ML19324C362
| ML19324C362 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/06/1989 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Chubb W AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML19324C363 | List: |
| References | |
| NUDOCS 8911160170 | |
| Download: ML19324C362 (23) | |
Text
-
e NOV 0 61989 ie Mr. Walston Chubb 3450 MacArthur Drive Murrysville, Pennsylvania 15668
Dear Mr. Chubb:
Your letter of October 24, 1989, to Commissioner Carr, Chairman of the U.S.
Nuclear Regulatory Commission (NRC), has been referred to me for reply.
The often repeated phrase "previously molten material" refers to the core material (fuel, cladding, control rods, guide tube fuel support structures) that was melted during the Three Mile Island Unit 2 (TMI-2) accident and subsequently cooled and resolidified.
It does not refer to the metallurgical sintering process used during fuel fabrication.
As you may not be aware of the degree of damage suffered by the core during, the accident on March 29, 1979, I am enclosing a copy of the most recent accident scenario developed by the licensee, GPU Nuclear Corporation, and by the U.S.
Department of Energy's (DOE's) contractor Idaho National Engineering Laburatory.
This scenario was included in the licensee's submittal on July 5, 1989 of the Defueling Completion Report.
Calculations simulating the accident suggest that.
a molten pool of approximately 50 percent of the original core material was formed 224 minutes into the accident.
Subsequent cooling resulted in the resolidification of the molten core, forming a substance that has been given the name corium.
I can assure you that the licensee, the NRC, and the DOE are continuing their efforts to understand the accident at TMI-2 and will continue this effort for some time.
As new data is collected, the accident scenario will undoubtedly be fur +.her refined; however, the evidence clearly indicates that melting and resolidification of the THI-2 fuel occurred during the accident.
Sincerely, C:2hlStdDj John F. Stolz, Director Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated i
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t Mr. Walston Chubb 3450 MacArthur Drive i
Murrysv111e, Pennsylvania 15668 8
Dear Mr. Chubb:
Your letter of October 24, 1989, to Commissioner Carr, Chairman of the U.S.
Nuclear Regulatory Commission (NRC), has been referred to me for reply.
The of ten repeated phrase "previously molten material" refers to the core material (fuel, cladding, control' rods, guide tube fuel support structures) that was melted during the Three Mile Island Unit 2 (TMI-2) accident and subsequently l
cooled and resolidified.
It does not refer to the metallurgical sintering process used during fue! fabrication.
As you may not be aware of the degree of damage suffered by the core during the accident on March 29, 1979, I am enclosing a copy of the most recent accident scenario developed by the licensee, GPU Nuclear Corporation, and by the U.S.
Department of Energy's (DOE's) contractor Idaho National Engineering Laboratory.
^
This scenario was included in the licensee's submittal on July 5, 1989 of the Defueling Completion Report.
Calculations simulating the accident suggest that a molten pool of approximately 50 percent of the original core material was formed 224 minutes into.the accident.
Subsequent cooling resulted in the resolidification of the molten core, forming a substance that has been given the name corium.
I can assure you that the licensee, the NRC, and the DOE are continuing their efforts to understand the accident at THI-2 and will continue this effort for some time.
As new data is collected, the accident scenario will undoubtedly be further refined; howaver, the evidence clearly indicates that melting and re olidification of the THI-2 fuel occurred during the accident.
Sincerely, l
r Jo n F. Stolz, Direc) r P oject Directorate' -4 l
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated 1
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ENCLOSURE
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i 2.0 POST-ACCIDENT FUEL DISP [R$10N o
This section provides a sum.ary discussion of the accident sequence as it I
relates to fuel material transport within the RV and from the RV to es-vessel locations. Included are sections which describe the mest l
likely supposition of the core accident scenario, the post-accident l
condition of the plant, and the fuel transport mechanisms within the RCS, I
RB, and AFHB. The bases for the following findings and conclusions are derived primarily from the results of visual examinations, analytical j
evaluations, and the experience and data derived from defueling i
operations.
Substantial core damage within the RV and subsequent attempts to cool the core provided the source material and initial pathway-t>y which fuel debris was transported into the RCS, R8, and AFHB. Because the plant i
systems required cooldown.. isolation, and water processing at various times during the )lant stabiluation and recovery periods, additional potential pathways entsted for insoluble fuel latterial transport.
However, the majority cf these pathways within the RB and the AFHB are defined by specific boundaries, filters, and/or flow restrictions, which significantly reduced any potential fuel transport. Of the total fuel debris available to be transported from the RV, it was conservatively estimatec that no more than 25 kg reached the AFHB locations, no more than 15 kg was relocated to the RB sump and various other RB locations, and no more than 230 kg was relocated throughout the RCS (see Table 2-1).
