ML20235G403
| ML20235G403 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 09/21/1987 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM TAC-44593, TAC-54601, NUDOCS 8709300026 | |
| Download: ML20235G403 (39) | |
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- DUKE PowEn GOMPANY P.O. DOX 33189 CHARLOTT72, N.O. 28242.
r EALU. TUCKER-retzessons i
(704) 073 4631 j;,
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= nuos. man raonuam<m September 21, 1987 U.S. Nuclear Regulatory Commission lt Document Control' Desk 7
Washington, D.C..
20755
Subject:
. McGuire Nuclear Station
' Docket. Nos. 50-369 and 50.-370 NRC.RAI Concerning Testing of Relief and Safety Valves (TACS 44593/54601)
Gentlemen:
By Mr. D.S. Hood's letter-dated May 27, 1987, the NRC Staf f requested additional information concerning Duke's. response on the subject dated March 11, 1985 and requested a response' within 45 days.
Additional time was needed to prepare a response.~as stated' in my letters of July 7, 1987 and August 20, 1987.
The re-
. quested'information is attached.
If you~ have further questions, pleese contact us through the normal ' licensing channels.
Very truly yours,
/~S$b {
1 B. Tucker JBD/215/jgc Attachment xc:
Dr. J. Nelson Grace Mr. Darl Hood Regional Administrator, Region II U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 101 Marietta St. NW, Suite 2900 Washington, D.C.
20555 Atlanta, Georgia 30323 Mr. W.T. Orders NRC Resident Inspector McGuire Nuclear Station
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9 8 REQUEST FOR ADDITIONAL INFORMATION REGARDING NUREG-0737 ITEM II.D.1 FOR MCGUIRE UNITS 1 AND 2 DOCKET NOS.:
50-369 AND 50-370 l
1 Previous responses, transmitted by W. O. parker's letters of June 30, 1982 and Novenber 1, 1982 and H. B. Tucker's letter of Juno 12, 1985, were based on analysis and reports prepared by a consultant.
Time-history loadings at supports, equipment, and piping components were not available.
From the information available, it was difficult to evaluate the supports.
To assure conservatism, maximum loadings without regard to sign were applied to the supports l
in the worst combination. Further. this analysis used a simplified linear elastic approximation to evaluate the inelastic response of Unit 1 supports shown to be overloaded by the maximum combined loads from the linear elastic analysis. This analysis was accomplished through iterative means involving many computer runs to try to get agreement between the linear assumptions and the non-linear behavior.
{
Engineering judgement based upon experience was used in this 1
analysis to approximate non-linear response of some components and to evaluate convergence of the response.
The piping was re-coded and additional analysis of the existing configuration was performed to obtain detailed time history results allowing more realistic support evaluations.
It became clear the only way to fully evaluate the simplified linear elastic model results and judgements would be to perform a detailed non-linear time history analysis.
In the process of refining the structural model, reviewing the analytical techniques. tnd waaluating the supports using detailed time history resuita, it was concluded the overloaded supports on Unit 1 can be eliminated, based on linear elastic analysis, by the removal of one support, the addition of one new support, and the minor modification of one other support.
It was decided that these modifications would be more cost effective than the time-consuming and costly non-linear time history analysia, documentation, verification, approval, and regulatory review process which may have been required to thoroughly prove the cccclusions drawn from the previous approximate analysis.
As a result of this linear elastic reanalysis and the removal of one support, addition of one srpport, and modification of one other support, some information provided in previcus responses iLor Unit 1 has changed.
The updated information is included in this submittal. All information previously provided on Unit 2 remains unchanged.
The support arrangements for Unit 1 and Unit 2 are equivalent.
For Unit 2, structural changes were made to meet allowable stress limits for the supports.
For Unit 1, this was not feasible as described in our previous
.e.
page 2 responses. Therefore, the criteria used for support
)
qualification on Unit 1 are difforent from'those used on Unit 2.
