ML20236G244
| ML20236G244 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/27/1987 |
| From: | Papanic G YANKEE ATOMIC ELECTRIC CO. |
| To: | Fairtile M NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20236G246 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM FYR-87-111, NUDOCS 8711020468 | |
| Download: ML20236G244 (57) | |
Text
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Telephone (617) 872 8100 TWX 710-380-7619 YANKEE ATOMIC ELECTRIC COMPANY g l 1671 Worcester Road, Framingham, Massachusetts 01701 YAuxme l October 27, 1987 FYR 87-111 l
United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Attention: Mr. Morton B. Fairtile, Project Manager Project Directorate I-3 Division of Reactor Projects, I/II
References:
(a) License No. DPR-3 (Docket No. 50-29)
(b) Letter, YAEC to NRC, FYR 82-11. "SEP Topic Assessment Completion," dated February 1, 1982 (c) Letter, YAEC to NRC, FYR 82-38, "TMI Item II.D.1, Safety and Relief Valves," dated March 30, 1982 (d) Letter, YAEC to NRC, FYR 82-72, "TMI Item II.D.1, Safety and Relief Valves," dated. July 1, 1982 (e) Letter, YAEC to NRC, FYR 82-82, "TMI Item II.D.1, Safety I and Relief Valves," dated August 1, 1982 )
(f) Letter, YAEC to URC, FYR 82-121. "TMI Item II.D.1, Safety I l
and Relief Valves," dated December 28, 1982 (g) Letter, YAEC to NRC, FYR 83-36 "TMI Item II.D.1, Safety and Relief Valves," dated April 1, 1983 (h) Letter, YAEC to NRC, FYR 84-41, "TMI Item II.D.1, Safety and Relief Valves," dated April 2, 1984 )
(1) Letter, NRC to YAEC, NYR 85-119 " Request for Additional Information on TMI Action Plan Item II.D.1, Performance Testing of Relief and Safety Valves," dated July 16, 1985 (j) Letter, YAEC to NRC, FYR 85-132, "TMI Item II.D.1, Safety and Relief Valves," dated November 22, 1985 (k) Letter, NRC to YAEC, FYR 87-138, dated July 23, 1987
Subject:
TMI Item II.D.1 Safety and Relief Valves l
Dear Sir:
l Reference (k) requested additional information on the performance testing ,
and design of our pressurizer safety and relief valves. The attachment to l this letter contains the requested information.
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- United' States Nuclear Regulatory Commission ' October 27. 1987'
.' Attentions; Mr. Morton B.'i Fairtile ' Project' Manager ' LPage 2 FYR 87-111
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I -If after reviewing this'information additional questions-or clarifications are needed, we suggest that you contact us to' discuss your
' questions.
I : Respectfully,.
YANKEE ATOMIC ELECTRIC COMPANY:
e G. Pa , Jr.
i' Senior Project Engineer Licensing CP/25.154
. Attachment cc: USNRC Region.I USNRC Resident Inspector YNPS E
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f-ATTACHMENT Response to Additional Questions for NUREG-0737 Item II.D.1
-QUESTION NO. 1 Your response to Question 10 (Reference 1) did not provide enough details on the design differences and similarities between the Yankee plant Pacific block valve and the Velan valves tested by EPRI to conclude the test valves were
- i. representative of the plant valves. Provide a more detailed evaluation of the differences and similarities of the plant and test valves. Also, on LDecember 28, 1982 (Reference 3), Yankee Atomic Electric Company (YAEC) submittal stated that the SMA-00-10 operator at the Yankee plant produces a torque of 62 ft-lbs,-10 percent greater than the required torque as calculated by the valve manufacturer. Since this is less than the minimum torque tested with the Velan valves (82 ft-lbs), it is the staff position that it is not adequate to conclude proper operation solely on manufacturer's calculations.
The problems encountered with Westinghouse gate valves on closing, which were traced to the calculations used to size the valve operator torque requirements, indicate the need to experimentally verify the adequacy of the block valve / operator combination. YAEC should provide test data to demonstrate the SMA-00-10 operator at the. Yankee plant is capable of providing adequate torque to close the block valves.
RESPONSE
The Yankee PORV block valve is a Pacific Valves, Inc., Figure No. 1550, 2",
ANSI 1500 lb., wedge disc gate valve. The similar valve tested by EPRI is a Velan Engineering Company, Model No. B10-30548-13MS, 3", 1500 lb. class, wedge disc gate valve. Both valves are stainless steel with stellite faced stainless steel discs and seats. An outline drawing of each valve is attached.
The wedge-type disc is a standard design offered by most gate valve manufacturers. There are other types of gate valve discs offered by various manufacturers that are designed te offer some distinction between manufacturers. However, the wedge disc gate is an industry standard.
