ML19290A097
| ML19290A097 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/10/1979 |
| From: | Duncan R ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| References | |
| TM-0954, TM-954, NUDOCS 7910010509 | |
| Download: ML19290A097 (11) | |
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COPY MADE ON OF DOCUMENT PROVIDED BY METROPOLITAN. EDISON COMPANY.:
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Supervisor, Document Control, NRC
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SUBJECT lfrILITY - IhBISTRY COMIT1 TEE ON REACTOR CORE CONDITION SU'StARY C hDRKING GROUP hEETINGS ON APRIL 5 APPROVED BY D
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ATTACHED FOR YOUR INFORTRTION ARE NOTES FTGiTHE FOLLOL(ING KORKING
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GRCUPS ON REACTOR CORE CONDITIONS HELD APRIL 5.
1.
ZIRCALOY - WATER REACTION AND }iYDROGEN INVENTORY 2.
CORE ENVIROMENT TDE HISTORY DETEPSIINATION AND CURREN. r THEINAL HYDRAULIC C0h7IGURATION 3.
PELLET CONDITIO'G
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l yIREE-MII,E ISL\\ND UNIT 2 UTILI1Y-INDUSTRY COSBlITTEE ON REACTOR COPdi CONDITION SGMARY OF h0RKING GROUP MEETING ON ZIRCALOY-h'ATER REACTION AND HYDROGDI INVENTORY 1
Held April S,1979 at BS1 Lynchburg, Virginia l.
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Compiled by R. N. Duncan, Director, F els Development Combustion Engineering, Inc.
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9 From the desk of R. N. Duncan Mr. James S. Tulenko
. Manager, Fuel Engineering Babcock 5 h'ilcox Lynchburg, Virginia
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THREE MILE ISLAND UNIT 2
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UTILITY-INDUSTRY COMMITTEE ON REACTOR CORE CONDITION
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SUMMARY
OF WORKING GROUP MEETING ON ZIRCALOY-WATER
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REhCTION AND HYDROGEN INVENTORY L.
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THE WORKING GROUP MEETING ON ZR-WATER REACTION AND HYRCOGEN INVENTORY WAS ATTENDED
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BY THE FOLLOWING:
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,f' R. H. DUNCAN, COMSUSTION ENGINEERING, INC.
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RALPH FREDERICKSON, BETTIS D. O. HOBSON, ORNL R. E. PAWEL, ORNL R. O. MEYER, NRC (PART-TIME)
C. J.' bAR0CH, B&W G. CLEVINGER, B&W A. LOWE, B&W
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V. DEMARS, B&W M. MONTGOMETY, B&W THE FOLLOWING.ARE HIGHLIGHT CONCLUSIONS AND RECOPl1ENDATIONS FOR FUTURE WORK RESULTING
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FROM THE WORKING GROUP DISCUSSIONS.
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1.
CONCLUSIONS
, _A.
FROM A,VAILABLE HYDROGEN INVENTORY DATA, THE EXTENT OF ZR-WATER REACTION WHICH OCCURRED AT TMI-2 APPEARS TO BE IN THE RANGE OF FROM 2S-35% OF AVAILABLE ZR CLADDING. EVALUATIONS BY DIFFERENT ORGANIZATIONS HAVE USED DIFFERENT INPUT-NUMBERS FOR ZR INVENTORY WITHIN THE CORE. THE FOLLOWING INVENTORY IS FROM
, B&W AND HAS BEEN USED TO NORMALIZE THE DIFFERENT EXTENT OF METAL-WATER REACTION ~
ESTIMATES: ZR-4 CLADDING - 45,000 LB. IN THE CORE. OTHER ZR COMPONENTS SUCH AS. GUIDE TUBES, INSTRUMENT TUBES, FUEL R0D END CAPS, FUEL R0D SPACERS AND BURNABIE POISON ELEMENT CLADDING, A TOTAL OF 7,000 LB. IN THE CORE. THE TOTAL INVENTORY OF ZR IN THE CORE IS 52,000 LB.