The remaining core inventory was retained in the RV.
The following discussion represents the basis for fuel transport dispersion at THI-2.
2.1 The Accident Scenario A postulated scen'ario of the accident was developed using currently aval14ble data from in-vessel and er-vessel defueling operations and the accident transient secuence information (References 2.1 and 2.2).
This data base included measurements from on-line instrumentation, visual observations, and supporting analytical studies as well as other i
experimental data from independent research facilities (Reference 2.3).
The accident can ba divided into the following five (5) phases:
i Phase 1. Time 0-100 Minutes: Loss-of-Coolant with the RCS Pumps Cperating.
Panse !!, Time 100-174 Minutes:
Initial Core Heatup and Degradation.
Phase 111 Time 174-224 Minutes: Degraded Core Heatup and Relocation.
Phase IV, Time 224-230 Minutes: Core Relocation to LCSA.
Phase V, Time After 230 Minutes: Long-Term Cooling of Degraded Core, i
2.1.1 Phase ! - Loss of-Coolant (0-100 Minutes)
The first phase of the accident is the time interval from the turbine trip until the A-Ioop RCPs were turned off at 100 minutes.
The RCPs provided 2-phase cooling to the core during this period, preventing core overheating and damage. During the first phase of the accident, the amount of water in the RCS decreased because the RCS makeup was insufficient to compensate for cooltnt loss through the PORV.
l 2-1 Rev. 0/0461P
_ _ _ _. ~. _.,..
7 2.1.2 Phase !! - Initial Core Heatup and Degradation (100-174 Minutes)
When the last two RCPs were turned off, at apprctimately 100 minutes, the top of the core was uncovered and coolant water separated into steam and 11guld phases. Temperatures in the upper regions of the core then increased more rapidly. The core liquid level dropped to approutmately the mid-core elevation at approutmately 140 minutes and fuel rod temperatures at the top of the core increased sufficiently (1100'K) to cause cladding rupture. During this period, the operators realtred that the PORV was open. They manua11y closed the pressurlaer block valve, thus limiting further loss-of-coolant and gaseous fission product release from the RCS to the RS. However, the block valve had to be cycled (i.e., opened and closed) frequently to maintain RCS pressure during this period.
Rapid outdation of the aircaloy cladding at the top of the core began at approutmately 150 minutes. The heat generated from outdation elevated fuel rod temperatures above the cladding melting point'(2100'K) developing a molten mixture of fuel, cladding, and some structural steel. This stature flowed downward and solidtfled around intact fuel rods near the coolant 11guld level interface. The responses of incore instrumentation and source range monitors indicated that a large region of partially molten core materials formed by 174 minutes, as shown in Figure 2-14.
It is conjectured that the first molten material to flow was a mixture consisting primarily of U0, steel, tircaloy, and 2
stiver, with some indium and cadmium. As this molten flow stopped at the coolant level interface, it formed a thin layer, or crust, which later supported additional molten material in the core region.
2.1.3 Phase !!! - Degraded Core Heatup and Relocation (174-224 Minutes)
Operation of the RC-P-28 at 174 minutes, for approutmately 6 I
minutes, resulted in the first major core relocation event when coolant was tireviated into the RV following core degradation.
Thermal-mechanical interaction of the coolant with the outdtzed and embrittled fuel rod remnants in the upper core regions is believed to have fragmented and collapsed these standing remnants and formed the upper core cavity and debris bed. The configuration is shown in Figure 2-Ib.
Af ter approutmately 25 minutes of further coolant heating and steam formation in the core, the ECCS was initiated at 200 minutes and subsequently filled the RV in 7 to 10 minutes. Studies of debris bed cooling indicate that final quenching of the upper core debris bed probably occurred during the last several minutes of this time period (Reference 2.4).
It is po:tulated that effective cooling of the molten core material was listted to the surrounding crust material. Thus, the amount of molten material in the central region likely continued to increase in size and temperature because of decay heat from retained fission products and lack of coolant flow through the damaged core. Calculations p
2-2 Rev. 1/0461P e
J I.. '
simulating the accident sug0est that a molten pool of i
approutmately 50% of the or ginal core materials was formed within the consolidated region by 224 minutes into the accident (Reference 2.5).
