ML18144A006
| ML18144A006 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/31/1985 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Varga S Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 85-615A, TAC-44622, TAC-44623, TAC-44633, NUDOCS 8511050123 | |
| Download: ML18144A006 (32) | |
Text
1..
\\.
e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIROINIA 23261 W. L. STEWART VICE PRESIDENT NUCLEAR OPERATIONS October 31, 1985 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. Steven A. Varga, Chief Operating Reactors Branch No. 1 Division of Licensing U. S.
Nuclear Regulatory Commission Washington, D.
C.
20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 ADDITIONAL INFORMATION RELATED TO NUREG-0737, ITEM II.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES Serial No.
E&C/TLG:acm Docket Nos.
License Nos.
85-615A 50-280 50-281 DPR-32 DRP-37 In our letter dated September 20, 1985, Serial No.85-615, we stated we would forward to you the requested information related to the above subject by October 31, 1985.
We noted that the information requested required input from our architect-engineer (Stone & Webster), our vendor (Westinghouse) and ourselves.
We were to review and approve the information and forward it to you.
In reviewing the information supplied by the architect-engineer and the vendor, it has become apparent that additional information and review are necessary to supply a complete response to your request.
A partial submittal (safety valve information only) is enclosed which includes the information for request numbers 2, 4, 5, 6b and 7.
The remaining information will be submitted to you no later than February 28, 1986. The need for additional time was discussed with the NRC Surry Project Manager on October 10, 1985.
Very truly yours, K2) t\\ a,J..;J ~
~ W. L. Stewart Enclosure
/----~-
' ~g1 !050t23es10~-1 -----,,
p R ADOCK 05000280 1
e Vrnou;,IA ELECTRIC AND PoWER COMPANY TO cc:
Dr. J. Nelson Grace Regional Administrator NRC Region II Mr. D. J. Burke NRC Resident Inspector Surry Power Station e
Mr. Harold R. Denton
e VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNIT NOS. 1 AND 2 ADDITIONAL INFORMATION RELATED TO NUREG-0737, ITEM II.D.I PERFORMANCE TESTING OF RELIEF AND SAFETY VALVES ENCLOSURE
e Question
- 2.
On page 4 of Ref~rence 3, it was stated that safety valve blowdown in excess of 5% had been analyzed in conjunction with the Westinghouse Owners Group program on safety valves and ~he results from these analyses showed no adverse effect on plant safety. Reference 3 also indicated that a discussion of the analyses was presented in the Westinghouse Report WCAP-10105 (Reference 4).
However, we were unable to find the aforesaid information in the Westinghouse report. Please provide documentati.on which demonstrates that the increased blowdown will not affect the safe operation of the reactor cooling system.
Response
The impact on plant safety of pressurizer safety valve blowdown in excess of 5% was evaluated for Surry Units 1 and 2.
The re$ults of this evaluation showed no adverse effects on plant safety.
Safety valve blowdowns in excess of that assumed in the Surry FSAR will have the following effects on the events in which safety valve actuation occurs:
- 1.
Increased pressurizer water level during and following the valve
- blowdown,
- 2.
Lower pressurizer pressure during and followi~g valve blowdown,
- 3.
Increased inventory loss through the valve.
e e
The impact of the increased relief valve blowdowns with respect to the above effects was evaluated for the Surry FSAR events in which safety valve actuation occurs, (i.e., Loss of External Electrical Load and Locked Rotor).
For the Loss of External Electrical Load event, results from sensitivity analyses performed for a 4-loop plant were used for the evaluation. These analyses investigated the effects of different blowdown rates on the event.
Similar results are expected for a 3-loop *plant.
The results of these analyses showed only marginal increases in pressurizer water volume and the maximum pressurizer water levels were well below the level at which liquid relief would occur.
The Surry FSAR analysis results show th/it a small -
increase in pressurizer water volume, due to increased safety valve blowdown, will not result in liquid relief.
The sensitivity analyses also showed that peak RCS pressures were unaffected by the increased blowdowns.
