ML20073H130
| ML20073H130 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 09/29/1994 |
| From: | Bernard J MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20073H135 | List: |
| References | |
| NUDOCS 9410050130 | |
| Download: ML20073H130 (13) | |
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C2 NUCLEAR REACTOR LABORATORY AN INTERDEPARTMENTAL CENTER OF MASSACHUSETTS INSTITUTE OF TECHNOLOGY O. K. HARUNG 138 Albany Street, Cambodge, Mass. 02139 4296 J. A. BERNARD, JR.
Director Telefax No. (617) 253-7300 Director of Reactor Operations Telex No. 92-1473-MIT-CAM Tel No. (617)253-4202 29 September 1994 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Response to Request Dated 08/31/94 for Additional Information Regarding Request for Adjustment of Facility Operating License No. R-37 for the Massachusetts Institute of Technology Research Reactor (MITR); Docket No. 50-20.
Gentlemen:
On 03/31/94, the Massachusetts Institute of Technology submitted a request that Facility Operating License No. R-37 be extended to April 24,2001. On 08/31/94, the U.S. Nuclear Regulatory Commission requested additional information concerning that proposed license extension. Enclosed is our response.
Sincerely, ik A l ohn A. Bernard, Ph.D.
Director of Reactor Operations MIT Research Reactor JAB /CRM Enclosure cc:
USNRC - Project Manager, NRR/ONDD USNRC - Region I-Project Scientist, Effluents Radiation Protection Section (ERPS)
FRSSB/DRSS ldlEj' 9410050130 940929 DR ADOCK 05000020 8
PDR g
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Ouestion #1 - Containment Building Surveillance and Testing The MIT Reseamh Reactor is equipped with a full containment. The major penetrations am the main and basement personnel air locks, the intake and exhaust ventilation ducts, and a number of small lines for CO2, helium, and compressed air. In addition, there is an air lock of sufficient size to accommodate large vehicles. This air lock, which is referred to as the ' truck lock,' has in recent years been downgraded so that it is used only when the reactor is in a secured condition.
The containment building is protected against both under and overpressure. The former is achieved through doubly redundant sets (i.e., two sets each with two breakers in series) of vacuum breakers. These open to admit air into the building in the event of an undegressum. Overpressum protection is provided by a pressure relief system that must be manually operated.
The material condition of the containment building is maintained below ground level through the use of sacrificial zinc anodes (cathodic protection system). Protection above ground is provided by periodic painting of the building.
There are a number of test procedures that are specific to the containment building and/orits penetrations. These are listed in Table 1-1.
MITR Technical Specifications TS #3.5, " Reactor Containment Integrity and Pressure Relief System," TS #4.2 " Containment and Pressure Relief System Surveillance," and certain parts of TS #4.3, " Reactor Control, Safety, and Radiation Monitoring System Surveillance" pertain to the containment building. The first of these specifies the allowed building leak rate (1% of the contained volume per day per pound of overpressure), the efficiency for iodine removal (10%) of the pressure relief system filters, the setpoint (0.1 inches of H2O) of the interlock that precludes reactor startup unless the building is slightly below atmospheric pressure, the setpoint (atmospheric pressure 0.1 psig above building) for the vacuum breakers, and the setpoint (3 psig) for the i
building overpressure scram. Specifications #4.2 and #4.3 specify surveillance frequencies.
Table 1-2 shows the MITR procedures that are used to fulfill the requirements of the Technical Specifications that relate to containment building surveillance and testing. A comparison with Table 1-1 shows that many of the procedures that are performed relative to the containment building and/or its penetrations are EDI technical specification requirements.
Results of the containment building surveillance and testing for the last five years have been as follows:
(1)
All surveillance and test requirements as specified by both the Technical Specifications and internal MITR Procedures (Tables ;-1 and 1-2) have been performed at the specified frequency. Moreover, all setpoints and/cr limiting conditions for operation were achieved.