5855R/12.159
1 The Velan valve tested by EPRI was successfully closed against a differential pressure of 2,470 psi with a torque of 82 ft-lbs. Using a standard calculational technique yields a required closing torque of 123 ft-lbs for ;
this same valve. The' calculated closing torque is 50 percent greater than l
that actually required by test. j i.
The PORV block valve at Yankee would be required to close against a differential pressure of 2,350 psi. Note the PORV at Yankee is designed to f 4
open at 2,400 psig and close at 2,350 psig. The block valve would only be required to function following a failure of the PORV to close. Using the same standard calculational technique yields a required closing torque of 53 ft-lbs. The PORV block valve motor operator produces 'a torque of l 62 ft-lbs, or a 17 percent margin over the calculated closing torque required.
l In summary, the Yankee PORV block valve will function based on the following:
- 1. The Yankee valve is similar in design to the Velan valve tested in the EPRI block valve test program.
- 2. The calculated closing torque for the tested valve is 50 percent greater than the actual closing torque required and verified by test.
- 3. The actual closing torque of the Yankee valve is 17 percent highar than the calculated closing torque, insuring valve closure. 1 Test data specific to the Yankee block valve is not available. f 1
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Diagram of the _ Pacific Valves Incorporated, Motor Operated, Bolted Bonnet Gate Valve, Model 1550 ,
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.0UESTION NO. 2 3
Provide'the maximum expected backpressure and bending moment for the Yankee plant PORV.
RESPONSE
The maximum backpressure is expected to be about 426 psia (Reference *). The attached Tables 2A and 2B provide the calculated moments at the safety valve and the powdr-operated relief valve inlets and outlets, respectively, due to the safety valve discharge conditions (Reference **). .i l
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- Stone & Webster Fluid Calculation 11986.18-NP(B)-002-FC, Revision 0, dated March 17, 1983
- Stone & Webster Stress Calculation 11986.18-NP(B)-01-XE, Revision 0, dated March 31, 1983 l
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i TABLE 2A M = Safety Valve Inlet Loads.Due to SRV Discharge 1
VALVE NUMBER
. Type ofL PRE SV-182 PR-SV-181 PR-SOV-90
- , Loading l(Safety Valve) .(Safety Valve) (Relief Valve)
Condition Loads'(Ft-lbs) Loads-(Ft-lbs)' Loads-(Ft-lbs)'
d'*a Mt Mt . Mb2 ' Mt
~ Mbt Mb2 Mb 'i- Mbt' Mb'2
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Thermal * -60 =-75 109 -18 -126 255 -3 -115 -36 <
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' Dead Load" -1 -28 .5 -31 O - 21 ' -1 -22'6 SSE (I)- 8 40 50 60 27 54' 164 127 211 SRV Discharge- 26 345 486- 53. -178 404- 161- 355 '97
- Where?
~SSE.(I) = Safe Shutdown Earthquake Inertia ;
Mt = Torsional Moment ~
L Mb- = Bending Moment- ,
.* . = Maximum Thermal Load Component with Sigri from -All Thermal-.
Conditions Analyzed Note: Dynamic loads such as SSE(I), and SRV Discharge are considered as i.
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YANKEE R0WE TABLE 2B Safety Valve Outlet Loads Due to SRV Discharge VALVE NUMBER Type of PR-SV-182 PR-SV-181 PR-SOV-90 .
Loading (Safety Valve) (Safety Valve) (Relief Valve) i Condition Loads (Ft-lbs) Loads (Ft-lbs) Loads (Ft-lbs) {
Mt Mb t Mb2 Mt Mb i Mb2 Mt Mb1 Mb2 Thermal * -63 81 -144 -183 -136 -540 25 37 -63 Dead Load 25 -1 -72 -21 3 39 31 -131 -21 SSE (I) 113 23 110 89 28 95 88 114 68 SRV Discharge 288 24 898 5 36 639 47 774 24 Miere SSE (I) = Safe Shutdown Earthquake Inertia Mt = Torsional Moment Mb = Bending Moment
- = Mr <imum Thermal Load Component with Sign f rom All Thermal Canditions Analyzed Note: Dynamic loads such as SSE(I), and SRV Discharge are considered as .