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TO ASSESS THE CONDITION OF THE TMI-2 CORE, THE EXTENT OF METAL-WATER REACTIO'l
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IS A MAJOR ELEMENT AND CAN OtlLY BE DEDUCTED FROM A DETAILED EVALUA HYDROGEN REACTION PRODUCT INVENTORY.
_ RECOMMENDATIONS A.
WHEN REPORTING ESTIMATES OF METAL-WATER REACTION EXTENT: BA
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i ZR INVENTORY SHOULD BE STATED.
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B.jDETAILS OF THE HYDROGEN INVENTORY NEED TO BE CAREFULLY LOOKED AT AND DOCO-MENTED Ill A MORE FORMAL MANNER. THIS EVALUATION SHOULD INCLUDE PO INVENTORIES UHICH HAVE NOT 8 KEN CONSIDERED IN SOME OF THE EVALUATIO EXAMPLES; HYDROGEN IN THE FORM OF HYDRIDES IN THE CLADDIllG AS A RESULT OF THE METAL WATER REACTION, HYDR 0 GEN DISSOLVED IN THE PRIMARY COOLANT AFTER THE
- METAL-WATER REACTION AND HYDROGEN INVENTORIES WITHIll THE PLANT, SUCH AS IN THE RADWASTE SYSTEM.
2.
CONCLUSION IN THE ASSESSMENT OF TEMPERATURES WHICH MAY HAVE BEEN REACHED, THE ENERGY _
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- DIRECTLY FROM THE METAL-WATER REACTION HAS NOT BEEN CO 8
RECOMMENDATION AN EVALUATION OF THE METAL-WATER REACTION RATES THAT MAY HAVE
$UBSEQUENT EFFECT ON LOCAL CLAD TEMPERATURES SHOULD BE EVALUATED, SINCE ALTER THE PICTURE RELATIVE TO THE CORE CONDITION.
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3.
CONCLUSION B&W EVALUATIONS OF INTERNAL FUEL R00 PRESSURE BUILDUP INDICA NAL PRESSURE COULD EASILY HAVE BEEN ACHIEVED. USING THE SYSTE AVAILABLE, B&W FEELS THAT 8.ALLOONI.NG AND A STRESS RUPTURE MODE BURST UOU WHEN THE CLADDIllG REACHED 1600 F USING LOCA TYPE HEATING RATES.
AT THE SLO *.ER HEATING RATES TYPICAL OF THAT THOUGHT TO HAVE OCCURRED AT THI-RUPTURES WOULD BE ANTICIPATED AT 1400 F CLAD TEMPERATUhE.
THE T10ST LIXELY REGION-
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- TO HAVE FUEL R00 DAMAGE IS THE UPPER PORTION OF THE CURE..THE UPPER PORTION i
OF THE CORE WAS UNDOUBTEDLY AT HIGH TEMPERATURE FOR TI,lE LONGEST TIME PERIODS.
ADDITIONALLY, FROM. THE FUEL ASSE!!BLY C00LAflT OUTLET TIiERM0 COUPLES READINGS, IT t
r APPEARS THAT THERE IS A POSSIBLE RADIAL DISTRIBTION OF FUEL DAMAGE AS WELL, WITH I
THE HIGHEST DAMAGED REGION TOWARD THE CORE CENTER.
j RECOMMENDATION THE OUT-OF-REACTOR TEST TO EVALUATE THE EXTENT OF FUEL GEOMETRY CHANGES WHICH MIGHT OCCUR NEAR THE TOP 0F A FUEL ASSEMBLY UHICH CONTAINS HEAVILY OXIDIZED ZR-4 CLADDING AtlD WHICH HAS UNDERGO'NE MULTIPLE HEATING AtlD COOLING CYCLES MAY BE WORTH j
4 PERFORMING. SUCH A SIMULATION WOULD'ALSO SHED LIGHT Oil THE POTENTIAL FOR EUTECTIC MELTING AT HIGH TEMPERATURES AT ZR-4/AUSTENITIC STEEL INTERFACES.
