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A REVIEW
OF THE
MITR-II DESIGN BASIS ACCIDENT
by
John Jay I Cauley
SUBMITTED TO THE DEPARTMENT OFNUCLEAR ENGINEERING
IN PARTIAL FULFILLMENT OF THEREQUIREMENTS OF THE
DEGREE OF
BACHELOR OF SCIENCE INNUCLEAR ENGINEERING
at the
MASSACHUSETTS INSTITUTE OF TECHNOLOGY
June, 1982
@John Jay McCauley 1982
The author hereby grants to M.I.T. permission to reproduce and to distributecopies of this thesis document in whole or in part.
Signature RedactedSignature of Author____
De rtment of Nuclear pgineering, May 25 1982
Signature RedactedCertified b
Accepted by
Aofessor David LanningThesis Supervisor
Signature Redacted
Chairman, Departmental Committee
ArchivesMASSACHUSETTS INSTITUTE
OF TECHNOLOGY
DEC 8 1982
LIBRARIES
2
A REVIEW
OF THE
MITR-II DESIGN BASIS ACCIDENT
by
John Jay McCauley
Submitted to the Department of Nuclear Engineeringon May 25, 1982 in partial fulfillment of the
requirements of the Degree of Bachelor of Science inNuclear Engineering
ABSTRACT
Recent investigation into the behavior of fission products releasedfrom the core of a nuclear reactor under accident conditions has indicatedthat the release fractions (from the core to the environment) have beensignificantly overestimated. The transport and release assumptions for theMITR-II design basis accident have been reviewed in light of this new in-formation, and changes have been proposed to certain aspects of the originalanalysis of the accident. Conservative estimates of the dose rates accom-panying the design basis accident have been made, as well as a best-estimateapproximation.
Also, the validity of channel flow blockage as the maximum credibleaccident (and, hence, the design basis accident) for the MIT reactor hasbeen re-established. This plays a major role in determining the core ini-tial fission product inventory available for release to the primary coolantsystem, and ultimately the atmosphere.
3
ACKNOWLEDGEMENTS
The author wishes to thank the following people for their assistance
in the production of this report; John Bernard, Lincoln Clark, Jr., Bill
Fecych, Joe Gehret, Professor Peter Griffith, Ed Karaian, Professor David
Lanning, Laura McCauley, Bill McDermott, Ara Sanetz, and Georgia Woodsworth.
4
TABLE OF CONTENTS
Page
ABSTRACT 2
ACKNOWLEDGEMENTS 3
CHAPTER 1 INTRODUCTION 6
1.1 Description of the MITR-II 6
1.2 Motivation for this Thesis 7
CHAPTER 2 THE MAXIMUM CREDIBLE ACCIDENT 10
2.1 Introduction 10
2.2 Evaluation of Loss-of-Coolant Accident 10Potential
2.2.1 Internal Initiating Events 12
Leading to Core Dryout
2.2.1.1 Experiment Failure 12
2.2.1.2 Formation of a CombustibleHydrogen/Oxygen Mixture 13
2.2.1.3 Primary Coolant Pipe Rupture 17
2.2.2 External Initiating EventsLeading to Core Dryout 20
2.2.2.1 Rupture of Vessels by a Pro-jectile Through a Beam Port 20
2.2.2.2 Rupture of Vessels as a Re-sult of an Earthquake 22
2.3 Evaluation of Channel Flow BlockageAccident Potential 23
2.4 Conclusion 28
CHAPTER 3 DESIGN BASIS ACCIDENT ANALYSIS 30
3.1 General Discussion 30
3.2 Fission Product Source Term For The DesignBasis Accident 31
3.2.1 Inhalation Dose Component 32
5
Page
CHAPTER 4
REFERENCES
APPENDIX A
APPENDIX B
APPENDIX C
3.2.2 Stack Gaseous Release Dose Component
3.2.3 External Gamma Dose Component
3.3 Fission Product Release and Transport
3.3.1 Release Fraction from the Fuel
3.3.2 Release Fraction from the Primary System
3.3.3 Attenuation due to Building NaturalProcesses
3.3.4 Release Fraction from Pressure Re-lief System
3.3.5 Summary
3.4 Calculation of Dose Rates
3.4.1 Inhalation Dose
3.4.2 Dose from Release of Gases from theStack
3.4.3 External Gamma Dose
3.5 Best-estimate Approximation of Dose Rate
CONCLUSIONS
4.1 Import of this Thesis
4.2 Applicability to Emergency Plan
DETERMINATION OF I RELEASE RATE TO THE EXHAUSTPLENUM
DETERMINATION OF 1311 CONCENTRATION IN THE PRIMARYCOOLANT
MAGNITUDE OF THE DEUTERIUM?OXYGEN EXPLOSION IN THED 20 REFLECTOR
34
34
36
38
39
43
45
45
47
47
48
48
51
53
53
55
56
58
62
64
6
CHAPTER 1
INTRODUCTION
1.1 DESCRIPTION OF THE MITR-II
The Massachusetts Institute of Technology Research Reactor (MITR-II)
is a five megawatt tank-type reactor, cooled and moderated by light water
and reflected by heavy water. Aluminum clad fuel elements, enriched to 93%
uranium-235, normally occupy twenty-four of the twenty-seven core positions,
the remaining spots taken up by solid aluminum dummy elements or orificed
sample assemblies. The H20 coolant passes upward through the core at about
2000 gallons per minute, undergoing a bulk temperature increase of about
eight degrees centigrade (at the normal operating power of 4,9 MW). This
heat is transferred to a secondary flow system via three parallel heat ex-
changers, and is subsequently dissipated to the atmosphere by two forced-
draft cooling towers. Immediately outside the primary coolant core tank is
the D 0 reflector, which serves to minimize the number of fission neutrons2
which escape from the active core region. A graphite reflector surrounds
the tank which houses this D20 reflector, and is in turn surrounded by 5.5
feet of concrete shielding.
The primary core tank normally contains a ten foot head of water above
the top of the core. It is constructed of 6061 aluminum alloy, and has
several penetrations for control rod drives, level and pressure sensors,
and flow outlet pipes. The core shroud is bolted to the tank support ring;
this shroud was designed to provide an annular space (between itself and
the core tank) through which the primary coolant flows downward before en-
tering the core.
The heavy water reflector tank, concentric with the core tank, is also
fabricated of 6061 aluminum alloy. It is normally filled with D20 to a
7
height corresponding to the top of the fuel elements. A pneumatically-
operated valve permits rapid 'dumping' of the heavy water volume to accom-
plish emergency shutdown. Approximately 140 gallons per minute are pumped
through the system to remove the heat lost to the reflector- this heat is
transferred to the same secondary flow system which cools the primary sys-
tem. A helium blanket-purge is smaintained on all free surfaces of the re-
flector system to inhibit corrosion and to prevent excessive degradation of
the high-purity D20. Fourteen re-entrant thimbles, hollow aluminum closed
tubes designed to optimize the flux available at the corresponding beam
ports and pneumatic sample systems, are welded to the inside of the re-
flector tank.
Tapered graphite stringers are stacked radially about the reflector
tank, providing additional reflection of escaping neutrons. An aluminum
membrane ensures that this graphite reflector region remain gas-tight. The
entire region is blanketed with helium to prevent excessive production of
Argon-41. Between the graphite reflector membrane and the 5.5 feet of con-
crete biological shielding lies the thermal shield (a 1.5 inch thick lead
annular cylinder with imbedded cooling coils, surrounded on both sides by
2 inches of steel); the thermal shield removes heat from the reactor
shields and provides the ultimate structural support for the reactor tanks
and the annular rings. (Ref. 1)
1.2 MOTIVATION FOR THIS THESIS-
The MIT reactor underwent substantial modification in 1974, and all
components that were not wholly redesigned and replaced were carefully in-
spected for wear. Six years of safe operation at five megawatts have
followed the completion of the modification, and the MITR-II (as described
in section 1.1) has provided useful irradiation services to -many institute
8
departments and local colleges, hospitals, and industries for the entire
period. Recent developments, however, threaten to adversely affect the
quality of service provided by the MITR-II. Stringent Nuclear Regulatory
Commission (NRC) legislation aimed at correcting deficiencies at commercial
nuclear power plants has not been adequately tailored to the non-commercial
activities of many research reactors. As a result, an increasingly higher
percentage of manpower at these research facilities must be dedicated to
the documentation of compliance with NRC regulations, leaving less -manpower
available for the performance of purely research-related activities.
One example of these potentially troublesome areas of regulation, of
particular concern to this thesis, is the administration of the emergency
plan for research reactors. The emergency plan details the actions to be
taken by the facility in the event of an accident which threatens the
safety of the local population. The NRC has recently endorsed a document
which establishes guidelines for the structure of the emergency plan -
revision 1 to the American National Standard ANSI/ANS 15.16. "Emergency
Planning for Research Reactors", outlines the appropriate actions, action
criteria, and training/surveillance requirements deemed necessary to estab-
lish a strong basis for the response to an actual emergency. (Ref. 2) The
actions specified in the case of the most severe accident class are very
extensive, probably beyond the capabilities (and budgets) of many research
reactors' staffs. If the potential for this most severe class of accident
at a given reactor facility does not exist, then application may be made to
the NRC for an exemption from the most severe accident class requirements.
It is the intent of this thesis to provide a foundation for the assess-
ment of the accident potential of the MITR-II, The point of departure of
this analysis shall be the analysis of the Design Basis Accident given in
9
the MITR-II SAR. Argument will be made that no accident of greater sever-
ity than the present DBA (channel flow blockage) is credible (chapter two).