The increased blowdowns did result in lower pressurizer pressure and increased RCS inventory loss, however, these had no adverse impact on the event and adequate decay heat removal was maintained.
These conclusions can also be applied to the Loss of Normal Feedwater/Station Blackout results.
For the Locked ~-otor event, increased safety valve blowdowns have little impact on the event.
As analyzed and presented in the Surry FSAR, the opening and closing of the safety valve occurs over a short time period (3 seconds).
As a resu1t, there is little change in either pressurizer level or RCS inventory.
Increased safety valve blowdowns would have no impact on peak pressure, peak clad temperature, or DNBR as these occur prior to the closing of the safety valve.
. Question:
e
- 4.
The magnitude of the maximum bending moment induced at the inlet flanges of the safety valves and PORVs during valve discharge were not provided in the* Surry transmittal.
Give the predicted values of the maximum moments for these valves so that the valve operability can be evaluated.
Response
Tables IA, lB, 2A, and 2B provide the calculated moments at the safety relief valve and power operated relief valve inlets due to the safety valve
- discharge loading condition. -
Type of_
Loading Condition e
SURRY UNIT 1 TABLE lA e
Safety Valve Inlet Loads Due to SRV Discharge Valve Number SV/1551C sv /255 lB SV /1551A Loads (Ft-Lbs)
Loads (Ft-Lbs)
Loads (Ft-Lbs)
Mt Mb1 Mb2 Mt Mb1 Mb2 Mt Mb1 Thermal*.':
1371 1043
-3951 780 804
-2892
-2245
-6360 Dead Load
-93 186
-530
-161 295
-907 44
-230 OBE (I) 69 131 15 87 139 54 155 41 DBE (I) 120 236 25 142 296 88 246 65 OBE (A) 150 14 35 103 13 35 78 45 DBE (A) 341 38 95 246 35 93 197 126 SRV Discharge 7094 1080 3251 9177 3252 2428 58 5166 Where OBE (I)
DBE (I)
OBE (A)
DBE (A)
Mt Mb
=
=
=
=
=
=
=
Operational Basis Earthquake Inertia Design Basis Earthquake Inertia Operational Basis Earthquake Anchor Movement Design Basis Earthquake Anchor ~ovement Torsional Moment Bending !foment
(
Reference:
Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note:
Dynamic loads such as OBE(I), DBE(!), OBE(A), DBE(A) and SRV Dischaige are considered as+/-. Mb2
-1293
-147 36 67 30 78 2305
- 7)
e SURRY UNIT 1 TABLE lB PORV Inlet Loads Due to SRV Discharge
_Type of Valve Number Loading Condition PCV/1455 PCV/1456 Loads (Ft-Lbs)
Loads (Ft-Lbs)
Mt Mb1 Mb2 Mt Mb1 Mb2 Thermal~~
-720 2022 90
-1318 1312
-679 Dead Load 31
-10
-211 42
-7
-224 OBE (I) 85 112 114 103 96 99 DBE (I) 137 224 221 194 184 196 OBE (A) 1 25 2
1 18 1
DBE (A) 2 60 5
3 43 3
SRV Discharge 6
28 0
6 47 4
Where OBE (I) =
DBE (I) =
OBE (A) =
DBE (A) =
Mt
=
Mb
=
(
Reference:
Operational Basis Earthquake Inertia Design Basis Earthquake Inertia
- 7)
Opera_tional Basis Earthquake Anchor Movement Design Basis Earthquake Anchor Movement Torsional Moment Bending Moment Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note:
Dynamic loads such as OBE(I), DBE(I), OBE(A),_DBE(A) and SRV Discha~are considered as+/-.
e SURRY UNIT 2 TABLE 2A Safety Valve Inlet Loads Due to SRV Discharge Type of-Valve Number
- i,oading SV/2551C Condition SV/2551A SV/2551B Loads (Ft-Lbs)
Loads (Ft-Lbs)
Loads (Ft-Lbs)
Mt Mb1 Mb2 Mt Mb1 Mb2 Mt Mb1 Thermalir
-1904 3902
-3114
-2002 2322 4177 2799
-6849 Dead Load 9
-106 48 1
-343
-724
-53
-292 OBE (I) 102 80 57 114 150 88 78 82 DBE (I).