This is consistent with the observed molten material found in the resolidtfied core mass, t'he CSA, and the loyer head regions (Figure 2-2),
i The interaction of the injected water with the upper debris bed during this period and the flow pattern of steam and gas entting the core through the upper plenum have been assessed. The j
observed damage pattern to the upper fuel assembly grid was 1
consistent with espected flow patterns, considering the location of the exit flow ortftces. Rapid outdation within the debris bed J
and the subsequent interaction between the upper grid structure and the high temperature gases entting the core at high velocity probably caused the observed limited damage.
l 2.1.4 Phase IV - Core Relocation to Lower Core Support Assembly (224 230 Minutes) i The second major core relocation event occurred between 224 and j
226 minutes, within about 100 seconds. This event was indicated r
by the RC5 pressure monitor, self-powered neutron detectors, and the source range neutron monitors.
It is believed that fatture of l
the supporting crust occurred in the upper and/or center region of i
the consolidated mass of molten core material, probably near the i
core periphery (1.5 meters from the bottom of the core) on the i
4 east side as shown in Figure 2-Ic. Visual inspections conducted during defueling indicated that the flow of molten core entered the core former on the east side and flowed around the core former l
and then down into the LCSA internals. Analysis of potential flow of molten core materials through fuel assembly location indicated
- i I
that all of the molten core material could have relocated into the LCSA internals and lower head in less than 1 minute through only one or two fuel rod assemblies.
j 2.1.5 Phase V - Long-term Cooling of Degraded Corv (af ter 230 Minutes)
Approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after the start of the accident, RC-P-1A was restarted and operated for approntmately one (1) week. This pump was replaced by AC-P-2A which operated untti Aprl) 27, 1979.
l There was no evidence of any additional major r91ocation of molten l
core materials into the LCSA and lower head after the second core relocation. Thus, the post-accident configuration of the core presented in Figure 2-It represents the final, stable, and coolable configuration for the matert Als in the core LC$A, and lower head regions. Detailed thermal analyses have evaluated the long-term cooling of the consolidated molten mass within the core region. Results of these studies suggest that cooling of this mass occurred over many days to weeks. It was also col;:1uded.
based on analyses and observations, '.'.at the RV maintained full Integrity during all phases of the accident sequence and the 2-3 Rev. 1/0461P 1
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l subsecuent defueling period. Therefore, only a small fraction of l
the original fuel inventory was relocated outside the RV and was contained althin selected RCS pathways.
j 2,2 Post Accident Condition of the Plant
)
i An accuratt determination of the post-accident state of the plant was required to understand the accident progression 4-d fuel transport i
mechanisms. Additionally, a thorough knowledge of the properties of the l
post-accident core deorts was necessary to anticipate the conditions to be encountered in defueling the RV and removing fuel from the RCS, RB, and support systems in the AFHB. Detailed ana'ysis of fuel including I
dispersion and general properties was also essential to completion of the final criticality assessment. This information was developed from i
several sources (References 2.6 through 2,11): visual inspections of RV internals, metallurgical / radiochemical examinations of samples acquired during the coursetof defueling, and readings from on-Itne instrumentation l
and experimental data developed from smaller-scale tests conducted at i
various independent facilities.
i The original core inventory included approximately 94.000 kg of UOg and 35,000 kg of cladding, structural, and control materials. Accounting for oxication of core materials Oring the accioent and for portions of the upper plenum structure that melted, the tctal amovat of post-accident core debris was estimated to be 133,000 kg.
l 2.2.1 Reactor Vessel Internals i
During the accident sfQuence discussed in Section 2.1, peak temperatures ranged from approximately 3100'K at the center of the core (molten U0 ), to 1255'K tmmediateiy soove the core and 2
723'K at het leg norrie eitwations. Approximately 5C'.' of the y
original core became eiten. Lower portions of three W carrte plates on the east side of the core melted and some of the molten
}
core material flowed into the core bypass region. ADeronimately 30,000 Le M elten materials flowed from the core to the core toass renton and throuch ins mier nal s. ADorosimateiv n-r 19.000 ko came to rett on the RV lower head. Tigure 2-3 illustrates the major RV components ann Ine post-accident configuration of the core.
l I
The post-accident condition of the upper plenum assembly, original core region, core bypass region, the UCSA, the LCSA, and lower head region are described in the following sections.
2.2.1.1 Upper Plenum Assembly The upper plenum assembly, which was removed intact, had two t
(2) damaged zones. Localtred variations of damage were evident in each rone. For example, in the Ilmited area above one fuel assembly, ablation of the stainless steel structure was observed; however, grid structures adjacent to the ablated zone appeared to be undamaged. In some regions, the once-molten grid material had a foamy texture, which occurs 2-4 Rev. 0/0461P
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when stainless steel ontdires near its melting point. A Once-molten mass close to this grid material appeiired to be I
unosidized, suggesting that some of the hot gases exiting the core were orygen deficient. The damage to the upper plenum i
assembly indicated that the composition and temperature of gases entting the core varied significantly within the flow stream. Only a small quantity of fuel debris was I.*asured within the plenum assembly.