These differences are discussed in the appropriate responses.
The re-analysis for Unit 1 is based'on 2% damping without yielding supports. As such, the stress results for Unit 2 envelope the Unit 1 stress results.
Unit 1 is currently in a refueling outage. The three modifications deceribed above will be completed during this outage. We believe this change and the corresponding elimination of need to evaluate yielding supports will significantly expedite the acceptance process.
We believe'that:
(1) the original approach of using a linear elastic iterative analysis to approximate the non-linear problem was sound; (2) the engineering judgements were reasonable; and (3) the system is/was functional per all requirements (codes, FSAR, NUREG-737, etc).
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_ _ _ _ _ _ _ _ _ _ _ _ _ __________________ _ _ _ __J
RESPONSE TO QUESTIONS TRANSMITTED 9/11/87 DAVE HOOD'S LETTER OF MAy 27, 1987 Page 1 Question 1 Question 11(a) asked the licensee to identify the computer programs used in the analyses and the metnods utilized in their verification. The response states only that the computer code SUPERPIPE was used for the piping analyses and that this code was verified in accordance with the methods in the McGuire FSAR, Section 3.9.2.3.
The following information will be required to complete the review:
a.
A discussion of the applicable FSAR require-ments and how activities (such as benchmark analyses) performed by the vendor and/or the licensee satisfy these requirements for all computer codes used in *he piping and r--
supports analyses.
b.
Clarification as to which groups (vendor /
consultant or licensee) utilized the codes.
c.
A description of the analytical methods and computer code (s) (if applicable) used for the analysis of piping supports.
RESPONSE
a)
The computer codes used in the analysis of the pressurizer safety and relief valve system, authors, description and method of verification of these computer codes follows:
Computer Programs Used in Analysis Code:
SUPERPIPE Author:
EDS Nuclear, Inc. (Now Impell) 220 Montgomery Street San Francisco, California 94101
==
Description:==
SUPERPIPE is a computer program for the structural analysis and code compliance evaluation of piping systems, with particular emphasis on Class 1, 2 and 3 nuclear power piping designed to meet the requirements of the ASME Boller and Pressure Vessel Code,Section III.
s The piping system may be modeled by straight pipes, curved pipes, elbows, reducers, tees, branch connections and other commonly-used piping attachments.
I..
r RESPONSE: (Continued)
Page 2 principal features of the program include the following:
1)
Static and thermal stress. analysis by the direct stiffness method.
11)
Frequencies and mode-shapes computation, e
using the subspace interaction method or the Q-R method.
iii)
Dynamic response analysis by the
{
,4 modal-superposition time-history analysis, 1
the direct-integration time-history analysis or the response spectrum analysis.
iv)
Seismic analysis of multiple-excited piping
.q systems by the " multiple-response-spectrum approach."
v)
Combination of directional and modal responses according to USNRC Regulatory Guide 1.92, vi)
Code compliance evaluation according to ASME Boiler Pressure Vessel Code,Section III, for Class 1, 2, and 3 nuclear piping (user's choice of code addendum).
i vii)
Determination of pipe-break locations.
I viii)
Fatigue damage computations.
ix)
Built-in library of standard material proper-l ties, cross-sectional properties, flexibility factors, and stress intensification factors.
x)
Restart options to store, recall, and modify piping geometry, and to perform analysis and code compliance evaluation in stages.
xi)
Options to print-out support loads and dis-placement summaries, nozzle and penetration load summaries, and code compliance summary.
xii)
Options to plot piping geometry and mode shapes.
Extent and Limitation of its application: All routines of the SUPERPIPE program are used as detaile6 in the description.
Verification:
The program has been bench-marked against the EDS program PISOL and other programs like NUPIPE and PIPESD.
This program has been verified by bench-marking to an ASME sample problem, by comparison to detailed analysis performed manually, by comparison to results achieved using similar programs, as described above, and by comparison to results achieved using the previous version of SUPERPIPE.
