1990-12 – NRC – NUREG CR-5653 – Recriticality in a Boiling Water Reactor (BWR) Following A Core Damage Event

1990-12-nrc-nureg-cr-5653-recriticality-in-a-bwr-following-a-core-damage-event

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MARTIN MARIETTA ENERGY SYSYEMS LIBRARIES
3 445b 0354S33 4
NUREG/CR-5653
PNI^7476
Recriticality in a BWR
Following a Core Damage Event
5
E
Prepared by
W. B. Scott, D. G. Harrison, R. A. Libby, R. D. Tokarz/PNL
R. D. Wooton, R. S. Denning, R. W. Tayloe, Jr./BMI
Pacific Northwest Laboratory
Battelle Memorial Institute
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Recriticality in a BWR
Following a Core Damage Event
Manuscript Completed: November 1990
Date Published: December 1990
Prepared by
W. B. Scott, D. G. Harrison, R. A. Libby, R. D. Tokarz, Pacific Northwest Laboratory
R. D. Wooton, R. S. Denning, R. W. Tayloe, Jr., Battelle Memorial Institute
Pacific Northwest laboratory
Richland, W A 99352
Subcontractor:
Battelle Memorial Institute
Columbus, OH 43201
Prepared for
Division of Systems Research
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555
NRC FIN 82930
NUREG/CR-5653
PNL-7476
RC, Rl, R3
3 4456 0324533 4

ABSTRACT
This report describes the results of a study conducted by Pacific
Northwest Laboratory to assist the U.S. Nuclear Regulatory Commission in
evaluating the potential for recriticality in boiling water reactors (BWRs)
during certain low probability severe accidents.
Based on a conservative bounding analysis, this report concludes that
there is a potential for recriticality in BWRs if core reflood occurs after
control blade melting has begun but prior to significant fuel rod melting.
However, a recriticality event will most likely not generate a pressure pulse
significant enough to fail the vessel. Instead, a quasi-steady power level
would result and the containment pressure and temperature would increase until
the containment failure pressure is reached, unless actions are taken to
terminate the event.
Two strategies are identified that would aid in rega1n1ng control of the
reactor and terminate the recriticality event before containment failure
pressures are reached. The first strategy involves initiating boration
injection at or before the time of core reflood if the potential for control
blade melting exists. The second strategy involves initiating residual heat
removal suppression pool cooling to remove the heat load generated by the
recriticality event and thus extend the time available for boration.
iii

SUMMARY
The U. S. Nuclear Regulatory Commission (NRC) initiated a program to
investigate the potential for recriticality in boiling water reactors (BWRs)l
as a part of its Accident Management Program. This report d~scribes the
results of a study conducted by Pacific Northwest Laboratory (PNL) to assist
the NRC in evaluating:
• The potential for recriticality in BWRs during low probability
severe accidents.
• The severe accidents that create a potential for recriticality.
• The possible consequences of the recriticality.
• The strategies for regaining control of the reactor when
recriticality is a potential concern.
Due to modelling and phenomenological uncertainties related to void
fraction, debris bed size, particle size, etc. and the lack of an analytical
tool capable of performing the complex analyses required to address the
complex interactions of these parameters, a conservative bounding analysis was
conducted. If the conservative bounding analysis indicates acceptable
consequences (e.g., no recriticality or a recriticality event resulting in a
benign pressure pulse that is non-threatening to the integrity of the
containment and accident management strategies can be implemented to
successfully prepare for and terminate the event), further, more sophisticated
model development and analyses may not be necessary to resolve the modelling
and phenomenological uncertainties. However, if the bounding analysis
indicates unacceptable consequences (e.g., a significant recriticality event
that creates a large pressure pulse and potentially fails the containment),
further research should be identified to resolve the uncertainties and
phenomenological issues.
The report concludes that there is a potential for recriticality in BWRs
during certain low probability severe accidents, but the recriticality event
will most likely not generate a pressure pulse significant enough to fail the
vessel. Instead, a quasi-steady power level would result and the containment
pressure and temperature would increase until the containment failure pressure
lFor a pressurized water reactor, core reflood is normally accomplished using
borated water supplies and recriticality is generally perceived not to be
very credible. However, for a BWR, reflood is normally accomplished using
unborated water; and recriticality is believed to be credible. Therefore,
this report addresses the potential for recriticality events only in BWRs.
2PNL is operated for the Department of Energy by Battelle Memorial Institute.
v

is reached, unless actions are taken to terminate the event. Two accident
management strategies are identified that would aid in terminating the
recriticality event before containment failure pressures are reached. The
first accident management strategy is to initiate boration injection at or
before the time of core reflood. The second accident management strategy is
to initiate residual heat removal (RHR) suppression pool cooling as quickly as
possible so as to remove some of the heat load and thus extend the time
available for boration.
The following sections summarize each of the recriticality evaluation
areas identified above.
Potential for Recriticality
Analyses of severe accident phenomenology in BWRs indicate that, while
the core is in the process of heating and melting, one of the earliest
components to melt and relocate are the steel blades that contain the B4C
control material. Because the control material is encased in stainless steel
blades that have a much lower melting point than the zirconium fuel cladding
and the fuel itself, as the core heats up, the control material may melt and
leave the core. If the core were to be reflooded following the relocation of
the control blades and prior to the relocation of the core into a rubble bed,
the possibility exists for the core to become critical again without an
adequate means of control.
Without adequate training and proper procedures, the operations staff
may be very surprised and confused to find that their actions to recover core
cooling may have entered them into anticipated transient without scram (ATWS)
conditions. This type of confusion or lack of training can cause the
operations staff to disbelieve instrumentation and alarms and to take
inappropriate actions. Recognition that this type of event can occur and
development of accident management strategies to handle these events can aid
in the prevention, mitigation, and termination of such events.
Accident Scenarios for Recriticality
Primarily using the NUREG-1150 risk study for the Peach Bottom plant as
the technical basis, severe accident sequences with the highest predicted
frequencies were selected to characterize the conditions that would likely
contribute to core melt scenarios where recriticality may be possible. The
accident sequences consisted of station blackout and ATWS events.
Station blackout is defined as the loss of all ac power, except that
which is powered through an inverter from the station batteries. In this
context, the station blackout involves the loss of both the normal ac power
source from the offsite electrical grid and the emergency ac power source from
the onsite diesel generators. From this point, station blackout events are
vi
divided into two groups based on the timing to core damage. The first group,
called short-term station blackouts (SSBOs), includes those accident sequences
where core damage begins within 1 hour of the initiating transient or event.