2.2.1.2 Core Region Figures 2-2 and 2-3 illustrate the end-state configuration of the original core region. A core vold or cavity entsted at the top of the original core region. Below that, a bed of loose debris rested on a resolidtfied mass of material that was supported by standing fuel rod stubs. The stubs were surrounded by intact portions of fuel assemblies. A previously molten, resolidified mass was encapsulated by a distinct crust of material in which other fragments and shards of cladding could be identified.
I The core void was approntmately 1.5 meters deep with an j
overall volume of 9.3 cubic meters. Of the original 177 fuel 1
assemblies 42 partially intact assemblies were standing at the periphery of the core void. Only two (2) of these fuel assemblies contained more than 90T. of their full-length cross-sections with the majority of fuel rods intact.
The other assemb1!es suffered varying degrees of damage ranging from ruptured fuel rods to partially dissolved fuel pellets t
surrounded by once-molten material.
The loose debris bed at the base of the core cavity ranged in depth from 0.6 to 1 meters and consisted of whole and fractured fuel pellets, control rod spiders. Ondfittings, and i
resolidified debris totaling approximately 26,000 kg.
Beneatn the loose debris bed was a large resolicified mass i
approximately 3 meters in diameter. This mass varied in depth from 1.5 meters at its center to 0.25 meters at its periphery and contained approximately 33,000 kg of core debris.
The i
center of this solid metallic and ceramic mass consisted of a mixture of structural, control, and fuel material that reached i
temperatures of at least 2800'K and possibly as high as 3100'K during the accident. The upper crust of this mass, which consisted of the same material and also reached 2800'K, contained intact fuel pellets near the periphery. The lower crust consisted of once-molten stainless steel, tircaloy cladding, and control rod materials resolidified in flow channels surrounding intact and partially dissolved fuel pellets. The thickness of this lower crust, based on initial video examinations, was estimated to be approximately.01 meters on the average. The resolidified mass was shaped Itke a funnel extending down toward the fuel assembly lower endftttings near the center of the core.
4 2-5 Rev. 0/0461P e
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The standing, undamaged fuel assembly stubs extended upward from the lower grid plate to the bottom surface of the i
resolletfled region of the once-molten materials. These stubs varied in length from approntmately 0.2 to 1.5 meters. The longer partta'. fuel assemblies were located at tne periphery of the resolidtfled mass. On the east side of the core, one (1) fuel assembly was almost completely replaced with Once-solten core material; this indicated a possible 4
relocation path into the LCSA and core bypass region for molten material. The stancing fuel assembly stubs and h ttpheral assemblies constituted about 45,000 kg of core debris.
2.2.1.3 Upper Core Support Assembly This region consists of vertical baffle plates that form the peripheral boundary of the core; hortaontal core former plates to which the baffle plates are bolted; the core barrel; and the thermal shield (Figure 2 3).
There are a number of flow holes in the baffle and core former plates through which coolant flowed during normal operations. On the east side of the core, a large hole approutmately 0.6 meters wide and 1.5 meters high, and entending across three (3) baffle plates and three (3) core former plates was discovered. Adjacent baffle plates on the east and southeast were warped polstbly as a result of the high temperatures and the flow of molten material in the bypass region.
It was concluded that molten core material from the core region flowed through the large hole in the beffle plates into the UCSA, circumferential1y throughout the UCSA, and downward through the flow holes in the core former plates into the LCSA at nearly all locations around the core.
The majority of the molten material appeared to have flowed into the LCSA on the southeast side through the hole in the baffle plate and through the southeast core former plate flow holes.
The circumference of the core retton (i.e., the area behind the baffle plates) contained loose debris throughout. The depth of debris varied from approntmately 1.5 meters on the east side to a few millimeters on the southwest side. There appeared to be a resc11dified crust on the upper horltontal surfaces of the three (3) bottom core former plates; this crust varted in thickness from approximately 0.5 to 4.0 cm.
l It is estimated that approximately 4000 kg of core debris was retained in the UCSA region. In the small annulus between the core barrel and the thermal shield, fine particulates were l
observed but no major damage to these components was seen.
l 2.2.1.4 Lower Core Support Assembly I
i The LCSA region consists of five (5) stainless steel structures. The structures vary in thickness from 0.025 to 0.33 meters with 0.0bo to 0.15 meter diameter flow holes.
l l
2-6 Rev. 1/0461P 1
T Some molten core material flowed through these structures and came to rest on the lower head. There was approntmately l
6000 kg of resolletfled material dispersed at various i
locations on the circumference of these structures.