(2)
The last five building pressure tests are discuned here. Each of the penetrations such as the air locks, ventilation ducts, and vacuum breakers contain two closure devices that are mounted in series. Accordingly, the building pressure test is run annually in two different configurations so as to test each closure device separately. The status of the affected i
devices in the two configurations is given in Table 1-3. The test is performed by first placing the building in the desired configuration and then pressurizing the building to 50 inches of 1120 (~ 2 psig). The volume of the air added to the building in order to maintain 50 inches of H2O over a period of several hours is measured as are temperatures throughout the building, relative humidity, the cycling of the reactor gasholders, and the 1
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Table 1-1 l
MITR Surveillance Tests Relevant to the Containment Building and/or Its Penetrations I#
l Title l Freauency I
r 6.1.2.1 Building Pressure Test Annual 6.1.2.2 Main Ventilation DamperInspection Semi-annual 6.1.2.3 New or Repaired Containment Penetration Leak Test As needed 6.1.2.4 Test of Vacuum Breaker Set Points Annual 6.1.2.5 Charcoal Filter Efficiency Test Annual 6.1.3.5 Building AP Indicator and Recorder Calibration Annual 6.1.3.6 Building Over-Pressure Scram Quanerly 6.1.4.3 Damper Closing Time Annual 6.2.1 Main Personnel Air Lock Gaskets Deflated Scram Quanerly 6.2.2 Basement Personnel Air Lock Gaskets Deflated Scram Quanerly i
6.3.4 Fan Interlocks and Alarms Semi-annual f
6.5.1 Cathodic Protection System Test Semi-annual 6.5.10.2 Vacuum Breaker Calibration Annual i
3.1 Startup Checklists (Startup Interlock Tests)
Done as part of startup checklists 7.1.5 Damper Accumulator Charging and Actuator Inspection Quanerly 7.4.2.1 Solenoid Valve Replacement Annual....-, _
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Table 1-2 MITR Procedures Used to Satisfy Technical Soecification Reanirements Relevant to the Containment Building and/or Its Penetrations Technical Specification MITR Procedure TS #
Requirement Frequency Procedure Frequency 3.5.2/4.2.l(a)
Building Leakage Annual 6.1.2.1 Annual 4.2.l(b)
New Penetrations As needed 6.1.2.3 As needed 3.5.3/4.2.3 Pressure Relief Annual 6.1.2.5 Annual 3.5.4/4.3. l(j)
Building AP S/U Interlock Prior to S/U Stanup Checklists Prior to S/U 3.5.5/4.2.2 Vacuum Relief Annual 6.5.10.2 Annual t
3.5.6 Overpressure Scram 6.1.3.6 Quanerly 4.3.2(e)
Containment AP Annual 6.1.3.5 Annual Calibmtion I
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m Table 1-3 Confieurations for MITR Buildine Pressure Test Device Configuration #1 Configuration #2 Main Intake Damper Open Closed Auxiliary Intake Damper Closed Open Main Exhaust Damper Open Closed Auxiliary Intake Damper Closed Open Main Personnel Air Lock Outer Door Closed Open Main Personnel Air Lock Inner Door Open Closed Basement Personnel Air Lock Outer Open Closed Door Basement Personnel Air Lock Inner Closed Open Door Truck Lock Outer Door Open Closed Truck Lock Inner Door Closed Open*
- Closed in recent years because truck lock use is restricted to times when the reactor is secured.,
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barometric pressure. From this information, both the observed building leak rate and the total leakage are calculated. These are then compared to the allowed values which are specified in Technical Specification #3.5. (Note: An allowed leakage of 1% of the contained volume per pound of overpressure per day equates to about 155 cubic feet per hour.) Data are taken every half-hour. Table 1-4 summarizes the final results (last set of data in each test unless otherwise specified) for the last five years. (Note: The 1994 building pressure test is currently scheduled for November 1994.) In each instance, the observed leak rate has been well below the allowed.
(3)
The cathodic protection system was replaced in its entirety in 1992. !