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. QUESTION NO. 3'
, Dresser Industries, in March 1976, recommended to Metropolitan Edison Company that'th'e PORV block valve be closed at pressures below 1.000 psig to prevent steam wirecutting of the PORV-seat and disk. Testing by Dresser.later showed Lthe 1,000 psig pressure ~1imit to be overly conservative and that the PORV as-designed was qualified.to system pressures of 100 psig. . . Below 100 psig,J the dead weight of the-lever on the pilot valve was-sufficient to' keep the' pilot valve'open. . Dresser. recommends that.if the plant is to' operate at pressures ,
below 100 psig, heavier springs be used under the main and pilot disks to ensure: closure. : Without the- heavier springs recomended by Dresser, the PORV might leak at system pressures below 100 psig. Dresser has also-stated that if a'PORV has-not leaked at low pressures, that the heavier springs.might'not be necessary. Has the. Yankee plant--PORV experienced.'any leakage during o 'startup or shutdown? If the' answer.is no, then the heavier springs are not necessary.' If the answer is v , then the heavier springs should be installed; and YAEC should.sts.a when the. heavier springs will be installed.
RESPONSE
The PORV is in service during startup and shutdown for low temperature
-overpressure protection. No leakage has been experienced by,the PORV.
Therefore, Yankee does not intend to install heavier springs in the valve.
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QUESTION NO. 4 Your response to our request for information stated the PORV control circuitry is located outside containment except for the PORV solenoid. However, there must be other equipment, if only cables, connecting the PORV to the components outside containment. The Nuclear Regulatory Commission staff has agreed that meeting tlie licensing requirements of 10CFR50.49 for this circuitry is satisfactory and that specific testing per NUREG-0737 requirement is not required. Therefore, verify whether the PORV control circuitry has been reviewed and accepted under the requirements of 10CFR50.49.
If the PORV circuitry has not been qualified to the requirements of 10CFR50.49, provide information to demonstrate that the control circuitry is qualified per the guidance provided in Regulatory Culde 1.89, Revision 1, j Appendix E.
As an alternative, the staff has determined that the requirements of NUREG-0737 regarding the qualification of the PORV control circuitry may be satisfied if one or more of the following conditions is met:
- a. .The PORVs are not required to perform a safety function to mitigate the effects of any design basis event in the harsh environment, and failure in the harsh environment will not adversely impact safety functions or mislead the operator (PORVs.will not experience any spurious actuations and, if emergency operating procedures do not specifically prohibit use of PORVs in accident mitigation, it must be ascertained that PORVs can be closed under harsh environment conditions).
i
- b. The PORVs are required to perform a safety function to mitigate the effects of a specific event, but are not subjected to a harsh environment )
as a result of that event.
- c. The PORVs perform their function before being exposed to the harsh environment, and the adequacy of the time margin provided is justified; subsequent failure of the PORVs as a result of the harsh environment will not degrade other safety functions or mislead the operator (PORVs will 5855R/12.159
_ _____-___a
not experience any spurious actuations and, if emergency operating procedures do not'specifically prohibit use of PORVs in accident mitigation, it must be ascertained that PORVs can be closed under harsh environment conditions).
- d. The safety function can be accomplished by some other designated equipment that has been adequately qualified and satisfies the single-failure criterion.
RESPONSE
The PORV at Yankeeqis not required to perfonr a safety function to mitigate the ef fects of, any(design b'esis event in a he.rsh environment.
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ThePORVQ-hetuatedQy'#hesolenoid. All of the, elect $ical components s .
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[normallyde-energized.' Thephibreofanyoftheseelectricalcomponents 4 i lyid[e;ntainmentwouldno$"rfNultinphpinadvertentopeningofthevalve.
, 'N Il the v'alve were opened hed 'tNDn the he.rshenvironment K ,
were formed, thegvalve g .
f}i 3couldbeclosedbysimplyremovingpkerfrom'theso1 Mold. All of the s, t W ccd,pcinents required to e this are lor.'ated outside 'of containment, in ardus
> < A i where a mild environment is maintained. LThere ore,'the PORV electrical components inside of de containment are 'not requir d to be lunlifled to i function in a harsh environment.
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It'should alvo be noted that the. instrumentation.'
n eeded to verify the positionh s,
of the PORV has been qualified as required by NUREG-0737.
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, QUESTION NO. 5 Provide more information on the verification of STEHAM and WATAIR by Stone &
Webster.- Provide comparisons of the results' for STEHAM and WATAIR f
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. calculations and EPRI/CE data to verify these codes are appropriate tools to 1 evaluate piping discharge transients.
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RESPONSE.
Attachments ~5A and 5B contain general descriptions and program verifications of the STEHAM and WATAIR computer programs, respectively. RELAPS/ MOD 1 and EPRI test'results are used to assist in the verification of the STEHAM computer program for Version 02, Level 01 in Reference 7 to Attachment 5A.
Hand calculations are used to verify the WATAIR computer program.
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ATTACEMENT SA /o[.:LU STTRAM l~ 1. CENERAI. DESCRIPTION The purpose of STIEAM (Ref.1) is to determine forcing functions en piping systems during stea=ha=ner transients for subsequent input to piping dy asic analysis.