In l
several places, reso11dified material completely filled the l
flow holes and columns of once-molten material were observed i
between the plates.
The largest accumulation of resolidtfled i
material appeared to have flowed into the LCSA from the east I
side of the core. Although most of the material was seen on i
the east to southeast side, many columns of resolidtfled l
material were also seen in the LCSA around the periphery of the core beneath the core bypass region.
j 2.2.1.5 Lower Head Region
}
The debris in the lower head region accumulated to a depth of 0.75 to 1 meter and to a diameter of 4 meters. The spattal distribution of the material was neither untform nor symetric. The surface debris had particle sites which varied f
from large rocks (up to 0.20 meters) to granular particles (less than 0.001 meters). The larger rocks, especially in the northeast and southwest regions, were located near the l
periphery. The debris pile was lower at the vessel center i
than at the periphery, vith granular or gravel-Itke mater 1al observed in the central region of the vessel. A large resolidified mass was identified between the loose debris bed
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l and the lower head of the RV. This mass was approutmately 0.5 r
meters thick in the center and 1.7 meters in diameter. A i
large tilff.1tko structure formed in the northern region from
(
once-molten tore material. The cliff face was approximately i.
0.38 meters high and 1.25 meters wide.
It was estimated that j
approximately 12,000 kg of loose core debits and 7,000 kg of aglomerated core debris relocated into the lower head.
l 2.2.2 Reactor Coolant System Durtny the accident, small quantitles of fuel debris (Table 2-1) anc fission products were transported throughout the RCS (see i
Figure F-4).
The largest RCS components operated during the l
i accident were the RCPs.
The RC-P-28 was the only pump which would j
respond to a " start" command 174 minutes into the accident. This pump was started and operated for approximately 6 minutes. The operation of this pump was the major driving force for the relocation of fuel from the RV. Coolant circulated th mugh the RV by this pump caused a rapid quenching of the highly osidtted, high temperature fuel which resulted in the fuel rods being physically shattered and rubbled.
y As the RCP operated, the flow of the 't" loop was in a " forward" (i.e., normal) direction. The flow rate through the RV was i
sufflctent to transport small amounts of fuel 'nto the
't" loop l
where a portic.. Of the fuel relocated into the "S" hot leg and settled out into the decay heat drop itne. The decay heat drop llee connects to the bottom of the horttontal section of the o-2-7 Rev. 1/0461P o
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"O" hot leg and was found to contain some fuel. presumably as a e
,.. s' result of the RC-P-28 operation (see Table 2-1).
The coolant continued to flow up the " candy cane" and depostted fuel material t
on the "B" OTSG upper tube sheet. The tube sheet acted as a
- strainer" for the collection of fuel transported outside the RV.
However a small quantity of fuel flowed down through the steam generator tubes and was deposited on the lower head of the *8" 0TSG and J-legs. As the coolant continued to flow. relatively san 11er quantitles of fuel were then deposited in the "B" reactor coolant pump and cold legs.
At approntmately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />. the RC-P-IA pump was stcrted. The operation of this pump deposited finely divided silt-like debris in the top of the "A" 075G and the bottom of the "B" OT5G doe to reverse flow in the "8 OTSG loop. RC-P-lA. which espertenced escessive pump vibration, operated for approutmately one (1) week and was replaced by RC-P-2A, which oH rated untti Aprl) 27, 1979.
This pump was shutdown because all pressurtzer level indicators failed.
Cold shutdown conditions (i.e.
RCS temperature below 100'C) were estabitshed on the evening of April 27, 1979. After all RCP i
operations were terminated, the system circulation and cooldown was achieved by natural convection / circulation heat transfer.
i This natural circulation continued into approutmately October i
1979. Eventually, there was insuffletent thermal driving head to maintain continuous natural circulation and a flow trans' ent in the RCS, referred to as the *B" loop " burp " began to occur frequently over a period of several months. This phenomenon I
occurred because the coolant in the "8 075G and '8" loop cold legs i
gradually cooled untti the density of this coolant increased I
sufficiently to initiate natural circulation flow in the "B" f
loop. The flow was sustained until the warmer fluid frca the RV displaced the cold fluid in the "B" OTSG and cold leg.
Repositioning of the coolant of different densities continued i
i until hydraulic balance was achieved.
Th1 coolant was then stationary for several days untti another " burp" occurred. This
[
repeated flow rate phenomenon was believed to have transported small quantitles of finely divided fuel debris from the RV to the
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steam generators and other RCS locations in both RCS loops.