Any unnecessary conservatism in predicting dose rates for the site environs
will be trimmed, and realistic approximations of these doses are calculated
along with the conservative design estimates (chapter three), The final
chapter will present the results of this research, in terms of dose rates
at the site boundary for the DBA, and establish the basis upon which exten-
sion of these results may be made to decide whether emergency plan require-
ments for the most severe class of accident are warranted.
10
CHAPTER 2
THE MAXIMUM CREDIBLE ACCIDENT
2.1 INTRODUCTION
As part of the safety analysis for the MITR-II, a maximum credible
accident was envisioned which led to the release of fission products to the
environment. This accident involved the melting of four fuel plates due to
coolant channel flow blockage, coincident with an instantaneous rise in
containment pressure to 2 psig. The goal of this chapter is to provide a
firmer basis for the selection of this accident as the design basis acci-
dent than was originally provided in the Safety Analysis Report,
The percentage of the fission product inventory available for release
from the four plates which melt is about 1.5%. (Ref, 3) It is probable
that, in the event of core dryout, a larger percentage would be available
for release than 1.5%. This chapter, then, focuses on the relative likeli-
hoods of the two scenarios, demonstrating that the potential for channel
flow blockage is significantly higher than for core dryout. Several pos-
sible paths leading to core dryout will be discussed, and design features
and administrative controls specifically implemented to avoid accident pro-
gression will be detailed. Comparison will be made to the paths leading to
channel flow blockage, still considered an unlikely scenario, but neverthe-
less more credible than core dryout.
2.2 EVALUATION OF LOSS-OF-COOLANT-ACCIDENT POTENTIAL
The design of engineered safeguards for the ITR-II focused on ensuring
that the core is always provided with sufficient coolant flow and volume to
prevent gross clad failure. A higher percentage of the fission product in-
ventory might be available for release in the event of core dryout than is
11
expected for the present design basis accident, so core dryout was of major
concern to the reactor designers. It is thus unlikely that this accident
will occur at MIT; however, there are accident sequences which may lead to
uncovering. *of the core, as discussed in the following two subsections.
The simultaneous rupture of the light water core tank, the heavy water
reflector tank, and the graphite reflector membrane would deprive the core
of coolant (Ref. 3) if the two water tanks ruptured below the height of the
core top. It is difficult to imagine how this accident might be initiated-
an explosion of an in-core sample assembly might do it, as might the pene-
tration of a massive projectile into one of the horizontal beam ports. In
general, two types of initiating events shall have to be considered: inter-
nal events, those events connected with'the design of the reactor and its
experimental appurtenances; and conversely, external events, those events
associated with off-site environmental or man-made abnormal occurrences:
Internal Initiating Events
- experiment failure
- explosion of hydrogen/deuterium--oxygen mixture in the vessel
- primary coolant pipe rupture
External Initiating Events
- beam port projectile
- earthquake
Other initiating events are deemed to be sufficiently improbable so as
not to merit consideration- airplane crash into the reactor containment,
for example.
12
2.2.1 Internal Initiating Events Leading To Core Dryout
2.2.1.1 Experimental Failure
In order to preclude any adverse effects on the safe operation of the
reactor as a result of experiment failure, strict administrative controls
are mandated in the Technical Specifications for the evaluation of an ex-
periment before it is implemented, The experiment is reviewed for possible
adverse reactivity, thermal-hydraulic, chemical, and radiolytic decomposi-
tion-induced effects. The aspects of this review which eliminate the pos-
sibility of tank rupture through experiment failure are summarized below
(Ref. 4):
1. "Metastable or other materials that could react to create a
rapid pressure rise shall be encapsulated. The capsule shall
be prototype tested under experimental conditions to demonstrate
that it can contain without failure an energy release equivalent
to at least twice the material to be irradiated or at least
twice the pressure that could be expected from any reaction of
these materials."
2. "... the quantity of material should be limited such that if the
maximum calculated energy release should occur, significant dam-
age to the reactor core will not result, assuming the material is
not encapsulated."
3. "No explosive materials (defined to include all materials that
would constitute Class A, Class B and Class C explosives as de-
scribed in Title 49, Parts 172 and 173 of the Code of Federal
Regulations) shall be placed in the reactor core or within the
primary biological shield, which, if completely detonated, could
cause any rearrangement or damage to the core."
13
4. "Materials whose properties (composition, heating, radiolytic
decomposition, etc.) are uncertain must be prototype tested."
These criteria are designed to ensure that the worst-case failure of
any experiment will not cause any substantial damage to either of the two
vessels - certainly not gross rupture leading to core dryout, Hence, the'
probability of major fission product release as a result of experimental
failure is deemed very small.
2.2.1.2 Formation of a Combustible Hydrogen/Oxygen Mixture
Hydrogen bubble formation and subsequent ignition -might cause an ex-
plosion large enough to damage the reactor vessel and internals, perhaps
leading to core dryout under extreme circumstances. Summarized here are
the design features and the administrative controls developed to eliminate
the potential of a hydrogen/deuterium explosion in the MITR-II vessels:
1. To prevent deuterium from reaching an explosive concentration
(calculated to be 6.87% under more severe conditions than anti-
cipated in the reflector region) in the helium cover gas which
blankets the D20 reflector system, a recombination system is em-
ployed. (Ref. 4) This system diverts a small flow of helium
cover gas through a chamber where palladium pellets catalyze the
2D 2 + 2D20 .reaction. So long as the flow rate of the system
is maintained between 1.5 and 8 cubic feet per minute and the
chamber temperature maintained above 50*C, the reaction will con-
tinue; these two parameters are monitored continuously. The he-
lium cover gas blankets the reflector tank, the dump and storage
tanks, and the blister tank used for medical irradiations; hence,
all these regions are purged of any gaseous deuterium that might
14
build up as a result of the fast flux to which the D 20 systems
are exposed, In addition, unavoidable leakage from this cover
gas system occurs, and subsequent replenishing is done with he-
lium of low hydrogen concentration. (Ref, 1).
2. Prior to construction, design analysis revealed the existence of
a point in the D20 region where deuterium -might collect and form
an explosive concentration, as shown in Fig, 2-1. To mitigate
this, the heavy water return line to the reflector tank is angled
to direct the flow into this stagnation point and thus eliminate
deuterium collection.
3. To avoid hydrogen collection in the light water medical shutter
tank, the system is purged at least weekly and the H2 concentra-
tion measured. If it exceeds 1%, the entire helium cover is
purged several times to ensure that clean helium predominates in
the system.
4. The reactor will be manually shut down by the operator (to <200
kw) if;
- the D2 concentration exceeds 2% in the reflector system,
with the recombiner out of commission, or if
- the H2 concentration exceeds 1% in the airspace above the
core.
Reactor shutdown to less than 200 kw slows the production of D2
and H2 by fast flux disassociation, the dominant mechanism.
5. The airspace above the core is continuously purged to prevent the
buildup of H2 and as a means of detecting unusual activity in the
primary system. Under normal operating conditions, then, there
is no potential for hydrogen buildup. In the event of detection
15
CORE
x
D 2 0
shutter
Figure 2-1
Stagnation Point 'X' in Reflector Region (not to scale)
rooo
16
of high radiation levels, however, solenoid valves on both the
entrance to and the exit from the airspace close, isolating the
volume. In this. instance, it is possible that hydrogen might
build up to an explosive concentration. This is protected a-
gainst by the written procedures available to the operator in
the event of a 'high radiation level core purge' alarm, as fol-
lows: (Ref. 5)
- The Abnormal Operating Procedure associated with this alarm
instructs that the reactor be shut down if the core purge
monitor indicates greater than 200 thousand counts per min-
ute (200K cpm), stemming the production of hydrogen in the
airspace.
- If the monitor indicates less than 200K cpm, the procedure
instructs that the alarm be reset to 200K cpm and the blower
be restarted, pending analysis of grab samples. Should the
monitor indicate greater than 150K cpm for 15 minutes, the
reactor will be shut down.
- If the reactor is operated above 200 Kw with the airspace
isolated, the isolated volume will be analyzed hourly to
determine the hydrogen concentration. If the concentration
exceeds 1%, the reactor will be shut down, as specified by
the procedure.
Based upon the system design and the comprehensive procedures
developed to deal with abnormal operating conditions, the pos-
sibilty of a hydrogen explosion in the airspace above the core
is deemed very small.
6. Appendix C estimates the energy which would be liberated in the
17
event of an explosion in the reflector region immediately below
the core. It is probable that this is a gross overestimate of
the magnitude of the energy release, (Ref. 6) Actual calcula-
tions estimating the stress and strain profiles in the aluminum
tank under these circumstances have not been performed, but it
seems unlikely that any damage will occur given the manner in
which the energy is most likely to be dissipated.
2,2.1.3 Primary Coolant Pipe Rupture
The rupture of the primary coolant inlet pipe at a level below the
core has the potential of leading to core dryout. Consider the three lev-
els of defense built into the design of the MITR-II which minimize the po-
tential for this accident scenario:
1. The design of the primary system represents the first barrier
against the rupture of the primary coolant inlet pipe. The flow
enters the vessel approximately 60 inches above the core and fans
out into an annular region which converges below the core proper.
This fact prevents the core from dryout should the inlet pipe
rupture above the core; no siphon effect is possible in this in-
stance.
Also, the bulk of the primary system below the core level is pro-
tected from gross damage by virtue of its location in the mech-
anical equipment room. This below-grade location provides the
additional shielding offered by the reactor floor against a mis-
sile of external origin, which would presumably enter above-
grade.