The bench-mark problems specified in NUREG CR-1677 have been evaluated using this program and the results have been transmitted to the NRC.
Code:
MCAUTO STRUDL page 3 Author:
McDonnell Douglas Automation Company i
Box 516 St. Louis, Missouri
==
Description:==
Large scale general purpose finite element program for structural analysis.
Extent and Limitation of its application: MCAUTO STRUDL is used to perform static elastic analysis of pipe supports.
Verification: MCAUTO STRUDL has been verified by comparison of the results with either hand calculations, closed form solutions found in standard text books or solutions from other programs.
Code:
EDSFLOW Author:
EDS Nuclear, Inc. (Now Impell) 220 Montgomery Street San Francisco, California 94104
==
Description:==
EDSFLOW is a computer program for the analysis of the thermal / hydraulic behavior of light water reactor systems subjected to postulated transients such as those resulting from loss of coolant, pump failure, or rapid depressurization.
It is a modified version of the RELAP4/ MOD 5 program developed for NRC by the Idaho National
' Engineering Laboratory.
Extent and Limitation of its application:
EDSFLOW is used primarily for time-history analysis of hydraulic forces on piping systems during rapid transient (i.e. safety valve blowdown).
It is also used for transient containment building subcompartment pressure / temperature analysis.
Verification: EDSFLOW has been verified by comparison to RELAP4/ MODS thermal / hydraulic predictions for two pressurizer safety / relief valve discharge transients and one PWR main steam hammer analysis.
Code:
ANSYS Author:
Swanson Analysis Systems, Inc.
- p. O.
Box 65 Houston, Pennsylvania 15342
==
Description:==
Large-scale finite-element program for structural, heat transfer and fluid-flow analysis.
ANSYS performs linear and nonlinear elastic analysis of structures subjected to static loads (pressure, temperature, concentrated forces and prescribed displacements) and dynamic excitations (transient and harmonic). The program considers the effects of plasticity, creep, swelling l
Page 4 and large deformations.
Transient and steady-state heat transfer analyses consider conduction, convection and radiation effects.
Coupled thermal-fluid, coupled thermal-electric and wave-motion analysis capabilities are available. Structural and heat transfer analyses can be made in one, two or three dimensions, including axisymmetric and plane problems.
Extent and Limitation of its application:
The ANSYS computer program is used to perform static elastic finite element analysis on pipe support baseplates.
Verification: The ANSYS program has been verified by a comparison of and test problems with analytical results published in literature and hand calculations.
Code:
BASEPLATE II Author:
Richard S. Holland Ernst, Armand, and Botti Associates,Inc.
60 Hickory Drive Waltham, Massachusetts 02154
==
Description:==
The program BASEPLATE II is a preprocessor /
postprocessor to the ANSYS computer code for the specific purpose of analyzing flexible baseplates.
Extent and Limitation of its application: The BASEPLATE II program is used to analyze support baseplates.
Verification:
CDC has verified BASEPLATE II in accordance with CDC's QA program utilizing a comparison of program results to hand calculations, published analytical results, or another program which has similar capabilities.
b) The computer codes listed above were used by Duke Power and a consultant (EDS) as shown below.
UNIT 1 UNIT 2 Group Code Group Code EDS and Duke SUPERPIPE EDS SUPERPIPE EDS EDSFLOW EDS EDSFLOW DUKE STRUDL DUKE STRUDL DUKE ANSYS DUKE ANSYS DUKE BASEPLATE II DUKE BASEPLATE II l
c)
Piping supports are analyzed and designed using elastic methods in accordance with standard engineering practice.
In general, hand calculation methods are I
used; more complicated structures may use the STRUDL l
computer program. Finite element analysis programs utilized particularly for anchor bolt / baseplate qualification include ANSYS and BASEPLATE II.
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F.
Question 2 pege.
The licensee's response to question 11(d) is considered inadequate. The following information should be provided:
a.
