ML19345G309
| ML19345G309 | |
| Person / Time | |
|---|---|
| Issue date: | 02/11/1981 |
| From: | Minogue R NRC OFFICE OF STANDARDS DEVELOPMENT |
| To: | |
| Shared Package | |
| ML19341B548 | List: |
| References | |
| REF-10CFR9.7 NUDOCS 8103180146 | |
| Download: ML19345G309 (12) | |
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o PRESENTATION TO THE COMMISSIONS OF THE NRC:
THE LOFT TEST PROGRAM BY R. MIN 0GUE RES OFFICE DIRECTOR The experiments proposed by Researci: Office in the remaining LOFT program have one common thread; that of providing data for the assessment and refinement of computer codes used to predict the behavior of commercial reactors under accident and anamoulous transient conditions. Each experiment is chosen to study a different area of concern pertinent to licensing issues.
Furthermore, these experiments take full advantage of the fact that LOFT is the largest reactor system available for safety research, and that it is unique in providing reactivity feedback effects and three-demensional fuel
. behavior effects.
We realize that differences between the LOFT reactor and commercial reactors cause atypicalities in behavior, h'owever, LOFT exhibits most of the physical
. phenomena which occur in the commerical plant and we believe that we can successfully account for the atypicalities through the medium of the computer codes.
Following, is.a list of safety related questions for which the Research Office believes LOFT is especially suited to provide answers, together with the list.of tests intended to provide those answers.
Issues and Areas of Concern How does a PWR behave upon a complete loss of feedwater? What are the piping loads downstream of the PORY when it operates under these conditions?
!4 the vendor-proposed procedure of opening the FORY to depressurize, in this situation, the correct or best procedure to prevent core damage? What is the effect of cooling a hot uncovered core of restarting the primary coolant pumps? Does natural circulation return wn?n the steam generator is refilled?
proposed Tests: Piggy-backed L9-1/L3-3 L9-1: Loss of all feedwater with delayed scram. Challenges and opens PORY under high primary system pressure. Steam generator dries out and natural circulation is lost.
L3-3: Lock open PORY to reduce primary pressure. Restart the primary coolant pumps when core temperature rises. Reflood steam generator secondary to study reinitiation of natural circulation.
Risk of Fuel Failure:
small Level of contamination in event of fuel failure:
small
r Issues and Areas of Concern Establish a scaling benchmark between LOFT and commercial PWRs. Provide data to. improve predictions of PWRs during rapid cooldown. Develop an understanding of the formation of a steam bubble in a pressure vessel upper plenum. These questions arose from the turbine trip incident at Arkansas Nsclear One cad the rapid _cooldown incident at St. Lucie.
r Proposed Tests: Piggy-backed L6-7/L9-2 L6-7: Turbine trip with stuck-on pressurized spray ara 4cuck-open atmospheric dump valve. Simulates Arkansas Nuclear One startup incident causing rapid cooldown.
L9-2: -Continue rapid cooldown causing. reactivity injection and upper plenum voiding as in St. Lucie incident.
Risk of Fuel Failure:
small Level-of contamination in-event of fuel failure:
small y
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Issues and Areas of Concern Different flow models must be used to predict large and small break LOCA behavior. Which models are applicable to intermediate break LOCAS? How effective is the ECCS for an intermediate break? How does a plant behave and what recovery procedure should be used in the event of an accumulator line break.. What is the clad temperature response during core uncovery following an intermediate sized break?
Proposed Tests: Piggy-backed L5-1/L8-2 L5-1: Accumulator line break - intermediate break size LOCA with no accumulator inj ection. Leads to...
L8-2: Sustained core-dryout during high decay heat level. HPIS and LPIS injection must serve to recover core.
Risk of Fuel Failure:
medium Level of contamination in event of fuel failure:
small L
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I Issues and Areas of Concern Following four operational transient tests run to date. everal areas of code et'ficiences have been identified. Once the appropriote i=provements are made and the codes reassessed it may be necessary to run a final test to ensure that the i=provements are not test specific. Plans will be made to do this te:t and any decision to delete it will be made in time to move the following test forward with in loss of facility time.
Proposed Test:
LA-10 (LSRG designation, to be changed) Add-on operational transient, to be identified and _ preplanned.
Risk of Fuel Failure:
Icw Level of contamination in event of fuel failure low g
o Issues and Areas of Concern Based on the first 2 large break LOCAs, best-estimate predictions of the Appendix K design basis LOCA suggest that no fuel will fail. Since no test in which a loss-of-offsite power is simulated has been done and since this factor is expected to significantly affect the phenomena which influence the peak clad temperature, this importar.t conclusion should be demonstrated, and it should be done using prepressurized fuel.