The analysis is based on tha =ethod of characteristics with finite dif ference approximations both in space and in time (Ref. 2). It calculates the ene-di=ensional transient flow responses and the flow-induced forcing functiens in a piping system caused by rapid operational changes of pipicg ec=penents, such as the stop valve and the safety / relief valve. Flev characteristics of piping ce=ponenes are mathematically for=ulated as boundary conditices in the program. These cceponents include the flow centrol valve, the step valve, the sa.fety/ relief valve, the steam manifold and the steam reservoir.
Trictienal effects are taken Into consideration in this program.
STEFJ.M accepts the follo' wing as imput: (1) the flow network representation of the piping systes, (2) the initial fiev conditions along the piping system, and (3) time-dependent flow characteristics of piping ce:ponents. Cutput censists of ti=e-histories of, flo.r pressures, tiev densities, flow velecities,
. inertia, and scuentum functions. Torces are vritten on tape for direct input to WPIPE-SW (ME-110) .
- 2. ?ROGRAM VERIFICATION The STEEAM :odel and a piping sche =atic of the sample problem are diagra==ed in Figure 1. STEEAM is verified by ccupcring the solution of this sa ple problem to the results for the sa=a problem obtained by an independent analytical appreach (RILA?S/ MOD 1, Ref. 3) as shom in Figures 2 6 ind by ec=parisen of predicted (per NUpIpE-SW, Ref. 6) versus ceasured support reactions (f:es I?RI
' dc OO
' 2.
test 1411, Ref. 3) as shown in Tigures 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 (Figures 13 and 14) and experimental measurements (Figure 15). Nedal force results are compared to hand calculations in Table 2.
The STEHAM generated forcing functiets and nodal pressures compara f avorably with the RELAP5/ MOD 1 predicted forcing functions and nodal pressures.
3.. REFERENCES
- 1. "Steamha=mer Analysis for Piping Systems", ME-167, (STEHAM)
Version 02, Level 01, created 83.047 by SVEC.
l l 2. " Verification of ME-167 STEEAM Computer Code", SVEC Cal. No. 574.554.1-N7(3) 090-73, by A. Esi and L. Loveridge, dated 3/7/83.
l
- 3. " Application of RELAP5/ MOD 1 for calculation of Safety and Relief Valve March 1932, by Discharge Piping Hydrodynamic Loads", Interim Report, Intermountain Technologies, Inc., Idaho Falls, Idaho, Project Laager R. K. House.
4 Progelhof, R. C. and Owczarek, J. A. , "The Rapid Discharge of a gas frem a cylindrical Vessel through a Nozzle", A1AA Journal, Vol. 1, No 9 Sept. 1963, PP. 2182-2184.
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'5. Protelhof, R. C. and overarek, J. A. , "The Rapid Discharge of a Cas freu a Cylindrical Vessel through an Orifice", ASME Paper No. 63-VA-10,
- 6. SVPIPE-SW, ME-110 V03, L14 (created 82.095), " Computer Code f or Stress Analysis of Nuclear Piping".
- 7. SWEC Calculation 574.551.1-NP(B)-082-FD, Rev."1, " Comparison of EPRI l SRV Test Results With Predicted Values from WATSLUG and STEHAM,"
l Dated July 18, 1984 i
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INPUT DATA TOR STIEAM
- TOTAL INSIDE TRICTION PIPE No. LENGTH (TC)_ DIAMETER (Tt) TACTOR l' 16.125. 0.408 0.0087 2 12.563 0.5054 0.0059 3 63.562 0.948 0.0077 VALVE CHARACTDISTICS OFFICE OPENING DISCEARGE FLOW AREA (Ft i TIME (Sec) COETTICI E T RATE (1bs/hr) 0.011 0.94 440,000 0.0253 UPSTRZAH STIAM CONDITIONS PRESSURI PRESSURZ' (PSIA)_ DENSITY ( ) ' RISE RATE ( )
2410. .7.156 262.5 DOWNSTREAM CAS CONDITIONS PRISSURZ (PSIA) -TEMPERATURE DENSITT (1bs)
Tc3 14.7 70 F (530 R)- 0.075 NOTES:
- SEE TICURE 1 TOR SKITCH OF SIZEAM MODEL
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?> u? r 2 NCCAL FORCT COMPMtISON Diameter D=0.25 ft Area A = rDs/g . o.oggog7u gt:
Nodal Force = (p +ev ajg ) A - p,,, A p= pressure 1b/f te J = density lb/fts V= velocity ft/sec .
9= gravitational constant 32.2 a 14.7 x 144 lb /ft:
p, ,,
At time t = 0.00650 sec.
