ML18144A006

Forwards Addl Info Re NUREG-0737,Item II.D.1 on Performance Testing of Relief & Safety Valves,Per 850920 Commitment. Response to Request Numbers 2,4,5,6b & 7 Also Encl.Remaining Info Will Be Submitted by 860228

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 October 31, 1985 NUCLEAR OPERATIONS Mr. Harold R. Denton, Director Serial No. 85-615A Office of Nuclear Reactor Regulation E&C/TLG:acm Attn: Mr. Steven A. Varga, Chief Docket Nos. 50-280 Operating Reactors Branch No. 1 50-281 Division of Licensing License Nos. DPR-32 U. S. Nuclear Regulatory Commission DRP-37 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 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

'/----~-

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p R ADOCK 05000280 1

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...* .. e e Vrnou;,IA ELECTRIC AND PoWER COMPANY TO Mr. Harold R. Denton cc: Dr. J. Nelson Grace Regional Administrator NRC Region II Mr. D. J. Burke NRC Resident Inspector Surry Power Station

e ENCLOSURE 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 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.

e e SURRY UNIT 1 TABLE lA Safety Valve Inlet Loads Due to SRV Discharge Type of_ Valve Number Loading Condition SV/1551C sv /255 lB SV /1551A Loads (Ft-Lbs) Loads (Ft-Lbs) Loads (Ft-Lbs)

Mt Mb1 Mb2 Mt Mb1 Mb2 Mt Mb1 Mb2 Thermal*.': 1371 1043 -3951 780 804 -2892 -2245 -6360 -1293 Dead Load -93 186 -530 -161 295 -907 44 -230 -147 OBE (I) 69 131 15 87 139 54 155 41 36 DBE (I) 120 236 25 142 296 88 246 65 67 OBE (A) 150 14 35 103 13 35 78 45 30 DBE (A) 341 38 95 246 35 93 197 126 78 SRV Discharge 7094 1080 3251 9177 3252 2428 58 5166 2305

(

Reference:

7)

Where OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Operational Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor ~ovement Mt = Torsional Moment Mb = Bending !foment

• = 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+/-.

- .. .. 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

(

Reference:

7)

Where OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Opera_tional Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor Movement Mt = Torsional Moment Mb -* 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 Condition SV/2551A SV/2551B SV/2551C Loads (Ft-Lbs) Loads (Ft-Lbs) Loads (Ft-Lbs)

Mt Mb1 Mb2 Mt Mb1 Mb2 Mt Mb1 ~2 Thermalir -1904 3902 -3114 -2002 2322 4177 2799 -6849 245i Dead Load 9 -106 48 1 -343 -724 -53 -292 -108 OBE (I) 102 80 57 114 150 88 78 82 101 DBE (I). 180 155 119 230 312 183 154 165 201 OBE (A) 88 48 41 209 80 87 68 44 8 DBE (A) 186 129 103 469 221 218 84 59 10 SRV Discharge 4268 4415 6451 9195 7969 8657 2189 9229 5889

(

Reference:

8)

Where OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Operational Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor Movement Mt = Torsional Moment Mb

-:r

= 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 Discharge are considered as+/-.

e SURRY UNIT 2 TABLE 2B PORV Inlet Loads Due to SRV Discharge Type of Valve Number

• Loading Condition PCV/2455 PCV/2456 Loads (Ft-Lbs) Loads (Ft-Lbs)

Mt Mb1 Mbz Ht Mb1 Mb2 Thermal* -711 1120 407 -972 -264 -1096 Dead Load -77 -1 227 -61 .o 104 OBE (I) 70 46 279 54 38 206 DBE (I)* 143 93 589 111 76 436 OBE (A) 0 1 0 0. 0 0 DBE (A) 0 1 0 1 2 0 SRV Discharge 8 17 5 2 15 7

(

Reference:

8)

Where*

OBE (I) = Operational Basis Earthquake Inertia DBE (I) = Design Basis Earthquake Inertia OBE (A) = Operational Basis Earthquake Anchor Movement DBE (A) = Design Basis Earthquake Anchor Movement Ht = Torsional Moment Mb = 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.

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. The PORV discharge conditions were not addressed.