In the short-term station blackout sequences, either the station batteries
fail or the high pressure coolant injection (HPCI) and reactor core isolation
coolant (RCIC) pumps independently fail early in the sequence. If the station
batteries are failed at the time of the station blackout, the loss of de power
will also fail the ability to depressurize the reactor and causes a loss of
vital instrumentation. If the batteries are available but the HPCI and RCIC
pumps fail, depressurization and vital instrumentation may be available.
The second group of station blackout sequences are the long-term station
blackout (LSBO) events, where core damage occurs after 1 hour, typically 9 to
12 hours after the initiating transient or event. Core damage occurs after
the station batteries are depleted. These two groups of station blackout are
essentially similar in consequences with the exception that core power is
lower in the long-term sequences due to decay. Once all electrical power is
lost, the ability to cool the core is lost. Water level decreases, core
temperature increases, and core damage results.
For both station blackout groups, in which core damage has begun, if ac
power is restored and unborated coolant injection is initiated within the time
window between the beginning of control blade melting to the beginning of fuel
rod melting, the potential for a recriticality event to occur may exist.
Since core damage will proceed from the central region of the core radially
outward, the potential for recriticality in the outer regions may occur at a
much later time than that for the central region and in fact may occur after
fuel rod collapse and debris bed formation within the central region.
Therefore, the recriticality time window was conservatively estimated to be
the time from the start of control blade melting to the time of vessel
failure. For short-term station blackout sequences, the time window is from
91 to 161 minutes long, starting 109 to 127 minutes after the initiating
event, respectively. For the long-term station blackout sequences, the time
window is approximately 118 minutes long, starting over 600 minutes after the
initiating event. It is estimated that between 12% and 1% of the time,
depending on the specific sequence, ac power will be restored and coolant
injection will be initiated within the recriticality time window.
An ATWS event occurs when, upon receipt of a scram signal following an
unspecified transient, the control rods fail to insert into the core due to a
mechanical failure of the rod control system. In these sequences, manual
insertion of the control rods is unsuccessful. In some ATWS sequences,
various systems used to recover from an ATWS (e.g., standby liquid control
system and the high pressure coolant injection system) fail and allow the
water level to drop until core overheating and damage begin. If coolant
injection is subsequently initiated, recriticality becomes possible. However,
for ATWS scenarios without boration (or inadequate boration), the containment
will eventually fail and core melt will occur regardless of the occurrence of
a recriticality event. If adequate boration does occur, the potential for
vii
recriticality is possible only if the boron concentration is diluted by
extended injection. Therefore, recriticality during an ATWS does not appear
to be the major concern, prompt termination of the ATWS is.
Consequences of Recriticality
Analysis showed that without the control blades, relatively high
reactivities are possible with standing fuel rods or over a broad range of
fuel particle sizes and fuel volume fractions for both unborated and fairly
heavily borated reflood conditions. The consequences of the potential
recriticality are important to estimate quantitatively for selecting the most
appropriate and effective accident management strategies. If the potential
consequences are very severe, it might be preferable not to reflood the core
if the only available water supply is insufficiently borated. If the
consequences are minor, the current procedures to reflood immediately upon
recovery with the maximum flow rate of water (borated or unborated) is a
necessary approach.
The primary concern is, of course, a super prompt-critical excursion
which would result in rapid disintegration of fuel, rapid molten fuel coolant
interaction, and the production of a large pressure pulse capable of directly
failing the reactor vessel. The analyses conducted in this study indicate
that the rapid disintegration of fuel is not likely under the conditions of
reflooding a hot core, which may or may not be degraded. Analysis also
indicates that a maximum power excursion produces a fuel enthalpy of 73 cal/g,
corresponding to a temperature rise of 1300°F in the fuel. Doppler feedback
is the principle mechanism for terminating rapid transients in low enriched
uranium-water systems and is adequate to limit the energetics of reflood
recriticality to a level below which the reactor vessel would be threatened by
a pressure pulse.
If the reactor remains critical following an initial excursion at the
time of reflooding (i.e., reflood is conducted without boration), it will
either enter an oscillatory mode in which water periodically enters and is
expelled from the core or it will approach a quasi-steady power level. In
either case, the average power level achieved will be determined by the
balance between the reactivity added and the feedback mechanisms. Based on
the analyses conducted in this study, a recriticality event is likely to
produce core power levels less than about 20% of normal power (and probably
not much more than 10% of normal power), but may be significantly above the
decay heat level (~2% after 15 minutes).
The main concern of remaining critical during and after reflood becomes
the increasing temperature of the suppression pool and the potential for
containment over-pressurization. Without the RHR system providing suppression
pool cooling and assuming that the power level is at 10% of full power,
analysis indicates that the containment will be over-pressurized in slightly
more than a half hour. With full RHR suppression pool cooling capacity
viii
utilized, the excess steam (i.e., that above the RHR capacity) to the
suppression pool would represent only 3% of full power. Under these
conditions, over two hours are available to shutdown the reactor before the
containment would be over-pressurized. In either case, if the reactor is not
shutdown, the containment will become over-pressurized and the suppression
pool will reach saturation conditions, which will cause the core cooling
systems to fail. This could subsequently lead to further core damage and a
direct release path to the environment. (This situation is true for the Peach
Bottom plant, which was used as the reference BWR. It is recognized that
newer plants may have low pressure systems that can operate under saturation
conditions).
To shutdown the reactor, without the availability of control blades
(which may have relocated from the c~0e), boration is required. Analyses
indicate that approximately 700 ppm B are required to ensure subcriticality
for all conditions, including standing fuel rods. The standby liquid control
(SLC) system, which is the primary method of emergency boration, is designed
to provide boration in one-half to two hours, depending on the flow rate of
the boration pumps. Therefore, the boration rate appears to be marginally
adequate to avoid containment over-pressurization, if it is initiated at the
same time as the core reflood and if the boration concentration is adequate to
terminate the reaction. The time allowed for boration is increased if RHR
suppression pool cooling is utilized at full capacity at the time of core
reflood. However, the use of RHR at full capacity in the suppression pool
cooling mode would require that reactor vessel coolant injection be performed
by another system (e.g., HPCI, RCIC, or low pressure core spray).
It should be noted that emergency boration, under the conditions
described above, does not prevent the occurrence of a recriticality event, but
rather, terminates the event and prevents any severe consequences from the
event. To prevent the potential for recriticality, the boration must occur
prior to core reflood. However, as stated previously, recriticality is not
expected to fail the vessel and the main concern becomes the continuance of
the event to the point of containment failure.
Recovery Strategies
Recovery of control in a postulated station blackout or ATWS event is a
primary concern. Two accident management strategies for terminating a
recriticality event are identified in this study and are described below.