I Force (1b)
Node Pressure Velocity Density Force (1b) (Hand No. fesia) (fos) (1b/ft3) _l,STl!l1rW4) Calculation) 1 22.523 0.0 0.23954 186.57 186.67 5 42.785 5.7843 0.24076 198.43 198.53 10 24.231 31.219 0.2s647 209.00 209.11 15 u7.003 78.172 0.25737 230.62 230.75 20 50.214 129.89 0.26979 257.84 257.97 25 52.095 159.43 0.27697 278.93 275.06 30 52.209 161.97 0.27742 276.09 276.23 35 52.168 162.21 0.27731 275.83 275.97
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- 2. -PROGRAM VERIFICATION J' ' The VATAIR Program is a computer code developed by SVEC. This program determines the dynamic forcing functica induced by the hydraulic transient with' trapped air.
- 3. REFERENCE SWEC Calculation 11986.18-NP(B)-002-FC, Rev.0, Pages 43-52
- conditions. report submitted August 1, 1982 (Reference 4). In the Overpressure R Protection Report submitted May 9, 1985 (Reference 5), the peak pressure and pressurization rate were revised to 2,665 psia and 56.5 psi /sec. Redo the thermal-hydraulic and structural analyses with.the new values or justify that-the results of the current analyses are conservative enough that they bound
RESPONSE
The thermal-hydraulic analysis performed is based on Safety Valves PR-SV-181 and 182 opening'during.an. upset condition. The pressure in the pressurizer is initially at 2,500 psia and rises at the rate of.34.2 psi /see to a peak pressure of'2,505. psia at which point both valves simultaneously opened. The
-following safety valve parameters were used in the.anslysis: j Dresser.31719A 2" Inlet x 4" Outlet Valve
-Orifice Area = 0.994 in Valve Opening Time = 0.015 seconds Valve Flow Rate = 130,000 lbs/hr (steam)
Following this analysis, PR-SV-181 and 182 were replaced with new valves with the following parameters:
2 l'. Dresser 31719A valves with No. 1 orifices (0.994 in ),
2' . Valve flow at 2,485 psig set pressure = 125,000 lbs/hr (nameplate) with 3 percent accumulation.-
i 3.. Valve flow at 2,560 psig set pressure = 129,896 lbs/hr (nameplate) with 3 percent accumulation.
l
.5855R/12.159
________=_
/1 The peak pressure of 2,665 psia in the Overpressurization Protection Report corresponds to an expected peak pressure.of 2615 psia at;the valve. The peak {
q Lpressure of.2,615 psia is a 1.6 percent increase over the nameplate pressure -l of-2,560.psig.1 Combining this with the 128 percent of rated flow noted in I Response 7 yields an increase in expected maximum flow of 30 percent. l This. gives approximately 30 percent higher ma:.imum flow than what had been usedLin the thermal-hydraulic analysis. With the same valve opening time of. .i l
0.015 seconds', the. fluid transient forces and related pipe stresses and pipe !
support loads would be expected to be approximately 30 percent larger.
The piping, pipe supports, and pressurizer. nozzles were evaluated for the s
- increased loads. The results of the evaluations of the pipe supports are included in the response'to Question 11. The results of the evaluations of j I
the nozzles are included in the response to Question 9.
The maximum pipe stress for Pressure + Deadweight + S/RV Discharge is {
22,386 psi. The allowable per the ASME B&PV Code, Section III-NC, is 1.8 S h or'28,620 psi. !
u i'
.The maximum pipe stress for Pressure + Deadweight + SRSS (S/RV Discharge +
' SSE) is.24,127.. psi. The allowable per the Systematic Evaluation Program (SEP) l
' Topic III Safety Evaluation Report
- is 1.8 S h r 28,620 psi.
I The effect of the increased pressurization rate of 34.2 to 56.5 psi /sec is i i
considered negligible for this steam discharge transient. y
(
a l
d i
- Letter, USNRC to'YAEC, dated July 16, 1987, NUREG-0825, Section 4.11, Seismic Design Considerations
_9_
. 5855R/12.159 l
L .
QUESTION NO. 7 Discuss the values of the important input parameters used in the EPRI/CE ;
verification calculations for STEHAM and WATAIR. Compare them with those used l in the plant-specific analyses. If they are different, provide justification that the differences do not invalidate the plant-specific analyses. The Dresser 31739A valve passed in excess of 118 percent of rated flow; justify use of the ASME-rated flow in the thermal-hydraulic analyses or provide.the l results for thermal-hydraulic and strubtural analyses which account for the !
larger flows seen in the tests. Compare the rated flow and test measured flows for the test valve used to qualify the Yankee plant PORV. If the test flows exceeded the rated flow for the test valve, provide the same information l for the PORV that was requested above for the safety valves.