2. Operator action could play a great role in arresting the leak
before core uncovery occurred, The loss of coolant could be iso-
18
lated in the equipment room in many instances. If not, the pri-
mary coolant storage tank inventory can be pumped to the core via
the auxiliary pump MM-2; should this volume of coolant be exhaus-
ted, a demineralized light water make-up storage tank on the re-
actor floor can provide additional coolant inventory to the stor-
age tank by gravity feed. If the loss of coolant has not yet
been secured when this make-up supply is also exhausted, primary
coolant collecting in the equipment room wet sump can be returned
to the system by means of a hose-and-strainer setup attached to
the inlet side of the auxiliary pump MM-2. Flow from MR-2 can be
routed either through the primary coolant outlet pupe or through
the in-vessel spray nozzles, Redundant differential pressure
level indicators read out in the control room to provide the oper-
ations staff with reliable information about the core tank level
(one sensor also indicates in the utility room).
3. Engineered safeguards represent the third line of defense protect-
ing against the uncovering of the core. Should the core tank
water level drop below the level of the inlet pipe, anti-syphon
valves, normally held shut by hydraulic pressure will open to in-
troduce air into the syphon flow path, breaking the syphon. Thus,
unless both syphon valves fail, the core level will not drop be-
low the level of the inlet pipe.
Should all other defenses and corrective action fail, city water
may be introduced as coolant to the core tank by -means of the
spray nozzles located in the reactor vessel. Two independent
nozzles supplied from the control room and utility room are each
capable of delivering sufficient flow to preclude gross core
19
failure. Thus, only in the case where both anti-syphon valves
and both spray nozzles fail, a very unlikely event, would gross
core damage result. (The second line of defense, operator ac-
tion, then becomes the determining factor in preventing core dry-
out.) The reliability of the emergency core cooling spray sys-
tem has been estimated at greater than 99.99%, per demand, The
reliability of the anti-syphon valves is probably greater, as
they operate in a passive manner- the absence of buoyancy and
positive pump pressure opens the valve,
The probability of core dryout as a result of a primary coolant
pipe rupture below the level of the core tank can be simply mod-
eled as the product of the probabilities of several independent
events:
P(core dryout via pipe rupture)
= P(initial pipe rupture) x P(_failure of both anti-
syphon valves) x P(ECCS failure),
Values for these probabilities can be found in references 7 and
8. The initial core rupture will occur with a probability of
10~9 per hour. The failure of one valve, per demand, has an ap-
proximate probability of 10-4; conservatively assuming that the
failure of two such valves is a dependent event implies a net
probability of 10-6 per demand. ECCS failure has recently been
assessed a probability of less than 10 per demand. In one year,
then, the probability of core dryout initiated by a pipe rupture
is roughly equal to 10- 16. It is clearly not a credible 'accident
for the lifetime of the reactor.
20
2.2.2 External Initiating Events Leading to Core "Dryout
2,2.2.1 Rupture of Vessels byza Projectile thr6ugh'a"Beam Port
The 5.5 feet of concrete shielding surrounding the graphite reflector
region would no doubt be highly effective in protecting the reactor inter-
nals from missiles of external origin. However, the horizontal beam ports.
which penetrate the reactor all the way to the reflector tank provide an
unobstructed path along which such a missile could cause great damage, po-
tentially leading to core dryout. There are three major reasons supporting
the contention that such an event (i.e,, simultaneous rupture of both the
core and reflector tanks) will not occur.
First, the probability of a projectile reaching the entrance to one
of the beam ports at the correct angle is very- small. If it originated
from outside the containment building, it would have to penetrate both the
.375 inch thick steel shell and the two foot thick concrete shielding wall
(or the .625 inch thick steel ceiling) and deflect in such a manner that
its final trajectory lay coincident with one of the ports. This is not a
credible set of circumstances. A more likely scenario would involve a mis-
sile originating inside the building -- a massive object dropped from the
crane, for example, glancing off a concrete block and striking the port
entrance with enough horizontal momentum to penetrate to the reflector
tank. The maximum height from which such an object could be dropped is
less than thirty feet, however, and it is extremely unlikely that the
object will have sufficient energy after its collision to penetrate far
and cause damage. Other missile sources within the containment building
appear to have neither the potential nor the probability of occurrence of
the crane missile accident, but it matters little, The level of protect-
ion afforded by the design of the system is great, as seen by the next point,
21
The next reason that rupture of a beam port by a missile is unlikely
lies in the construction of the beam port. All the ports have either plugs
or solid concrete in them to attenuate the beam that would otherwise pass
freely to the occupied areas of the reactor floor. The ports which house
the pneumatic tube irradiation facilities are filled with concrete around
the tube; the ports which are used for beam manipulation utilize plugs and
sleeves (both of which are stepped to reduce streaming) with only a small
diameter hole through which' the beam may pass unattenuated. In addition,
gasketed cover plates and additional concrete shielding has been installed
at the entrance to most of the beam ports at the reactor face. The stepped
plugs and this additional concrete shielding represent a .formidable barrier
against the penetration of a missile to the end of a beam port.
Finally, the re-entrant thimbles which are welded to the inside of the
reflector tank are located just below the core tank (for all the radial
ports), rather than pointing directly at the core tank; this reduces the
chance that a projectile will rupture the core tank after rupturing the re-
entrant thimble. The two semi-radial ports, 6RH1 and 6RR2, are located
twelve inches above the remaining ports; their thimbles point slightly off-
center at about the core midplane. 6RH2 is protected by several large con-
crete shielding blocks, however, and 6RH1 (which houses the two inch pneu-
matic tube facility 2PHl) is filled with concrete surrounding its tube.
Extra precaution was taken for these two ports because of the position of
their respective thimbles.
In summary, the design of the MITR-II beam ports minimizes the poten-
tial for simultaneous rupture of the reflector and core tanks by eliminat-
ing all unnecessary paths along which a projectile could travel through the
22
use of stepped port plugs in beam ports and concrete fill in tube ports.
The concrete shielding employed to reduce the dose to personnel working on
the reactor floor provides additional protection against a projectile ac-
cident. These features, combined with the improbability of a projectile
of sufficient energy achieving a direct hit upon the entrance to one of the
ports, represent a more than adequate means of protection against this ac-
cident scenario.
2.2.2.2 Rupture of Vessels as a Result of an Earthquake
The possibility of severe core damage as a result of an earthquake was
addressed in a study presented in part in section 15.8 of the SAR. The fol-
lowing three conclusions of this study are relevant to this discussion:
1. The Design Basis Accident for the Cambridge area would probably
have a maximum acceleration of about 0.2 g.
2. The maximum response acceleration of the reactor is only slightly
greater than 0.2 g, as little or no amplification occurs in the
rigid foundation.
3. The reactor can withstand forces corresponding to 3.4 g of vert-
ical acceleration and 5.1 g of horizontal acceleration, which in
turn corresponds to the yield stress of the tank at 9500 psi.
It is not considered credible that the primary and reflector vessels
rupture under the forces caused by a Cambridge-area Design Basis Earthquake.
A pipe rupture might occur, and this has been previously considered (section
2.2.1.3). Thus, an earthquake has little potential, direct or indirect, for
leading to core dryout.
23
2.3 EVALUATION OF CHANNEL FLOW BLOCKAGE ACCIDENT POTENTIAL
The channel flow blockage accident is initiated by the presence of for-
eign material in the fuel element inlet plenum. These materials might pre-
vent proper passage of coolant through the coolant channels and thus deprive
the element of adequate heat-removal capability, leading to plate melting.
In view of its consequences and relatively high potential, a form of this
accident has been chosen as the design basis accident. [The largest piece
of material able to pass through the fuel element inlet nozzle and lie
flush across the bottom of several fuel plates does so, and is held there by
inlet coolant pressure.] Four plates are affected, and it is assumed that
these plates are entirely melted.
There are two possible sources of foreign material which might lead to
channel flow blockage; these are materials that originate inside the core
tank, and materials that originate outside the core tank. Examples of the
former are polyethylene tubing used for helium connections to the drives
and asbestos gaskets used at in-core flanges. The latter encompasses many
possible materials-pieces of rubber gloves, pieces of poly viewing windows,
coins, dosimeters, and bolts are just a few of the many possible items which
could potentially be dropped into the core during refueling or maintenance.
Great care is taken when maintenance evolutions or refuelings are in
progress to both prevent the introduction of foreign material into the core
tank and to visually ascertain that in-core materials have not degraded to
the point where flaking or cracking might occur. In-service inspection of
the core tank is performed periodically to provide an extensive assessment
of the condition of core components. These levels of surveillance are ex-
pected to be sufficient in guarding against the channel flow blockage acci-
dent scenario.
24
Operating experience at other reactor facilities has indicated that the
potential for this accident scenario is significant. On 12 December, 1961,
the Engineering Test Reactor experienced a flow blockage accident which af-
fected six fuel elements; the damaging material was a Lucite sight box, sim-
ilar to those in use at MIT, which escaped detection prior to the Materials
Testing Reactor on 13 November, 1962; the cause was channel flow blockage by
a piece of the seal tank gasket material. Two weeks later, pieces of for-
eign matter identified loosely as "red rubber" were discovered in four fuel
elements of the MTR. The vigilance engendered as a result of the recent
fuel melt led to early detection and prevention of a similar accident.
(Ref. 9)>
The second indication that the chanfiel flow blockage accident scenario
has a greater relative potential with respect to the core dryout accident is
found in the operating history of the MITR-II. In several instances foreign
material has been either discovered in or introduced into the reactor ves-
sel. The three major occurrences of note are summarized below:
9 October 1975: While performing a procedure designed to measure the
flow through each fuel element position, the operator
of the flow meter dropped his dosimeter into the core
tank, The dosimeter was swept into the coolant out-
let pipe and, after passing relatively unharmed
through a main primary pump, lodged at the inlet to
one of the heat exchangers. The primary system was
taken apart in sections after a search revealed that
the dosimeter had not landed on top of the fuel or
the core support housing assembly; the dosimeter was
discovered when the inlet valve to the heat exchanger
was taken apart. (Ref. 10)
25
26 April 1976:
17 August 1981:
Any object swept into the coolant outlet pipe (a rela-
tively high probability with the pumps running) will
not be capable of initiating a channel flow blockage
accident, because of the filtering effect of the pri-
mary heat exchangers. In order for the object to reach
the fuel inlet plenum, its size and shape must be such
that no significant blockage can occur.