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1 Table 1-4 Summary of MITR Buildine Leak Rate Tests Configuration #1 Configuration #2 Observed I
Allowed Observed Allowed Observed Total Allow ~i i Total Observed Total Allowed Total
' leakageO)
Leak Rate Izakage Leak Rate leakage (l) leak Rate leakage Ixak Rate Year (cu ft/hr)
(ca ft)
(cu ft/hr)
(cu ft)
(cu ft/hr)
(cu ft)
(cu ft/hr)
(cu ft) 1989(2) 102.3 i 2.9 580 i 19 161 925 94.9 3.3 540t20 164 901 1990 30.4 i 7.0 257i47 156 965 38.7 i 4.7 235 i 15.8 159 985 1991 45.415.1 303 62 154 1420 32.9 4.3 181 23 155 800 1992 45.1 3.6 338 i 30 157 1136 98.0 7.8 426i35 156 725 1993(3) 31.8 i 6.9 165 i 36.4 153 790 16.7 7.0 42 i 15 153 398 Notes:
(1)
The allowed total leakage varies from year to year and with configuration because tests are of different durations. The test is run until the error analysis of the data shows that the measurements are reliable. IIence, durations vary.
(2)
The 1989 data was satisfactory. Yet, the observed leak rate was higher than nonnal. Refurbishments were subsequently made to the inner gasket of the basement personnel air lock, to the diaphragms of the pressure relief system's valves, and to one of the vacuum breakers. Data shown are from data set #12 for the first configuration and set #11 for the second.
(3)
Data for the configuration #2 of the 1993 test is given for data set #6 rather than the final data set. The latter showed a near-zero leak rate which was probably caused by a non-uniform temperature distribution in the building. The 1993 test ran late and the sun rose after the sixth data set and began heating the building.
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Ouestion #2 - Core Comnonent Insoection Core component inspections are performed quanerly in accordance with PM 7.4.4.2, "In-Service Inspection of Primary Core Tank and Fuel." A copy of that procedure is attached as Appendix A. Virtually all of these inspections have resulted in negative findings. That is, no deficiencies were identified. This is to be expected because (1) all core components and the core tank were new as of 1973; (2) material certifications were (and are) required for all in-core components as well as for those of the primary, D 0, and shield systems; and (3) primary coolant 2
v,ater samples are analyzed prior to a reactor startup if shut down for more than sixteen hours and j
ine coolant chemistry is maintained to strict standards.
The following, all minor, items have been noted on the core component inspections conducted during the past five years:
1.
Proximity switches, which are reed switches that are activated by a magnet, are one of several indicators used to sense the ' full-in' position of the reactor's shim blades. These switches require replacement at the rate of one to two per year. (Note: The switches are inexpensive, are easy to replace, and are not required for reactor operation.)
2.
Electro-magnets are used to connect the reactor's shim blades to the shim blade drives.
These magnets, which are physically immersed in the reactor coolant, are electroplated to prevent corrosion. Nevertheless, the surfaces do eventually corrode. The present set of electromagnets was installed in 1988. One was replaced in 1994 because of surface corrosion. The others remain in excellent condition.
3.
In 1988/1989, rust was noted in the upper portions of the core tank. This problem was summarized in the MITR FY89 Annual Report to the U.S. Nuclear Regulatory Commission and we quote from that repon:
"The upper shield access ring is a lead filled steel weldment supported by the upper shield ring situated above the upper core tank (see Figure 2-1).
The inner cylindrical surface of the upper shield access ring is clad with a thin layer of 304 stainless steel and the other surfaces are protected by epoxy paint. The gasket which seals water from entering the interface between the upper shield access ring and the upper core tank had deteriorated and allowed rust to form on the non-stainless surface of the shield ring. The remedy for this situation was to remove all rust from all surfaces with a grinder and wire brush. All non-stainless surfaces on the upper shield ring, upper shield access ring, and the top shield lid were then primed and repainted with epoxy paint."
This problem has not been observed since the above corrective action was taken.
