View source for B12634, Forwards Response to NRC 861212 Question 5 Re NUREG-0737, Item II.D.1 Concerning Testing of PORVs & Safety Valves at Facility,Per Util & 870803 Telcon.Flow Rate Equal to Theoretical Flow for Fluid Transient Analysis

{{Adams
| number = ML20237H387
| issue date = 08/10/1987
| title = Forwards Response to NRC 861212 Question 5 Re NUREG-0737, Item II.D.1 Concerning Testing of PORVs & Safety Valves at Facility,Per Util & 870803 Telcon.Flow Rate Equal to Theoretical Flow for Fluid Transient Analysis
| author name = Mroczka E, Werner R
| author affiliation = CONNECTICUT YANKEE ATOMIC POWER CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF ADMINISTRATION &amp; RESOURCES MANAGEMENT (ARM)
| docket = 05000213
| license number = 
| contact person = 
| case reference number = RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM
| document report number = B12634, NUDOCS 8708170180
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 26
}}

=Text=
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                                                                                                                                       'l I
CONNECTICUT YANKEE AT O MIC POWER COMPANY                                                              1 I
B E R L I N, CONNECTICUT P.o. Box 270
* HARTFORD, CONNECTICUT 06141-0270
                       ]l'J3][                                                      August 10,1987                                       j 1
Docket No. 50-213                                      {
L                                                                                                    B12634                              l l                                                                                                                                         l l
1 U. S. Nuclear Regulatory Commission                                                                          j Attn: Document Control Desk                                                                              ..
Washington, D. C. 20555                                                                                      )
l

==Reference:==
(1) F. M. Akstulewicz letter to E. 3. Mroczka, " Review of                            )
NUREG-0737, Item II.D.1, Performance Testing of Relief and                         !
Safety Valves for Haddam Neck," dated December 12,1986.
(2) E. 3. Mroczka letter to U. S. NRC Document Control Desk,                          !
                                                        " Relief Valve and Safety Valve Testing," dated July 9,1987.
I Gentlemen:
l
!                                                                   Haddam Neck Plant Relief Valve and Safety Valve Testing In Reference (1), the NRC Staff requested that Connecticut Yankee Atomic Power Company (CYAPCO) provide additional information regarding NUREG-0737, Item !!.D.1, the testing of power operated relief valves (PORVs) and safety valves at the Haddam Neck Plant. Reference (2) provided this information and stated that a response to Question No. 5 would be submitted by September 1, 1987.
In accordance with that commitment, the response to Question No. 5 is provided in Attachment 1.            This letter will also confirm a statement made during a telephone conversation with the NRC Staff on August 3,1987.
During the telephone discussion, the NRC questioned whether the theoretical flow rate was used for the fluid transient analysis. It was stated that the ASME rated flow for the pressurizer safety and relief valves is equal to 90% of the theoretical flow rate but that 111% of the rated flow was used for the fluid transient analysis of the safety and relief valve discharges. Thus, the flow rate                            i is equal to the theoretical flow for this calculation.
8708170180 870810 Oh PDR          ADOCK 05000213                                                                                           '
p                             PDR                                                                                       7(

                                                                                                                      )
U. S. Nuclear Regulatory Commission B12634/Page 2 August 10,1987 1
Please contact us if you have any questions.
Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY 1
                                                         . M cm o                                                 i E. J. Mro                 O Senior Vice- resident
_         >   .b-x~
By: R. P.'WErner Vice President Attachment I
cc:   W. T. Russel, Region 1 Administrator                                                                     -
F. M. Akstulewicz, NRC Project Manager, Haddam Neck Plant                                                j J. T. Shediosky, Resident inspector, Haddam Neck Plant                                                    {
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                                                                      - . _ _ - _ _ _ _ _ _ _ _ _______m___-_~

7.
Docket No. 50-213 B12634                                   .1 l
1 1
l Attachment 1 Relief Valve and Safety Valve Testing                                            l Response to Question No. 3                                               .) 1 I
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August 1987 l
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7 q
4 Question'5:                                                                          ,
1 The discussion on the thermal hydraulic analysis of the PORV and safety valve piping system presented in Reference' 1 (response to Question 9) has been reviewed.       Additional information is still required in order to
    . _ complete     the  evaluation    of    the~ analysis.-     Provide   the   following information
: a. Explain' in detail how the thermal-hydraulic anclysis program'           ;
                   "STEHAM" was verified to ensure that - it was adequate for the safety valve /PORV piping application.       Provide a comparison of     !
the calculated results using the STEHAM program for an EPRI test condition with the EPRI test results for the same test.
l
: b. Discuss the piping model used for the thermal-hydraulic analysis.          l 1
Identify the important parameters used in the analysis such as node spacing, valve flow area, pressure ramp rate, choked flow             ]
i locations and discuss the rational for their selection.
Response 5
: a. Attachment    5A    contains   a   general  description    and   program verification of the STEHAM program.        RELAP5/ MOD 1 and EPRI test results are used to assist in this verification of the STEHAM computer program for Version 02, Level 01 and Interim Version.(to Version 01, Level 02) in Reference 7 of Attachment 5A.
: b. The node diagram for the STEHAM thermal-hydraulic analysis is shown in Figure       1,  " Schematic of STEHAM Computer Model for Connecticut Yankee Pressurizer". The pressurizer is modeled as a reservoir (RESVR) with pressure variation.          Piping connections are considered by using CHEST subroutines in the program.
Time steps used in STEHAM are determined by the node spacing, the local fluid particle velocity, and the sound speed to satisfy the