180 155 119 230 312 183 154 165 OBE (A) 88 48 41 209 80 87 68 44 DBE (A) 186 129 103 469 221 218 84 59 SRV Discharge 4268 4415 6451 9195 7969 8657 2189 9229 Where OBE (I)
DBE (I)
OBE (A)
DBE (A)
Mt Mb
-:r
=
=
=
=
=
=
=
Operational Basis Earthquake Inertia Design Basis Earthquake Inertia Operational Basis Earthquake Anchor Movement Design Basis Earthquake Anchor Movement Torsional Moment Bending Moment
(
Reference:
Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note:
Dynamic loads such as OBE(I), DBE(I), OBE(A), DBE(A) and SRV Discharge are considered as+/-. ~2 245i
-108 101 201 8
10 5889
- 8)
Type of
- Loading Condition Thermal*
Dead Load OBE (I)
DBE (I)*
OBE (A)
DBE (A)
SURRY UNIT 2 TABLE 2B e
PORV Inlet Loads Due to SRV Discharge Valve Number PCV/2455 PCV/2456 Loads (Ft-Lbs)
Loads (Ft-Lbs)
Mt Mb1 Mbz Ht Mb1 Mb2
-711 1120 407
-972
-264
-1096
-77
-1 227
-61
.o 104 70 46 279 54 38 206 143 93 589 111 76 436 0
1 0
- 0.
0 0
0 1
0 1
2 0
SRV Discharge 8
17 5
2 15 7
Where*
OBE (I)
DBE (I)
OBE (A)
DBE (A)
Ht Mb
=
=
=
=
=
=
=
(
Reference:
Operational Basis Earthquake Inertia Design Basis Earthquake Inertia
- 8)
Operational Basis Earthquake Anchor Movement Design Basis Earthquake Anchor Movement Torsional Moment Bending Moment Maximum Thermal Load Component with Sign from all Thermal Conditions Analyzed Note:
Dynamic load_s such as OBE(I), DBE(!), OBE(A), DBE(A) and SRV Dischar_ge are considered as+/-.
e e
Question
- 5.
NUREG-0737, Item II.D.1 requires the qualification of the control circuits of the plant specific PORVs for design-basis transients and accidents.
However, such qualification does not have to be submitted for this review, if it has already been included in the submittal to fulfill the requirements of 10CFRS0.49.
Verify whether the in-plant PORV control circuits have been included in the 10CFRS0.49 review or provide the necessary evaluation to demonstrate that the requirement of NUREG-0737, Item II.D.l concerning control circuitry has been met, if the PORV circuitry was not reviewed under 10CFRS0.49.
Response
. The in-plant PORV control circuits have been reviewed and included in the 10CFRS0.49 and the Regulatory Guide 1.97 reviews.
Included in these.
reviews and qualified are the MOVs, SOVs, limit switches, associated cables and electrical interfaces. A Regulatory Guide 1.97 implementation schedule has been submitted which includes work which will provide redundant limit switches and safety valve position indication consistant with the Regulatory Guide. -
! I'
~
e e
Question:
- 6.
The thermal-hydraulic analysis of the safety/PORV Inlet and discharge
_piping was discussed on pages 15 and 16 of Reference 3.
In the discussions only information concerning safety valve discharge conditions was provided.
addressed.
The PORV discharge conditions were not
- a.
Provide a discussion of the thermal-hydraulic analysis of the PORV discharge conditions and give the input parameters used in the analysis.
- b.
Since the ASME Code requires the derating of the safety valves to 90 percent of actual flow capacity, the safety.valve analysis should be based on flow equal to or higher than 111 percent of the valve flow rating unless a different flow rate can be justified.