A routine inspection of the core tank revealed the pre-
sence of foreign material upon the core support housing
assembly. The material was removed by the use of a
small recirculating pump and suction tube. It was ten-
tatively identified as a piece of asbestos gasket mate-
rial ("Ankorite"), which-is used at several points in
the core, The piece measured 1 inch by 3/4 inch by
1/16 inch, with a slight degree of activity, primarily
65 Zn. The entire core was inspected for further evi-
dence of foreign matter, and none was found. The re-
actor was started up without incident. (Ref. 11).
While performing the procedure governing the transfer
of spent fuel from the storage ring to the storage pool,
an operator-trainee dropped a 3/4 inch eyebolt through
the twelve inch viewing port. The reactor lid was re-
moved and the bolt located on top of two central fuel
elements. It was removed, and an inspection of the en-
tire core was performed. No damage was revealed, and
the reactor later started up uneventfully, (Ref. 12).
26
It is important to emphasize at this point the concern of the reactor
staff over these incidents. Great care is taken to avoid the inadvertent
dropping of objects into the core, and comprehensive inspections are per-
formed regularly to investigate the level of deterioration of potentially
hazardous in-core components such as gaskets and polyethylene tubing. The
response to each of the accidents described here has been excellent, from
complete visual examination of all potentially affected components to
guarded monitoring of core water and core purge activities during the sub-
sequent startup. With each incident, the following practices were r?-
emphasized:
-- "People working in or over the core tank must remove loose arti-
cles from outer pockets, or must securely tape them. This pre-
caution is written into existing procedures involving such work,
but the intent must be observed for all work where the possibility
of dropping materials into the system exists, whether covered by
procedure or not.
-- Articles which can be knocked into the tank inadvertently must not
be left on the open lid or on the edge of the upper shield when
the lid is off.
-- During maintenance in or over the core tank or at other points
where stray materials can enter the system, care must be taken to
prevent this from happening under all normal and forseeable acci-
dent conditions. Such precautions can include screens to catch
articles, means to secure them if dropped, piece counts or inven-
tories of materials and tools after a job is complete, etc. Such
precautions should take into account the potential for articles
being swept around with the primary water if the pumps are ON or
27
for falling through open natural convection valves if the pumps
are OFF." (Ref. 11)
It is not expected that the presence of foreign materials in the core
will go undetected and lead to channel flow blockage. However, in light of
the several incidents at MIT and the accidents at the Materials Testing
Reactor and the Engineering Test Reactor, it must be recognized that this
accident scenario is certainly credible. An accurate assessment of the pro-
bability of this accident cannot be made, because of the wide variety of
circumstances under which foreign matter might be introduced into the core,
as well as the difficulty in predicting whether or not this material is
overlooked. However, it can be stated with reasonable certainty that,
should incidents of the nature described in the Unusual Occurrence Reports
continue to occur, an object might eventually escape detection and some de-
gree of flow blockage result.
28
2.4 CONCLUSION
This report supports the source term as defined in the MITR-II SAR's
design basis accident analysis, namely the failure of fuel plates as a re-
sult of the maximum credible accident of channel flow blockage, The impor-
tant characteristics of the accident (e.g. number of affected plates, lo-
cation of failure in core, degree of plate damage, etc.) are discussed in
chapter three.
It is the reasonable conclusion of this chapter that the probability
of core dryout is several orders of magnitude smaller than the probability
of channel flow blockage. This is best illustrated by comparing the paths
to each accident; the required combination of several component failures
and operator errors necessary to lead to core dryout is far less likely
than the single event of undiscovered introduction into the core of foreign
material which would lead to channel flow blockage. The depth of defense
incorporated into the design of the reactor primary system (including ves-
sel, piping and engineered safeguards) effectively precludes a core dryout
scenario, but can not offer complete protection against the channel flow
blockage accident.
The major conclusion of this report regarding the maximum credible
accident is that channel flow blockage represents the proper point of de-
parture for the design basis accident analysis, as determined similarly at
the time the SAR was written. A second conclusion which can be drawn is
that the level of protection against the channel flow blockage could possi-
bly be enhanced. Several schemes recommended in the reports describing the
incidents (both at the MITR and other facilities) have been tried with
limited success (e.g., screens placed over the natural convection valves
inhibited operators from verifying the valves open or closed). The possi-
bility of decreasing further the frequency of the events described in the
29
previous section should be investigated thoroughly, both in terms of admin-
istrative controls and system design. The present design and controls have
proved themselves extremely effective, but it is possible that a small in-
vestment at this time could lead to a further decrease in the likelihood of
the flow blockage accident.
30
CHAPTER 3
DESIGN BASIS ACCIDENT ANALYSIS
3.1 GENERAL DISCUSSION
It is the intent of this chapter to review the Design Basis Accident
analysis as presented in the MITR-II Safety Analysis Report (.SAR). Atten-
tion will be focused upon the assumptions concerning the source terms for
and the release fractions of radionuclides to the environment, rather than
on the analytical methods employed to determine dose rates around the site.
(The analytical methods used in the MITR-II SAR are those generally accepted
by the Nuclear Regulatory Commission as best modelling radiological condi-
tions following a reactor accident.)
Based upon previous reactor accident experience and recent evidence of
significantly lower fission product volatility, the release fraction for
iodines from the reactor coolant system is expected to be about a factor of
ten lower than stated in the SAR. It is not expected that a comparable re-
duction in the nobel gas release fraction is applicable - accident exper-
ience tends to support the validity of present assumptions concerning the
behavior of these isotopes.
New dose rates have been calculated given the proposed reduction in
iodine source terms. In general, the dose rate can be broken down into
three components: dose resulting from building leakage, dose resulting from
gaseous stack release, and external gamma dose. The contribution to each of
these from the iodines is evaluated, For instance, the inhalation dose is
dominated by the iodines, and is therefore substantially decreased; scatter-
ing doses are not dominated by any particular nuclide, and are thus less
sensitive to the proposed reduction in the iodine release fractions.
Finally, more realistic values for all the parameters involved in de-
31
termining the radiological consequences of a DBA will be estimated, and
dose rates .calculated for these conditions. While it is not appropriate to
use these values in a design analysis because of uncertainty, it is inter-
esting to note the difference between the realistic and the conservative
estimates; it lends further reassurance as to the safety of the MITR-II de-
sign.
3.2 FISSION PRODUCT SOURCE TERM FOR THE DESIGN BASIS ACCIDENT
In the event of a design basis accident at the HITR-II, it is conser-
vatively assumed that the active portions of the four affected fuel plates
will completely melt. In actuality, the creation of voids in the affected
channels would lead to increased noise on the nuclear instrumentation and
the initial release of fission gases would trip the core purge monitor,
thus alerting the operator-in-charge to an abnormal condition; reactor
power would then be lowered and the melting arrested long before total
melting occurred. Because it is difficult to predict at what point the
melting would be stopped, however, total failure is assumed for purposes of
calculation. Until a detailed analysis has been performed modelling the
behavior of the fuel under DBA conditions, demonstrating that fuel melt
cannot progress beyond a certain point under any credible conditions, the
assumption regarding total fuel melt is a necessary conservatism. (Ref. 3)
Another necessary conservative assumption is that the channel flow
blockage occurs in the 'hottest' (i.e. highest local heat generation rate)
element in the core. A conservative value for the radial power peaking
factor, the ratio of the maximum to average volumetric heat generation rate
in the core, was chosen as 1.5. Operating experience has shown that 1.5 is
a reasonable value; typical values range up to about 1.35 for recent core
loadings. (Ref . 3, 13)
32
3.2.1 Inhalation Dose Component
At this point, we have four plates, producing about 1.5% of the core
power, undergoing complete failure and releasing the same fraction of the
total core fission product inventory to the primary coolant system, What is
needed next is an approximation of this total fission product inventory;
the simple calculations involved are detailed in reference 14, " Calcula-
tion of Distance Factors for Power and Test Reactor Sites ", One assump-
tion made here which might have a large effect on the inhalation dose source
term was that all radionuclides had reached their saturated activities. This
assumption is valid for isotopes whose half-life is far exceeded by the time
the reactor has spent at full power. For the MITR-II, the time spent at pow-
er varies up to about 105 hours, given the standard Monday-Friday operating
cycle employed in recent years. After 105 hours at full power, all the io-
dine (which is the dominant contributor to the inhalation dose) isotopes
have effectively reached saturation levels with the exception of 131I; with
a half-life of 8.05 days, 131I will only have reached 31% of its saturated
activity, as shown below:
qt = qs-[1 - exp(,-ATO)-]
where qt is the reactor inventory of some isotope in curies,
qs is the saturated activity of the isotope in curies,
X is the disintegration constant for the isotope, in sec- 1 , and
To is the time the reactor has spent at full power, in sec.
For the specific case of 1 3 1i, the disintegration constant is 9.96x10 7
sec-l. Assuming 105 hours at full power, To is equal to 3.78x105 seconds.
Thus, the ratio of actual activity to saturated activity is 0.313, Due to
the buildup of 131- from previous weeks, however, the actual 131I activity
33
would be about 70% of tle saturated activity. Table 5.3.2.2-1 of the SAR
would then read as Table 3.2.1-1 as below.