2 Courant stability criteria.                          Since the minimum node spacing used in STEHAM was about 1.4 ft, the time steps were in the order of one millisecond.
1 The valve flow areas for relief valves (A0V-570 and A0V-568) and safety         valves      (SV-584-586)                  are  2.9 in2 and 2.545 in 2, respectively.           Variable pressure ramp rates were considered by inputting pressure time histories in the pressurizer as shown in Table 1, " Pressurizer Time Histories".                                Choked f lo.w locations usually occurred at area changes such as at a reducer valve or conmon header.            These were considered by the STERAM program as                                   j would be appropriate.
TABLE 1 PRESSURIZER TIME HISTORIES Case 1 Time               0.0     0.65         1.65                2.65       3.65        4.65           5.65        6.65 (sec)
Pressure   ~~2,285.        E[319.       E!359.             27469.      2,499.      2,581.        2 ,~ 65 6~~ ~fiEi6-(Psia)    _________                   _ _ _ _______     ,________             ___                  _ _ _
Densit       6.536                      6.946
______6.699_                     7.257       __7.646     8.138_ _ _ _ _ 8.586
__          8.586 (1b/ft )
Cases 2 and 3 l
Time            0.0                0.988             1.988           2.988 (sec)
Pressure       2,500               2,581             2,650           2,650 (psia)
Density        7.652                8.138            8.586           8.586 (1b/ft3)
Case 1: Two relief valves open, discharge steam.
Case 2: Three safety valves open in sequence, discharge steam.
Case 3: Three safety valves open in sequence, discharge steam with two relief valves already open.
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l c-ATTACHMENT 5A
_5TEEAM
,                    1. GENERAI. DESCRIPTION
!                                                                                                             i The purpose of STEEAM (Ref.1) is to determine forcing functions on piping systems during stea=ha==er transients for subsequent input to                 j i
piping dynamic analysis.                                                            l' The analysis is based on the method of characteristics with finite difference approxi=ations both in space and in ti=e (Ref. 2). It calculates the ene-
                                                                                                            .I dimensional transient flow responses and the flov-induced forcing functions         I l                         in a piping system caused by rapid operational changes of piping ec=ponents, i'
l                         such as the stop valve and the safety / relief valve. Flow characteristics l                         of piping co=ponents are mathematically for=ulated as boundary conditiens in the program. These ce=ponents include the flow control valve, the stop valve, the safety / relief valve, the steam nacifold and the stes:1 reservoir.
Frictional effects are taken into consideration in this progra=.
STEEAM accepts the follo'ving as input: (1) the flow network representation of the piping system, (2) the initial flev conditions along the piping system,       i and (3) ti e-dependent flow characteristics of piping ec=ponents.      Output 1
censists of ti=e-histories of flow pressures, flow densities, flow velocities,      ;
inertia, and me=entum functions. Forces are written on tape for direct input         l l
to WpIPE-SW (ME-110) .                                          ,                   i 1
: 2.       PROGRAM VERIFICATION The STZEAM model and a piping schematic cf the sa=ple problem are diagra:=ed
                                                                                                               )
in Figure 1. STEEAM is verified by ce=pering the solution of this sa:ple        :I problem to the results for the sa:e problem obtained by an independent analytical I
approach (REI.A.PS/ MOD 1, Ref. 3) as shown in Figures 2 6 'and by ce=parison of predicted (per WpIPE-SV, Ref. 6) versus ceasitred support reactions (from EPRI i

