ML20245H538
| ML20245H538 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 10/31/1988 |
| From: | Ellis G, Johnson E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19297H470 | List: |
| References | |
| WCAP-12029, NUDOCS 8902280468 | |
| Download: ML20245H538 (7) | |
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. WESTINGHOUSE CLASS 3 WCAP-12029 TROJAN NUCLEAR PLANT REACTOR VESSR VERTICAL SUPPORT LOADS OCTOBER 1986 AUTHOR:
E. R. JOHNSON b'
APPROVED:
p G. R. ELLIS, MANAGER STRUCTURAL ENGINEERING AND PIPING TECHNOLOGY WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR AND ADVANCED TECHNOLOGY DIVISION P.O. BOX 355 PITTSBURGH, PENNSYLVANIA 15230 f'
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8902280468 890223 E'
DR ADOCK O 4
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i ABSTRACT
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As a result of a review of the design of the Trojan Nuclear Plant reactor pressure vessel supports, the NRC has requested certain information on the design basis loadings and analysis methodology used by Westinghouse. This request is in the form of three questions, the answers to which are provided hereirs.
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I TROJAN NUCLEAR PLANT REACTOR VESSEL VERTICAL SUPPORT LOADS Question 1 What is the basis' for the auxiliary line break areas for accumu1ator and
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surgeline breaks?
&nswer 1 i
Surge 11ne is 14-inch, Schedule 160 pipe, and Accumulator is 10-inch, Schedule 140 pipe. The flow areas are 98 square inches and 60 square inches, respectively. The gagigated vertical support loads for these breaks i
are [
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, respectively. The smaller accumulator break is expected to produce higher vertical loads since the decompression j
of the annulus region between the core barrel and reactor vessel is asymmetric for this break which is located on the cold leg. This asymmetric decompression causes lateral loads on the reactor vessel system which leads to rocking motion and results in vertical support loads.
i Question 2 Provide a description of the dynamic system model used to determine reactor vessel support loads, including all the applied loadings.
Answer 2 Following a postulated pipe rupture at the reactor vessel nozzle, the reactor vessel is excited by time-history foroes. These foroes are the combined effects of three phenomena:
(1) reactor internal hydraulic forces, (2) reactor cavity pressurization forces, and (3) reactor coolant loop mechanical loads.
Depressurization waves propagate from the postulated break location into the reactor vessel through either a hot leg or a cold leg nozzle.
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For a postulated break in the cold leg, the depressurization path for waves entering the reactor vessel is through the nozzle which contains the broken pipe and into the region between the core barrel and reactor vessel. This region is called the downcomer annulus. The initial waves propagate up, around, and down the downcomer annulus, then up through the region circumferentially enclosed by the core barrel; that is, the fuel region. In the case of a postulated break in the hot leg, the waves follow a dissimilar depressurization path, passing through the outlet nozzle and directly into the upper internals region, depressurizing the core, and entering the downcomer annulus from the bottom exit of the core barrel. Thus, for a hot leg break, the downcomer annulus is depressurized with very little difference in pressure access the outside diameter of the core barrel. For breaks in both the hot leg and cold leg, the depressurization waves continue to propagate by reflection and translation through the reactor vessel loops.
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Time-history values of the pressure, mass velocity, density, and other thermodynamic properties within the RPV (all of which are computed by the MULTIFLEX code), are utilized in the determination of the applied vertical and lateral loads on the reactor vessel internals.
In the case of a cold leg break, the region of the downcomer annulus close to the break depressurizes rapidly but, because of restricted flow areas and finite wave speed (approximately 3000 feet per second), the opposite side of the core barrel remains at a high pressure. This results in a net horizontal force on the core barrel and RPV. As the depressurization wave propagates around the downcomer annulus and up through the core, the barrel differential pressure reduces, and similarly, the resulting hydraulic forces drop. A hot leg break produces less horizontal force because the depressurization wave travels directly to the inside of the core barrel (so that the downcomer annulus is not directly involved) and internal differential pressures are not as large as for the cold leg break. Since the differential pressure is less for a hot leg break, the horizontal force applied to the core barrel is less for the hot leg break than for a cold leg break.
The applied hydraulic blowdown loads on the reactor vessel internals are caused by the propagation of acoustic decompression waves into the vessel region. In a postulated loss-of-coolant accident, these waves originate at the break in a primary coolant loop. The MULTIFLEX computer code calculates the hydraulic transients within the entire primary coolant system; it considers subcooled, transition, and two-phase (saturated) blowdown regimes. The MULTIFLEX program employs the method of characteristics to solve the conservation laws, and assumes one-dimensionality of flow and homogeneity of the liquid-vapor mixture.
For pipe breaks located outside the primary shield wall (such as reactor coolant pump outlet, accumulator line or surgeline breaks) there are no loads caused by reactor cavity pressurization.
The reactor coolant loop mechanical loads are calculated based on a fixed reactor vessel. For the guillotine type breaks in the primary loop (such as the reactor coolant pump outlet) these loads are equal and opposite to the normal operating loads at the break location which are released at the time of the break. These loads are applied as step loads on the reactor vessel l
dynamic model. For breaks in attached branch piping (such as the i
surgeline), these are the combination of the thrust load from the reservoir on the loop side of the break location and the jet impingement load on the l
loop from the reservoir on the branch piping side of the break location.
These loads are applied as step loads on the reactor vessel dynamic model.
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l Question 3 1
Provide information on the frequency response at the reactor vessel vertical' supports, which can be used to assess strain rate effects.
Answer 3 The seismic response of the reactor vessel systems is determined. from a The dominant frequencies for the vertical linear response spectra analgsgs; for the lower vertical mode and [
supportigags,are[
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]
for the lowest horizontal / rocking modes. The time history characteristics for the LOCA loadings can be estimated from the attached figure.
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Page 4 of 5 t.
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9 a,c,e M?oo c.
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B
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c time (seconds)
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Trojan Plant Double Ended Pump Outlet Nozzle Break i
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