 2
density (kg/m^3)
~
I pl(T) = 1000.1 + 0.0026863 T- 0.0054424 T* + 1.232410 ' T' pv = 0.7 heat capability (W/kg C) (Ref. McGuire's Thesis)
_1120.9 T + 2.2210' ' ^'
7 cpclad(T) -
3675 2712.6 i
4
" " EI hg = 268210' hf 444.4610' hfg : hg - hf
~
's '
, effective heat transfer coefficient between fuel and coolant ,
I '
U(Xe) :i!Xe<0.0, / dgap ,( dgap l
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! dfuel. _delad
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- _ _ . +
i ikfuel kclad kerud kl ' ; kfuel kelad kerud (Xe kv + (1 - Xe) k]) , ,
i 3
t_ scram = 2.0 Reactor scrams in 2 seconds U(0.0) - 1.18 10 U( 1.0) - 28.73 .
l P(t.t_ scram') . if t' t,_ scram, 1.53,0.0k 11152 l
assume FC power changes linearly after is instrument delay time l
j z 1. 300 dt 0.01 tj = (j - 1) dt Initial conditions l
Two ;69 11w, 4193 Two X e , ; 0.0 -
(this is the hot channel outlet temeperature corresponding to the steady-state average
\
core outlet temeperature of 60 C)
Tf,:Tw.59S"*)
o Tf, - 106.128 fuel temp increases upon loss of flow U(0.0) Area l l
Solve for energy equations !
I lTfj i , l P, tj ,t_, scram ; - U Xej _ i PArea t Tfj _ i - Twg ,jl-t J
!Mfuel cpfuel Tfj_ ,) + Mclad cpelad(Tfj , ,1 Tfj dt i 11w -
j j 11wj j_3 + Vcharmel-(Xe j - _ i pvs- j(1 - Xe _3) j _pl(Tw j{ UIXe _ , yArea-(Tf ,3 - Tw)_ ,)
/ I
! Twj Hw; _ 3 l
if;Hw j_3 5 hf, ,106 Xe 6 4193 /
la
/ iHw hf Hw hf I illHwj_g ihf,0.0,ifi -- j _ ,- - - .1.0, - - j , ,
,1.0 :
- g hfg hfg '
i l
l-
2 s~.
Q,,I Results ils , , , , ,
110 TT L
105 I I I I I gg O 0.5 1 1.5 2 2.5 3 j (b)
Figure 1 Fuel Ternperature f
'L i i i i i 80 Tm) 70 - -
g I t i l 1 0 0.5 1 1.5 2 2.5 3
', (5)
Figure 2 Coolant Temperature N. ,
i O Fission converter maximum power durine an overpower transient l
t_ shut 60.0 (time to fully close the CCS) '
l
\
The power changes as a function of shutter position is obtained from Kiger's thesis closing rate Rate : (constant speed) t_ shut 1 11 - 12 t1 = 18.293 t2 = 41.707 Rate Rate i P_ shut (t) z if titi,i78 - t- }.10',if 1E12, 74 - (t - t1)- 103 ,2 10'
[ \ 11 / [ (12 - t!). ,j k .: 0. 60 t g =k 100 , g , , ,
]
P,,sbut (tg )
I
, o e I e c 0 10 20 30 40 50 60
'k [$ )
Normalize the power to 20 kW P(t,t_ shut) ':i t 5t_ shut + 1,if t 51.0,20 10', ,200 P_ shut (t - 1) 1 78 10' }, y 20 , , , ,
13 -
P, tg,1,,,she ,
l 10 -
3000 l -
3 -
l l
0
! ' I I O 10 20 30 40 50 60 !
(b)
L 1
l l
l
. .. . _ _ . . i
j l
1 l
8 J
, l
?
l Fission converter power with reactor at 100 s period and CCS closing i
i ,
P_nc(t,t_ shut) : P(t,t_ shut) exp .
+
100 I 30 g ; g j
i l
20 i P_nc t ,t_sbus .,
.. g .. .
10 i
1 0 40 50 60 0 10 20 30
'k (b) !
i 15. 22 t, e i P_ncI t,,t_ shut i 1000 22.325 22.483 22.643 22.603 22.965 22.494 21.748 20.984 maximum fission converter power during this transient is 23 kW i
i l
( :
3.mJ a .W-4 a 24.4 44_Jh._a_ .e,u.aums.AA. J.. 4 a*%..mm_ 4 h 4 4-w, ,4 a.W-ama+- . _ -__.m J 4+_4_A~,4-Mam_%..udawaA.auE e_ . .
- m_ h. a. .ad.w. _m.. 4 ese,- A c.-aa _,6aau-O l
I i
i APPENDIX H O ,
1
!O
, s
- MIT NRL Fissi:n Conv:rter Group l O
m_am To: FCB Group a cc: File Fmm: Kent Riley Does: 12/04/98 Re: Estimates of dose rates inside FCB medical room with failure of one shutter MCNP calculations of the dose rate inside the medical room for all possible shutter configurations are currently being completed. Though these calculations are stillin progress, this memorandum provides an estimate for the dose rate at the patient position with the reactor operating and the failure of one shutter (two shutters closed). MCNP calculations with each shutter closed (and the other two open) have been completed. The values in the table below have been estimated by attenuating the single shutter closure results by the amount appropriate for the '
other shutter.
O Dose Rate at Patient Position Condition of FCB Shutters (MITR-il at 5 MW) (Rom /hr)
CCS Open, Water and Mechanical Shutters Closed 0.1 Mechanical Shutter Open, Water Shutter and CCS Closed 15 Water Shutter Open, CCS and Mechanical Shutter Closed 0.2 1
Although the MCNP results for single shutter closures are statistically accurate to '
10% or better, the uncertainty in the above numbers is much higher. The attenuation of two shutters together can be different than the product of their attenuation when acting alone. It is also important to remember that the above ,
numbers are calculated in the path of the direct beam; general area room dose rates and dose rates outside the beam path will be much lower. Nevertheless, the ,
I table above can be used for guidance in identifying failure scenarios that might prohibit access to the room. Failure of the mechanical shutter may prohibit access O
! 1
- Decsmb:r 4,1998 .
l to the room (without a reactor scram). However it is important to remember that if
- the mechanical shutter can even be partially closed, dose rates will be dramatically reduced. In any case, a reactor scram will sufficiently reduce dose rates to permit t room entry. j 1
l I
l
\
f P
i l
1 i
l 2
_, . - - - - - - - . - -