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 I'

~ 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.

e Table 3 HAXIffiJM STRESS LEVEL

SUMMARY

SURRY UNIT 1 Maximum Point Calculated Allowable Equation Criteria No. Stress (psi) Stress (psi) 8 s Sh 465 8053 15990 p + SDl ~ 1. 0 9N sp + SDL + SOBET ~ 1.25 sh 465 10308 19188 9F s p + SDL + (SfiBET + s6cc); ~ 1

• 8 sh 200 28604 28782 10 STH <SA= ( 1. 25 SC + 0 . 25 Sh) 300 36769 27435 11 s p + SDL + STH ~ Sh + SA 3.00 39595 43425 Note: It is acceptable to exceed EQ(lO) for thermal expansion stress as long as 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 sp + 5DL ~ 5H 248 8187 15990 9N s 5 2 248 12988 19188 p + DL + SOBET ~ 1. SH 9F Sp+ SDL + (S2DBET +szocc )\ < 1.8 SH 38 22797 28782 10 STH ~ 1. 25 Sc+ 0.25 SH= SA 38 37666 27435 11 sp + 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 1

111

~. S p

= Longitudinal Pressure Stress (psi)

SDL = Deadload Stress (psi)

Operational Basis Earthquake Stress (includes anchor mov~ents)

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.

e TABLE 5 SURRY UNIT I RESTRAINT H-906 Calculate Allowable Item Stress/Load Stress/Load Reference Snubber 4186 lbs 10350 lbs Corner & Ladda Catalog Tube Steel Axial Stress 0 psi 21600 psi AISC Manual of Steel Bending Stress 17798 psi 21600 psi Construction (7th ED)

Shear Stress 1207 psi 14400 psi ~ ~ 0.6 Fy ls - 0.4 FY Integral Welded Attachment (Lug)

Bending Stress Negligible 11874 psi AISC Manual of Steel Shear Stress 2093 psi 7760 psi Constr,uction (7th ED)

~ = 0.6 Fy S = 0.4 FY Critical Weld 0:27 in. Fillet 0.31 in. Fillet AISC Manual of Steel (Required) (Provided) Construction (7th ED)

Drill co-Maxi-Bolt Anchor Bolts Tension 5983 lbs* 10142 lbs S'WEC:-STS-ACll-1 Shear 1057 lbs* 8000 lbs Baseplate Stres~ 24949 psi 27000 psi AISC Manual of Steel Construction (7th ED)

FB = . 75 FY Reference 5

• Note: Bolt interaction= 0.72 < 1.0 e

-TABLE 6 SURRY UNIT II RESTRAINT H-900 Actual Allowable

• It~ Stress/Load Stress/Load Reference Snubber 18000 lbs 21000 lbs Corner & Ladder Catalog Tube Steel Axial/Bending AISC Manual of Steel Interaction 0.234 l Construction (7th ED)

Shear Stress 427 psi ,14400 psi ~ : 0.6 FY

~ - 0.4 FY Integral Welded Attachment (Trunnion)

Bending Stress 5294 psi 11874 psi AISC Manual of Steel 3226 psi 7760 psi* Construction (7th-Ed)

F = 0. 6 FY

~ = 0.4 FY Critical \oi'eld

• 0. 16 in. Fillet 0.28 in. Fillet AISC Manual of Steel (Required) (Provided) Construction (7th ED)

Drillco-Maxi-Bolt Anchor Bolts Tension 8029 lbs: 16593 lbs S\oi'EC-STS-ACll-1

• Shear 4401 lbsr 11259 Baseplate Stress 11889 psi 27000 psi AISC Manual of Steel Construction (7th ED)

FB = 0.6 FY Reference 6

• Note: Bolt interaction= 0.87 < 1.0

~eferences:

1. 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."

2. 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. SWEC EMD Calculation - 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.

e e

7. 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.

8. SWEC EMD calculation: 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 r.nnJT DATA :OR ~ATSt:G

• TOT.AI. n?SIDE  :!UC'!'ION

?!P! ~o. u::NGTH (!t) D L ~ (:t) FAC':'OR l l6.l25 0.408 O.Ol.5 2 l.2.56.3 0.5054 0. Ol.S 3 6.3 *.562 0.948 0. 0 l.'.3 ORinCI OPINING OISCRARG? FT...OW

~

A~~~ (~_.2)

... ~

..:....;. (S ec ) COE:T!C!~ RAT! (lbm/sec) 0.0253 O.Ol.S 0.805 120.83 tTPSn!AM STL\M CONDITIONS PRl:SstnU:

P!U:SSUR.E (?SL\) R.!SE ?_o\T! (!SI)

lee 269C. ~o.

