ML20006A560
| ML20006A560 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/18/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20006A559 | List: |
| References | |
| NUDOCS 9001290113 | |
| Download: ML20006A560 (4) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION I
RELATED TO IN-PLANT SAFETY RELIEF VALVE DISCHARGE TEST RESULTS LICENSE CONDITION 2.C.(4)
DETROIT EDISON COMPANY t
WOLVERINE POWER SUPPLY COOPERATIVE, INCORPORATED i
FERMI-2 DOCKET tk. 50-341 j
1.0 INTRODUCTION
In 1975, the NRC identified concerns regarding Boiling Water Reactor (BWR) Mark l
I containment system design.
These concerns were based upon test results showing increased dynamic loads from safety relief valve (SRV) discharges.
The SRVs are mounted on the main steam lines inside the dry well, with discharge pipes routed into the suppression pool.
The scenario was postulated as follows:
When an SRV is actuated to provide overpressure protection for a primary system, steam is discharged from the primary system thrhig,h the SRV into a discharge line that leads to the pressure suppression pool.
Air initially exists in the line and is compressed by the influx of steam.
The water column at the end c,
.i the litm which i$ submerged in the pcci is expelled first through T-Quenthers mountui st the submerged end of the line. This water coluran is followed by the corrpressed air, which forms one or more air bubbles in the pool.
Each pubble undergoes oscillatory t;xpansion and contractions as it rises to the surface of the pool.
Following the air-clearing phase, steam is injected into the pool U
through the qw ncher.
fhe steam-water interfaces formed at the quencher during this phase is stable as long as tne local pool temperature terrains Dinow the normal boiling 7emperature of 212*F.
In swaary, the discharge of _ both the air, which was in the SRV line, and the steam into the suppression pooi produces hydrodynamic loads on the containment structure, piping, and equipment,,
The key parameters affecting the loads and the pool temperature gradients have been identified th*
4h generic testing.
However, concerns have been expressed that there is enough.incertainly about the interdependence and quantitative effects of plant-specific variables that confirmatory testing should be conducted in plants in which these parameters are substantially different from those previously tested.
The generic approach was accepted by the NRC in NUREG-0661, y'
" Safety Evaluation Report for Mark I Containment Long-Term Program," dated July 1980 and in Supplement No. I dated August 1982.
The guidelines for the in plant tests were provided in NUREG-0763, " Guidelines for Confirmatory In plant Tests of Safety Relief Valve Discharge for BWR Plants," dated May 1981.
The generic analysis was to be applied to each Mark I containment plant through the use of a Plant Unique Analysis (PUA).
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2-This process was undertaken for the Enrico Fermi Atomic Power Plant Unit 2 (Fenti-2) as part of the initial 91 ant licensing. The NRC accepted the Fermi-2 PUAcontingentuponconfirtaationbyin-planttestoftheconservatismofthe
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load reduction factor used in the calculation of SRV water jet impinger +nt and air bubble drag loads. The Fermi-2 Operating License included License Conditien l
2.C.(4) requiring these in-plant tests and submission of an analysis of the test results to the NRC staff within six rnonths of the con:pletion of the testing.
2.0 fACKGROUND In order to satisfy the licensing condition. Detroit Edison conducted a series of in-plant SRV discharge tests for Fermi-2 on March 12, 1987. The test was preceded by shakedown tests performed on March 11, 1967. The test matrix consisted of two shakedown tests, four single valve actuations (SVA), and four consecutivevalveactuations(CVA). The tests followed the general guidelines provided in NUREG-0661 and NUREG-0763.
The test program focused on n+asurement of the following:
1.
Peak suppression pool boundary pressures durirg SRV discharge line (SRYDL) air clearing and stearn discharge due to a singit SRV actuation under normal water livel in the submerged section of the discharge line and under both cold and hot conditions of the line.
l 2.
Pressure ruegnitude and frequency content of the T-Quencher air bubble pressure transients.
3.
Kater and cir clecting reaction 1rjeds on the SR\\CL and T-Quercher l
- Suppcris, 4
Suppression chatber structur>1 respense inclucing tmvs shell c+.r$rane stresses due to a singic SRV discharge (colo urf tot pipe).
Fout typts of instrim,entation were used for sensing and neas. iring the parannters.
The instru tnts are pressure transducers, stain gauges, accelerometerh vtsistaice ter.perature detectors and Uisting plant system thermoccuples.
Twelty presicre transducers were inst 211ed to measure the torus shell, SRYDL and T-Quencher air bubble and internal pressures. IMnety-eight etair gauges were instelled on the containment, submerged structures and piping to measure representative strein data.