TABLE 3.2.1-1
DOSE TO CRITICAL ORGAN PER MEGAWATT REACTOR POWER
(Total Release of Inventory)
Iodine Dq cmegarads duies D d
Isotope A't curie MW AT 7 . W
131 1.48 1.757 x 10 26004
132 0.0535 3.81 x 104 2038
133 0.40 5.63 x 104 22520
134 0.025 6.58 x 104 1645
135 0.124 5.10 x 104 6324
It is important to note that this table is now valid for operating
periods of up to 105 hours. It is not expected that the operating schedule
be changed in the near future, but if a longer period of full power opera-
131tion was desired, the data for I would be non-conservative. (90% of the
saturation value would be reached in 27 days of full power operation, for
131example, and this would mean 33515 megarads/MW due to I.) Given the
current operating schedule, the table above provides a conservative esti-
mate of the core iodine source term under all conditions. However, the 30%
reduction in the 131I source term does not warrant the risk of underestima-
ting the consequences of the Design Basis Accident if a monthly schedule
were ever adopted. Thus, the saturated inventory of 37148 Megarads/MW will
be used for 131I in the DBA analysis, the lower value calculated above will
be used in the realistic estimate of area dose rates.
No other assumptions regarding this source term are debatable. Thus,
no changes to the q/P term are proposed for the design basis accident ana-
lysis.
34
3.2.2 Stack Gaseous Release Dose Component
The source. term 'Q'. for this mode of release is a linear function of
the source term 'q/P1 used in calculating the inhalation dose component.
The q/P term was reviewed in the previous section, and no changes to it
were advocated. Accordingly, no changes to the source term 'Q', used in
calculating the gaseous release dose component, are advocated here.
3.2.3 External Gamma Dose Component
The gamma ray source strength So is linearly dependent upon the total
gamma ray emission energy for a given radionuclide. The data on gamma ray
energies for the fission product nuclides of concern in the MIT reactor
DBA analysis were taken from a 1957 reference. Improvements in radiation
detection technology, most notably the development of the Ge-Li detector
family, have resulted in vastly improved gamma spectrum resolution. Cur-
rent values for the average gamma ray energy were :calculated for each nuc-
lide, and the S 0 term was updated to account for the higher resolution.
These values are given in Table 3.2.3-1. The effect of the change in aver-
age gamma ray energy values is not linear; it is seen in the linear and
energy attenuation coefficients deep within the scattering calculations.
The magnitude of the induced error is not dealt with here; it is recommend-
ed that a computer code with the ability to scalculate scattering dose comp-
onents (e.g. CRAC-II) be used to make that approximation.
35
TABLE 3.2.3-1
External Gamma Dose Parameters
Isotope - Eav,(Mev) .$ (Mev/sec-Mw)
Iodine
131 0.38 3.5.3x1014
132 0.76 2,82x10 15
133 0.58 1.21x10 15
134 0.84 2,05x1015
135 1.13 2,13x10'5
Xenon
131m 0.164 1.51x101 2
133m 0.233 1.16x101 3
133 0.081 1.67x10'4
135m 0.527 2.98x1014
135 0.262 4.87x1014
Krypton
83m 0.010 negligible
85m 0.151 6.53x101 3
87 0.881 2.13x101 4
88 0.731 8.92x101 4
36
3.3 Fission Product Release and Transport
The previous section defined the fission product inventory in the core
available for release under design basis accident conditions. The goal of
this section is to trace the path of these fission products between their
point of release in the core and the environment. The mechanisms which
serve to impede the migration of the fission products through the reactor
vessel and the containment building will be discussed, and appropriately
conservative values for the release fractions for the dominant isotopes will
be proposed. Realistic values for these release fractions will be deter-
mined, to provide input to the best-estimate calculations of section 3.5.
Figure 3.3-1 illustrates the transport of fission products from the
molten fuel to the atmosphere. The melted aluminum, the primary coolant,
and the containment volume all act as barriers against the release of act-
ivity to the envirinment. Release fractions can be defined for all of the
possible barriers, as follows;
Ff, the release fraction from the fuel, is defined as the ratio of the
activity released into the primary coolant system to the activity in-
itially present in the four melted fuel plates.
F , the release fraction from the primary coolant, is defined as the
ratio of the activity released into the containment atmosphere to the
activity present in the primary coolant.
Fb, the release fraction due to removal processes within the contain-
ment atmosphere, is defined as the ratio of the activity available for
release from the building (to the environment) to the total activity
released into the building from the coolant.
Fpr, the release fraction from the building pressure relief system,
is defined as the ratio of the activity released into the stack ef-
37
ENVIRONMENT
REACTOR BUILDINGFe F
~CORE TANK
PRIMARYF COOLANT F6
PRESSURE
CLDLRENE
CONTAINMENT
Figure 3.3-1: Fission Product Transport Model
38
fluent to the activity available for release from the building.
In the original SAR analysis, two groups of fission products were con-
sidered: noble gases and iodines. Conservative release fractions were set
according to the generally accepted reduction factors for each of the two
groups. Table 3.3-1 gives the values of release fractions chosen:
TABLE 3.3-1
Original (SAR) Release Fraction Values under DBA Conditions
Noble Gases Iodines
F 100% 100%
F 50% 10%p
Fb 100% 50%
Fpr 100% 10%
It is important to note that Ff does not reflect the fact that only a
small number of fuel plates (four) have melted; Ff represents the fraction
of the fission product inventory held by those plates which is released to
the coolant. (Let the percentage of fuel that melts be denoted by F fm, to
avoid confusion. Ffm is not a transport-dependent parameter; it is fixed,
by the initial assumptions regarding the DBA, at approximately 1.5%. Refer
to Section 3.2 for a discussion of how F was determined.)
New values for the parameters listed in table 3.3-1 have been pro-
posed, based primarily on reactor accident experience and recent research
into fission product chemistry. These proposed values are discussed in
sections 3.3.1 through 3,3.4.
3.3.1 Release Fraction from the Fuel
Some attenuation of fission products by the fuel and cladding material
is expected, due to deposition upon and trapping within the molten metal. A
39
series of experiments was conducted by EG&G investigating fission product
release from failed light water reactor (LWR) fuel rods; it was concluded
that "fuel melting appears to reduce iodine release [from the fuel]" and
that "revision of the regulatory guides should be considered" (Ref, 19).
Furthermore, an extensive literature survey conducted by Nuclear Safety
Associates (Ref. 18) concluded that an attenuation factor of 2 due to core
materials represents a conservative assumption for pressurized-water react-
ors (PWRs).
Despite these conclusions, the release fraction from the fuel plates
for the MITR-II DBA will again be assumed to be 100%, The vast difference
in design of MITR fuel from LWR fuel does not permit direct extension of
the above results. Also, melting of the fuel was chosen as the DBA because
it was presumed that melting would result in the greatest release of fis-
sion products from the fuel; if this is not the case, as the EG&G study in-
dicates, and credit was taken here for this effect, it might be objected
that channel flow blockage leading to that percent of melting for which re-
lease of fission products was greatest should be the basis for design. A
sufficient amount of data does not exist which would justify decreasing the
release fraction from the fuel (Ff) from 100%. Accordingly, Ff is assumed
to be 100%.
3.3.2 Release Fraction from the Primary System
Reference 18 proposes a conservative value of 50 for the attenuation
of iodines by the primary coolant envelope (the surfaces and coolant) under
degraded-core accident conditions at a PWR, It is anticipated that the
actual attenuation may be as high as 10000, which was what was observed
during the first 24 hours following the accident at TMI-2,
40
The reason for the large discrepancy between the accepted values of
release fractions used in licensing and siting analysis and the observed
data from TMI-2 (and other facilities, as well reference 18) lies in the
chemical form of the iodine. At atmospheric pressure at temperatures below
600 *C, the dominant form of iodine in a cesium-iodine-hydrogen-oxygen en-
vironment is cesium iodide (rather than elemental iodine), assuming there
is sufficient cesium with which the iodine can react (Ref. 17,18). Account-
ing for fission product yields and decay schemes, it is seen that there are
about seven gram-atoms of cesium for every gram-atom of iodine, assuring
that all of the available iodine will react. Cesium iodide is extremely
soluble; 44 grams can be absorbed in 1 milliliter of water at room temper-
ature (Ref. 19). In this manner iodine is bound to the primary coolant
system of the reactor, even under the severe conditions experienced at
TMI-2.
In extending these conclusions to the MITR-II, four statements. can
be made supporting the conservatism of assuming a factor of 50 attenuation
of iodine (corresponding to FP = 2%) by the primary envelope:
1. The entire core inventory of iodine, about 1,2xI06 curies, when
bound in the form of cesium iodide, has a mass of approximately
3.86x10-6 grams. Given the solubility of cesium iodide mentioned
above (44 g/ml water at room temperature), it is apparent that the
cesium iodide inventory of the four melted fuel plates will read-
ily become trapped in the primary coolant system.
2. The conservative value of 50 for iodine attenuation by the pri-
mary system proposed in reference 18 was evaluated for a PWR
under accident conditions. The operating temperature and the
shutdown temperature of the PWR are, of course, substantially
41
higher than those encountered at the MIT reactor. In general,
fission product release decreases with lower temperatures, even
considering the decreased pressure.
3. The fact that the MITR-II primary system is open to the atmosphere
represents a possible increase in gaseous release relative to the
PWR of reference 18, but it does not significantly affect the as-
sumptions regarding cesium iodide release. The stepped concrete
reactor lid limits the bulk of the non-gaseous fission product
escape to the core purge system, which is automatically isolated
by solenoid valves under accident conditions of the magnitude of
the DBA. Also, despite the fact that large volumes of coolant
were blown down into the respective containments of TMI-2 and
Crystal River Unit 3 (thereby eliminating the attenuation effects
of the closed loop primary system), release fractions from the
coolant were still on the order of 10~4 or less (Ref, 18).