A complete discussion of the load case combinations used for the pipe support analyses for both Units 1 and 2.
b.
A complete discussion of the codes and standards governing the pipe support analyses (note that referral i
to the FSAR is not considered adequate).
c.
A complete discussion of the methods used to combine transient loads with other dynamic loads such as SSE, OBE, etc.
RESPONSE
a.
Support load combinations and allowable support member stress criteria used for Unit 2 supports are shown in Table 2.1.
These load combinations and stress criteria were used for both Duke Class A and non-code supports.
The original support qualification for Unit 1, based on SRV thrust loads without NUREG 0737 considerations, l
also utilized the support load combinations and allowable member stress criteria in Table 2.1.
An additional evaluation performed for Unit 1 considers the NUREG 0737 thrust load in the following load combination and stress criteria for pipe support analysis as described in the previous response:
Load Combination:
pressure + Weight + Thermal Displacement + SRV Thrust (New)
Stress Criteria:
(primary Stresses)
< Yield Stress at Operating Tempera-ture
Page 6 b.
The following codes and standards govern pipe support design for Units 1 and 2:
1.
MSS SP-58, 1967, Pipe Hangers and Supports Materials and Design 2.
ANSI N45.2, 1971, Quality Assurance Program Requirements for Nuclear Power Plants 3.
ANSI N45.2.11, 1974, Quality Assurance Require-ments for the Design of Nuclear power Plants 4.
AWS A2.4-79, Symbols for Welding and Nondestructive Testing 5.
AWS A3.0-1980, Welding Terms and Definitions 6.
AWS D1. 0-69, Code for Welding in Building Construction 7.
AISC Manual of Steel Construction, 7th edition 5
(1969) 8.
MSS SP69, 1966, Pipe Hangers and Supports Selec-tion and Application c.
The blowdown loads were not combined with other dynamic loads for Unit 1.
The blowdown loading is highly localized and has a very short duration. The probability of peak seismic loads I
occurring simultaneously with peak blowdown loads is extremely small even in the unlikely situation of both events occurring simultaneously. A seismic event is not an initiator for an event that results in safety valve discharge.
The blowdown loads were combined with eeismic loads for Unit 2.
Question 3 Question 11(e) requested information regarding the results of the structural analyses.
The licensee's response did not provide sufficient information. The following should be provided:
a.
Tabulated data showing the maximum stresses compared to the Code allowable values (ASME or ANSI, as applicable) for both upstream and downstream piping for both units.
This should include data for all load cases and complete Class 1 data (Eq. 10, 11, usage factors, etc.) where applicable, b.
Tabulated data showing the maximum stresses compared to allowable values for all pipe supports (upstream and downstream) for all load cases for both units. As an alternative, support capacities may be listed with the appropriate maximum capacity for the individual load cases.
Page 7
RESPONSE
UNIT 2 a.
. Analysis results show that upstream piping is acceptable in accordance with NB-3600 of ASME Section III code as shown in Table 3.1.
Downstream ANSI B31.1 pipe meets the ASME Section III Class 2 allowables as shown in Table 3.2 All support loads are within the stated allowable loads as shown in Tables 3.3 and 3.4.
l Unit 1 The Unit i revised analysis is based on 2% damping and Unit 2 uses 1% damping. The support arrangement is equivalent to j
the Unit 2 support arrangement and the pipe geometry is the l
same. Due to the addition of one support, removal of one support and modification of one support, all supports are qualified for the blowdown loads. Therefore, there is no need to approximate yielding supports and the results of the blowdown analysis are similar for both units.
Unit i revised analysis results are enveloped by the Unit 2 analysis results presented in Tables 3.1 and 3.2.
b.