Four acceleror+ters were installed to neasure the torus shell and torus-attached piping response to SRV discharge loads.
Two temperature sensors were used to monitor the SRVDL vent line penetration end wet well SRVDL ternperature.
Upon corapletion of the test ptogram, Detroit Edison submitted to the NRC a l
report dated November 13, 1987 describing the test procedures, test instrurentation, and the results as discussed above. The report entitled
- Final Test Report:
In-Flant Safety Relief Valve Discharge Test - Enrico Fermi Atomic l
Power Plant, Unit 2" was prepared by Hutech Engineers Inc.
l
3-After a preliminary review of the above report, the NRC requested additional information and clarification in the following areas:
1.
Sequence of testing 2.
Testing temperature j
3.
90-90 vs 95-95 statistical results 4.
Analytical results used for comparison with test data Detroit Edison provided the requested informaticr. by letter dated August 12, 1988.
300 EVALUATION The NRC staff has con.pleted review of the report and the additional information.
The following is a sumery of the evaluation.
Regarding testing sequence and temperature, the scope of the Femi-2 in-plant test was limited to confirmation of the SRV discharge methodology used in the Fermi PUA, and this was mainly to address the issue that the discharge configuration at Termi-2 is geometrically different froir configurations tested previously. Suppression pool thermal mixing tests were not perforr+d since the pool terperature response to SRV transients descrited in the PVA denionstrates corrpliance with the required pool temperature limits.
NUREG-0703 recorcnends testing under norrnal discharge conditions and that plant specific tests generally r.ct include leaking valve actuations (LVA).
For inadvertent testing under LVA conditions, load changes should be quantified on a generic bacts.
For Fermi-2 all but one of the SVA discharge tests were perferrmed with e tail pipe teitperature o# appioxiretel" 21 W which is above the range of nors.cl phnt nperation tersperature.
This was indicative cf a 1eoHrg velve. Corrpred to the one test perforc4d with the katient temperature, the SRV tests under the leaking valve condition tend to increase the derninant tuttle frequency tcward the fundertental frequency of the suppression pool b.
chanber ard retults in a conserv6tive shift of the discharge lehds. Therefore, L
the eleveted tail pipe terapcrature cet ditions are accepteble, i
Ecgarding the test and analysis results, the test data were statistically l
ar.a'yzed to obtain a 90-90 probability value which was then coinpared with the I
results obcained in the PUA.
in response to a request for 95-95 probability l
values, the licensee concluded that the quality of the test data is such that
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therr.agnitudeofthechange(90-90vs95-95)wouldbesmall. Margins in the PUA results cor. pared to the test 90-90 data are appreciably high. Therefore, 95-95 values are expected to be bounded by the PUA results.
L The critical issues being addressed are whether the tests were sequentially and properly perforned and the proper pareceters characterizing the actual SRV hydrodynamic loading were accurately measured and cornpared to the analytically estirttted values used in the PUA reports. This is to be done within the I
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l guidelines of NUREG 0661 and NUREG-0763. The key parameters selected for this purpose and the results of the cornparison are as follows:
1.
Peak Pressure - The measured peak T-Quer.cher bubble pressure and the torus stell pressure are less than 50% of predicted values. The test data were analyzed for a 90-90 probability value, i.e., 90% confidence that 90% of measured results will be less than the seat pressure. T-Quencher bubble frequencies showed good correlation witi the predicted values, j
2.
Reaction loads - The measured water and air clearing reaction loads on the SRV discharge line and T-Quencher supports were about 20% of the analytically predicted values and test conditions.
3.
Strains - The sneasured strains on the torus shell, torus su) port structures, internal structures, and piping corpared favora)1y with the predictedvalues(about10f-40%oftheanalyticallypredictedvalues),
i 4
Zero Period Acceleration (ZPA) - The peak sneasured zero period acceleratiori is well below the analytically predicted response at each location.
In addition, the sneasured coupled system acceleration (torus shell/ piping systems) are less than 50% of the analyzed clean shell resper,se.
4.0 C.ONCLUSIONS Detroit Edison has satisfitd the license condition of perfoming an in-plant test for confirmitig the SRV discherge loads in accordance with the quidelines
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l provid(d in !!UREG-0E01 and NUREC-0763.
Ensed on the 'icetisee's set.mitve'is, the W staff concludes th6t the Fermi-S plent d'41 gun enaly:1s losids are conservative ard the safety tr.argin in the y
design of the primary containmer.t system fcr SRV tiischarge loads is edequate.
b Nted: January.18, 1990 Principal Gntributor:
C. Hernmer l
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