What is the effect of initiating a large break LOCA from off rtrmal condition?
The effect of clad mounted thermocouples on the peak clad temperature has been studied, and, in the case of the LOFT large break LOCAs, it has not been found to be a very significant factor. Some people, well-respected in the reactor safety field, continue to disagree with this conclusion. Since either side can be supported by different separate effects tests, a test resolving this question should be run in LOFT.
L2-5: Large break LOCA at off-normal initial conditions with loss-of-offsite power leading to primary coolant pump rundown and delayed ECC injection.
(LA-1) Central fuel bundle prepressurized to 350 psi, and instrumented with new clad-imbedded thermocouples.
Pe.sk clad temperature predicted to fall below yield point.
Risk of Fuel Failure:
moderate Level of contamination in event of fuel failure:
medium m
Issues and Areas of Concern The first intermediate break LOCA has yet to be run.
If this test uncovers unexpected behavior, pheonomena or code problems, a second test may be needed. This test is scheduled 8 months after the first, to. permit sufficient time to analyze the results and plan the second.
Any decision to delete the test will be made in time to move the followir.g test forward with no loss of facility time.
Proposed Test:
LA-2( LSRG designation, to be changed) Add-on intermediate break, to be identified and preplanned. (possibly a break in the pressurizer surgeline).
Risk of Fuel Failure:
low Level of contamination in event of fuel failure low to moderate s
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r Issues and Areas of Concern How does a plant respond during a scram-initiating event followed by a failure of the nor:al scram mechanisms? ( anticipated transient without scra:n, ATAS).
In order to assess the neutronic feedback aspects of codes used to predict ATAS behavior, a well instr =4nted nuclear system :~.ast undergo a planned ATAS and the resulting data obtained.
Proposed Test L9-3:. A multifailure test in which one of the failuras is the nor:al scram mechanicci.
Initiating failure would be a loss-of-feedwater.
Risk of Fuel Failure:
low Level of contamination in event of fuel failure:
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Issues and Areas of Concern P
Following a small~ break LOCAs run to date, several areas of code deficiencies have.been identified. Once appropriate improvements are made and codes reassessed, it may be necessary to run a final test to ensure the improvements are not test specific. Plans will be made in time to move the following test forward, with no loss of facility time.
Proposed Test:
LA-9 (LSRG designation, to be changed) Add-on small break LOCA, to be identified and preplanned.
(could be a pumps on/off related test).
Risk of fuel Failure:
low
- Level of contamination in event of fuel failure:
low to medium 1
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Issues and Areas of Concern The first ATWS test has yet to be run.
If this test uncovers unexpected behavior phenomena or code problems, a second test may be needed. This test is scheduled 6 months after the first, to termit sufficient time to analyze the results and decide on proceeding with the second. Any decision to delete the test will be made in time to move the following test forward with no loss of facility time.
Proposed Test:
LA-3 (LSRG designation, to be changed) Add-on intermediate break, to be identified and preplanned.
Risk of Fuel Failure:
low Level of contamination in event of fuel failure:
low to moderate e
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Issues and Areas of Concern By how much must the ECC injection be delayed and to what temperatures must the fuel cladding rise before significant clad ballooning and burst occur?
What is the appearance of a 15x15 bundle in which massive ballooning and burst has occurred, and how effective is the ECC in reflooding the bundle.
Does pellet washout occur in this case? What is the fission product dispersal for this situation? Do the surface clad thermocouples affect the outcome of this test?
a L2-6: Large break LOCA with loss-of-offsite power and extended ECC delay.
fuel bundle is prepressurized to 600 psi, and instrumented with clad-imbedded thermocouples. ECC will be delsyed to insure ballooning and burst occur generally throughout the central assembly.
Risk of Fuel Failure:
high Level of contamination in event of fuel failure:
medium A
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Issues and Areas of Concern In the course of a multifailure transient which leads to core uncovery.
Nat does severe core damage proceed? How does fuel coolability change?
To what degree does the fuel disperse? What is the fission product release and where do the fission products dispers,' to? What questions will arise from the Degraded Core Cooling Rule Mak.Ng Hearings which can be addressed by a degraded core cooling test in '0FT? Where else can relevant data be obtained of fuel damage behavior in a large core.
L8-4: A multifailure transient to be determined will lead to ore uncovery.
Fuel failures will be allowed to progress to some point to be defined in the light of the DCC rule making hearing.
Risk of Fuel Failure:
high Level of contamination.in event of fuel failure:
high 5
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