In sunnary, there were two (2) methods of transport of fuel to ex-vessel locations. The primary tr.tnsport method was a sequential operation of the RCPs: RC-P-28. RC-P-IA, and RC-P-2A.
The secondary transport method was attributed to the burping" i
phenomenon during natural circulation. Table 2-1 provides an estimate of the quantity of fuel relocated into the RCS during the accident sequence and resulting thermal hydraulic phenomenon t
(References 2.12 through 2.14).
I 2.2.3 Reactor Building l
Reactor coolant was discharged from the RCS through the PORV located on top of the pressurizer. The PORV discharges to the RCOT which is located in the basement of the R8 (see Figure 2-5).
2-8 Rev. 1/0461P l
l l
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The RCDT contains two (2) safety components: a rettef valve which discharges to the R8 sump and a rupture disk which discharges to the R8 floor adjacent to the RCDT cubicle. Both safety devices were believed to have performed their respective safety functions. The rupture disk was subsequently found in an open or ruptures condition, as espected.
If the retlef valve had initially operated during the pressure butidup in the RCOT, it would be espected to restat after the rupture disk opened, thereby 4
sintatzing any continuous release to the R8 sump via that pathway.
At approntmately 138 minutes into the accident, the operators realtred that the PORV was not closed and they manually closed the pressurifer bicck valve. Further loss of coolant and gaseous fission product release from the primary coolant system to the R8 was essentially terminated. However, the block valve had to be cycled repeatedly to maintain system pressure. This cycling of the block valve permitted the transport of fission products, noble gates, and small quantitles of fuel through the pressurizer and PORV into the RCOT and subsequently into the R8 through the rupture disk' discharge.
l The MU6P System was operated during the accident and recovery period. The MULP System inlet piping is fed from the RCS on the suct1on s1de of the RC-P-1 A.
The f1rst mejor components in this system are the letdown coolers which are located in the basement of the R8 (see Figure 2-5).
Thus, some fuel was transported into the letdown coolers and associated piping.
In summary, a relatively small quanttty of fuel (see Table 2-1) was released to the RB as a result of the accident due to the operation of. the PORV and the MULP System (References 2.13 through 2.17).
2.2.4 Auxiliary and Fuel Handling Buildings A small quantity of fuel was transported to the AFHB during the accident.
The majority of this material was transported through the MULP System and into the RCBTs. This system is fed from the RCS cold leg side of the "A" loop through tne letdown coolers and discharges into the AFNB via the RCSTs. Although this system communicates through a large number of the cubicles in the AFHB, only a small amount of fuel was transported into the system as
~
indicated by the fact that very little fuel was measured in upstream components such as the block ortftce, MU6P destnera11rer filters MJ&P domineralizers, and the makeup filters.
The block orifice is the normal pressure reduction device for flow rates up to 45 gpm through the Mu&P system. The block ortftce and its isolation valve became blocked during the accident; subsequently, they were bypassed. As a result, very little fuel was measured in the block orifice and its associated piping. The letdown flow was directed to the letdown filters and purification dominera11 ers at very low rates during the accident and was then
~
5 2-g Rev. 1/0461P
l
~
routed to RC87 "A" and the makeup tank. Letdown flow was lost several times during the accident due to flow blockage. More than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the unttlation of the accident, the purification demineralizers also were bypassed and letdown was directed to RCBT
'8".
Due to the flow blockage of the letdown coolers and restrictions in the block orifice, fuel transport to the filters, domineralizers, and ACSTs was listted.
L Another potential pathway for transport of fuel to the A8 was through the Seal Injection System. The Seal Injection System i
return line, which is downstream of the reactor coolant pump i
seals, rec 61ves reactor coolant pump seal return water. As a t
result of this, potential trace amounts of fuel may have been transported to the Seal Injection System.
RCSTs A, 8 and C also contained fuel as a result of their use i
during the accident, interconnection with the 44P System, and as i
a result of RCS water processing and removal of water from the RS sump and the AB sump.
In sumary, a relatively small quantity of fuel was transported into the AFNS (see Table 2-1), principally through the RCSTs and the MU&P System. Some of this fuel may have further relocated j
into other systems as part of the post-accident water processing and cleanup activities (References 2.13 and,2.14).
2.3 Fuel Transport and Relocation Due To Cleanup Activities As a result of the accident sequence and resultant cleanup activities, a small, but measurable quantity of fuel was transported into the various plant systems, tanks, and components. These cleanup activities were a I
necessary part of restoring conditions in the plant and significantly I
assisted recovery operations in meeting defueling compietton objectives.