DOWNS~ GAS CONDITIONS PRESSURE (?SL\) D~S!T! C;"1~)

l;. 0.099i5

~CT::S:

e*

_COTOFF ctTOF: I!mGRATION

-~E '!'llOtm.C'! T~

SJ 433 Hz 0.0009 S&c. O.S Sec. l0%

P!r! TO'IAL OUTSIDE

• SECTION u:NGTH (Ft) DUHITDt. (IN) ~nnc:n:ss (nl) ilt!Gcr! (~~ 1Ft) l 4 .* 73 8.62.5 0.906 . 74. 71 2 12 * .31 6.625 0.864 53 .* 16-3 12.43 6.625 0.28 18.97 4 69.0 12.15 0.688 sa. 60 5 l.l 12. 75 l * .s -

6 l.O 8.625 0.322 2S.53 1 O.S3 6.625 0.432 :S.57 6

EaoT * !COLD

• YOONG' s ~0Dut.US OF P,!P_!

• 28. 3 x 10 PS!

. ,JI;.

"1 ,.u

! ~

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• I.A

~

VUUl. 0

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-r l 0 OISTAIICU "'OM ~ 1 % D VOSEl.;

i.. I s.l~/flll.llJ' VAi.Vi I.II

• OCWNSTM:.t.M A"IA CHANU 1.1 * ~,<< l:XIT r:xt

• Tfl.AIUM tD81 ~ sa.ue 1

. , . _ l'O"CUMI fl'UNCT10N .OUtlC'T'ION,

• D !Jl'tl T'!!T ( ~ , ;

SIGMINT NO, WAT'S'9UG MODEL ,1111 NUMIU I

• il'l!!SUIIIIZER T'O 1.4 I

Z

• L.o i"O L.G  ! -

3

• 1.G T'Q t.1 FIGURE_ 3A.3.A

• I WAiSL.UG MOOEl. CF EPR!

SAMPt..E PROBL.EM *

- - - - ~ - *1

.. I E] D.o*

IOG UO MIi

@ 120 JO 121. LEGEND, O* NODE NUii Bf 11 JII

• SUPPORT I 110 1111 140 A* MAIi POIHJ e

40 60 145 1:50 EJ 2i.1* E]* IPIU TlH CRH SI HG~ENT ~O.

10 411

© Ill IH Q

• NUPIP(-IW MODEL PIP( SECTION NO.

<D DO 6 180 I

N

@ 240 O"I I

2.D'

, UID WE 30/'8 ;}--~==il~--1110

@ 0 1D 110 1811 180 IH uo 2,:s SUPPORT 2 0 1 e

411.4 SUPPORT 4 SUPPORT~ WE 34/35 rn

\Al, 3Z/l3 FIGURf 5A. 3.A- 2 NUPIPE-SW MODEL EPRI SAMPLE. PROBLEM

D0,000 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ,

0.0 r e

.A w

u

-D0,000-0:

0 IL N

J-N I

-..J ffi~ -100,000 I (!)

Ill II)

' I

-ID0,000 .

200,000* L-----..I------L-----L------r:&--------1~----------........

0.0 0.1 0.f 0.J

---

'0.4 TIME tSEC)

LEOEND FIGURE 3A. 5.A- 5 WATSLU8 COMPARISON OF SEGMENT 2

---

llfLAP II/MOD I FORCINO FUNCTION

1)0,000 r

-.J IJJ 80,000 u

a:

~

If) l-I z N Ill S0,000 00 :I:

I I!)

Ill en 0.0 --

-20,000

-10,000 L------IL-------'1....----_...----=-------------------....._____,

,0.0 0.1 O.l 0.1 0.4 TIME (SEC)

LEGEND FIGURE 5A. 3.A - 4 WATSLU8 COMPARISON OF SEGMENT 3

---

RfLAPG/MOO I FORCING FUNCTION

e e

~-

10,:co---------------------

--* 4C,OOO

-z~

t,~

0

(,J C

'II I

I 1M 0 a: I

...a: I I

0 I

a. tl a.

= '"

..."'z -~

en_ I .

lo&,

I 0

lo&,

en

-eo.ccx,

• 12Q.i:)QO~--~......i------......

0.1 ----~----....J o.z 0.4 TIME (SEC) 1.!GENO ----

NUP111( *SW

---

EJl"I i!ST 111£SUL1'S

) FIGURE ~A.3.A *'

COMPARISON OF SEGMENT~, I SUPPORT ~EAC710N I

~

.-,t , . ,

, , . ... 1 e '

160,000---------------------,

80,000

- 119 0

.Q z

Q I-0 C:

Lu -80,000 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 FIGURE 3A.3A-~

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EPRI TEST RESULTS COMPARISON OF SEGMENT 2 SUPPORT REACT ION