RESPONSE
The input parameters used in the EPRI/CE verification calculation for STEHAM are presented in Attachment 5A which is part of the response to Question 5. ;
In this verification, EPRI Test 1411 was simulated. The important input .
parameters used in the Yan'kee plant analysis are presented in Attachment 7A.
As can be seen from both Attachments 5A and 7A, the input data are very ,
close. Hand calculation methods were employed for the verification of the k 3
WATAIR program.
The Dresser 31719A valves used in the Yankee plant were tested by Yankee Atomic and the results indicated that the valves could pass up to 128 percent of rated flow. This higher value will be considered in the thermal-hydraulic j response in Question 6. "
The increased flow rate for the PORV is not significant because the maximum PORV thermal-hydraulic forces occur when the rupture disk opens.
1 l
5855R/12.159 L _ __ _ _ - _ _
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Cuc l.I imT,4L WEssume = 117ps,4 lum 4L Taxpr 47vec = Slo *R N
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g QUESTION NO. 8' The load combinations analyzed did not include a. load combination that
^ combined the occasional icad with the. Operational Basis Earthquake (0BE).- The
~
load combinations analyzed will bound this case' if the Safe Shutdown Earthquake (SSE)= response spectrum is,twice that assumed for the OBE. Discuss whether'this is true.'. The description on the lumped' mass spacing.of the structural model did not give enough details to adequately determine whether
.the' structural response was: adequately' calculated. Provide more. details, (i.e., node spacing, time step, etc.).
~ RESPONSE An OBE.has not been developed for Yankee, Nuclear Power Station. Load l combinations for piping and pipe supports, which include seismic loads, have
' lbeen1 reviewed and approved by the NRC staff as part of the review for
- Systematic Evaluation Program' Topic III-6, Seismic Design Considerations.*
' ~
C The node: spacing of the' piping model is clearly shown on the. pipe stress.
isometric.provided in Yankee Atomic Electric Company's 1985 submittal **. The
-seismic analysis of the piping was performed using amplified response spectra; I therefore, time step is not applicable..
Further details of the piping analysis are available in the SWEC calculations which~are available at Yankee Atomic Electric Company in Framingham.
1 I
i 1
'* Letter, USNRC to YAEC, dated July 18, 1987, NUREG-0825, Section 4.11,
)
Seismic Design Considerations )
1
.. . . . __ A
QUESTION NO. 9 Pressurizer nozzle loads during safety valve and PORV discharge were not discussed. Compare the calculated and allowable loads for the pressurizer nozzles.
RESPONSE
The pressurizer nozzles were evaluated for the load combination of Pressure + Dead Load + Thermal + SRSS (Earthquake + S/RV Loads).
The results are as follows:
Stress Allowable 16,101 psi (Pm) 40,050 psi (1.5 S m) 21,134 psi (Pm + Pb ) 48,060 psi (1.8 S m)
I i
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l 5855R/12.159
. QUESTION NO. 10 The table included in your response to our request for information indicated the low temperature overpressure protection transient analyzed with WATAIR considered two relief valves opening. Since the Yankee plant has only one'
-PORV, this does not seem correct. What is the correct information for the table?
RESPONSE
The table submitted previously was incorrectly labeled for Case 4 - LTOP (Low-Temperature Overpressurization Protection). It should be one relief valve
, , , instead of two since the Yankee plant has only one PORV. Attached is the revised table.
i i
I 5855R/12,159 I
i.
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T a f h a A 5) 0,4 ) ) )
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PS I
LS AY CL I A y m TN l a I A s e R u t CE f i
som eea s
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e g
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9, 9
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t 't vvb h n oh al 7 1 7
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- 2
1 I-l QUESTION NO. 11 The licensee's responses did not. provide a complete' discussion of the supports-analyses.; They simply stated that the supports were analyzed by Teledyne Eng'neering'i Services to the AISC Code (seventh edition)'. Complete details: of .
these analyses should be provided. This information should include, but not
.necessarily be limited to: s a.- A complete description of the analytical techniques and computer codes' 3 (if applicable) used.
- b. A complete. description of the load-cases analyzed and the methods used to combine them.
c.. Tabulated'results comparing maximum stresses and allowable values for all
< supports and alll load' cases recoma nded in the EPRI submittal guide.
- d. Assurance thatithe factors of safety required by IE Bulletin 79-02 are-maintained for alil load case. combinations for-all supports using concrete
. anchor bolts. The required factors of safety are 4.0 for wedge-type anchors and 5.0 for sleeve-type anchors.
e.- Documentation. clearly showing appropriate verification of any computer codes used in~the supports analyses.
RESPONSE.
- a. The majority of supports analyses were performed using hand techniques. j STAAD-III GTSTRUDL, and BASEPLATE II were used for some support _
analyses. Further details of the support analyses are contained in the calculations, which are available at YAEC's office in Framingham.