In the RS, the majority cif the post-accident fuel material relocation from cleanup and defueling operations was attributed directly to the transfer of RV components. Major components have been removed from the l
RV which contained relatively small quantitles of fuel. These components, which are currently stored in various R8 locations, include the RV head, upper plenum assembly, internal RV structures (e.g.,
endfittings. LCSA grid plates, distributor plates, grid forging), and contaminated equipment / tools.
In all cases, these components and i
equipment were physically cleaned and decontaminated to the extent practical and surveyed for fuel content before storage. Some additional small amount of fuel material was relocated to the R8 basement as part of I
tool flushtr,g and building decontamination activities. In each case, the i
effect of this fuel material relocation is quantified as part of the fuel measurement activities reported herein.
l l
L l
f 2-10 Rev. 1/0461P l
0
_ _., - - _ _... - -, _,...., _ _... _. - _. _ _ _ _ ~
V 3
Subsequently, they were bypassed. As a result, very little fuel debris was measured in the block orifice and its associated piping. The letdown flow was directed to the letdown filters and purification demineraltrers at very low rates during the accident and was then routed to RCBT "A" and the makeup tank. Le'utown flow was lost several times during the accident due to flow blockage.
More than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the initiation of the accident, the purtftcation demineralizers also were bvpassed and letdown was directed to RCST "B".
Due to the flow e, lockage of the letdewn coolers and restrictions in the block orifice, fuel transport to the filters, dominera11rers, and RCBis was Ilmited.
Another potential pathway for transport of fuel debris to the AB was through the Seal Injection System. The Seal Injection System return line, which is downstream of the reactor coolant pump l
seals, receives reactor coolant pump seat return water. As a result of this, potential trace amounts of fuel debris may have been transported to the Seal Injection System.
RCBis A, B.'and C also contained fuel debris as a result of their use during the accident, interccnnection with the MULP System, and i
as a result of RCS water processing and removal of water from the RB sump and the AB sump.
In summary, a relatively small avantity of fuel debris was i
transported into the AFHB (see Table 2-1), principally through the RCSTs and the MULP System. Some of this fuel debris may have further relocated into other systems as part of the post-accident i
mater processing and cleanup activities (References 2.13 and 2.14).
2.3 Fuel Transport and Relocation Due To Cleanup Activities As a result of the accident sequence and resultant cleanup activities, a e
small, but measurable quantity of fuel debris was transported into the various plant systems, tanks, and components. These cleanup activities l
were a necessary part of restoring conditions in the plant and significantly assisted recovery operations in meeting defueling completion objectives.
In the RB, the majority of the post-accident fuel material relocation i
from cleanup and defueling operations was attributed directly to the transfer of RV components. Major components have been removed from the RV which contained relatively small quantitles (<10 kg) of fuel cabris.
These Components, which are currently stored in various RB locations, t
include the RV head, upper plenum assembly, internal RV structures (i.e.,
l endfittings, LCSA grid plates, distributor plates, grid forging, etc.),
i and contaminated equipment / tools.
In all cases, these components and equipment were physically cleaned and decontaminated to the extent i.
practical and surveyed for fuel content before storage. Some additional small amount of fuel material was relocated to the RB basement as part of L
tool flushing and building decentamination activities.
In each case, the l
effect of this fuel material relocation is quantified as part of the fup!
[
measurement activities reported herein.
2-10 Rev. O!0461P
4-In the AFHB, the primary cause of fuel debris relocation from cleanup operations was mater processing through the RCBis, MWHT, SRSTs, and 505 monitoring tants. Additionally, fuel debris material nay have relocated into the FHB Spent Fuel Pool "A" as part of fuel canister transfers from the RV. While every effort was made to flush residual fuel material from the external surfaces of the defueling canisters, a small quantity of uncontained fuel waterial may have been transferred into the "A" fuel pool as part of handling and movement of over 300 defueling canisters.
Post-defueling cleanup activities are expected to reduce the amount of residual fuel and ensure subtriticality.
P i
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t 2-11 Rev 0/0461P
.~
4 TABLE 2-1
[
ET-ACC : DENT EST1k ATED EX-VES$tt i
FUEL 4ATERI AL 1 STRIBJTION l
(References 2.12 througi 2.17) i Reactor Coolant Systes j
Kiloernes i
'A" $1de l
Not Ley.........................
1 I
OT5G upper Tube Shee t..................
1 Tube Sundle.......................
3 3
Lower Need;......................
1 J-Legs..........................
1 Reactor Coolant Pumps..................
2 i
Cold Legs........................
1 t
"B" Side Not Leg.........................
8 Decay Heat Drop Line................... 30 OTSG Upper Tube Sheet.................
125 Tube Bundle......................
9 Lower Head.......................
I i
J-Legs..........................
