ML20237L306
| ML20237L306 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/30/1987 |
| From: | Doug Garner, Kachmar M, Wengerd M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19292H668 | List: |
| References | |
| WCAP-11523, NUDOCS 8708200167 | |
| Download: ML20237L306 (17) | |
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IEESTDGE00SE CIAS8 3 IOCAP-11523 RESPCREE 'IO NRC QUESTIGE N HE IDCA BERADLIC PORCES ANALYSIS OF BEAVER VALIEY ICIElR SD&ICH, WIT 2 Prepared for NBC Review in Ceojunction with Docket Number 50-412 JUNE 1987 bY D. C. Garner M. P. Fa h r M. R.15engard WESTDKHXEE EIECTRIC 00RIORATICE NOCIEAR ENERGY SYSTEMS P. O. BOX 355 PITISBURCH, PENNSYLVANIA 15230 8708200167 870017 PDR ADOCK 05000412 p PDR
1 WESTINGHOUSE CLASS 3 8%g * . -
This report is submitted by the Westinghouse Electric Corporation on ,
jf .
behalf of Duquesne Light Company's Beaver Valley Power Station, Unit 2 g -
(BVPS-2). Review of this report by the Nuclear Regulatory Commission '
(NRC) is requested in conjunction with the submittal of WCAP-9735 and WCAP-11004 on the BVPS-2 Docket, Number 50-412. The intent of this transmittal is to answer the questions of the NRC with respect to their review of the Multiflex 3.0 code and its application to the LOCA hydraulic forces analysis of BVPS-2.
The questions addressed are those which were discussed in a telephone conversation in which representatives from the NRC, Duquesne Light Company, and Westinghouse Electric Corporation participated.
The specific objectives of this report are: \
- 1) To provide the reasoning behind the use of the Multiflex 3.0 code, as opposed to the Multiflex 1.0 code, in the LOCA hydraulic forces analysis of BVPS-2. An explanation of the postulation of the simultaneous breaks at the hot and cold leg connections of the 8" bypass line and the refinements incorporated into the Multiflex 3.0 version of the. code will be given,
- 2) To identify the limiting line break at BVPS-2 with respect to LOCA hydraulic forces based upon the final break locations detailed by " '
Section 3.6 of the BVPS-2 FSAR,
- 3) To compare the results of the Multiflex 3.0 BVPS-2 results (WCAP-8784 Addendum 1) with the results of analyses of other 3 loop ,-
plants analyzed with the Multiflex 1.0 code. These comparisons will provide a basis for the NRC to apply engine.ering judgement in an effort to reach the conclusion that the BVPS-2 core geometry can be maintained in a "coolable" configuration under faulted loading conditions.
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WESTINGHOUSE CLASS 3 packaround The Multiflex 3.0 computer code (reference 1) was used in the analysis of the Beaver Valley Unit 2 nuclear plant. At the time of the analysis it was judged that the code enhancements of the 3.0 version would provide lower structural loads for a number of postulated break locations. This appeared to be particularly desirable for an analysis of a postulated simultaneous break of a reactor coolant system (RCS) eight inch bypass line. This break, postulated to occur on the basis that the support / restraint system for this line was not designed for pipe rupture loads, assumed that a rupture of the bypass line would result in the simultaneous guillotine break of the line attached to the RCS loop piping at both the hot and cold leg connections. The effect of such a break on the magnitude of the LOCA hydraulic forcing functions had not been previously established and the use of the advanced beam model (Multiflex 3.0), which was known to yield better results, was considered prudent.
Multiflex 1.0 versus Multiflex 3.0 The Multiflex computer program is an engineering design tool that is used to analyze the coupled fluid-structure interactions in a pWR system and to compute the transient hydraulic forcing functions on the reactor pressure vessel internals. The thermal hydraulic portion of the Multiflex code is based upon a one dimensional homogenous model which is expressed in a set of mass, momentum, and energy conservation equations. These equations are quasi-linear first order partial differential equations solved by the method of characteristics.
Multiflex considers the interaction of the fluid and structure simultaneously, whereby the mechanical equation of vibrations is solved by the use of the modal analysis technique. A detailed description of tre Multiflex 1.0 code is provided in reference 2. The i
development of the Mel+.iflex 3.0 code was undertaken to improve the consistency between the Multiflex code and the structural programs and 2
WESTINGHOUSE CLASS 3
.5 to further refine the original beam Multiflex model. The advanced Multiflex 3.0 program /model contained the important features detailed below; a complete description of which can be found in reference 1.
The structural non-linearity in a fluid-structure interaction can be taken into account by the pseudo-force method which solves the structural dynamic equation with the non-linear term incorporated in the external force term. Time dependent modal analysis has been ~~ ---~
demonstrated to be capable of solving a structural non-linear boundary problem by the use of the pseudo force method (also known as modal superposition). These techniques have been incorporated into the Multiflex 3.0 advanced beam model so as to include linearized boundary ----
conditions in the modal analysis, and the difference between the non-linear and the linearized boundary conditions in the pseudo-force term. Thus, an intermediate pseudo-force method has been developed which considers the non-linearities in the pressure vessel / core barrel structure.