. - .. 1 b.- The'following load combinations were evaluated: J
- 1. D + 0C0
- 2. D - 000 5855R/12.159 J
- 3. D +.T m + 000 + FL
!" J4. D+TMIN'-.000: ..FL 5 '. :D+E 6.. D -- E '
7..
D-+ T. MAX + S SS( + 0C0) + FL s .
- 8. : D.+ TMIN - S SS + OCC) - E -
where D- = deadweight TMAX = algebraic maximum' thermal load TMIN = nigebraic minimum thermal load OCC = total S/RV' loads !
.FL
= friction load E. = seismic loads based.on'NRC. spectrum
-)
All supports were'thenl evaluated for the algebraic maximum and minimum. !
'from the above load combinations.
Member stresses / loads were compared to the allowables from the,AISC
' Manual of Steel Construction, Seventh or Eighth-Edition, or manufacturer's allowable loads.as-appropriate.
1
- c. The maximum ratio of actual to allowable is provided for-each support in Table /11-1.
- d. All concrete expansion anchor b'lts o were evaluated for the maximum and minimum load combination from Item b above. Allowable anchor bolt loadsE l
.were be ed on the manufacturer's published ultimate loads with a factor of' safety' of four 'for wedge-type anchors and a factor of, safety of five for sleeve-type anchors.
i Anchor bolt interaction ratios are provided in Table 11-1.
o ;
5855R/12.159 1
-)
, i i
-} ;
i' p e. All computer programs used by Teledyne Engineering Services (TES) were l verified and documented as required by the TES Quality Assurance Program. The TES Quality Assurance Program meets the requirements of 10CFR50, Appendix B, 10CFR21, ANSI N45.2, ASME BPVC Section III, and' NCA-4000.
All computer programs used Yankee Atomic Electric Company have been verified as required by Yankee Atomic Electric Company's Engineering Manual.
c.
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5855R/12.159
TABLE 11-1 Support Maximum Ratio Maximum Anchor Bolt No. Actual / Allowance Interaction Ratio Notes BRL-H-3 B 0.74 0.96 All Loads (D, TH, FRIC, SSE, SRV)
Multiplied by 1.3 l
BRL-G-10B 0.52 0.40 q BRL-R-27 0.65 0.17
-BRL-SNB-21 0.94 0.52 j BRL-SNB-22 0.72 0.81
-3RL-SNB-23 0.86 0.21 BRL-SNB-24 0.86 0.16 BRL-R25/R26 0.52 0.48 BRL-A-10 0.99 0.97 BRL-A-11 0.92 0.85 BRL-C-1 0.97 --
Pipe Strap Compared ;
With Level C Allowance !
BRL-G-2 0.95 -- Pipe Strap Compared With Level C Allowance BRL-G-3 0.89 --
.BRL-G-11A 0.14 0.20 5855R/12.159
~
q
-QUESTION NO.' 12a LThe stress analysis information provided seems to indicate that the piping' systems are adequate; however, additional information is required to confirm ,l structural adequacy. 'The following information should be'provided:
a.: Clarification of'the system geometry and components. The response'to
. Question 13(d) in-the. licensee's November 22, 1985 submittal
'(Reference 1) mentions both pressurizer and relief' tank. connections.;
however, tDe isometric. sketch included in this submittal and in the 'i April 1, 1983 (Reference 6) submittal do not show a relief tank. These sketches show a rupture disk at the end-of the discharge pipe.-
2 RESPONSE'-
The isometric, sketch in the April 1, 1983 submittal is correct. The two ,
i pressurizer. safety valves and the relief valve all-discharge into a common -;
pipe. 'This piping is cut and. terminated with a rupture disk which will-rupture and discharge-into the containment if any one of the three valves 1 lift. There is no relief tank inside containment. !
A,small 1/2" line is provided around the rupture disk. This line is installed to keep'the line drained, and to direct any valve leakage to a tank outside of
. containment. This tank served as a relief tank for the pressurizer safety and relief valves at one time, but no longer.
l l
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5855R/12.159 1
L' I
i 1
j 4 q
l CUESTION NO. 12b-p
~A discussion describing whether actual stiffness or (artificial) stiffnesses j' were:used'to represent. supports'.and nozzle connections. A justification for the'use of artificial stiffnesses should be provided if this approach was used.
RESPONSE-i
' Supports, anchors, restraints, as well as connections to'the equipment, are
'a modeled as elastic springs in the NUPIPE-SW piping model. (Reference 1)
'These supports are generally assigned generic stiffness values from EMD 80-02
-(Reference 2) for evaluation of pipe stresses, pipe support loads, and equipment reactions. These stiffness values'are shown on Table 1. Guidance .