4. Since iodine release from the coolant system of a reactor is time-
dependent, much of the literature reviewed in reference 18 spoke
in terms of release rates rather than gross release fractions.
Reference 18 defines a " 24 hour decontamination factor " (one-
day DF) as " the ratio of the amount available as input to any
portion of the system divided by the output from that portion of
the system during the first 24 hours." The one-day decontamina-
tion factor for iodine from the primary coolant system of a PWR
was estimated to be between 50 and 10000.
Appendices A and B of this report have estimated the release rate
131 131of I from the reactor and the primary coolant I concentra-
42
tion, respectively. These results can be combined to yield an
equivalent one-~-day DF for the MITR-II primary coolant system for
iodine, under steady-state operating conditions. If 0,552 pCi of
l3I are released (through the sweeping action of the core purge)
from a primary coolant inventory of 5.678 mCi, the one-day DF is
4calculated to be slightly greater than 10 . This is comparable
to what was estimated by reference 18 for a PWR. It is not ex-
pected that this factor be greatly reduced under DBA conditions
(though some reduction will probably occur), due to the isolation
of the core purge system and the high solubility of cesium iodide.
This further demonstrates the appliability of the results of ref-
erence 18 to MITR-II accident analysis.
Given that a factor of 50 was conservatively proposed as the attenua-
tion factor for iodine from the primary coolant system of a PWR, and given
that the system differences between the MIT reactor and a PWR either have
little effect on or reduce the expected release from the MITR-II with re-
spect to the PWR, an attenuation factor of 50 for iodine release from the
MIT reactor primary coolant system is justified and considered conservative.
This corresponds to a release fraction of 2%.
No reduction in the present release fraction for noble gases can be
justified at this time, though accident experience has indicated that it
may be a factor of 100 lower than presently assumed for PWR licensing. A
release fraction of 5% rather than 50% will be assumed for the purposes
of best-estimate calculations. This is reasonable especially in light of
the assumption of 100% release of noble gases from the fuel, which is also
conservative but difficult to justify changing.
43
3.3.3 Attenuation due to Building Natural Processes
Attenuation by natural processes within the containment of a PWR redu-
ces the iodine concentration by a factor of 10 in the worst case, according
to the analysis of reference 18. It is expected that the actual attenua-
tion will be a factor of 10-20 higher. For the MITR-II, the dominant mech-
anisms by which this removal is accomplished are plate-out on containment
structures and gravitational settling. The removal of non-gaseous fission
products by any method is characterized by two time constants - one is on
the order of several seconds, when coolant - atmosphere interactions and
rapid aggregation and subsequent settling take place; the other time cons-
tant. is on the order of one hour, and represents the slower interaction of
fission products with materials encountered on the migration route to
building leakage points. The faster process is about equally'effective
as the slower in attenuating fission product escape (Ref. 16,18),
For the case of the MITR-II, credit can be taken for most of the re-
duction due to the 'faster' process, and some of the reduction due to the
'slower' process. The 'fast' process is driven partially by the temperature
difference between the coolant and the containment atmosphere, and partially
by the rapid formation of macroaggregate particles which are more suscep-
tible to gravitational settling. This aggregation occurs at the point of
release because it is there where the maximum particle concentration occurs;
at later times, diffusion has scattered the particles and aggregation is
less likely. The coolant - atmosphere interactions driven by the tempera-
ture difference between water and air will be less pronounced in the MITR-II
than in the PWR because of the substantially lower temperature difference
experienced at MIT, but some vapor condensation does occur and reduce the
44
expected release fraction from the coolant.
The 'slow process-which occurs during the first hour or two following
an accident, reflects deposition of fission product particles upon the ceil-
ing, floor, and walls of the containment building and upon whatever surfaces
exist in the stream of the escaping particles. The magnitude of this effect
depends on the distance a particle must travel between point of release from
the coolant and point of release from the containment; it is fair to assume
that the smaller radius of the MIT reactor containment with respect to that
of a PWR will cause a reduction in the effectiveness of this process; some
deposition will certainly occur, though, as fission product particles must
migrate at least the approximate 35 feet from the reactor pool to the con-
tainment inner wall (Ref. 1,16,18).
If a conservative value for attenuation due to both processes is 10,
and the two processes are about equally effective, then each process has
an associated attenuation factor of vTU, or about 3. Assuming a one-third
reduction in effectiveness for each process due to the above MITR-II system
considerations implies a net attenuation due to both processes of about 4.
It is probable that these effects account for a significantly greater re-
duction than anticipated here, but the processes are highly variable; a fac-
tor of 4 is considered the minimum expected attenuation, corresponding to a
release fraction of 25%.
Again, no reduction in the noble gas concentration is anticipated due
to building processes. The fact that they are gaseous and inert implies no
interaction with other materials and no significant gravitational settling.
Thus, even for the best-estimate calculation, Fb for noble gases will be
assumed to be 100%.
45
3.3.4 Release Fraction from Pressure Relief System
If pressure relief is initiated by the operations staff in the event
of an emergency, it can reduce the radiation exposure to the general public
by allowing controlled, filtered venting to replace uncontrolled, unfiltered
leakage as the primary means of escape of radionuclides. The pressure re-
lief filters are designed to remove 99% of the iodine from the effluent air
escaping through the system; replacement of the filters will be performed if
the observed efficiency for iodine release drops below 95%. Experience with
the filters has demonstrated a minimum efficiency of about 99.4%, dating
back to 1975. If the system were initiated tomorrow, it would be at least
99% efficient, so a release fraction from the system of 1% will be assumed
for the best-estimate calculations. However, if the filter efficiency tests
indicated a gradual decrease in effectiveness at some time in the future,
the replacement would not evolve immediately. It is then conceivable that
the efficiency might drop to as low as 95% in the interim between tests.
An efficiency of 90% is then not over-conservative for the purposes of
analysis, and the corresponding release fraction of 10% should continue to
be assumed. This added attenuation is, of course, only applicable to the
air that exits via the pressure relief system in the reactor exhaust system;
leakage from the building is not attenuated in this manner, and this is re-
flected in the dose calculations of chapter four.
As noble gases are unaffected by the charcoal filters in the pressure
relief filter bank, no attenuation of these gases occurs, and none is as-
sumed.
3.3.5 Summary
Accident experience has indicated that the release fractions used in
46
current dose rate estimates are grossly overestimated. The accident at TMI
Unit 2, for instance, released a factor of 1000 less iodine than WASH-1400
had predicted as its lowest estimate for that class of accident (Ref. 18).
About 0.01% of the long-lived fission product inventory was released to the
environment following the accident at the SL-1 facility in Idaho, in which.
20% of the core was destroyed (Ref. 9), SL-1 did not benefit from the pre-
sence of a containment building; releases would certainly have been even
less had one been utilized.
Accidents at the Westinghouse Test Reactor and the Materials Test Re-
actor involving limited fuel element failure also resulted in activity re-
leases far below those which had been anticipated. No iodine was detected
in the effluent following the destructive testing of the SNAPTRAN 2/10A-3
reactor, while releases from destructive testing at the SPERT-1 reactor were
restricted to less than 0.01% of the iodines and 10% of the noble gases
(Ref. 3,9). All accident data available supports the premise that present
release fraction estimates are over-conservative.
Table 3.3-2 summarizes the proposed release fraction parameters as de-
termined in the previous sections of this chapter:
TABLE 3.3-2
Proposed Release Fraction Values under DBA Conditions
Noble Gases Iodines
F 100% 100%
F 50% 2%p
Fb 100% 25%
Fpr 100% 10%
47
3.4 Calculation of Dose Rates
3.4.1 Inhalation Dose
The dose rate to the thyroid gland in time 't' is given by
equation 3.4.1: (Ref. 14 )
(Eq. 3.4.1): D RPKcAU
tDs_ ).j (t)dt
all AT i p i 0iodines
where D is the dose rate in rads,
R is the breathing rate in m3 /sec,
P is the reactor power in megawatts,
K is the net fraction of the core inventory available
for release from the building,
c is a geometric correction factor accounting for the
size of the observed pressure wake,
A is the cross-sectional area of the building normal
to the wind direction, in m2 and
u is the average wind velocity in m/sec.
Values of Dx and -S depend only upon the isotope, and are
AT P
discussed in section 3.2.1. X(t) is the leakage rate from the
building, and it is derived in section 5.3.2.1 of the SAR.
It has been proposed that the value of 'K' be reduced by
a factor of ten from the original analysis. The proposed value
of 'K' is given by:
K = Ffm - Ff F p * Fb
= .015 -1 .02 & .25
= 7.5x10-5
(Eq. 3.4.2):
48
With this proposed value of K' and the other parameter values taken
from section 5.3.2.2 of the SAR, the total dose to the thyroid received in
two hours by a person standing in the path of the cloud generated by build-
ing outleakage is calculated to be 0.028 rads,
3.4.2 Dose from Release of Gases from the Stack
Table 3.4.2-1 gives the maximum ground level concentrations (for the
dominant isotopes) averaged over two hours, based on the proposed release
fraction out the stack of 7.5xl0-6 for the iodine isotopes, The format of
this table follows closely the format of Table 5.3.2.3-1 of the SAR, as the
calculations proceeded in an identical manner.
-6The release fraction for iodine of .7,5x10 was arrived at as the pro-
duct of Ffm, Ff, Fp, Fb, and Fpr; all mechanisms for the removal of fission
product radionuclides are in operation for this release path.
A person standing at the point of maximum ground level concentration
for two hours will receive less than 1.5% of the permissible annual expo-
sure for unrestricted areas as specified by Appendix B of 10CFR20. This
corresponds to an additional dose of less than 10 millirem in two hours.
