ML20031B959
| ML20031B959 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/30/1981 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8110060275 | |
| Download: ML20031B959 (7) | |
Text
_,
Malling Addross Alabama Power Company
=6 600 North 18th Street Post Office Box 2641 Birmingham. Alaaama 35291 Telephone 205 783-6081 F. L. Clayton, Jr.
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Alabama Power
!!e soutivn ehmtre sg!cm September 30, 1981 4
Docket No. 50-364-g.
C Director, Nuclear Reactor Regulation OCT 6 198 ll. S. Nuclear Regulatory Comission T, Q u,ouw,J5,
.6 Washington, D. C.
20555 t
O Attention: Mr. S. A. Varga 9
l yI 7
Dear Sir:
Joseph M. Farley Nuclear Plant - Unit 2 NPF-8 License Condition 2.C(21)(d)(1)
In accordance with the requirements of the Farley Nuclear Plant -
Unit 2 Operating License NPF-8 [Section 2.C(21)(d)(1)], Alabama Power Company is participating in the EPRI PWR Safety and Relief Valve Test Program. This program is designed to test Power Operated Relief Valves (PORVs) and Safety Valves which are representative of the valves currently being used or planned to be used on Pressurized Water Reactor (PWR) facilities in the United States. The valves are being tested under steam flow, liquid flow and steam to liquid transition flow conditions representative of the expected conditions resulting from design basis transients and accidents. The test program will also provide piping load data to be used to verify piping load calculational models.
1 All required testing of the ten relief valves which are part of the overall EPRI PWR Valve Program has been completed. During all tests, the l
valves opened and closed on demand.
Following the testing, the valves l
were disassembled and inspected. No damage was observed that would affect future valve performance. The PORV design used on Farley Nuclear Plant -
Unit 2 was included in the EPRI program, thus its qualification has been demonstrated. The next revision of the EPRI PWR, Safety and Relief Valve Program Interim Data Report will include all approoriate relief valve data obtained. Current schedules call for submittal of this revision to NRC on December I, 1981.
The testing of the safety valves i: still underway with the Crosby safety valva (similar to the valve used on farley Nuclear Plant - Unit 2) scheduled to be tested in the next few weeks.
It is expected that all safety valve tests will be completed in early 1982.
Test data demonstrating the qualification of the Farley Nuclear Plant - Unit 2 safety valve 4111 be provided to the NRC when available.
00 0
Director, Nuclear Reactor Regulation September 30, 1981 U. S. Nuclear Regulatory Commission Page 2 To comply with the license requirement to demonstrate the qualifi-cation of the associated piping, Westinghouse has performed a Thermal Hydraulic Analysis for the Farley Nuclear Plant - Unit 2.
A copy of this analysis is provided as Attache:ent 1.
Due to the time constraints, this analysis was performed using preliminary piping data obtained from the testing of the small Crosby safety valve (3K6) to verify the piping code used to analyze the Farley Nuclear Plant - Unit 2 discharge piping. The initial comparisons show good agreement between calculations and tests.
The loadings calculated and used in the Farley Nuclear Plant - Unit 2 analysis are thus consistent with test results received to date. A complete thermal hydraulic analysis will be performed when the final test data from the testing of the Crosby 6M6 safety valve (representative of the Farley Safety Valve) is available.
If you have any questions, please advise.
Yours very tr h O
f F. L. Clayton, Jr.
FLCJr/ JAR:de Attachment cc: Mr. R. A. Thomas (vdattachment)
Mr. G. F. Trowbridge (w/ attachment)
Mr. J. P. O'Reilly (w/ attachment)
Mr. E. A. Reeves (w/ attachment)
Mr. W. H. Brad #]rd (w/ attachment) f f
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Description of the Themal Rydraulic Analysis performed _for the Joseph M. Farley Unit II Nuclear Power Plant When the pressurizer pressure reaches the set pressure (2500 psis f0F safety valva and 2350 psia for relief valve) and the valve opens, the high-pressure steam in the pressuriar forces the water seal through the valve and down the pipingssystent to"the pressurizerfelief tank.
For each pressurizer safety and relief piping system, an analytical hydraulic inodel was developed to represent the condition descrfbed above. The piping from the pressurizer nozzle to the relief tank nozzle was modeled including the pressurizer which was rudeled as a reservoir containing steam at constant temperature and pressure. The relief tank was modeled as a sink which contains a steam and water
- mixture, A Westinghouse proprietory computer progra": Was used -to pufom the.
transient-hydraulic analysis of the system. All three safe-ty valves were assuned to cpen simultaneotsly while the relief valves remain closed Similarly, two relief valves opened simultaneously while the safety valves were closed. These cases umbrella all pemutations that may possibly control the design of the system.
l Fluid acceleration inside the pipe generates reaction forces on all
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segments of the line which are bounded at either end by an elbow or band. Reaction forcer, resulting from fluid pressure and twentum varia-tions were calculated. These forces were then expressed in tems of the fluid properties available from the transient !tydraulic analysis.
Unbalanced forces were calculated for each straight segment of pipe from the pressurizer to the relief tank. The time histories of sese forces were then stored on tape to be used for the subsequent structural analysis of the pressurizer safety and relief lines.
A set of as built isometric drawings were received and comprad with as designed infomation.
It was concluded that dimensions of piping segments and support locations were consistent.
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l Comparison.of Thermal flydraulic Calculations with available EPRI test data _
A Westinghouse proprietary computer program that yields similar results as the program used in the Joseph M. Farley Unit II thennal hydraulic analysis was used to model the EPRI test arrangement. The particular
~
series of tests of interest included the Crosby 3K6 spring-loaded safety valve which-hae arr ASME rated flow. of 213.000 Lbm/hr Even though a 6M6 Crosby safety valve having an ASE rated flow of 420,000 lbm/hr is installed on the Joseph M. Farley Unit II plant, the comparison of calculated versus test data for the 3K6 Crosby Valve will illustrate the capab_ility of the thermal hydraulic program.
The following stat.ements can be cade based on the comparative evaluation of the preliminary and limited amounts of data' received from EPRI (ie. data for water slug discharge tests 512 and 526):
(i) The initial inertia pressure peak just downstream of the valve PT08, see the test instrumentation diagram - Figunt 1) for both EpRI cases was 150-180 psi. Thennal h'ydraulic calculations i._
show approximately 170 psi.
1 (2) Testss512 and 526 indicate peak forces of 13 and 24 kips respectively for the first verticalirun downstream of the valve (WE 32/33), Calculations predict 13 kips. Figures 2 and 3 illustrate the predicted time history forcing functions and the force results of test 512 respectively.
(3) Thermal hydraulic calculations show a peak force for the second horizontal run downstream of the valve of 30 kips. Both tests show roughly 6 kips at location WE 30/31. This intended con-servatism can ba attributed to the analytical methods used to assure support design adequacy in the discharge piping.
(4)
The loads calculated are the loads acting on cach straight run of pipe whereas the tesrt loads-are theeloads transmitted thru the piping arrangement to the sensors. Differences between the calculations and test forces can be attributed to this.
Initial comparisons show good agreement between calculations and tests.
The loadings calculated and used in the J. M. Farley Unit 2 analysis are thus consistent with test results received to date.
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