1 from Reference 2 is.also utilized'in providing c<erload margin to protect equipment nozzles and also for the use of judgmental or calculated stiffnesses ,
1 in special cases.
The generic stiffness values in EMD 80-02 are considered to be conservative i
-and reasonable values for seismic evaluation of piping systems.
The values are reasonably stiff and were arrived at based upon engineering
~
judgment, experience, and evaluation of common supports. ;
The generic'stiffnesses assure satisfactory pipe stress values and support loads for thermal growth and differential anchor movements. Adequate dynamic responses are assured by using conservative input load designations, load
< combinations, and low damping assumptions.
REFERENCES l -
- 1. Stone & Webster Engineering Corporation Computer Program NUPIPE-SW. .)
i.
L 2. Engineering Mechanics Division (EMD) Memorandum, EMD 80-2, " Stiffness
' Representation of Supports, Anchors, and Restraints for Pipe Stress i
Analysis and Support Design." l 5855R/12.159 ,
f w _ --_ .
s.< j f- -
" ]
TABLEJ. 1
[ SUPPORT STIFF:iESS REPRESENTATIVE VALUES
- SUPPORTS AND RESTRAINTS 1
Transnational Rotational {
Nominal Pipe Size. Inches Stiffness ib/in- Stif fness in-lb/ rad. .
l 1ess than 6" 2X105 1X107 6" 'o t 14" 1X100 1X108 greater than 14" 2X106 to 5X106 1X109 SHOCK SUPPRESSORS (Snubbers) l Hydraulic Total Load Range (1b) Stiffness 1b/in less than 5000 1X105 5000 - 32000 2.5X105 - 4.5X105
>32000 -' 72000 OX105 greater than 72000 1.0X106 - 1.5 X106 Hechanical*
Total Load RanRe (1b) Stiffness 1b/in less than 650 9X103
-650 - 6,000 6X104 i
> 6,000 - 15,000 - 2.4X105
>15,000 - 50,000 7X105 7 50,000 - 120,000 + 1X106
- 1. Mechanical Stiffness are based upon Pacific Scientific Data.
- 2. For Snubbers in parallel, double stiffness values.
An chors *
- Transnational Rotational Anchor Type Stif fness ib/in S tif fness in-lb/ rad.
Welded structure. 1X106 - SX106 2X108 - 1X109 Major Equipment connec tions 1X109 1X10ll
-(steam generator, pumps l penetrations)
Fabricated Weldments (Manifolds, condensers, etc. 1X107 1X109 In-line Anchor 1X108 1X1010
- For all points coded AMCHOR, HUPIPE assigns values of 10**8 for transnational and 10**10 for rotational stiffnesses. In order to override these values, follow the ANQIOR card with a RLSTRAINT card and put the desired stif fnesses on that RESTRAINT card.
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- -QUESTION NO. 12c A discussion of the earthquake spectra used in the seismic analyses. It is understood that two different spectra were developed during efforts to respond to Systematic Evaluation Program (SEP) issues. The licensee should clarify i
whether the more severe "NRC" spectra was used in their responses to the subject NUREG. If the more severe earthquake spectra were not used in the results included in the submittal, the licensee should provide a comparison of the differences in stress levels (for both piping and supports) that would result from the inclusion of the higher spectra.
RESPONSE
The more severe NRC spectra was used for the seismic analysis of the S/RV piping.
5855R/12.159
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. QUESTION NO. 12d ,
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.'A more detailed' discussion'of the mass lumping criteria used in the piping.
i analyses.- For example, this discussion should address whether the analysis o . ;>i used" automatic mass' lumping or.whether minimum spacing requirements were hand
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-calculated,.'how many. modes were found below the 400 Hz cutoff, etc. The discussion should provide enough detail to: permit verific'ation that enough' modes were calculated t'o achieve accurate modal superposition results.
W RESPONSE '
.q..
The location of the lumped masses are-preselected byi the analyst. Typically, the-model will contain at least three mass points between restraints active in' the samed 'irection. Sufficient modes were analyzed (cut off modes = 144) to ensure.that the system response' included all significant dynamic loadings.
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If the loading for the: axial support were less than the maximum segment inertia j force':then the support' design load would have been increased-to the maximum-
-segment inertia force. Stone & Webster Calculation 11986.18-NP(B)-003-XF,
. Revision 0,' dated March 13,: 1983 contains this structural analysis.
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-5855R/12.159
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j; v k QUESTION NO. 12e A piping system isometric should be included which clearly shows support i
I locations, types, and directions.
RESPONSE
EES Drawings 80023-PI-1041, Sheet 1 of 2, and 80023-PI-1041, Sheet 2 of 2, are attached.
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i c APERTURE CARD /HARD COPY AVA' LABLE FROM RECORD SERVICES BRANCHM ~
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