Tabulated data for support capacities versus support loads along with anchor bolt safety factors (as applicable) are as shown in the following tables:
Table 3.3 - Unit 2 S/R Capacities vs. Loads / Duke Class A Supports (Upstream of Relief Valves)
Table 3.4 - Unit 2 S/R Capacities vs. Loads /Non-Code Piping Supports (Downstream of Relief Valves)
Table 3.5 - Unit 1 S/R Capacities vs. Loads / Duke Class A Supports (Upstream of Relief Valves)
Table 3.6 - Unit 1 S/R Capacities vs. Loads /Non-Code Piping Supports (Downstream of Relief Valves)
L'*
page 8 I
Question 4-In the response to question 11(e) it is stated that the stresses in the 12" tee exceed allowable values but.
that the stresses will redistribute and strains will be small. The licensee should clarify whether these conclusions were based on engineering judgement or if a specific analysis exists to demonstrate these~results for this component.
-A supporting discussion should be included.
RESPONSE
. Based on the previous approxusato analysis, the conclusion that stresses in the 12" tee would redistribute and strains would be small was based on engineering judgement. As the tee is loaded, yielding would begin in local areas due to structural discon-tinuities. Calculated stress included an SIF of 2.17-l based on fatigue evaluation.
The tee would continue to carry more load, with negligible strains, until more general yielding began to occur with stresses approximately'twice as high. As generalized yielding begins, the ability of the tee to carry additional moment rapidly decreases. The tee would deform slightly and the forces would be carried by the stiffer nearby components. The deformation would be very small and would not affect flow through the pipe.
Based on the results of the revised analysis, the 12" tee is not overstressed. Maximum stress ratio is 0.214.
Question 5 The licensee should clarify whether any support loads on supports using concrete anchor bolts increased to the point where the required factors of safety in IE Bulletin 79-02 are now exceeded.
These factors are 4.0 for sleeve type anchors and 5.0 for wedge type anchors.
RESPONSE
i The anchor bolt safety factors as defined by IE 1
Bulletin 79-02 are as follows:
5.0 for self-drill I
anchors, 4.0 for sleeve and wedge anchors. No self-drill anchors are used on the subject supports on either Unit 1 or 2.
q l
page 9 Anchor bolt safety factors are given in Tables 3.3 and 3.4 for Unit 2 supports and in Tables 3.5 and 3.6 for Unit 1.
The safety factors listed are based on the applied load..For the non-code pipe supports where the anchor. bolts controlled the capacity, the capacity is based on maintaining the following safety factors: 4 for Normal, 3 for Upset, and 2 for Faulted. The justification for the use of these safety factors is based on the extremely short duration of the discharge event. There is one support on the Unit 1 Class A piping where the bolt safety factor is less than 4.0.
The bolt safety factor is 3.69.
Again, the justification for the acceptability of this is the extremely short duration of the discharge event. All Unit 2 Class A support bolt safety factors are greater than 4.0.
Question 6 The June 12, 1985 response to question 6 was confusing.
A comparison between the maximum predicted woments on the safety valve and PORV and the maximum moment applied to the valves during the EPRI/CE tests was requested. The response by Duke power relative to the safety valves was " Allowable lateral bending moment was taken to be 298,000 in-lb.
The allowable bending moment is an envelope of the maximum moments from the EPRI tests that most closely 4
represent the McGuire S/RV System. The actual end moment was found to be 295,000 in-lb".
It is not clear where the actual end moment came from.
State what the maximum predicted moment on the safety valve is, including the effect of thermal, gravity, seismic and valve discharge loads. Also provide the corresponding corparison for the PORV.
The statement "The operability of the PORVs and the stop valves was also verified by qualification analysis" is not acceptable.