The reactor core tank is surrounded by the heavy water (D 0) reflector tank which is in 2
turn encased by the graphite reflector. Ilence, it is not possible to inspect or observe the D O 2
reflector tank on a routine basis. The inner surface of the D 0 tank is, of course, exposed to heavy 2
water. The quality (pli, chloride, conductivity) of that water is monitored and a deuterated mixed-bed ion column is used to maintain its purity. Hence, no corrosion would be expected on the inner surface of the D 0 tank. The outer surface adjoins the graphite reflector which is kept under a 2
blanket of inen (helium) or non-reactive (CO ) gas. Ilence, no corrosion would be expected on 2
the outer surface of that tank.,
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Three inspections have been conducted of the graphite reflector and the exterior surface of the D O tank. One was performed in 1986 and the others in 1987 and 1989. In all cases, access 2
was achieved by first removing one of the experiment ports that extends downward into the graphite region and then installing a periscope (military surplus)in the vacated space. In addition, special tools were constmeted for obtaining samples. The procedure used was as follows:
1.
The 3GV2 experimental port was removed to obtain access to the normally scaled graphite region.
2.
A periscope was inserted in the 3GV2 opening.
3.
A visual inspection was made of the graphite, the exterior of the reflector tank, and the exterior of the 3GV2 thimble.
4.
Photographs were taken through the periscope.
5.
The periscope was removed.
6.
Several samples of graphite were taken.
7.
Scrapings of the surface of the reflector tank were taken.
8.
The 3GV2 thimble was reinstalled.
9.
The reflector region was resealed.
The first inspection was performed over the period 1 December 1986 - 29 December 1986.
At that time, the exterior surface of the reflector tank was found to be uniformly covered with a conosion layer. This layer appeared to be only a few mils deep. Samples showed it to be a white powder that had a granular consistency. It was possible to make a scratch in this layer by using an aluminum tool. A y-spectrum showed that the following nuclides were present: Cr-51, Co-58, Co-60, Fe-59, Zn-65, and K-40 (in approximate order of abundance). Cr, Fe, and Zn are all used as alloying elements in 6061 aluminum - the material from which the reflector tank was made. Further analysis of the samples showed them to be aluminum oxide. A small area of the reflector tank surface was then polished so as to remove all traces of the oxide layer.
The second inspection of the graphite region occurred about a year later, late in 1987. The original oxide layer was still present. The area of the tank that had been polished was examined and no visible changes were noted. In particular, there was neither any visible change to the polished area nor any oxide build-up on it.
The results of both inspections were provided to the MIT Reactor Safeguards Committee (MITRSC) and to several members of the MIT faculty who specialize in corrosion. One of these was Prof. Ronald M. Latanision who is an internationally recognized expert on the subject. The conclusion of the MITRSC was as follows:
"It is believed that this oxide layer formed shortly after the initial operation of the MITR-II in 1975. A small amount of moisture may have been present in the graphite reflector region and it would have condensed on the outer wall of the reflector tank. In any event, the layeris stable and not growing. Also, it is only a few mils thick."
A third inspection of the graphite region was made in December 1989. The findings were identical to those in 1987. No further inspections have been made or are planned. It should be noted that this oxide layer may have a beneficial effect in that it may passivate the metal surface. -
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Ouestion #3 - Review of Radiation Damage Mechanisms to Aluminum l
The review to which we referred on page seven of the original submission (03/31/94) was performed by the faculty and staff who are mvolved in the planning for the reactor's relicensing.
As such the review was in the form of an extended discussion with the principal participants being Professors liarling, Lanning, and Meyer, all of MIT. Several students were assigned special i
problems on the topic of radiation-induced damage mechanisms to aluminum. As part of thesc projects, they prepared term papers and identified useful references in the literature. One of those students, Mr. Jyh-Tzong Hwang, produced a well-written report entitled "A Study of MITR-II Core Tank Aging for Relicensing Consideration" and we are enclosing a copy as a separate l
document. Please recognize that it is a student report and not a published paper.