The first recommended strategy is to borate at the time of reflood for
core damage events where control blade material may have relocated from the
core or not inserted. The first boration alternative in BWRs is to use the
SLC system, which is normally borated for response to ATWS events.
ix
If the SLC system is unrvailable or fails, numerous alternate methods of
boration could be considered. These alternative methods include connecting
the SLC tank to the HPCI turbine-driven pump suction using temporary
connections (such as firehoses and appropriate fittings) for injection into
the core, or boration of the injection water supply (i.e., the condensate
storage tank). In the latter strategy, a large quantity of sodium pentaborate
would be stored in a location convenient to the tank and equipment and
procedures would be in place 18 quickly borate the water supply to the
appropriate concentration of B. Since the condensate storage tank is the
normal suction source for the HPCI and RCIC system, temporary connections
would not be necessary to use these systems for boration injection.
Other options include depressurizing and using low pressure systems.
However, all low pressure systems are presently dependent on ac power, which
may not be available in some scenarios (e.g., continued station blackout). A
low pressure system pump that is not completely dependent on the normal ac
power supplies (e.g., turbine-driven low pressure pump, de-powered pony motor
for pumps, dedicated diesel generator, etc.) could alleviate this concern.
The second accident management strategy involves suppression pool
cooling. For heat removal in ATWS events, full RHR suppression pool cooling
is established as quickly as possible. RHR is capable of removing more than
7% of full core power. This strategy, while usually applied to ATWS events,
is equally effective for severe accidents where control material may have
relocated from the core. The use of RHR would greatly extend the amount of
time available to terminate the recriticality event and in so doing prevent
the containment from failing. Such a strategy presumes the operability of the
RHR system and the availability of ac power supplies. In addition, using the
RHR system in the suppression pool cooling mode requires that another system
be used for injection into the reactor vessel.
Impact of Implementing Strategies
The effect of the above accident management strategies on the
probability of containment failure due to over-pressurization was investigated
to determine the benefit of implementing these strategies. As stated earlier,
the dominant accident sequences for Peach Bottom are short-term and long-term
station blackout events, with core damage frequencies of 4.5E-6 per reactor
year and 1.7E-6 per reactor year, respectively. The potential for
recriticality following station blackout exists if ac power is restored and
unborated coolant injection is initiated within a time window of potential
recriticality. It was estimated that between 12% and 1% of the time,
depending on the specific sequence, ac power would be restored and coolant
lin private communications with a BWR plant, the authors verified the
existence of alternate emergency boration procedures and boron supplies to
borate to the levels necessary to limit recriticality.
X
injection would be initiated within the recriticality time window. For shortterm
station blackout, the probability of recriticality was estimated to be
5.6E-7 per reactor year. For long-term station blackout, the probability of
recriticality was estimated to be 6.9E-7 per reactor year.
Based on present operating philosophies and guidance it was assumed that
the operators would not immediately borate and initiate RHR suppression pool
cooling at the time of core reflood. Thus, the probability of suppression
pool saturation and containment over-pressurization in about one half hour is
the same as the probability of a recriticality event occurring. Again, this
probability is 5.6E-7 per reactor year for short-term station blackout and
6.9E-7 per reactor year for long-term station blackout.
If the above accident management strategies were implemented at a plant,
a recriticality event could be terminated prior to reaching saturation
conditions in the suppression pool and in so doing avert containment failure.
The probability that the above accident management strategies fail to avert
containment failure is estimated in this report. Since the primary means of
boration (i.e., from the SLC system) may only be marginally adequate if the
excess steam to the suppression pool is greater than that generated when the
reactor is generating about 10% power, which may occur if RHR suppression pool
cooling fails, failure of either accident management strategy was assumed to
eventually result in containment failure. The probability of boration failure
was estimated to be 5.0E-2, based on the NUREG/CR-4550 ATWS analysis value for
operator failure to initiate boration within a very short time frame
(approximately 4 minutes). It is assumed that the boration concentration when
successfully injected is adequate to terminate the reaction. The value for
RHR suppression pool cooling failure was also estimated to be 5.0E-2, assuming
ac power was restored and the dominant failure is operator failure to
immediately establish adequate RHR suppression pool cooling.
If the accident management strategies were implemented, the probability
of a short-term station blackout event, followed by a recriticality event, and
the event not being terminated prior to containment failure was estimated to
be 5.6E-8 per reactor year. For long-term station blackout, the probability
of the accident management strategies failing to avert eventual containment
failure was estimated to be 6.9E-8 per reactor year.
The results of the analysis are provided in Table 1. These results
indicate that implementation of the accident management strategies suggested
in this report should provide approximately a factor of 10 reduction in the
potential for a recriticality event to cause containment failure (and
subsequently further core damage).