3.4.3 External Gamma Dose
Fission product radionuclides which plate out within the containment
building as a result of 'natural processes' still contribute to the gamma
dose resulting from scattering. Attenuation credit, therefore, can Onl& be
taken in this instance as the product of Ff and F ; the overall release
fraction is Ffm * Ff o Fp, or about ,0003 for iodines and about .0075 for
noble gases.
Since the analysis of the external gamma dose component assumes uni-
form distribution of the released fission product inventory within the
TABLE 3.4.2-1
Ground Level Concentrations from Gaseous Stack ReleasesMax Ground Ratio Max
Isotope [gs/P] (Ci/MW) Release Fraction Released Source Conctrn.*(ICi/cc) to Perm.t
Iodine -6 -10131 25100 7.5x10- 0.94 3.J7xlq .00071
6 -10132 38100 7.5x10 1.43 4.80x10 .000004
-6 -10133 56300 7.5x10 6 2.11 7.12x10 .00041
751 6 -10134 65800 7.5x10 2.47 8.32x10 .00003
6 -10135 51000 7.5x4Q6 1.91 6.45x10 .00015
Xenon-9
131m 250 .0075 9.4 3.16x10 .000002-8
133m 1340 .0075 50.2 1.69x10 .00001-7
133 47100 .0075 1770 5.96x10 .00045
1 3 5m 15300 .0075 573 1.93x10~7 .00147
135 46700 .0075 1750 5.90x10 .00135
Kryptonm -8
83 4170 .0075 156 5.26x10 .00040
85m 12900 .0075 484 1.63x10 .00037
87 23400 .0075 878 2.96x10 .00338
88 31900 .0075 1200 4.04x10 .00461-------------------- ----------------------------------------------------------------- -
* ground level concentration averaged over two hour period
t ratio of maximum ground level concentration to permissible concentrations based on two hour exposure
50
containment, the time-integrated source term Si comes out of all the scat-
tering integrals, which implies a linear relationship between Si and the
calculated dosage levels;?, Si is given in reference 3 by:
TSR-(Eq. 3.4.3-1): Si - P ' exp(--; t) dt
Ci
PSR-= ~RL [1 - exp(-XiT)]
XiC
Thus, for a given exposure time (two hours in this analysis, for ex-
ample), Si is equal to a constant times SRi, which is the released source
strength. SRi is, in turn, simply the product of the initial source
strength So (fixed for a given isotope) and the overall release fraction
for the external gamma dose component. Superimposing all these linear re-
lationships implies that external gamma ray dose varies linearly with the
release fraction for a given isotope.
The change in iodine transport assumptions did little to alter this
component of the dose. The two-hour gamma dose at the nearest point of
public occupancy was calculated to be 0.6 rads, as compared to 0.9 rads
calculated in the original analysis. More generally, the dose rates from
the original analysis, illustrated in a graph of radiation level versus
distance from the building in chapter five of the SAR, have been conser-
vatively re-estimated to -63% of the original values. Figure 5.3.3.7-1 of
the SAR, "External Gamma Doses .Outside Containment", properly represents
the gamma dose for the proposed release fractions if the ordinate is scaled
down by a factor of 0.63.
51
3.5 Best-estimate Approximation of Dose Rates
The assumptions governing this approximation are:
1. Fpr is equal to 1% for iodine, corresponding to the minimum
filter efficiency of 99%.
2. Fission product activities are assumed to be those reached after
105 hours of full power operation, rather than saturated values.
(Many have reached saturation at 105 hours, of course.)
3. The building leak rate during the past four years has not exceed-
ed 0.6% per psig overpressure per day. It will then be assumed
that leakage is 0,6%, There are two effects of this assumption;
the inhalation dose drops, first,'because of the decreased leak
rate, and second, because flow past the pressure relief filters
drops slower, resulting in greater iodine removal from the system.
4. The scattering calculations presented in the SAR ignore attenua-
tion of the fission product gammas in air. Also, scattering is
assumed to take place in the direction that will maximize dose.
-These assumptions are unnecessary.
5. The photon fluxes are consistently over-estimated to facilitate
calculation.
6. Fp and Ffm have been overestimated for the noble gases. A fur-
ther reduction of 10 can be expected because of only partial
fuel failure.
On the basis of these assumptions, the dose at the exclusion area
boundary for two hours is approximately 70 millirads. Again, the major
contribution to the dose comes from the gamma source in the building.
52
Throughout this analysis, it has been assumed that the building pres-
sure rose to 2 psig above atmospheric at the same time channel flow block-
age occurred. If an initial overpressure of 0.5 psig were assumed, the in-
halation dose due to building outleakage (with the pressure relief system
in operation) would be less than 2 millirads, under the proposed assumptions
regarding the release fractions presented in this thesis. This illustrates
the conservatism of the 2 psig overpressure assumption.
53
CHAPTER 4
CONCLUSIONS
4.1 Import of this Thesis
This thesis has provided a strong argument in support of designating
channel flow blockage as the maximum credible accident for the MIT research
reactor. The levels of defense incorporated into the design and procedural
control of the facility result in extremely low probabilities for accidents
of greater magnitude - in particular, core dryout.
The analysis of the channel flow blockage accident presented in the
1970 Safety Analysis Report was reviewed and two general flaws noted.
First, the release fractions chosen to characterize the accident were over-
conservative for certain isotopes, leading to over-statement of the result-
ant dose rate levels. Second, necessary conservatism in the analysis tend-
ed to propagate; while a twenty percent margin of safety applied to one
parameter is certainly reasonable, the effect of combining four such par-
ameters, each with a twenty percent safety margin, could result in an error
of more than 100%. There are several instances in the SAR analysis where
this occurs, and little can be done to rectify the situation; an inspection
of constituent assumptions reveals no gross overconservatism in the safety
margins. The re-assessment of the release fractions was a much more tract-
able problem.
Other minor problems with the original analysis existed, The reference
from which gamma ray energy data was drawn predated semicinductor detector
technology, and varying degrees of inaccuracy were found when this data was
compared with that from a current reference. For calculations involving
directly the average or total gamma ray energy of a nuclide, correction was
10 54
trivial; for calculations involving gamma ray energy-dependent parameters
buried deeply within the early stages of the analysis, corrections were not
made. In one calculation (-external gamma dose component), it was found that
the average linear and energy attenuation coefficients were determined by
finding the coefficient's value at the average gamma ray energy, an inappro-
priate approximation. Three isotopes were chosen, and their actual average
attenuation coefficients calculated for purposes of determining the magni-
tude of the error; it appears to be between 10 and 30% in both directions.
This error propagates through the entire calculation, but was unfortunately
discovered too late to allow a complete re-evaluation of the results.
Rather than repeat the cumbersome analytical methods with new values for the
flawed data, it is recommended that a nutrferical solution be sought, taking
advantage of the recent development of computer models of fission product
transport and release.
It is probable that the dose rates predicted by this thesis are also
over-conservative, but not enough data exists yet to substantiate further
reductions in release parameters. The following was achieved in terms of
better approximating the radiological consequences of a MITR-II design ba-
sis accident:
1. The conservative estimate of the two-hour inhalation dose due to
building leakage (with the pressure relief system in operation)
was reduced by a factor of ten, from 0.280 rads to 0.028 rads.
2. A 47% reduction in the conservative estimate of the maximum ground
level concentrations due to the release of gases through the stack,
was determined to be conservative. The concentration dropped from
2.52% permissable to 1,34% permissable,
3. The conservative estimate of the gamma ray dose at the point
55
of closest public occupancy was reduced from 0,9 rads to 0.6 rads.
The dose due to the gamma ray source inside the containment was
reduced at all points outside the containment by the same amount,
about 33%.
It is concluded that lower estimates of the radiological consequences
of this design basis accident than those proposed in this thesis are not
conservatively justified at this time.
4.2 Applicability to Emergency Plan
The channel flow blockage accident described in this report represents
a reasonable point of departure for evaluating the level of emergency plan-
ning required by the MIT reactor. The dose rates calculated in chapter 3
can be compared to the radiation action levels- specified in the " Proposed
American Nuclear Society Emergency Planning for Research Reactors " (Ref.
2), to determine the level of readiness that is warranted.
It is recommended that a computer model of the external gamma dose
source term be utilized to verify ( or correct, if necessary) the calculated
dose rates at the exclusion area boundary, before exemption from certain
emergency plan requirements is sought; some errors were discovered in the
analysis of this component, which, in light of the reduction to the inhala-
tion and stack gas components proposed in this thesis, has become the dom-
inant contribution to the public dose.
56
References
1. MIT Nuclear Reactor Laboratory Staff, " Reactor Systems Manual ",
MITNRL-004, January 1980.
2. Proposed American Nuclear Society Emergency Planning for Research
Reactors ", Draft 2 of ANSI/ANS 15.16, November 29, 1981.
3. " Safety Analysis Report for the MIT Research Reactor (MITR-II) ",
MITNE-115, October 22, 1970.
4. " Technical Specifications for the Massachusetts Institute of Tech-
nology Research Reactor ", Appendix A to Facility License No. R-37,
July 23, 1975.
5. " Abnormal Operating Procedures for the MIT Research Reactor ", March
25, 1981.
6. interview with Professor Peter Griffith, May 1982,
7." Reactor Safety Study - An Assessment of Accident Risks in U.S.
Commercial Nuclear Power Plants ", WASH-1400 (NUREG-75/014) October
1975.
8. Park, Chang Kue, Reliability Analysis of the Emergency Core Cool-
ing System of the MITR-II ", Nuclear Engineer's Thesis, May 1982,
9. Thompson, T.J. and J.G. Beckerly, (Eds,), The Technology of Nuclear
Reactor Safety, Vol. I, MIT Press, Cambridge, Massachusetts (1964).
