ML19253B803
| ML19253B803 | |
| Person / Time | |
|---|---|
| Issue date: | 08/28/1979 |
| From: | Tong L NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| Shared Package | |
| ML19253B800 | List: |
| References | |
| NUDOCS 7910220257 | |
| Download: ML19253B803 (15) | |
Text
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PLAN OF SMALL BREAK AND REACTOR TRANSIENT RESEARCll PF:ESENTED AT ACRS ECCS SuncoMMITTEE MEET!f4G AucusT 27 AND 28, 1979 BY L. S. I0f1G B
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1149 343
SCOPE OF SMALL BREAK AND REACTOR TRANSIENT RESEARCH (A)
RESEARCH ON BEHAVIOR OF LWR's DURING ANOMALOUS TRANSIENTS (INCLUDING POSTULATED AND EXPL_nATORY ACCIDENTS)
(B)
DEVELOP CAPABILITY (BOTH ANALYTICAL AND EXPERIMENTAL)
TO SIMULATE WIDE-RANGE OF POSTULATED TRANSIENTS AND ACCIDENT CONDITIONS (INCLUDING SMALL, LARGE, AND MEDIUM BREAKS)
NOTE: My oral presentation added:
"Our current plan is to modify the existing facilities to meet a basic requirement that all tests must be meaningful and useful."
e
.ag 1149 344
APPROACilES OF SMALL Br
- AND REACTOR TRANSIENT RESEARCil EllGINEERINGANALYSIS EXPERIMENTATION Review existing data (in test facility and Develop and justify the scaling criteria and e
in reactor) and infonnation (vendor's sub-rationale, identifying questionable area mission, NRC audit & ACRS recommendations) needing parametric study
- and identify the weakness of existing codes Plan experiment to meet the needs, including
& improvements required e
Inf nnation instrumentation requirement and facility Analyze reictor transient and abnonnal oper-y
+
e ational events and associated consequences 7
modifications Request based on physical behavior of plants (e.g II.
Establish test matrix an.1 expected results Establish the nature and extent of new phenoinena to be studied and the data required Construct & modify test facilities e
Perfonn detailed sensitivity study by t
computer codes Testing and data acquisition y,
7,
- _. Improve or develop models in the codes Validate engineering analysis results by e Data evaluation and error analysis experimental data and/or code calculations e Physical interpretation of experimentally Predict reactor behavior during various Informa tion measured separate effect and system integral transients k
behavior when considered in tenns of ' reactor Discover any safety deficiencies, such Organize the data into a useful format such as as unforeseen potential reactor accident, e
poor control, improper operation pro-correlations, models, non-dimensional groups cedures, etc.
+
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Capability of analyzing wide range Characterization of system response of postulated transients and accident during various postulated transients conditions in LWRs for use of simulating LWRs for confinning our h
aut'iting Licensee's calculations understanding and exploring further postulations NOTE: My oral presentation added:
"A facility project like Semiscale should be responsible for both Engineering Analysis and Experimentation."
APPROACHES - ENGINEERING ANALYSIS A E EXPERIMENTATION IN ESSENCE, ENGINEERING ANALYSIS IS NOT ONLY TO PREPARE COMPUTER INPUT AND TO READ PRINTOUT, NOR IS EXPERIMENTATION LIMITED TO BUILDING HARDWARE AND OPENING VALVES.
BOTH WRSR STAFF AND CONTRACTORS MUS* CONCENTRATE ON THE ANALYSIS AND UNDERSTANDING OF PHYSICAL PHENOMENA WITH OR WITHOUT A COMPUTER CODE.
1149 546 s.eg
~
e, SCALING CRITERIA Preservation of Physical Phenomena Small Break Large Break Correct time history of system press ure Power / volume = constant Power / volume = constant of flashing 2-4 fluid Break area / volume = constant Break area / volume = constant Fluid inventory and distribution Initial loop temp. L core AI maintained Correct history of pressure Elevation and flow resistance Flow length and resistance simulated coefficient simulated gradient throughout the system initial Core AT, RC pump flow and Core ouench during blowdown N/A loop resistance maintained Refiood heat transfer Core height maintained Core height maintained
' Natural circulation rate, lleight of steam generator and N/A fraction of core uncovery,,
loop geometry maintained Reactor thermal-hydraulic transient Both primary and secondary of N/A steam generator simulated.
Secondary coolant tenperature and level simulated.
Auxiliary FW systen simulated Downcomer size, geometry and rate of ECC Bypass N/A counter-current steam flow maintained F-'
Downcomer swelling and time Wall heat flux / volume = constant Wall heat flux / volume = constant
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constant of plant thennal transient sg)
'u Time of core uncovery'in S.B.