xi

TABLE 1. Recriticality Analysis Results for Station Blackout Sequences
Probability of Probabi 1 ity of
Probability of Containment Failure Containment Failure
>< Core Damage Without Strategies Using Strategies ..... Sequence (per reactor year} (per reactor year)l (per reactor year} ..... SSBO 4.5E-6 5.6E-7 5.6E-8 LSBO 1. 7E-6 6.9E-7 6.9E-8 I TOTAL 6.2E-6 1.25E-6 1.25E-7 I lThis is also the probability per reactor year of a recriticality event. ABSTRACT SUMMARY 1.0 INTRODUCTION 2.0 POTENTIAL FOR RECRITICALITY CONTENTS 2.1 BWR CORE MELTDOWN PHENOMENOLOGY 2.1.1 Control Blade Melting 2.1.2 Core Geometry Changes . . . . . . . . . . . 2.2 CORE MELT CALCULATIONS ••.• 2.2.1 MARCH Code Calculations 2.2.2 Results of Calculations . . . . . . . . 2.3 DOMAIN AND POTENTIAL CONSEQUENCES OF RECRITICALITY 2.3.1 2.3.2 2.3.3 2.3.4 Domain of Critical Configurations Excursion Analysis .•.•••••••• Debris Bed Dryout Power Limits ••••. Containment System Effects •••• 2.4 REFERENCES . 3.0 ACCIDENT SEQUENCES 3.1 STATION BLACKOUT (SBO) . . . . . i i i v 1.1 2.1 2.1 2.2 2.4 2.11 2.11 2.15 2.37 2.38 2.50 2.53 2.63 2.66 3.1 3.1 3.1.1 Loss of Offsite Power Event Tree Headings 3.11 3.1.2 Short-term Station Blackout (SSBO) • • • • . • • • . 3.14 3.1.3 Long-term Station Blackout (LSBO) . • • • . • • • 3.16 3.1.4 Recriticality Potential Following Station Blackout 3.19 3.2 ANTICIPATED TRANSIENT WITHOUT SCRAM 3.2.1 ATWS Event Tree Headings .•.. 3.2.2 Dominant ATWS Sequences ••.••.•. 3.2.3 Recriticality Potential Following ATWS 3.3 REFERENCES ....•.........••• xiii 3.28 3.28 3.34 3.37 3.39 CONTENTS (contd) 4.0 STRATEGY DESCRIPTIONS ........•• 4.1 REFLOOD BORATION STRATEGY FOR CORE MELT EVENTS 4.1.1 Strategy •...••..•••.•. 4.1.2 Justification ...••........ 4.2 HEAT REMOVAL STRATEGY FOR CORE MELT EVENTS 4.2.1 Strategy 4.2.2 Justification 5.0 CONCLUSIONS .•....• 4.1 4.1 4.1 4.2 4.3 4.3 4.3 5.1 5.1 RELATIVE TIMING OF CONTROL BLADE AND FUEL ROD MELTING 5.1 5.2 CORE GEOMETRY CHANGES OCCURRING DURING MELTING AND CORE REFLOOD • • . • • • • • • • • . • • • • • • • 5. 2 5.3 NATURE OF THE REACTIVITY TRANSIENT . • • • . . . • . • . 5.2 APPENDIX A: ~ CALCULATIONS • • APPENDIX B: SPREADSHEET INPUT APPENDIX C: NITAWL INPUT . . . . . . . . . . . . APPENDIX D: XSDRNPM INPUT APPENDIX E: MCDAN INPUT . APPENDIX F: DANCOFF CORRECTION FACTORS APPENDIX G: INSTRUMENTATION CONSIDERATIONS xiv A.1 B .1 C.1 D.1 E.1 F .1 G .1 TABLES Table 1. Recriticality Analysis Results for Station Blackout Sequences xii 2.1. Summary of Key Accident Events 2.12 2.2. Early Melt Station Blackout Events (Case PBEM2) 2.30 2.3. Calculated Maximum k. for Spherical Particles in Water 2.40 2.4. Maximum Calculated k. for Pellet Equivalent Particles in Water • 2.46 2.5. Calculated k• Values for Fuel Rods at Assembly Pitch in Water 2.48 2.6. Calculated Critical Masses for Spherical Particles and Pellets in Water •••••.••••• 2.7. Capacity of BWR Makeup Pumps •••••..•. 3.1. Short-term Station Blackout Dominant Cutsets 3.2. long-term Station Blackout Dominant Cutsets 3.3. Recriticality Time Windows for Station Blackout Sequences 3.4. Recriticality Potential for SSBO Dominant Cutsets 3.5. Accident Management Strategies Implemented for SSBO Dominant 2.49 2.62 3.17 3.20 3.21 3.23 Cut sets . . . . . . . . . . . . . . . . . . . . . . . . . 3. 25 3.6. Recriticality Potential for LSBO Dominant Cutsets 3.26 3.7. Accident Management Strategies Implemented for LSBO Dominant Cutsets . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.27 3.8. Recriticality Analysis Results for Station Blackout Sequences 3.29 3.9. ATWS Dominant Cutsets 3.38 XV TABLES (contd) Table Page A.l. Calculated k00 for Spherical Particles in Water (0 PPM lOB) A.l A.2. Calculated k00 for Spherical Particles in Water (200 PPM lOs) A.2 A.3. Calculated k00 for Spherical Particles in Water (500 PPM lOB) A.3 A.4. Calculated k00 for Spherical Particles in Water (1000 PPM lOB) A.4 F .1. Dancoff Factors for Spherical Particles in Water . . . . . . F .1 xvi FIGURES Figure 2.1. DF-4 Cross Section 2.2. BWR Flow Areas with Loss of Structures (ORNL) 2.3. Sketch of Reactor Vessel and Core at Start of Melting 2.4. Volume of Lower Plenum Debris (ORNL) 2.5. TMI-2 End-State Accident Configuration 2.6. Logarithmic Plot of the Cumulative Weight Distribution for 2.3 2.5 2.6 2.7 2.9 TMI-2 Core Debris Grab Samples . . . • . . . . . . . • . . 2.10 2.7. Mixture Level as a Function of Dimensionless Time (t- tu)/tau for Cases PBTBO and PBTBS . • . . . • . • • . 2.14 2.8. Grand Gulf Accident Event Timing 2.9. Peach Bottom Accident Event Timing 2.10. Average Core Temperature Change, (T- To), as a Function of Dimensionless Time After Core Uncovery, (t - tu)/tau, for 2.16 2.17 Group 1 • . . . • . . . • • • • • • . . • • . • . . . • . . 2 • 18 2.11. Average Core Temperature Change, (T- T0), as a Function of Dimensionless Time After Core Uncovery, (t - tu)/tau, for Group 2 . . • • • . . . • • • • . . . • • . • • . • . • . • 2.19 2.12. Average Core Temperature Change, (T- T0), as a Function of Dimensionless Time After Core Uncovery, (t - tu)/tau, for Cases PBTBS, PBTBO • . • • . • . • • • • . • • • • . • . . . . 2.20 2.13. Fraction Fuel Rods Melted as a Function of Dimensionless Time After Core Uncovery, (t - tu) /tau, for Group 1 . . • . . . . . 2. 21 2.14. Fraction Fuel Rods Melted as a Function of Dimensionless Time After Core Uncovery, (t - tu)/tau, for Group 2 • . • . 2.22 2.15. Fraction Control Blades Melted as a Function of the Average Core Temperature for All Cases . . . • . . • . . • . . . . 2.24 2.16. Fraction Control Blades, Channel Boxes, and Fuel Rods Melted as a Function of Dimensionless Time After Core Uncovery, (t - tu)/tau, for Case PBTBO . . . . • • . . • . • . . 2.25 xvii FIGURES (contd) Figure 2.17. Fraction Control Blades, Channel Boxes, and Fuel Rods Melted as a Function of Dimensionless Time After Core Uncovery, (t - tu)/tau, for Case PBTBS • • • • • . • • • • • • • • • 2.26 2.18. Fraction Control Blades, Melted by Radial Region as a Function of Dimensionless Time After Core Uncovery, (t - tu)/tau, for Case PBTBO • . . . . . • • . . • • • • • • • . • • • • • • • • 2.27 2.19. Highest Elevation of Unmelted Control Blade as a Function of Core Radius and Time After Core Uncovery, (t- tu), for Case 2.20. 2.21. 2.22. PBTBO • • . • • • • • • • • • • . . • • • • • . • · • • Vessel Pressure versus Time for Peach Bottom Early Melt Station Blackout (PBEM2) ...••••.••••••• Reactor Vessel Water Level versus Time for Peach Bottom Early Melt Station Blackout (PBEM2) .••••.••.• Core Temperature versus Time for Peach Bottom Early Melt Station Blackout (PBEM2) •••.•••••.