Discuss the pertinent input data used in the therma_l-hydraulic analysis to demonstrate that the above derating requirement was met. I
~
Response
6b.
The safety valve input data used in the thermal-hydraulic analysis did not explicitly account for the ASME code derating factor, however, the slug flow analysis performed by using WATSLUG code is considered to be conservative by more than the derating factor.
The slug after passing through the valve had a density set equal to 6 lb/ft 3 due to flashing based upon the energy in the 400°F slug upstream of the valve.
The downstream slugs are treated as square-edged slugs which is conservative, and the slugs from each safety valve are combined into a larger square-.edged slug which is also conservative.
The overall conservatism of our analysis (fluid transient and stress. analyses) is demonstrated by Figure 3A. 3A. 6 of Attachment A to be significantly larger than 11 percent.
Attachment A contains a general description and verification of the WATSLUG program against RELAP 5/MOD 1 and EPRI test results.
The allowable stress used is
- 1. 8 sh.
This is conservative for SRV
- discharge loads case which includes DBE load-ing.
Question:
- 7.
The discussion of the piping stress analysis on pages 17 t.hrough 20 of Reference 3 did not include the stress evaluations for the safety/PORV piping and supports and other pertinent data.
The following informa-tion is required in order to complete the evaluation of the piping stress analysis.
- a.
Give numerical values of the cutoff frequency and total analysis time for safety valve and PORV discharge used in the NUPIPE computer analyses.
- b.
Were the same load combination equations and stress. limits given on page 19 of Reference 3 used to evaluate both the piping upstream and downstream of the safety valves and PORVs, and explain why the inlet piping (Class 1) and the discharge piping stresses (Class 2/3) are evaluated by the same standard.
- c.
Provide a comparison between the worst stresses and allowable stresses for the inlet and discharge piping to show that the piping system is adequate for all.loading combinations.
- d.
Provide the same comparison for the piping support stresses as in item c above.
- e.
Provide a copy of the thermal-hydraulic and stress analysis report for both the inlet and discharge piping.
Response
7a.
The cut-off frequency and mode, and the total integration time for the NUPIPE *computer analysis for the safety valve discharge loading condition for Surry Power Station Uni ts 1 and 2 are as follows:
Unit 1 Unit 2 Cut-off frequency (HZ) 502 543 Cut..:off mode 196 181 Total integration time 0.9 0.9 (seconds)
Reference 1
2 The analysis for the safety valve discharge loading included both t.:.e inlet and discharge piping of the safety as well as relief valve piping from the pressurizer nozzle to the pressurizer relief tank nozzle.
7b.
The applicable code per Updated Final Safety Analysis Report (LTSAR)
Table 4.1-9 for the piping analysis, USAS B31.l.l
- 1955, "Code for Pressure Piping," does not contain any piping classification such as Class 1 and Class 2/3 piping.
Accordingly, the piping, both upstre3m and doi.rnstre_am of the safety valves and POR\\'s, is evaluated to a sing:e set o_f load combination equations and allowable stress limits.
The
. loading combinations and allowable stress limits utilized in the analysis are con.sistent with UFSAR and Westinghouse reco1111Dendations.
(References 3 and 4) e e
7c. Tables 3 and 4 provide a comparison of maximum stress level for Surry Units 1 and 2 at any location in the piping system (between the pressurizer nozzle and the pressurizer relief tank nozzle), to the code allowable limits. The comparison indicates that the piping system meets the allowable stress limits for all loading combinations.
7d. Tables 5 and 6 provide a list of attributes for two representative restraints associated with the PSRV system.
The calculated stresses/loads for the various restraint components of the system are compared to their respective allowables. These allowables used are those specified in the AISC Manual for Steel Construction (7th Ed, 1970) and manufacturer allowable limits.
The comparison ind1cates that the stresses/loads for these two restraints, which are typical of others within the PSRV system, are within allowable.
stresses/loads permitted.