10. Allen, George C., " Dosimeter Dropped into Core Tank ", (UOR 75-2)t
January 29, 1976.
11. Clark, Lincoln, " Foreign Material in Primary System ", (UOR 76-2),
May 7, 1976.
12. Bernard, J., G. Hughes, and S, Dubnik, " Improper Handling of MITR-II
Equipment ", (UOR 81-3), September 15, 1981.
13. MITR-II Reactor Physics File: review of data on core power profiles
through March 1982.
" Calculation of Distance Factors for Power and14. DiNunno, J.,J. et al.,
57
Test Reactor Sites ", TID-14844, March 23, 1962.
15. " Technical Bases for Estimating Fission Product Behavior During LWR
Accidents ", NRC report, March,1981.
16. Hilliard, R.K. et al., " Removal of Iodine and Particles from Contain-
ment Atmospheres by Sprays - Containment Systems Experiment Interim
Report ", BNWL-1244, Battelle Northwest Pacific Institute (1970).
17. Osetek, D.J., et al., " Fission Product Release Signatures for LWR
Fuel Rods Failed During PCM and RIA Transients ", EG&G Idaho, Inc.,
1979.
18. Rodger, Walton A., " The Second Cross ", Energy Progress, March 1982.
19. " Handbook of Chemistry and Physics ", 61st edition, Chemical Rubber
Company Press, 1981.
58
APPENDIX A
131Determination of I Release Rate to the Ekhaust Plenum
The MIT Radiation Protection Office continuously monitors the exhaust
air leaving the containment building both before and after the main bank of
exhaust filters. A small volume of air is drawn through a pancake tube-
type detector and gross count rate is indicated remotely in the control
room. In addition, the air is drawn across particulate and charcoal fil-
ters which are subsequently counted for a more detailed breakdown of the
radioisotopes being released. The data presented in this appendix reflects
the sampling of the charcoal filters alone.
Figure A-1 describes the mechanics of the sampling operation, The
charcoal filters are placed in the flow path at time t=0, and removed at
131time t=t with an accumulated I activity of A microcuries. The fil-
ters are then allowed to decay until time t=t C, when they are counted for
about 25 minutes. Output from the MITR RPO computer is thus A microcuries,
the measured activity of the filters after tfr days of exposure and
Ct - tfr) days of decay. It is desired to calculate the steady-state re-
lease rate of the 131I, given A tnr, and t
The relationship between A and A is given by:m 0
(eq. Al); A = A exp -A(t - t )],in o c fr
where X is the disintegration constant for 1311; tc and tfr are de-
fined above.
131If the steady-state Cinstantaneous) release rate of I is A, the re-
lationship between A and A is given by:0
fr(eg. A2): A = I [A - exp(:-At*)]-dt*
0o
59
Activity(pCi)
A 0
Am -
0 tfr tc time (hours)
Figure A-1
131Charcoal Filter I Activity as a Function of Time
60
where t* is a dummy variable of integration.
Integrating equation A2 to solve for A and using this result in equa-0
tion Al, we derive a direct- relationship between the desired value A and
the known parameters A , t, tf, and A:
(eq. A3): A = -AA' expAX'Ct - t f})] .[exp(-X'tf) - 1-
Two such samplings of charcoal filters were performed recently, and
the following data was recorded:
-lt (hrs.) t (hrs.) A (1Ci) .(hrs.. )fr c _
-3Case 1: 168 266.27 0.3383 x 10 3.59 x 10
Case 2: 168 316,12 0.2489 x 10-3 3.59 x 10-
3
3
Substituting these values into equation A3, we find A:
-6Case 1: A = 3.81 x 10 Ci/hour
-6Case 2: A = 3.36 x 10 pCi/hour
It is necessaiy to clarify the exact meaning of this number. 'A' re-
131presents the instantaneous deposition rate of I upon the charcoal fil-
ters in the exhaust stream sampling facility. The flow rate of air past
these filters is maintained at 30 liters/minute, while the actual building
5 5exhaust flow is about 1.415 x 10 liters/minute, Thus only t30/(1.415x10 )
x 100] percent of the escaping 131I is actually detected- allowing for this,
we find that the averaged actual continuous release rate for 131 is
0.017 pCi/hour, or 2.84 pCi/week,
Notes: All references within this appendix to 'escaping' or 'released' re-fer to activity reaching the exhaust plenum, before the exhaust fil-ter banks and before the fast-acting damper which isolates the
building on receipt of an abnormally high effluent radiation level.
61
Notes: (continued) It is important to know the actual dates when each of
the samples was taken, counted, etc. The time corresponding to t=0
as illustrated in figure A-1 was 1300 on 12 March, 1982 for case 1
and 1300 on 19 March, 1982 for case 2. The filter removal times or
the times at which the samples were counted can be determined with
the data (t c and t f) given on the previous page.
Two charcoal filters, one from each of two parallel sampling systems,
were counted together. Each charcoal filter has an efficiency of
50% for the collection of iodine, so the net efficiency is 2 times
50 equals 100%.
The normal procedure in measuring the activity of these charcoal
filters requires that the filters be replaced once weekly, early
Friday afternoon. Thus, no .distinction can be made between steady-
state releases while the reactor is operating and latent releases
while the reactor is shutdown, To obtain an estimate of the acti-
vity which is released while at full power, the charcoal filters
were removed and replaced one Monday (17 May, 1982); this effect-
ively separated the week into its operating period and its shutdown
period, and allowed a comparison between operating period and shut-
down period releases, It was found that iodine release at power
constitutes about 80% of the weekly release to the stack. This is
most likely a result of the higher water temperature causing greater
evaporation, as well as aeration of the iodine species by -the rapid
flow of coolant upwards through the core. Given this data, it can
be said that the maximum operational release rate fraction of iodine
from the coolant is about 0.552 pCi/full-power day.
62
APPENDIX B
131Determination of I Concentration in the Primary Coolant
The MIT Radiation Protection Office (RPO) analyzes a primary coolant
sample each week to monitor fission gas levels and the presence of activa-
24 16tion products such as Na and N. For the purposes of this report, 100
milliliter samples were taken in lieu of the usual 10 milliliter samples,
to decrease the amount of counting time required to achieve reasonable
statistical accuracy. Samples were taken on five consecutive days (Monday
to Friday) to quantify any variation of the concentration with time. The
samples were counted at varying times after they were taken, and the decay
time therein is accounted for.
Raw data from the RPO computer is given i-n gross activity of the sam-
ple. A geometrical correction factor f must be applied to compensate for
the larger than normal volume of coolant. If 't' is the time the sample is
allowed to decay, 'A ' the measured activity of the sample, and 'X' them
131 131disintegration constant for I, then the concentration of I in the
primary coolant is given by:
(eq. B1): C - A - exp( -Xt) -f (pCi/ml)100 m g
The data is summarized below:
ADay t (hrs) m (pCi)
Monday, 3/22/82 6.00 0.154x10 2
Tuesday, 3/23/82 52.57 0.172x10 2
Wednesday, 3/24/82 200.36 0.121x10 2
Thursday, 3/25/82 174.48 0.143x10-2
Friday, 3/26/82 151.68 0.161x10 2
average over week 0. 150x10 -2
63
The volume of the primary coolant flow system is about 1000 gallons, or
3.785x106 fil. At .0015 pCi per ml, the gross coolant activity of 1131 is
5.678 mCi.
64
'APPENDIX C
Magnitude of the Deuterium-Oxygen Explosion in the D70 Reflector
In order to estimate the energy released by a deuterium-oxygen explo-
sion in the stagnation volume (.see Figure 2-1) in the heavy water reflect-
or region, we first use the ideal gas law to obtain the number of moles of
gas contained in the bubble:
(Eq. Cl): P * V = n - R T
P, the pressure within the bubble, is found by adding the
hydrostatic head pressure of D20 above the bubble to the
pressure above the reflect'or pool (app. atmospheric):
(Eq. C2): P = Pa +pgh (lb/ft-sec2 )
= 68161 + (68,98) (32.2) (2.5)
= 73714 lb/ft-sec2 = 1.082 atm.
V, the volume of the bubble, is approximated by considering
its shape to be a right circular cylinder with a radius of
7 inches and a height of 0.8 inches. V is then 0.071 ft3,
or about 2 liters.
R, the universal gas constant, is 0.082 liter-atm/mole-*K.
T, the temperature of the gas in the bubble, ranges from
20 0C at shutdown up to 50 *C at full power; the lower temp-
erature implies higher gas density, and thus represents the
worst case. T must be expressed in degrees Kelvin: 293 *K
n, the number of moles of gas, is the desired quantity.
Solving for n given the above values, it is seen that 0.09 moles of
gas will be present in the bubble. If the gas is in its most potentially
65
explosive state, there will be two molecules of D2 for each molecule of -
02, and no other gases present, Thus, there will be '0,06 mnoles of D 2 , and
0.03 moles of 02; the reaction will then be described as-
(0.06)D2 + (0,03)02 + (0.06)D20
It is conservatively assumed here that all the gas present reacts,
The heat of fusion for D20 is given in the 61st edition of the Handbobk of
Chemistry and Physics as 1518 calories/mole D20; thus, the energy released
in the explosion is:
E = (0,06 moles) - (1518 cal. /mole)
= 91 calories
= 381 watt-seconds
a 285 foot-pounds,
This is roughly equivalent to dropping a ten pound weight from the
top of a three story building. Since this energy will be dissipated among
the core tank, the heavy water in the reflector, and the reflector tank,
it is not expected that this explosion will result in gross rupture of the
primary vessel.