4.
CONCLUSION THE TIME SEQUENCE OF FUEL ROD PERFORATION COULD BE ESTABLISHED IF GASEOUS FISSION PRODUCT ACTIVITY RELEASE DATA WERE AVAILABLE AS A FUNCTION OF TIliE. DETAILED ACTIVITY RELEASE DATA WAS NOT AVAILABLE TO THE WORKING GROUP.
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RECOMMENDATION EVALUATE FISSION PRODUCT ACTIVITY RELEASE INDICATIONS (BOTH GASES AND SOLUBLE FISSION PRODUCTS) AS A FUNCTION OF TIME.
5.
CONCLUSION THE QUESTION OF GUIDE TUBE INTEGRITY IS CRITICALLY DEPEt! DENT ON THE EXTENT.0F ID COOLING WHICH ftAY HAVE BEE,N AVAILABLE DURING HIGH TEMPERATURE PERIODS. THE GUIDE
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TUBE IflTERNAL FLOW GEOMETRY VARIES WITH SEVERAL FUEL ASSEMBLY CONFIGURATIONS.
WITHIN THE CORE THERE ARE ASSEMBLIES WHERE THE GUIDE TUBE: CONTAIN CONTROL RODS, CONTAIN BURtlABLE P0lS0ij RODS OR ARE EMPTY.
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REC 0ft4ENDATION O.
', EVALUATE FUEL ASSE!ELY COOLANT GUTLET TEMPERATURE DISTRIBUTIONS TG ESTABLI THERE IS A PATTERN, NOT ONLY OF POSITION IN CORE BUT OF THE PARTICULAR CONFIGU-RATION WITHIN THE GUIDE TUBES, E.G., EMPTY VS. RODDE0.
OVERALL, IT BECAME APPARENT THAT -TQE14 EMBERS OF THE WORKING GROUP WERE NOT AWARE OF CERTAIN KEY INFORMATION PRIOR TO THIS MEETING. EXAMPLES INCLUDE: THE CONTAIN-MENTOXYGENANALYSESWHICHWEREUSEDTODETERMINEHYDR0dENREMOVEDFROMTHE CONTAINMENT BY A POSSIBLE FIRE.0R EXACT TIttIt'.G OF THE FIRST MAJOR BURST OF FISSION PRODUCT ACTIVITY FROM THE CORE. A MO,RE SYSTEMATIC CATALOGUING OF FACTS WO PROVE THE ABILITY'0F THE WORKING GROUPS TO STATE CONCLUSI0NS AS TO MOST PROBABL CONDITION OF THE CORE.
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THREE-MIk,EISLANDUN'IT2
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UTILITY-INDUSTRY COMMITTEE OR REACTOR CORE CONDITION
SUMMARY
'0F WORKING GROUP MEETING ON CORE ENVIRON 11ENT,,
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TIME HISTORY DETERiINATION AND CURRENT THERMAL HYDRAULIC CONFIGURATION f
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WORKINC GROUP MEETING i
4/5/79 Summarized are a list of obse' vations : hat were discussed l
r for working group along with the information that might be defined from th~ese observations.
'[.tslitional Observations _
,'2' Reasons 1.
Evid'ence of boiling.
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-Noise on high TC's..,,
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History of ex-core current.
- 2. ~ Rate of water level drop.
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Present current readings of 3.
Distributi$n of gamma emitters.
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excores.
4.
Gas ' solubility "spmginess".
N.
Gas rate emissions.
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Noise on pressure sensors.
5.
Movable in-core operation.
5.
Gamma distribution in-core.
6.
Ifove an APSR (nominally 257.
6.
Will tell if guide tubes are withdrawn).
intact.
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Calculate temp. distribution.
7.
Guide to TC indication.
8 Correlate SPND and TC data 8.
Rate of heatup vs core location ~.
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during transients.
9.