Upper plenum volume and distance Internals and nozzles in upper as,
~~a and steam binding in L.B.
between the top of core to the plenum to be simulated top of exit nozzle maintained
COMPARISON OF SMa'l BRfAK AND LARGE BREAK LOCAs Small Break LOCA large Break LOCA 2
2 Sample break s,1ze
'0.02 ft 4.0 ft SignificantI[eatSource Decay heat (Stored heat only Stored and decay heat in early stage)
Significant lleat Sink Break Flow and lleat Transfer thru Break flow and ECC water S.G. to Secondary side lleat Transfer in S.G.
Ppri > Psec. IIpri + II ec Psec> Ppri. Ilsec~* IIpri s
AFW Significant AFW Insignificant Primary Side Pressure liigh pressure matriained because fast depressurization by blowdown of slow drair.;ng Flow behavior in Primary 1.
Stratified flow 1
Bubbly or droplets dispccsed flow Side 2.
Separation of non-condensibles 2.
Ilomogeneous with finw at high spot 3.
Gravitational. force control 3.
Momentum control 4
Core may uncover by flashing 4
Core emptied and recosered quickly 5.
Pressurizer effect significant 5.
Pressurizer has less effect ECCS 1.
Charging pump & llPSI 1.
Accumulator most effective 2.
Effectiveness depends on the 2.
Effectiveness depends on the pressure for initiatA n of initiation. pressure and location injection of injection 3.
In. cold leg break LOCA, core may 3.
In cold leg break LOCA, there
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have to be partially uncovered may be steam binding and LCC to vent steam thru loop seal, bypass to slow down reflooding.
1.
AFH f,riatural Circulation for wot u
Plant recovery S.G.
1 Accumulator 7. Reflooding a
2.
Manual open all PORVs to lower the O
2.
Continuous tp51 or Ri rg pressure for lipSI, Accumulator, LPSI L RllR when steam dump is not available 3.
Test of emergency procedures
MODIFICATIONS IN SEMISCALE FOR SMALL BREAK TEST SCALING DEFICIENCIES IN MOD 3 MODIFICATIONS la WALL AREA TO VOLUME RATIO TOO LARGE IN DOWN-1.
ADD If1SULATIori TO ALL THESE COMER, CORE BARREL, UPPER AND LOWER PLENUMS LOCATIONS 2.
TOO SMALL A BhcAK OPENING AND THE SIZE OF 2.
DISCHARGE FLOW RATE AT OPENING WILL 15E PRE-CAllBRATED SURGE LINE IN J,CALING DOWN A SMALL BREAK LOCA 3.
IWO DISSIMILAR STEAM GENERATORS 3.
REPLACED BY SIMILAR STEAM GENERATORS 4.
NO SECONDARY SIDE PLANT TRANSIENT SIMULAT10fl 4.
ADD A CLOSED LOOP ON SECONDARY SIDE WITH PROGRA1MED COiPOL 5.
NO TWO EQUAL ACTIVE RC LOOPS AND PUMPS 5.
ADD A MATCHING PUMP IN' INTACT t
LOOP 6.
UPPER PLENUM SIMULATION NEEDS MORE STUDY 6.
DE-ENTRAINMENT EFFECT IN. UPPER PLENtlM COULD DE OBTAINED FROM
,C CCTF P
NOTE: My oral presentation added:
"I suqqest EGLG invite a group of independent consultants to review tiie scaling criteria and w
modifications.
It would be nice also to invite ACRS consultants and vendor experts to narticipate in t h i c, review."
END STATE OF ACCIDENT De,
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S.mgle Failure Failures or Core lntact
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+- All Safety Features Work Operator Errors r
Total Dreak Sizo NOTE: Hy oral presentation added:
"The End State is a function of time in which the core stays in an inciequate cooling condition. The surface ABCDE represents the end states of long time stay. Studying end state is for-(1) Confinning or exploring the ways and means to turn around the end state for plant recovery, and (2) Dethnnining the acceptable time lag for operator's remedial actions to limit the degree of core damage during the period of inadequate core cooling."
8/14/79 TESTING MATRIX LOFT Semiscale TitTF FLECllT TLTA PKL CCTF SCTF TPTF 2250 2250 2250 60 1000 500 80 80 2300 Technical Subjects For Testing psi psi psi psi psi psi psi psi psi 1.
Integra! System Tests a.
Natural Circulation Loop geometry & component effects X
X X
X X
System Voiding & S.G. condensation X
X X
X X
X Effect of non-condensibles X
X I
X X
ECCS injection & location X
X X
Effect of break locations X
X X
Effect of upper plenum size 1 exit nozzle height Effect of running RC pumps part time into X
X accident b.
Primary Loop Cooling Mechanisms Break flow cooling, w/o S.G.
X X
X Effect of RilR delay, w/o S.G.
X BWR jet pump effect X
c.