7e. The thermal-hydraulic and stress analysis for the pressurizer safety and relief valve piping system has been performed in accordance with the ASA B31.1 Code for Pressure Piping 1955 Edition, as noted on page 19 of our October 31, 1984 submittal. Surry Power Station is an ASA B31.1 plant and unlike an* ANSI l?Jl.7 plant, Code for Nuclear Power Piping,does not require formal stress reports.* As noted, the analysis has been performed, however we do not have the reports as have been requested.
Equation 8
9N 9F 10 11 Note:
It as s p s p s p e
Table 3 HAXIffiJM STRESS LEVEL
SUMMARY
SURRY UNIT 1 Point Criteria No.
+ SDl ~ 1. 0 Sh 465
+ SDL + SOBET ~ 1.25 sh 465
+ SDL + (SfiBET + s6cc); ~ 1*8 sh 200 STH <SA= ( 1. 25 S + 0. 25 C
Sh) 300 s p + SDL + STH ~ Sh + SA 3.00 Maximum Calculated Allowable Stress (psi) Stress (psi) 8053 15990 10308 19188 28604 28782 36769 27435 39595 43425 is acceptable to exceed EQ(lO) for thermal expansion stress as long the requirements of EQ (11) a re met.
e Table 4 MAXIMUM STRESS LEVEL
SUMMARY
SURRY UNIT 2 Maximum Point Calculated Allowable Equation Criteria No.
Stress (psi) Stress (psi) 8 s + 5DL ~ 5H p
248 8187 15990 9N s
+ 5DL + SOBET ~ 1. 2 SH 248 12988 19188 p
9F Sp+ SDL + (S2
+sz )\\ < 1.8 SH 38 22797 28782 DBET occ 10 STH ~ 1. 25 Sc+ 0.25 SH= SA 38 37666 27435 11 s p + 5DL + STH ~SA+ SH 38 41617 43425 Note:
It is acceptable to exceed EQ(lO) for thermal expansion stress as long as the requirements of EQ(ll) are met.
~.
~
Notes for Tables 3 and 1111 S
= Longitudinal Pressure Stress (psi) p SDL = Deadload Stress (psi)
Operational Basis Earthquake Stress mov~ents)
(includes anchor SDBET
= Design Basis Earthquake Stress (includes anchor movements)
SOCC = Safety Valve Discharge Time History Loads SC= Allowable Stress at Ambient Temperature= 18750 psi SH= Allowable Stress at Maximum Operating Temperature= 15990 psi at 470°F (See Note 3 below)
STii = Thermal Stress (psi)
- 2.
The pipe stress reanalysis and pipe support modifications associ-ated with this design change are based on allowable stresses as provided in ANSI B31. l Code-1973 Edition.
This later Code used a different method liQ__ establish allowable stresses and incorporated test data performed subsequent to the ASA B31. l-1955 (including Code Case N7)
Code.
- 3.
The allowable stress used is lower of SA376 TP316 material (piping upstream of safety and relief valves) at 650°F and SA312 TP304 (piping downstream of safety and relief valves) at 470°F.