Zr-water reaction heat 9*
Calculate heat rate of Zr-H O contributa.on to heating of pellet.
2 reaction.
10.
Identify how various delta P &
10.
Blocking.
flows go around.
11.
Materials & energy balance 11.
Get better thermal hydraulic input.
sequence Better history e.g., quench t3hk pressure 12.- Temp. distribution from SPND's 12.
Distribution of cooling flow.
(measure resistance).
13.
Better estimate-od cote flow.
13.
How much bypass?
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Outside of core measurement -
Detcirmine how much convective
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cooling.
Determine if we can turn electrical heat.
on second pump.
14.
Thermocouples telling, truth?
14.
Distribution of core flow.
Flow streaming?
1 Good core flow balance - what 15.
Same.
are resistances around the 2221 134
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,ionni Observations Reasons pelta T across' core from TC's 16.
True core flow.
' ',~ &nd RTD's.
Change core flow (pump off) or 17.
Same.
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' Modulate flou.
18.'hlowthroughOTSG.
D.P. across steam generator.
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- . More analysis of SPND data.
19.
Core temp.'
i Use of qdick transient code
- 20. ' Distribution of heat generation.
/.0 (CRAC, strip-down of FRAP) l 21.' Delta P across core.
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21.
Core dislocation.,
22.
Measure pump power & coolant 22.
Get Delta P across core,
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temp.
,23.
Analyze gamma in containment.
_23.
Core uncovering.
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24.
Mass balance on primary coolant 24.
Get quench tank heatup rate.
7 system.
'25.
In natural circulation -
- 25. ' Determine heat contribution.
TC reading Imbalance of constants.
L 26.
Analyze for boiling.
26.
Determine boiling which determines gas control required.
27.
Maintain H over pressure 27.
Recombine oxygen.
to keep 10 cc/kg.
28.
Experiments on hydrogen 28.
Low temp. Zr H 0.
2 evolution (can do it -
Is it self liuu. ting?
Johnson says we have some).
Have cracked cladding?
29.
Correlate radial and axial history.
30.
Follow future transients with TC and other instruments, e.g. SP.ND's.
31.
Temperature stability of RTD's.
31.
Accuracy of temperature.
32.
Look for a' lumina counting for 32.
Determine failures of SPND's aluminum activity, of lumped burnable.
33.
More data on PR and 33.
Access,to system and radionnalysis of primary coolant.
configuration changes.
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. THREE-MIL 5 ISLAHD UNIT 2
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UTILITY-INDUSTRY'COMMITTBE.OR REACTOR CORE CONDITION
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SUNIIARY OF.WORl'ING GROUP MEETING ON PELLET CONDITIONS Held April 5, 1979 at B&W Lynchburg, Virginia Compiled by Gary Thomas EPRI k
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Three-Mile Island Unit 2 UtilithIndustryCommitteeonReactorCoreCondition
.b Summary of Working Group Meeting on' Pellet Conditions l
The working group meeting on P,cliet Conditions reached the.
following highligh_t conclusions and recommendations for future work
. re,sulting from the Working Group discussions.
1.
Because of the limited opportunity'that fuel melting could have occurred, based on heat tra,nsfer boundary limitations, little active concern should be spent on this area.
Based on noble gas fission produqt release, fuel peak temperatures 2.
would have been in region of $1700-18000C.
3.
'The high core ATs (and' calculated APs12 psi) in some regions indicate significant mechanical fuel and/or cladding redistribution leading '
'to locally high flow resistances.
'4.
For purposes of' determining fuel conditions, new primary coolant sample analyses are not needed under current conditions because i
these samples may not b'e representative of bulk coolant chemistry 4
(sample line cannot be cleared first because of activity levels and upstream filters may be plugged - preventing proper sampling
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of some solids).
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Some suggested tests
- Can neutron sources be used for subcritical tests?
- Can single control rod be moved in region of hottest assys?
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- Thru use of existing self-powered neutron detectors and ex-core monitors, can information on axial / radial fuel relocation be obtained thru the above tests?
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