Secondary Loop Cooling Mechanisms S.G. performance & control X
X X
X X
tX Secondary simulation of reactor transient l
sw.
sC) d.
Symptoms of small break & reactor transient Diagnostic display & detecting instrumentation X
X w/4 Leak signal and location X
LJ]
i Foreign facility tests as backups.
8/14/79 TESTING MATRIX LOFT Semiscale TitTF FLECllT TLTA PKL CCTF SCTF TPTF i
2250 2250 2250 60 1000 500 80 80 2300 Technical Subjects For Testing psi psi psi psi psi psi psi psi psi 1.
Integral System Tests (continued) e.
Plant Recovery Techniques X
Adjust ECC activation pressures X
X Manual opening of PORVs X
X X
Return to natural circulation X
X Operator intervention X
X f.
Nuclear feedback in_, reactor transient X
including ATWS in long term 2.
Separate Effect Tests a.
Pressurizer and relief valve operation X
X b.
lleat Transfer in S.G. (UTSG or OTSG)
X X
Primary condensation and CCfL X
Secondary liquid level effect on ll.T.
X X
X Auxiliary FW inlet locations X
-* c. lleat Transfer in Uncovered Core Mixture level swelling X
X X
X X
X X
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Steam cooling with moisture & flow X
X X
X X
X X
X X
X X
Stagnant steam natural convection t/4 LJ7 rt)
TESTING MATRIX LOFT Semiscale TitTF FLECllT Ti T3 PKL CCTF SCTF TPTF i
2250 2250 2250 60 1060 500 80 00 2300 Technical Subjects for Testing psi psi psi psi psi psi psi psi psi 2.
Separate Effect Tests (continued) d.
Flow blockage in core X
X Reflooding Under water X
X e.
Two phase flow patterns X
X Slug & stratified flow in horizontal hot leg X
X Mixture level swe'lling in downcomer & rore X
X X
f.
Valves Olscharging rate of relief / safety valves X
X (Some data expected from Japan and FRG) t e
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PLANNED PKL-TESTS FOR SMALL BRCAKS 10/79 - 6/80 Corresponding items in
- of Test Goal for 20/3D Technical Coordination Tests at 500 psia Tests Detennination of Committee Agreement (6/26/79)
Natural circulation single 4
Heat transfer coefficient in SG la phase sub-cooled with and without break and ECC Natural circulation rate 2d Mixed surface development All influence of increasing phase 2c separation on heat transfer
& circulation 2-phase energy transport 6
Ileat transfer in SG under la dering increasing void condensation & counter-flow 2d fraction with and without conditions 2f non-condensibles Cooling behavior 2
Ileat transfer in the partly 2b of partly covered bundles covered bundles during 2-phase flow l
Transient shutdown in 7
Influence of primary side in-lb small break LOCA for plant jection & break flow on energy 2d recovery transport in primary systen.
lleat transport in an SG apk functioning as a heat sink Transient shutdown in 6
Like above Ib g,a medium break LOCA for plant lleat transport in an SG func-2d ty, tioning as a heat source as, recovery
~
CODES AVAILABILITY FOR PREDICTI IACTOR TRANSIENTS AND SMAll BREAK CODE LAB WilEN STRENGTilS AND WFAKNESSES AVAILABLE TRAC-PIA LASL
,e NOW Small Breaks: Excessive Running Time. (6lirs.)
Consequences of collapsing nodes to reduce running time not yet assessed.
Trips and controls not adequate for Reactor Transients analyses.
RELAP-5 INEL NOW Small Breaks:
Insufficient assessment as of now, concerning : mall break capability.
Trips and controls not adequate for Reactor Transients analyses.
TRAC-PF1 LASL December 1979 Good, fast running code, capable of addressing PWR small breaks and reactor transients.
(PWR)
Trip _s and controls as per RETRAN.
December 1980 (BWR) 4>=
sf)
RETRAN BNL NOW Not adequate for small breaks.
fjk (from Good for Reactor Transients (PWR and DWR), not requiring multidimensional neutronics EPRI)
Teedback. Good trips and c ontrol logic. Long running time.
ty3 RELAP-3B BNL NOW Adequate for many transients. Not for small breaks. Control loq1c not as good as that of RETRAN. Long running time.
RAMONA-III BNL NOW Good for many BWR transients and accidents. Not for small breaks. Fast running.
(BWR)
(from Some control system by December 1979; Complete controls by September 1980.
Norway) i'IRT BNL NOW For PWR transients, not for Small Breaks.
Improvement needed in S.G. modeling.
Fast running time. Controls and trips not as gogd as in RLiHAN, i
RELAP-4/
INEL NOW More adequate than RELAP-4/ MOD 6 for PWR small break analysis. Long MOD 7 running time.
Inadequate controls and trips for reactor transients.
s.
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1149 356