Item Snubber Tube Steel e
Axial Stress Bending Stress Shear Stress Integral Welded Attachment (Lug)
Bending Stress Shear Stress Critical Weld Drill co-Maxi-Bolt Anchor Bolts Tension Shear Baseplate Stres~
Reference 5 TABLE 5 SURRY UNIT I RESTRAINT H-906 Calculate Stress/Load 4186 lbs 0 psi 17798 psi 1207 psi Negligible 2093 psi 0:27 in. Fillet (Required) 5983 lbs*
1057 lbs*
24949 psi Allowable Stress/Load 10350 lbs 21600 psi 21600 psi 14400 psi 11874 psi 7760 psi 0.31 in. Fillet (Provided) 10142 lbs 8000 lbs 27000 psi
- Note:
Bolt interaction= 0.72 < 1.0 Reference Corner & Ladda Catalog AISC Manual of Steel Construction (7th ED)
~ ~ 0.6 Fy ls - 0.4 FY AISC Manual of Steel Constr,uction (7th ED)
~ = 0.6 Fy S = 0.4 FY AISC Manual of Steel Construction (7th ED)
S'WEC:-STS-ACll-1 AISC Manual of Steel Construction (7th ED)
FB =. 75 FY
e
-TABLE 6 SURRY UNIT II RESTRAINT H-900 It~
Snubber Tube Steel Axial/Bending Interaction Shear Stress Integral Welded Attachment (Trunnion)
Actual Stress/Load 18000 lbs 0.234 427 psi Bending Stress 5294 psi 3226 psi Critical \\oi'eld Drillco-Maxi-Bolt Anchor Bolts Tension Shear Baseplate Stress Reference 6
- 0. 16 in. Fillet (Required) 8029 lbs:
4401 lbsr 11889 psi Allowable
- Stress/Load 21000 lbs l
,14400 psi 11874 psi 7760 psi*
0.28 in. Fillet (Provided) 16593 lbs 11259 27000 psi
- Note:
Bolt interaction= 0.87 < 1.0
--- Reference Corner & Ladder Catalog AISC Manual of Steel Construction (7th ED)
~: 0.6 FY
~ - 0.4 FY AISC Manual of Steel Construction (7th-Ed)
F = 0. 6 FY
~ = 0.4 FY AISC Manual of Steel Construction (7th ED)
S\\oi'EC-STS-ACll-1
FB = 0.6 FY
~eferences:
- 1.
- 2.
SWEC EMD calculation No. 14937.03-NP(B)-003-XF, Rev. O, dated Novl!mber 23, 1984, "Time-history analysis for the pressuri*zer safety and relief valve piping."
SWEC EMD calculation No. 14937.03-NP(B)-001-XF, Rev. O, dated November 27, 1984, "Time-history analysis for the pressurizer safety and relief valve piping."
- 3.
UFSAR for Surry Power Station - Units 1 and 2, Section lSA.
- 4.
Review of pressurizer safety valve performance as observed in the EPRI Safety and Relief *valve Test Program, dated June 1982,
- WCAP 10105 -
Westinghouse Corporation.
- 5.
12846.22-NP(B)-Z-630-036 Virginia Power-Surry Unit No. 1, "Pipe Support Calculations for:
Pressurizer Safety and Relief Tank Piping MKS-124Al and A2 Problem 630," dated February 28, 1985.
- 6.
SWEC _EMD -calculation -
14937.03-NP(B)-001-ZB-Oll, Surry Unit No. 2, "Pipe support calculation for:
Pressurizer Safety and Relief Tank Piping, DCP-84-72," dated March 2, 1985.
- 7.
- 8.
e e
S\\¥EC EMD Calculation:
12846.22.NP(B)-030-X12 Rev. 2, dated February 26, 1985, "Pipe stress analysis of the pressurizer safety and relief system with time history -
pipe stress Problem 630," computer run number R2586008, Job No.: 3132, dated December 6, 1984.
14937.03-NP(B)-002-XE, Rev. O, dated December 6, 1984, "Stress Analysis for Pressurizer Safety and,Relief System Piping ~ Problem 2000" computer run number R2586012, Job No.:
2842 dated November 13, 1984.
~AFETY EVALUATION QUESTIONS
~
t,1:REG Oi 3 7 SURRY UNITS 1 & 2
- ATSLUG ATTACHMENT A
- l. G4neral Description The-pu'rl)oa~ of W,1.TSLOG (Ref. l) is to detem~e forcing functior.s on piping 1ystems during vater slug discharge events for subsequent i~put to piping dyu&m.ic &A&l.ysis.
- The analysi* ~ baaed upon rigid' body motion of the generally subcooled water slug and id~ gas representations of the steam or air using rigid column theory to facilitate tracking the several water-ste.am or vater-air interfaces.
The driving force i.s the ste.am pressure between the valve and the slug, less friction and other losses, and back pressure.
Density changes due to-poasible local f~shing of the-water slug are c~nsidered.