The fluid-structure interactions in the downcomer annulus region are effective only when the core barrel displaces or deforms relative to the pressure vessel. The original Multiflex 1.0 beam model was ..
developed based upon conventional modal analysis for the absolute displacements, under the assumption that the vessel is fixed. A relative modal analysis technique has been incorporated into the Multiflex 3.0 advanced beam model which takes into account vessel motion. Consequently, it is found that the formulation of the conventional nodal analysis can be utilized without any changes, but the structural input data must be computed by using relative modal analysis. The downcomer annulus region has been represented by a one-dimensional piping network that is equivalent to two-dimensional fluid-structure interactions. The basic nodalization rules used in formulating the network downcomer model are provided in references 3 [
and 4. These rules have been applied to develop a network model of the uniform annulus of the Fritz-Kiss Shaker Experimental Facility.
The computed in-water frequency agrees very well with the 3
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WESTINGHOUSE CLASS 3 experimentally measured value, thus verifying the method of network formation. The utilization of a network downcomer model in the Multiflex 3.0 hydraulic forces calculation, produces a more realistic representation of the decompression waves propagating throughout the downcomer annulus region.
In order to make the Multiflex computer program applicable to a complex thermal-hydraulic-mechanical system, large numbers of hydraulic legs (regarded as pipes of arbitrary length and flow area) and flexible walls are required. In transitioning from the original Multiflex 1.0 beam model to the advanced beam model (3.0), the only hydraulic piping model which had to be modified was that of the . _ .
downcomer annulus region; the hydraulic models simulating other portions of the reactor coolant system remain unchanged. Figure 1 shows the network of legs which represent the downcomer annulus in the advanced beam model (network downcomer) of a three loop neutron pad plant. The corresponding three loop downcomer annulus hydraulic model (known as a conventional downcomer) utilized with the original beam model is provided in Figure 2. A comparison of the two downcomer annulus representations leads to the following observations:
"" ~'(a,c)
The structural surfaces interfacing with the fluid in the downcomer annulus region are the outer surfaces of the neutron pads and core
_ barrel, and the inner surface of the reactor vessel. The fluid -
j flexible walls in the advanced Multiflex model. The interface (a,c) modeling between the flexible walls and the hydraulic pipes in the original beam model is as given in Figure 4.
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WESTINGHOUSE CLASS 3 Identification of Limitina Break Traditionally, for LOCA hydraulic forces considerations, breaks located between the reactor pressure vessel inlet (RPVI) nozzle and the reactor coolant pump (RCP) discharge have been the locations of the most limiting breaks. This observation holds true for BVPS-2 with respect to hydraulic forces produced from a postulated LOCA.
In Section 3.6 of the BVPS-2 FSAR the use of Leak-Before-Break concepts, along with the application of the WHIPJET program, has (
demonstrated that " fluid leakage from a postulated defect at the highest stress location concurrent with minimum materials properties in a high energy line can be detected well before the rupture of the pipe." From the WHIPJET program it was demonstrated that fluid leakage could be detected before a pipe rupture in the RCS piping and RCS branch line piping with a nominal pipe diameter of six (6) inches or greater. Therefore, the next most limiting break for LOCA hydraulic forces considerations at BVPS-2 will be determined as a break of a line with a diameter of less than six inches located somewhere between the RPVI nozzle and the reactor coolant pump discharge. At BVPS-2, the next biggest line size connected to the RCS has a nominal diameter of four (4)~ inches.
However, it is known that the severity of the peak LOCA hydraulic forces increase as the postulated break location approaches the inlet nozzle. Therefore, a single line break with a four inch diameter (12.6 in 2 break area) anywhere al'ong the cold leg would be less limiting than a postulated 12.6 in 2
break of the RPVI nozzle.
At SVPS-2 there is also the possibility of a pipe break propagation scenario where the rupture of the pressurizer spray line could interact with and cause a rupture of the PORV line. The pressurizer spray line is a Schedule 160, four inch diameter pipe connected to the cold leg of the reactor coolant system (RCS). This pipe is a liquid carrying line. The PORV line is a Schedule 160, three inch diameter 5
s WESTINGHOUSE CLASS 3 pipe which is a vapor line connected to the pressurizer. In turn, the pressurizer is connected to the hot leg. A simultaneous break of both of these lines (pressurizer spray and PORV) is similar to a break of the bypass line, connected between the hot and cold leg of the RCS, as ,
previously analyzed for BVPS-2. The total break area of the .