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    .                                                                                                    q i
1 2.
1 test 1411, Ref. 3) as shown in Figures 7-10.       The input for this problem is         '{
                                                                                                           )
indicated in Table 1.
A second problem is defined'in Figures 11 and 12 (Ref. 4 and Ref. 5).
STEHAM pressure time-history results are compared to analytical 'results Nedal (Figures 13 and 14) and experimental measurements (Tigure 15).
forca results are ccupared to hand calculations in Table 2.
The STEEAM generated forcing functiets and nodal pressures compara f avorshly with the RELAPS/ MOD 1 predicted forcing functions and nodal pressures.
: 3. REFERINCES
: 1.  "Steamha::cer Analysis for Piping Systema", ME-167, (STEEAM) t Version 02, Level 01, created 83.047 by SWIC.
l
: 2.  " Verification of ME-167 STEEAM Cemputer Code",'SWEC Cal. No. 574. 5 54.1-NP (B)-     I i
090-73, by A. Esi and L. Loveridge, dated 3/7/83.                                  f i
1
: 3.  " Application of RELAPS/ MOD 1 for calculation of Saf ety and Relief Valve Discharge Piping Hydrodynamic Loads", Interim Report, March 1982, by                1 Intermountain Technologies, Inc., Idaho Talls, Idaho, Project Manager R. K. House.                                                                       a i
Progelhof, R. C. and Owezarek, J. A., "The Rapid Discharge of a gas 4                                                                                      i j
from a cylindrical Vessel through a Nozzle", A1AA Journal, Vol. 1, No 9, Sept. 1963, PP. 2182-2184.
4 e

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''                Proselhof, R. C. and Owezarek,       J.'A., "The Rapid Dischstge of a Gas 5.
                                                                                                .j from a cylindrical ,Vassel through an ''rifice", ASME Paper No. 63-WA-10.
: 6. NUPIPE-5W, ME-110, V03, L14 (created 82.095), '' Computer Code f or            ,
l Stress Analysis of Nuclear Piping".                                            il
: 7. SWEC Calculation 574.551.1-NP(B)-082-FD, Rev.'1,        " Comparison of EPRI SRV Test Lesults with Fredicted Values from WATSLUG and STEHAM,"
Dated July 18, 1984 l                                                                                                 1 l                                                                                                 l
: 1.                                                                                                A l
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TA31.E 1                                      .
l IhTUT DATA FOR STEHAM *                          'j l
INSIDE              FRICTION               I TOTAL l
PIPE NO. LENGTH (Ft)_         DIAMETER (F Q       FACTOR i
lj 1          16.125               0.408              0.0087 i'
0.5054             0.0059 j    ._     _
2          12.563 3          63.562               0.948'             O.0077 i
VALVE CHARACTERISTICS
!                       0FFICE           OPENING        DISCHARGE       FLOW 2
!                       AREA (Ft 3       TIME (Sec)     COEFFICIENT     RATE (1bs/hr) l-                                                                         440,000 0.0253            0.011           0.94 UPSTREAM STEAM CONDITIONS PRESSURE PRESSURE (PSIA)       DENSITY (     )   RISE RATE (     )
2410.                7.156              262.5 DOWNSTREAM GAS CONDITIONS l
PRESSURE (PSIA)        TEMPERATURE     DENSITY (1bs) l                                                                        Ft3 l
14.7             70 F (530 R)      0.075                j l
l l
l          NOTES:
* SEE FIGURE 1 FOR SKETCH OF SIIHAM MODEL l
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TABLE 1                                !
i j                                             NODAL PORCE COMPARISON                      ;
1 1
Diameter    D=0.25 ft                                         i Area A = rDa/4 = 0.0490874.ft8 Nodal Force = (p +av afg ) A - patm A p=     pressure 1b/fta l                           s=     density lb/f t3 i
V=     velocity ft/sec                   -
g=     gravitational constant 32.2 p, , , = 14.7 x 144 1.b /fta At time    t = 0.00650 sec.
Force (Lb)
Node Pressure     Velocity      Density    Force (1b)  (Hand
_No. fosia)         (fos)    (Ib/fta)      (prEgAM) Calculation) 1  42.523         0.0         0.23954     186.57    186.67       !
5  42.785         5.7843      0.24076     158.43    198.53       I 10  4a.231        31.219       0.2n647     209.00    209.11       i 15  47.003        78.172       0.25737     230.62    230.7)       i 20  50.21u       129.89        0.26979     257.84    257.97 25  52.095       159.43        0.27697     274.93    275.06 30  52.209       161.97        0.27742     276.09    276.23 35  52.168       162.21        0.27731     275.83    275.97 l'

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e           LEGEND NUPIPE-3W
              --- EPRI TEST RESULTS F I G l, N E 8 COMPARISON OF SEGMENT 2                     l I
SUPPORT RE ACTION
                                                                                                                           ~
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LEGEND NUPIPE-sw
             --- EPRi TEsr REsutis                                                                                          .
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i                                                                              COMPARISON OF SEGMENT 3                        l SUPPORT RE ACTION                              l

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ll            THE FORCE PEAKS AND THE                      ;
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