Raving recourse co the control volume theory, the subsequa~t 1egment force calculation i1 carried cue.
'!'he* input consists of complet~ piping system geomet-:-y, pipe dimensions, valve flow characteristics, valve opening time, detail upstre.am steam conditions, and initial downstream steam o~ air conditions, while the output contains forcing functions for each piping segment baaed upon flow velocities, pressures, and de:;siciea during the water slug diach.arge event.
Forces ara written on cape for direct input to NUPil't-S'J (MI-llO).
(Ref. 2).
- 2.
Program Verification The WATSLUG model of the test problem is di.&grammed in Figure 3A.3.A-l.nd the mr?I?E-SW model is diagrm:mied in Figure 3A.3.A-2. WATS!.OG is veri!ied by com;,ar:!.ng tbe solution of this test problem co the results for the same problem obtained by an inde?endenc analytical approach (RE!.Al'5/~0D l, ~ef. 3) as shown in Figures 3A.3.A-3 and 3A.3.A-4 and by c01:parison of ?redicteci ve:-sus measured support reactions.
NUPn'E-SW (ME-llO) ge~eraced support reactions due :o the ~AIS1.0G forcing functions vere compared'with ~e:-:::e~:a:
me.asurements from & test run cf this problem (:::PRI Test 908, Ref. 3) as shown in Figures 3A.3.A-5 and 3A.3.A-6.
The WA~SlUG generated forcing functions and tbe resultant NUl'!PE-~J support re.actions co!Zll)are favorably vith the RD.A.PS/MOD l pt'edicted forcing tunctio~s and the !PRI me&sured suppot't reactions, respectively.
- 3.
References----
- l.
\\JA.!SLOG" CME-212) computer code by J. S. Hsieh and 0. A. Van ~yne, Ver. 0, Rev. 3, December 1982 &nd the related documentation calculation 576.4iO.l-NP(B)-038-FD, Rev. 2, '"water Slug Disc~rge i~
P1pi~g System (~AISLOG) - PTeproduction Version 3", dated ~.arch J, l9S:.
- 2.
NU?IP!-SW, ME-llO, V03,I.l4 (created S2.095), "Computer code :or Scre~s Analysis of Nuclear Piping",
- 3.
"A;,plication of RI:..APS/~OD 1 for calculation of Safety and Re lie: Va.:se
- >ischarge Piping Hydrodynamic Loads", **nce:-im Report, ~rch 1962, ':,~*
!~ce:-mou~ta1n !echnologies, lnc., Idaho Fal:s; :daho, Prcjec: !".a.~ager R. K. Bouse.
- I
?!P! ~o.
l 2
3
~CT::S:
r.nnJT DATA :OR ~ATSt:G
- TOT.AI.
n?SIDE u::NGTH (!t)
DL~
l6.l25 0.408 l.2.56.3 0.5054 6.3 *.562 0.948 ORinCI OPINING A~~~ (~_.2)
~
(S
)
~...
ec 0.0253 O.Ol.S
(:t)
- 0. Ol.S
- 0. 0 l.'.3 OISCRARG?
FT...OW COE:T!C!~
RAT! (lbm/sec) 0.805 120.83 tTPSn!AM STL\\M CONDITIONS PRl:SstnU:
P!U:SSUR.E (?SL\\)
(!SI)
R.!SE ?_o\\T!,
269C.
DOWNS~ GAS CONDITIONS PRESSURE (?SL\\)
l;. ~o.
D~S!T! C;"1~)
0.099i5
- lee
_COTOFF
-~E SJ P!r!
SECTION l
2 3
4 5
6 1
e*
ctTOF:
'!'llOtm.C'!
433 Hz TO'IAL u:NGTH (Ft) 4.* 73 12 *.31 12.43 69.0 l.l l.O O.S3 0.0009 S&c.
OUTSIDE DUHITDt.
8.62.5 6.625 6.625 12.15
- 12. 75 8.625 6.625 I!mGRATION T~
O.S Sec.