2 pressurizer spray line and the PORV line is 14.7 in ,
The major difference between the bypass line analysis and the scenario proposed above is the fluid conditions of the line connecting to the hot leg side of the RCS. The fluid condition of the bypans line connection at the hot leg is liquid, while the fluid cond:Ltion et the (
PORV line is vapor. From the standpoint of LOCA hydraulic fbrces, a break of a liquid filled line would produce loads on the reactor vessel and internals that would be more severe than an identical break 2
of a vapor filled line. On this basis, a postulated 14.7 in break ..
of a RPVI nozzle or a postulated 14.7 in2 break of the bypass line is more limiting than the forces that would be produced from the proposed simultaneous break of a pressurizer spray line and a PORV line with a total break area of 14.7 in 2, ,
Thus, based upon the information in FSAR Section 3.6 and FSAR Figure 3.6N-1 (Figure 5 of this report), application of Leak-Before-Break concepts and the WHIPJET program, and the relative '
location in the cold leg the limiting break can be determined. The limiting break for LOCA hydraulic forces considerations at BVPS-2 with [
respect to total peak vessel and barrel horizontal forces, is a postulated 14.7 in 2 area break of the reactor pressure vessel inlet '
nozzle. This break is given as location 1 in Figure 5.
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WESTINGHOUSE CLASS 3 1 DMW LOCA Loads Evaluation The analysis described in reference 5 on Beaver Valley Unit 2 reported four break locations in the evaluation of,LOCA forces. As an indicator of the relative loads involved, Table 1 lists the total t
herizontal load on the reactor core barrel for each break. Analysis of these loads with Multiflex 1.0 would have resulted in a 30%
increase in loads as indicated in Figure 6. In order to demonstrate that an analysis performed with Multiflex 1.0 and utilizing " Leak Before Break" criteria would produce acceptable results, it will be shown that the currently estimated limiting load is substantially less than the Table 1 loads addressed previously .
Application of the " Leak Before Break" concepts to Beaver Valley Unit 2 results in the elimination of breaks in the main reactor coolant This eliminates all breaks given in Table 1, i .e., the piping. '
reactor vessel inlet nozzle, steam generator inlet nozzle, steam -
generator outlet nozzle, and bypass line. The limiting break now occurs in a 4 inch line with a break area of 12.6 sq. in in a .
location which is postulated to cause a simultaneous break in the pressurizer spray line, resulting in a total break area of 14.7 square inches. For a given break size, the break location known to produce the maximum barrel horizontal load is at the reactor vessel inlet nozzle, i.e., the barrel loads for a 14.7 sq.in. break of the inlet nozzle bound all other 14.7 sq.in. breaks with respect to barrel .
loads. This is shown explicitly for the bypass line break location in Figure 6. Thus the bounding break is now postulated to be a 14.7 sq.
in. break of the reactor pressure vessel inlet nozzle. Using the correlation in Figure 6, it is shown that if a Multiflex 1.0 analysis wereperformed,thepeakbarrelporzontalloadwouldbeestimatedto beapproximately( ]fo'ra15sq. in. inlet nozzle break.
As indicated in Table 1, the previous analysis addressed a range of (a.c) barrelloadsuptothesteamgeneratoroutletbreakloadof[ ]
lbs., well above the current estimate, similarly, the vertical I
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WESTINGHOUSE CLASS 3 internals loads and the vessel loads calculated with Multiflex 1.0 for a RPVIN break of 80 sq.in. or less can be shown to be smaller than those calculated with Multiflex 3.0 for the 150 sq.in. RPVIN break.
Thus, the effect of employing " Leak Before Break" concepts on the ,
location and size of the limiting break more than compensates for the 30% increase in loads associated with using Multiflex 1.0 instead of Multiflex 3.0.
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WESTINGHOUSE CLASS 3 References
- 1) Takeuchi, K., et al., "MULTIFLEX, A Fortran-IV Computer Program for _
Analyzing Thermal-Hydraulic-Structure System Dynamics (III) ---
Advanced Beam Model," WCAP-9735 Rev.1 (Proprietary) and WCAP-9736 (Non-Proprietary).
- 2) Takeuchi, K., et al., "MULTIFLEX, A Fortran-IV Computer Progrtim for Analyzing Thermal-Hydraulic-Structure System Dynamics," WCAP-8708-P/A (Proprietary) and WCAP-8709-A (Non-Proprietary), September, 1977.
- 3) Takeuchi, K., "One Dimensional Network for Multi-Dimensional Pressure Wave Propagation with Hydro-Structural Interactions," Trans.
Am. Nuc. Soc., 30, 210 (1978).
- 4) Takeuchi, K., "One Dimensional Network for Multi-Dimensional Fluid Structural Interactions, " Nuc1 Sci, and Enc., 71, 322 (1979).
- 5) Jenkins, H. E., " Dynamic Analysis of Reactor Pressure Vessel for Postulated Loss-of-Coolant Accidents: Beaver Valley Power Station Unit 2," WCAP 8784, Addendum - 1, July 1982.
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WESTINGHOUSE CLASS 3 Table 1. Horizontal LOCA Loads on Reactor Core Barrel. Ref. 5 Break Description Peak Horizontal Load 150 Square Inch Reactor Vessel Inlet Nozzle Break F 5.24 Square Foot Steam Generator Outlet Nozzle Break 5.24 Square Foot Steam Generator Inlet Nozzle Break 8 Inch Bypass Line Break With , ,,
0.253 Square Foot Area (a,c)
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