••••• 2.23. Melt Fraction versus Time for Peach Bottom Early Melt Station 2.28 2.31 2.32 2.33 Blackout (PBEM2) • . . . . . . . . • • . • • • . . • • • • 2.34 2.24. Fraction of Core in Lower Head versus Time for Peach Bottom Early Melt Station Blackout (PBEM2) . • • • • • • . • • • • 2.35 2.25. Lower Head Debris Temperature versus Time for Peach Bottom Early Melt Station Blackout (PBEM2) • • • • • • . • 2.36 2.26. Calculated ~ for U02 Fuel Particles in Water with 0 PPM lOs 2.41 2.27. y8lculated ~for U02 Fuel Particles in Water with 200 PPM B • • • • • • • • • • • • • • • • • • • • • • • 2. 42 2.28. t8lculated ~ for U02 Fuel Particles in Water with 500 PPM B • • • • • • • • • • • • • • • • • • • • • • • • • • • • 2. 43 2.29. t8lculated ~ for U02 Fuel Particles in Water with 1000 PPM 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2. 44 2.30. Envel~Be of Calculated ~s for uo2 Fuel Particles in Water with B Concentrations of 0, 200, 500, and 1000 PPM • • • 2.45 xviii FIGURES ( contd) Figure 2.31. Steady-State Dryout Heat Flux and Bed Quench Heat Flux Data Compared With Steady-State Lipinski Model (P=0.1 MPa, e=0.4) (From References 2.17 and 2.18) . • • • . • . • • • • 2.55 2.32. Lipinski Correlation of Debris Bed Dryout Heat Flux (Ref. 2 . 18) . . . . . . . . . . . . . . . . . . 2.56 2.33. Fraction of Core Enclosed in Debris Bed as a Function of Bed Diameter . . . . . . . . . . . . . . . . . . . . . . . 2.57 2.34. Debris Bed Dryout Power Density Limit at 15 and 200 psia for 0.1 and 0.5 inch Particle, FA= 2.0 • • . • . • • . 2.58 2.35. Ratio of Debris Bed Power to Core Operating Power 2.60 2.36. Coolant Boiloff Rate From a Critical Debris Bed 2.37. Containment Over-pressure Time as a Function of Energy to Poo 1 • • • • • • • • • • • • • • • • • • • • 3 .1. Peach Bottom Loss of Offsite Power Event Tree 3.2. Station Blackout Event Tree 3.3. Peach Bottom Case B ATWS Event Tree . . . . . . . . . . . . . . 3.4. Simplified ATWS Event Tree xix 2.61 2.65 3.3 3.15 3.30 3.35 1.0 INTRODUCTION This report addresses the potential for a recriticality, following a low probability severe accident and subsequent reflooding of the fuel in a Boiling Water Reactor (BWR). It provides scenario sequence definition and accident management strategies that could be used to mitigate or terminate the postulated recriticality events. Analyses of severe BWR accidents indicate that, while the core is in the process of heating and melting, one of the earliest components to melt and relocate are the steel blades that contain the B4C control material. If the core were then to be reflooded following relocat1on of the control blades, the possibility exists for the core to become critical again without an adequate means of control. This study•s objective is to explore the likelihood of recriticality and the consequences thereof. If a significant likelihood of recriticality exists, accident management procedures would be warranted which could prevent recriticality or mitigate the consequences of recriticality, in order to return the plant to a safe stable state. Due to modelling and phenomenological uncertainties related to void fraction, debris bed size, particle size, etc. and the lack of an analytical tool capable of performing the complex analyses required to address the complex interactions of these parameters, a conservative bounding analysis was conducted. If the conservative bounding analysis indicates acceptable consequences (e.g., no recriticality or a recriticality event resulting in a benign pressure pulse that is non-threatening to the integrity of the containment and accident management strategies can be implemented to successfully prepare for and terminate the event), further, more sophisticated model development and analyses may not be necessary to resolve the modelling and phenomenological uncertainties. However, if the bounding analysis indicates unacceptable consequences (e.g., a significant recriticality event that creates a large pressure pulse and potentially fails the containment), further research should be identified to resolve the uncertainties and phenomenological issues. In order to establish whether the development of accident management strategies to control recriticality is necessary, some important questions must be addressed. • Is recriticality credible following the initial stages of severe core damage? • How likely is recriticality? Is it a factor in risk dominant sequences? How long is the time window for recriticality? Are recovery actions likely to occur in the time window? 1.1 I • Would existing equipment and procedures result in the appropriate control actions without the need for additional accident management procedures? The determination of the most appropriate accident management strategies requires resolution of an additional issue. • What are the potential consequences of recriticality? Is an excursion possible which has the potential to disrupt fuel and fail the vessel and/or containment? Would a quasi-steady power level be developed? If so, at what level? Primarily using the NUREG-1150 risk study for the Peach Bottom plant as a technical basis, the sequences with the highest predicted frequencies are used to characterize accident sequences which would be likely to result in core damage. The most likely recovery mechanisms that could arrest these sequences are then identified. It is assumed that if the cooling water systems can be recovered, the operators would use the systems to restore core cooling as quickly as possible. The reflooding of the core with the attendant potential for initiating recriticality would then be a concern. The accident management strategies discussed in this report focus on the control of recriticality by means of soluble poison addition and containment heat removal. The offered strategies are not developed in the degree of detail that would be required for operating procedures for a specific plant. It is recognized that the development of specific, effective procedures are most appropriately accomplished by the plant staff. To provide the needed analysis of existing information, Section 2.0 of this report addresses recriticality; 3.0 sequence definition; 4.0 strategy description; and 5.0 conclusions and recommendations. 1.2 • 2.0 POTENTIAL FOR RECRITICALITY This chapter addresses the potential for recriticality. It begins with core melt phenomenology, continues with core melt calculations, provides potential recriticality considerations, and concludes with potential consequences of recriticality. 