6 Reactor Coolant Pumps.................. 20 s
Cold Legs........................
7 Pressurizer...........................
12 Reactor Buildine l,
R 8 Ba s e me n t / S ump.......................
5 Reactor Coolant Drain Tank...................
0.1 l
L Letdown Coolers.........................
4 l
Core Flood System........................
1 Auxiliary / Fuel Handitne Buildincs Nakeup and Purification System.................
6 1
Seal Injection System......................
1 i
l Reactor Coolant Bleed Tanks A, 8, and C.............
15 l
Wa s t e Di s pos a l Li qui d Sy s t em..................
1 9
f k
f'%
2-12 Rev. 1/0461P
i l
e (a) Hypothesteed eere confleutetten het prior to pump tennelent et F3GURt 2-1 174 minutes.
1 y;
k j fi use, e *
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e-m am i
j sc ene l
r l
S) Hypothestaed sore config.
y,,, g Wretion het efter pump M
,,, gems trenelent et 174 minutes.
llg,7,,**'
o I
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(c) Hypothestaed core eenfig.
gg' i
urstlen during major core o s relocation event during aweesin g%*,88 W j.
224 226 minutes.
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2 13 Rev. 0/0461P
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f FIGURE 2-2 POST ACCIDENT ISTIMATJD CORT MTERIAL DI5_TRIBUTym 4
(STimit0 l
M DESCRIPTIOR QUmsitTY (KC) 1 Upper Debris Bed 26.000 j
2 Resolidified Mass-33.000
[
3 Intact Assea611e5 45.000 l
4 LCSA (loose debris 6.000
}j/
ky I
and resolidified l
j mass) i f'
q j
5 tower Need (Icose 12.000 l
y g
3 E
j y debris and 7
resolidified mass) 7,000 l
j l
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6 UCSA (loose debris 4.000 l
j q
l and resolidified i
[
l E
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f TOTAL -
133.000 g
I S
g l l l
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M i
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FIGURE 2 3 o-TMI 2 Core End State Configuration i
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.------.--__.-------m---._,,-,,--._,_-,n,--maw,~,v-,-------.,-,,-we,---.,-_..m,_,,w.m p, o-mmp.p-,,,ww,n,ww>mmw,,w.m_,,,w,y.,
h FIGURE 2-4 a.
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88 ge e
v-2-16 Rev. 0/0461P
c TMI-2 REACTOR BUILDING BASEMENT
~'
LEAKAGE COOLERS 8
STAIRS NO. 2 E
j Y-
- 'ac.m.g T'
,~
LEAKAGE TRANSFER ELEVATOR-
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Ch51rman'Carr: L. [ FOR SIGNATURE OF:
- GRN CRC NO: 89-1172 gyj d, f 3/0 / y 0
ROUTING: c DESC: 0 EXPLAIN-WHY NRC.IS CALLING THE SINTER-CAKE FOUND Bernero,JNMSS V IN'.THE' DAMAGED TMI-2 REACTOR "PREVIOUSLY MOLTEN Russ11, RI i MATERIAL": U-GDATE: 10/27/09 ASSIGNED.TO: CONTACT: 4
- NRR' Murley n.
SPECIAL INSTRUCTIONS OR REMARKS: ..7r NRR' RECEIVED:- OCTOBER 27,,1989 ACTION: .{DR,PR?VARGAl y,4 'NRR' ROUTING: MURLEY/SNIEZEK . PARTLOW CRUTCHFIELD MIRAGLIA s GILLESPIE .i M SSBURG G. gg g 3, 1c ACTION l L DUE TO NRR DIRECTOR'S OFFICE 1 1 BY li 1 A p l 1*. c e p g
- 4 &g
- y.
j i v& ,g~ L. ' f. - l. OFFICE OF THE SECRETARY i o . n f '* ' ' ' CORRESPONDENCE CONTROL TICKET i-a p I ! PAPER NUMBER: ' CRC-89-1172 LOGGING DATE: Oct 27,89 ) 1 ACTION OFFICE: EDO 1 1 1 r AUTHOR:' 'Walston Chubb .) ' AFFILIATION:-- PA (PENNSYLVANIA) ) LETTER DATE:- Oct 24 89 FILE CODE: y I
SUBJECT:
Reporting'that Uranium Dioxide can be sintered'in
)
steam at temperatures as low as 1200C..
1 i,
ACTION:.
' Direct Reply i
. DISTRIBUTION:
None p,
SPECIAL'~ HANDLING: None e
NOTES:
1 j
'DATE DUE:
Nov 10 89 1
SIGNATURE:
DATE SIGNED:
AFFILIATION
- J j
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1 l.
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