(IN)
~nnc:n:ss 0.906 0.864 0.28 0.688 l *.s 0.322 0.432 l0%
(nl)
EaoT * !COLD
- YOONG' s ~0Dut.US OF P,!P_!
- 28. 3 x 106 PS! ilt!Gcr! (~~ 1 Ft)
. 74. 71 53.* 16-18.97 sa. 60 2S.53
- S.57
.,JI;.
"1
~
WATD
~
0 VUUl.
OISTAIICU "'OM ~1%D VOSEl.;
i.. I s.l~/flll.llJ' VAi.Vi I.II
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SIGMINT NO, WAT'S'9UG MODEL,1111 NUMIU I
- il'l!!SUIIIIZER T'O 1.4 Z
- L.o i"O L.G 3
- 1.G T'Q t.1
,.u r
- I.A
-r l FIGURE_ 3A.3.A
SAMPt..E PROBL.EM *
-1..,
0 I
I N
O"I I
10
<D 6
JO JII 40 60
© 411 Ill DO E] D.o*
IOG UO MIi 120
- SUPPORT I 121.
110 1111 140 145 1:50 IH 180 UID EJ 2i.1*
- LEGEND, O
Q
- NUPIP(-IW MODEL PIP( SECTION NO.
240 2.D' 0
WE 30/'8 ;}--~==il~--1110 110 1811 180 IH 1D uo 2,:s SUPPORT 2 SUPPORT~
rn
\\Al, 3Z/l3 0
411.4 1
SUPPORT 4 WE 34/35 FIGURf 5A. 3.A-2 NUPIPE-SW MODEL EPRI SAMPLE. PROBLEM
- - - - ~ -
- 1
.. I e
e
I N
-..J I
~
D0,000 -----------------------------------------,
0.0 r
.A -
w -D0,000-u 0:
0 IL N
J-ffi -100,000
~
(!)
Ill II)
-ID0,000 I
200,000* L-----..I------L-----L------r:&--------1~----------........ ------'
0.0 0.1 0.f 0.J 0.4 LEOEND WATSLU8 llfLAP II/MOD I TIME tSEC)
FIGURE 3A. 5.A-5 COMPARISON OF SEGMENT 2 FORCINO FUNCTION e
.J -
IJJ u
a:
~
If) l-I z
N Ill 00
- I:
I I!)
Ill en 1)0,000 r
80,000 S0,000 0.0
-20,000
-10,000 L------IL-------'1....----_...----=-------------------....._ ___ _,
,0.0 0.1 O.l 0.1 0.4 LEGEND WATSLU8 RfLAPG/MOO I TIME (SEC)
FIGURE 5A. 3.A - 4 COMPARISON OF SEGMENT 3 FORCING FUNCTION
)
e e
~-
10,:co---------------------
4C,OOO
~
t,~
--z 0
I I
I
(,J C
0 I
1M I
a:...
I a:
I 0
I
- a.
tl
- a. =
en_ -~
I.
z lo&,
- I 0
lo&,
en
-eo.ccx,
- 12Q.i:)QO~--~......i------...... ----~----....J 0.1 o.z 0.4 TIME (SEC) 1.!GENO NUP111( *SW EJl"I i!ST 111£SUL1'S FIGURE ~A.3.A *'
COMPARISON OF SEGMENT~,
SUPPORT ~EAC710N I
I J
~.-,t,.,
,,.... 1 e
160,000---------------------,
80,000 0
119
.Q -
z Q
I-0 C:
-80,000 Lu a:
I-a:
0 Q.
Q.
Cl)
N
-160,000 I-z Lu
~
C, Lu (J)
-240,000 EPRI PEAK
-320,000 400,000"'-----i......-------------...__-__
0.1 0.2
- o. 3
- o. 4 TIME (SEC)
LEGEND NU PIPE-SW EPRI TEST RESULTS FIGURE 3A.3A-~
COMPARISON OF SEGMENT 2 SUPPORT REACT ION