2.1 BWR CORE MELTDOWN PHENOMENOLOGY During a severe accident, the neutron absorbing control rods and control blades are expected to melt before the fuel rods. This occurs because the control materials are contained in metallic structures which have lower melting points than the oxide (U02) fuel rod material. Thus, the control rods and fuel rods will become separated during the core melt; and reflooding of the core has the potential to result in recriticality. For a pressurized water reactor (PWR), reflooding is normally accomplished using borated water supplies; and recriticality is generally perceived not to be very credible. However, for a boiling water reactor (BWR), reflood is normally accomplished using unborated water; and recriticality is believed to be credible. Therefore, this report addresses the potential for recriticality events only in BWRs. Core heatup commences when the core becomes uncovered. The timing of core uncovering, heatup, and melting may occur over tens of minutes or a few hours, and depends on the nature of the accident sequence. There are a number of phenomenological issues or areas of uncertainty in this core melt process. These include: • • • • • The melting and relocation of the control blades, the fuel rod cladding, and the fuel rods. The effects of the core melt and relocation on steam generation, flow blockages, steam-zirconium reactions, and hydrogen generation. The effects of core melt and relocation on the surrounding structures (such as the core plate and the core baffles) and on the lower head structures (such as the lower head itself, control rod guide tubes, instrument tube penetrations, and drain lines). The potential for recriticality as a result of changes in core geometry due to melting and the damage that occurs during core reflood. The coolability of a damaged or molten core, assuming reflooding can be accomplished. 2.1 These issues have varying degrees of importance, depending on whether the interest is in risk assessment, accident management, or recriticality. All of the issues are important, although not of equal importance, to the analysis of severe accidents and their consequences. A lesser number are important to the development of in-vessel accident management strategies. For the assessment of BWR recriticality only the following issues are important: • The relative time of control blade and fuel rod melting (separation of the control blades from the fuel rods is what makes recriticality possible). • The core geometry changes occurring during melting and core reflood (the reactivity of the damaged core depends on the debris mass, fuel particle shapes, and porosity). • The nature of the reactivity transient (the ability to manage the recriticality event depends on whether it is a core-damaging or explosive transient event or is a benign event, which gradually increases to higher power levels). 2.1.1 Control Blade Melting The first issue deals with the timing of control blade melt and relocation. Two experiments have been performed that support early control blade melting (i.e., control blade melt and relocation prior to fuel rod melting): the DF-4 experiment (Ref. 2.1) by Sandia National Laboratory and the CORA 16 experiment (Ref. 2.2) by KfK (Karlsruhe, FRG). Figure 2.1 is a sketch showing the cross section of the DF-4 experiment and the arrangement of the fuel rods and the simulated channel box and control blade. The fuel length was about 19 inches. The CORA 16 experiment was of a comparable scale. Both experiments showed melting and relocation of the control blades to the bottom, leaving standing fuel rods behind. The CORA 16 experiment indicates the effective melting and relocation temperature of the control blades is approximately 2280°F, which is about 270°F below the stainless steel melting point (2550°F). This decreased effective melting temperature is due to alloying reactions with the B4C neutron absorber. Calculations by Ott (Ref. 2.3) for the DF-4 experiment confirm that an assumed reduction of the control blade melting temperature by 200°F (to 2350°F) is necessary to explain the observed timing of the control blade melt relocation. Ott used a specially modified version of the BWRSAR code in his calculations. Special modifications were required because, in the experiment, nearly all of the heat losses were in the radial direction into the zirconia shroud. These radial heat losses are typically negligible and are not modeled in the normal full-sized core model. The modifications allowed the special experimental geometry to be accurately modeled. The DF-4 experimenters 2.2 BWR CORE 00 cocc 00000000 0000000~ oo'=l•oooo COOOC'CjO coooccco cooo~ooc cooooo~c oooooooc 00000000 00000000 ooo•oooc oooo•oco 00000000 00000000 00000000 oooocc~:: oooooo=o cocoocco oooo•ooc ooo•ooc~ jOOOOOC::: oooooocc CCOOOOC'C -....... FUEL ROO ZIRCALOY CAN WALL STAINLESS STEEL SHEATH B4C POWDER ZIRCONIA SHROUD '- CERAMIC ZIRCONIA FIGURE 2.1. DF-4 Cross Section 2.3 reported an 11 extensive11 blockage was formed at the bottom from the relocated control blade and channel box melt. However, the blockage had little effect on the steam-zirconium reaction. The relocated materials continued to oxidize with about half the hydrogen produced after the melt relocation. In conclusion, the DF-4 and CORA 16 experiments confirm the early melt relocation of the control blades. Also, the melt relocation temperature is believed to be somewhat (200-300°F) below the melting point of stainless steel. Because of the small size of the fuel assemblies used in the DF-4 and CORA 16 experiments, these experiments may not provide a complete picture of the melt and relocation phenomena associated with typical nuclear power plant fuel assemblies. The experiments provide little information on whether relocated blade melt would accumulate at the bottom of the core on the core plate or would simply pour through the existing flow holes. The extent of relocation would affect the location of blade remnants and other issues such as hydrogen generation, core slumping into the lower head, the lower head failure mode, and in-vessel accident management strategies. The planned FLHT-6 experiment in the NRU reactor will be 12 feet long and may give a better picture of length effects on blade and channel box melt relocation. 