There is one item in Mr. Hwang's report that may cause some confusion. On page 10 of l
the report, a set of empirical relations aie given that fit experimental data obtained by Weeks et al.
for an aluminum component that was removed from the High Flux Beam Reactor (HFBR) at ORNL. These equations are useful for assessing expected changes in mechanical properties. They are not necessarily useful for determining absolute values of a given property. For example, the j
empirical relation for the elongation (elasticity) is:
%E = 10.7 - 0.69dQin *10-23)
(;)
Mr. Ilwang uses this relation to estimate the %E for the MIT Research Reactor's core tank and concludes (see p.13, middle of page) that, if the reactor were relicensed at 10 MW and then operated for 20 years, the %E would be slightly below the minimum ASME-recommended figure of 10%. While this is an arithmetically correct conclusion, it is irrelevant to the MIT Research Reactor. To see this, use Equation (1) to calculate the %E for unitradiated material. In that case,
$ty is 0.0 and the %E is 10.7%, a figure that is only slightly above the recommended minimum.
l Evidently, the material used for the study from which Equation (1) was derived had an unusually low value of elasticity at the outset. The %E for the material (6061-T6 wrought aluminum alloy) from which the MITR ~ core and reflector tanks were made ranges from 12%-17% (Source:
llandbook of Engineering Fundamentals, O. W. Eshbach and M. Souders, eds.,3rd edition, Wiley Engineering Handbook Series, p.1383.) The value of Equation (1) to the MITR is that it provides a means for estimating the change in elasticity as a result of radiation damage. If one assumes that the present license is extended to the year 2001, and that the power level remains at 5 MW, then the 22 2
maximum expected thermal fluence to the core tank would be 3.22 10 n / cm and the change in
%E from the unirradiated state would be 0.22%. This would not be significant.
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Ouestion #4 - Reactor Con'rol and Radiation Detection Electronics The maintenance of all reactor systems, including the reactor contml and radiation detection electronics, is an on-going activity. The reactor audit process, which was described in our original submission dated 03/31/94, is used to identify systems that warrant upgrading or replacement. For example, in the mid-1970s, the frequency of scrams that resulted from instrument malfunctions was judged to be excessive. The solution was to replace the fission and ion chambers that provided signals to the nuclear safety system with integral lead chambers. The pon plugs that held these chambers were also replaced or refurbished at that time. Similarly, maintenance support for the temperature recorder for the various process systems was judged to be excessive in the mid-1980s and that recorder was replaced. Table 4-1 lists recent upgrades to the reactor control and radiation detection electronics. Also listed are upgrades that are in progress and ones that are planned. For those that are in pmgress, the expected completion date is given.
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Table 4-1 Uncrades to the Reactor Contml and Radiation Detection Flectmnics A.
Comoleted Uccrades (1990-1994) i 1.
Auxiliary Core Purge Radiation Monitor (1990) i 2.
Core Outlet Temperature Recorder (1992) 3.
CO2/IIelium Gasholder Level Recorder (1992) 4.
Radiation Monitor Recorder for Effluent Radiation Detectors (1993)
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Radiation Monitor Recorder for Interior Radiation Detectors (1994) 6.
Differential-Pressure Flow Transmitters for the Secondary, D20, and Shield Coolant Flows (1994) 7.
Flow Recorder for the Secondary, D20, and Shield Coolant Flows (1994) 8.
Equipment for the Calibration of Temperature Sensors (1994) 9.
Toxic Gas (Ammonia) Detection System (1994)
B.
Uogrades in Progress l
1.
Stack Area Radiation Monitor (1164) i 2.
Effluent Radiation Detectors (2B5) l 3.
Expanded Closed-Circuit TV Surveillance System for Containment Building l
Entrances and Interior (2B5) i C.
Planned Uocrades 1.
Redesign and replacement of nuclear safety system. (Funding approved; contract to l
be issued 10S4) l 2.
Procurement of new fission chambers for the new safety system. (Funding approved; contract to be issued 10S4) 3.
Replacement of recorders and sensors for primary flow, AT, and thermal power.
(Funding requested) 4.
Replacement and possible upgrade of emergency power supply batteries. (Under discussion) 1... ~.
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O Anoendix A PM 7.4.4.2, "In-Service Inspection of Primary Core Tank and Fuel" i
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