2.1.2 Core Geometry Changes This section discusses the issues relating to the effects of core geometry changes which occur during melting and core reflooding. Figures 2.2 to 2.4 illustrate some of the theoretically possible types of fuel rod and core material rearrangements which could occur with core melting. These conceptual core conditions are based primarily on considerations of the material volumes and their possible relocations. Figure 2.2 illustrates that the potential water volume in the assembly could progressively increase from 58% to 67% by removal of the control blades and channel boxes and clad (by melting). Fuel rearrangements which increase the water content could have a higher neutron multiplication constant. Figure 2.3 is an illustration of how rod bowing could lead to fuel rearrangements. Figure 2.4 illustrates that relocation of the whole core in the form of a debris bed, with a porosity of 40%, would fill most of the lower head. The illustrations do not take into account the core damage which is expected to occur during the reflood process. Because the core is severely overheated at the time of blade melting, reflood would be expected to result in fracturing and shattering of the fuel. A number of experiments have been performed which provide information on fuel rod shattering and the types of debris beds which might form after reflood. The principle experiments are those of Chung (Ref. 2.4) and Katanishi (Ref. 2.5). Two of the Severe Fuel Damage experiments (SFD Scoping Test and SFD-1) were also water-quenched; however, no specific evaluation of those tests for information on rod shattering is apparent. The Three Mile Island Unit 2 (TMI-2) accident also provides relevant information. 2.4 12 IN. 12 IN. - - i ;rooooooo 000000001 ,oooocooo '00000000 00000000 00000000 IOOOOI3000 000~0000 j000130000 0000~000 ,ocoooooo OOOCOCOOI lo0o00o0o0o0o0o0c 0oo0o0o0o0c0o0~j0! I I J 0o0o0o0o0o0o0o0:l 00000000000000~0~ 100000000 000000001. 1 ooo~ooooi 0000~000 I C·~OO~OOO 000~0000 00000000 00000000 1 ooooooo~j OOOOOOOOP, jOOOOOOOO 00000000 -~-- 'fooO"oooO"o- oooooooojl 'jO10O0O0O0O0O0O0O0 0000000000000000~1 l ooool3ooo ooo~oooo 00000000 0000~000 1 loooooooo ·oooooooot 1oooooooo oooooooo 1 ,oooooooo oooooooo, I l l 0o0o0o0o0o0c0o0o .0o0o0o0o0o0o0o0o11 IOOOOOOOC 00000000~ 'looo~oooo oooo0oooU 00000000 00000000~ loooooooo oooooooo~ jfOOOOOOOO OOOOOOOOf. ~oo~~~~o_? __ <:.':?.:'~C:.o~~ I ,-cocaa-;-o-o--ooacooool 00000000 00000000( llgggg~ggg ggg~gggg,, 00000000 0000~000 100000000 OCOOOCOOI (OCCCOOCO 000000001 ,cccccooo cccocooo, i I .ococccco coo~cooo 1 •ccococco coooocoo, !cccoocco ccoooooc· 1 cccooc~c occo0oool CCC~~OCC OCC~OOOC( loccccocc ocoococo 1 '=ccc=ccc cocooooo. ;..c._.c_ ccoccc ccoooooo~ _____________ ~ CORE HEATUP LEADS TO PROGRESSIVE RELOCATION OF UNIT CELL COMPONENTS • ALL STRUCTURES INTACT FLOW AREAS FUEL ASSEMBLY INTERSTITIAL 0.440 -0.1-40 0.580 STRUCTURE CROSS SECTION 0.420 • LOSS OF CONTROL BLADE FLOW AREAS • FUEL ASSEMBLY INTERSTITIAL 0.440 -0.18-3 0.623 STRUCTURE CROSS SECTION 0.377 LOSS OF CANISTER WALLS FLOW AREA 0.670 STRUCTURE CROSS SECTION 0.330 FIGURE 2.2. BWR Flow Areas with Loss of Structures (ORNL) 2.5 • N 0'\ FUEL ROD BOVING AS CONTROL RODS 1\EL T JET PUJ\P-- CORE PLATE CONTROL ROD DRIVE TLBES BLADES -· ·--TED CONTROL BLADES - •• ·~ SO/\E BOXES FIGURE 2.3. Sketch of Reactor Vessel and Core at Start of Melting • HEIGHT OF DEBRIS
~ too
..J
::J
~
::J 5 u
2
10 • I
101
FIGURE 2.6.

SAMPLE J,
2
6 v 1
0 3 0 ~
A 6 0 c
0 7 8 0 c 8 0~
~ 9 <> 10 0 ~R
0 11 0
0 c
v
<>
0 0
c
0 v
cQ.
~ &
<> v
0
v
5 t02 2 6 t03 2 5 104
PARTICLE DIAMETER (microns)
Logarithmic Plot of the Cumulative Weight Distribution
for TMI-2 Core Debris Grab Samples
2.10
Based on this information, shattered fuel rods would be expected to form
under-moderated debris beds.
Based upon the above discussion, there is a significant potential for
fuel rod shattering and debris bed formation when an overheated core is
reflooded with water. The shattering appears to be related to the extent of
oxidation of the fuel rod cladding. It is also expected that debris beds,
formed from shattered fuel rods and quenched core melt, will not be optimally
moderated (i.e., they will be under-moderated and thus not a recriticality
concern). Heat transfer aspects of debris bed criticality will be discussed
in Section 2.3.3.
2.2 CORE MELT CALCULATIONS
The results of a number of computer code calculations of BWR core melt
are presented in this section. This section provides:
• A discussion of the MARCH computer code analysis of the melt
behavior of the control blades and control rods.
• Information on core melt timing for different accident sequences.
• Information on the relative timing of control blade and fuel rod
melting.
The principle use for this information is to assess the potential for BWR
recriticality. Thus, aspects of the evaluation which might be important to
other issues such as hydrogen generation, core coolability, core relocation,
or vessel failure are not emphasized. The accident scenarios considered
include primarily those for which MARCH code results were available from
previous Battelle work on NUREG-1150. Additional MARCH code calculations were
performed for station blackout scenarios, and these results were used to
provide more detailed information on the time of control blade melting.
2.2.1 MARCH Code Calculations
Table 2.1 lists Peach Bottom and Grand Gulf core melt accident scenarios
for which MARCH code results were available from previous Battelle work on
NUREG-1150. Also listed are a number of BWR scenarios for which MARCH
calculations were performed to address station blackout scenarios specifically
for this study. The calculations performed specifically for this study used a
more recent version of the MARCH code (version V194) than was used in NUREG-
1150 (i.e., version V192). The difference in code version accounts for the
differences between sequences PBTBUX and PBTBO, which are similar cases.
2.11
.N ……
N
Case
GGTB1
PBTBUX
PBTB2
PBTC3
GGTPI
GGTC
PBTW
GGTBS
GGS2E
PBAE
PBV
GGTQUV
PBTBO
PBTBS
PBEM2
TABLE 2.1. Summary of Key Accident Events
Time Time of Temp Time of Time of
Constants Core at Blade Rod
Tau Alpha Uncovery Uncovery Melt Melt
min °F/min min °F
33.9 21.8
18.8 42.6
34.1 23.5
16.1 49.7
42.9 20.4
22.0 60.5
71.0 18.7
18.6 39.8
9.6 83.4
13.8 124.0
17.9 44.7
16.8 52.0
18.8 42.6
34.1 23.5
18.8 42.6
min min Descriptions
GROUP 1 (NUREG-1150)
483.0 571.9 552.0 579.0 Grand Gulf station blackout, late melt, no ADS
67.0 577.4 109.0 134.0 Peach Bottom station blackout, early melt, no ADS
527.0 559.4 601.0 616.0 Peach Bottom station blackout, late melt, no ADS
34.0 850.6 53.0 58.0 Peach Bottom ATWS, no ADS
1536.0 322.8 (a) 1645.0 Grand Gulf, open valve, no RHR
90.0 463.2 111.0 117.0 Grand Gulf, ATWS, no ADS
2620.0 316.7 (a) 2748.0 Peach Bottom, no RHR
GROUP 2 (NUREG-1150)
51.0 577.4 82.0 85.0 Grand Gulf station blackout, ADS at top of core
6.0 593.7 (a) 28.0 Grand Gulf small LOCA, no makeup
1.5 1197.0 15.8 12.0 Peach Bottom large LOCA, no makeup
3.1 1061.0 24.0 27.0 Peach Bottom LOCA outside containment
47.0 579.2 (a) 103.0 Grand Gulf transient, no makeup, ADS at 2 ft
RECENT MARCH V194 CALCULATIONS
66.0
530.0
65.0
567.0
564.0
567.0
113.0
649.0
127.0
120.0 Peach Bottom station blackout, early melt, no ADS
716.0 Peach Bottom station blackout, late melt, ADS at 2 ft
132.0 Peach Bottom station blackout, early melt, ADS at 2 ft
(a) Information not available.
– ….
The NUREG-1150 accident scenarios in Table 2.1 are roughly divided into
two groups, made partially for plotting convenience. The primary physical
difference between the groupings is the vessel water level during core heatup
and melt. For the cases in Group I, the water level remains close to the
bottom of the core. For the Group 2 cases, the water level is well below the
core during core heatup. Generally, the sequences in Group 2 involve pipe
break LOCAs, stuck-open valves, and cases where the automatic depressurization
system (ADS) is activated. Figure 2.7 illustrates the differences in water
levels as a result of ADS activation. Results are shown for Peach Bottom
station blackout sequences PBTBO and PBTBS. For the PBTBS sequence in which
the ADS is activated when the water level falls to 2.0 feet., the water level
quickly falls well below the core. Generally, less metal-water reaction is
predicted for the low water level cases.
The results in Figure 2.7 are displayed in terms of the dimensionless
time after core uncovery:
(t – tu)/tau,
where,
t = accident time, min
tu = time at start of core uncovering, min
tau = “boildown time constant,” min
and
tau = rho x A x H x HFG/QDK
where,
rho = water density, lb/ft3
A = vessel water area, ft2
H = active core height, ft
HFG =water heat of vaporization, Btu/lb
QDK = decay heat at start of core uncovery, Btu/min.
It is seen that the coolant water levels in Figure 2.7 prior to ADS activation
are quite similar when displayed in this manner. Use of the dimensionless
time parameter has been found to be convenient, and it will be frequently
used in the following discussions to display the code results.
The MARCH calculations for NUREG-1150 were performed using the source
term code package (STCP) or V192 version of the code. Although control blade
melting was predicted for these calculations, actual control blade relocation
was not modeled. The melted control blade nodes were assumed to remain inplace
after melting. Since the control blade nodes have relatively low heat
capacity compared to the rest of the core, this assumption has little effect
,
2.13 i
TBO A TBS
15 tf…..
STATION ELAO/tau
FIGURE 2.7. Mixture Level as a Function of Dimensionless Time (t – tu)/tau for Cases PBTBO and PBTBS
10
,
on the heatup of the core. However, for criticality evaluations it is
desirable to model control blade relocation in addition to melting.
The more recent MARCH calculations in Table 2.1 were performed using
version V194. The V194 version of the code contains a number of modeling
enhancements, including a BWR control blade relocation model in which melted
blade nodes fall either (input option) to the core plate below the core or
into the water in the lower head. If there are solid control blade nodes
above the melted nodes, the solid nodes are assumed to fall downward and
replace the melted nodes. As before, the control blade melting temperature is
specified by input.
Version V194 of MARCH also contains enhanced capability to calculate
heat transfer and metal-water reaction during reflooding of a degraded core.
The improved models were found to be necessary to explain the thermal behavior
of the TMI-2 core. In addition, examination of the TMI-2 core debris
indicated the core melting temperature used in the NUREG-1150 MARCH calculations
(4130°F) was unrealistically low. Based on the TMI-2 data, a core melting
temperature of 4870°F is more representative. The higher melting temperature
was also used in the BWR core heatup calculations for the more recent
MARCH calculations listed in Table 2.1.
All of the BWR scenarios in Table 2.1 are unmitigated meltdown accidents.
That is, no makeup was assumed to be available after the start of core
uncovering. The major potential for recriticality occurs if the core is
reflooded with water in the time window between blade melting and fuel rod
melting. No additional calculations were performed in the present study for
BWR core reflood scenarios. Thus, the MARCH calculations have been used
primarily to indicate that there is a time window during which a potential for
recriticality exists due to the melting of control blades. Based on the
Version V194 MARCH calculations, this window ranges from 5 to 67 minutes
depending on the nature of the transient event.
2.2.2 Results of Calculations
Figures 2.8 and 2.9 display key event times for the MARCH calculations
listed in Table 2.1. The event times range from minutes to over two days.
Figures 2.10, 2.11, and 2.12 show the average core temperature from the MARCH
calculation displayed as a function of the dimensionless time parameter
discussed above.
Figures 2.13 and 2.14 are plots of the core melt fraction. Melting
generally starts between 1 and 6 time constants after the start of core
uncovering. Some of the larger melt start times can be explained in terms of
the cooling from ADS activation or to the use of a higher assumed core (fuel
rod) melt temperature. In general, however, there seems to be no simple
correlation to pinpoint the time core melting starts. The best that can be
2.15

.N ……
0’1
START UNCOVERY x START FUEL MELT I VESSEL FAILURE +
GGTQUV I X I +
GGTPI
GGTC I xl +
GGS2E ~ +
GGTBS I xl +
GGTBl
I
0
X I +
I
10
L I
20
TIME (HOURS)
X I
I
FIGURE 2.8. Grand Gulf Accident Event Timing
I
30
+
START UNCOVERY x START FUEL MELT I VESSEL FAILURE +
PBTC3 xl+
PBAE 4 +
PBTW X I +
PBTBO I X I +
!’> PBTB2 I X I +
……
‘-J
PBTBS X I +
PBV 4 +
PBTBUX X I +
0 10 20 30 40 50
TIME (HOURS)
FIGURE 2.9. Peach Bottom Accident Event Timing
.N …..
(X)
u.
~ I-I
~
+
‘iJ
GGTB1
GGTPI
6

PBTBUX
GGTC
0
A
PBTB2
PBTW
+
<>
PBTC3
PBTBO
4000~——————————————————~——–~
el v i>”‘<>“lo o ~
~ e e i’~ Q+JAO+ + 121. 0~ 0 <9 • • 'V 0~ + 0,._, <9
• 0 ~0
• • + ‘iJ~ + +
• 1000 t- • + ~ <>~
• ,. -Y. <>,
… ‘V
<>
o,.,
-1000 I I I I I I I I I I I I I I I I I I I I I
0 1 2 3 4 5 6 7 8 9
TIM:. (t – tul/tau
FIGURE 2.10. Average Core Temperature Change, (T- T0), as a Function of
Dimensionless Time After Core Uncovery, (t – tu)/tau, for Group 1
10
u.
.N
~ ……
1.0
+
t”