Page 10
RESPONSE
I Unit 2 SAFETY VALVES i
COMPARISON OF C-E TEST VALVE MOMENTS TO PLANT-SPECIFIC ANALYSIS I
EPRI TESTS ANALYSIS 1
Lateral Moment (1)
Maximum Lateral Moment (in-lb)
(in-lb) i Test Number 908 1406 i
Valve Inlet 400,000 330,000 131,604(2) 149,208(3)
{
l Valve outlet 298,780 286,800 261,545(3) 266,832(3) l l
POWER OPERATED RELIEF VALVES COMPARISON OF C-E TEST VALVE MOMENTS TO PLANT-SPECIFIC ANALYSES EPRI TESTS ANALYSIS i
Lateral Moment (1)
Maximum Lateral Moment l
(in-lb)
(in-lb)
I i
Test Number 47-CC-3S Valve Inlet /
39,000 16,080(2) 36,588(3)
Outlet NOTES:
(1) Moments are perpendicular to the plane of pipe configuration.
(2) This is the vectorial resultant of the two orthogonal moments.
Load combination is weight + pressure + blowdown.
(3) This is the vectorial resultant of the two orthogonal moments.
Loading combination is weight + pressure +
thermal + blowdown.
(4) The Unit i reanalysis results are enveloped by the Unit 2 results.
)
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.Page 11 The loading combination that most closely resembles the:
.EPRI Test is weight + pressure + blowdown.
The EPRI j
test does not predict the maximum allowable load on the-i valve, it only confirms acceptability of the measured.
load. Loading combination weight + pressure + blowdown
+ thermal is presented since itois the maximum load case.
i Seismic loads are qualified in the original analysis and are not considered here for the reasons stated in 4
the response to question 3a.
Valve operability is insured since the analysis moments essentially match or are less than the test moments.
Question 7 In the response to question 11(e), it is stated that the relief tank nozzle is overstressed in the axial direc-
)
tion but that the stresses will redistribute and be 3
carried by a support located two feet above the nozzle.
The licensee should clarify whether these conclusions were based on engineering judgement or if a specific analysis exists to demonstrate these-results for this component. A supporting discussion should be included.
RESPONSE
Based on the previous approximate analysis, the
)
' conclusion that the support two feet above the nozzle I
would carry the load was based on engineering judgement..As the nozzle yields its ability to carry additional load is small.and the load would be supported by the stiffer support close by which has the capacity to carry the excess load from the nozzle.
Based on the new analysis, the nozzle is not over-stressed. The ratio of the nozzle load to allowable j
nozzle load is 0.531.
l Question 8 In the response to question 11(e), it is stated that the results of the analyses for McGuire 1 indicate that the system is functional and that this is sufficient to demonstrate compliance with NUREG 0737, item II.D.1.
The licensee should clarify its interpretation of the NUREG requirements to support the position that complete Code (ASME or ANSI) compliance is not required.
a.
page 12
RESPONSE
The purpose of NUREG 0737, Item II.D.1 is to maintain primary coolant loop pressure boundary integrity.
This can be insured by qualification of the safety and relief valves and the Class 1 piping upstream.
Additionally, the downstream piping must be checked to insure that it doesn't adversely affect safety and relief valve operation. The acceptance criteria insures primary coolant loop pressure boundary integrity and a flow path for safety and relief valve discharge.
Discharge from the safety and relief valves is normally routed through the discharge pipe to the reactor drain tank. This tank is provided with rupture disks as overpressure protection that allow the discharge to collect in lower containment if the tank capacity is exceeded.
All portions of the upstream and downstream piping were shown acceptable per the criteria defined in the l
response to Question 3.
l l
Modifications to Unit 1 similar to Unit 2 are not I
necessary or practical because:
J.
Simpler modifications will be made that are adequate to assure functionality of the system and meet the described criteria for pipe stress and l
support design.
2.
The results of the analysis indicated that the system will function for a safety / relief valve discharge event demonstrating the adequacy of McGuire Unit 1 SRV and PORV piping and supports in accordance with NUREG 0737, Item 11.D.1.A.
3.
On 9/2/86 McGuire Unit 1 safety relief valve, INC-1, lifted while the unit was in Mode 3, Hot p
Standby. The system pressure and temperature were 2375 psig and 530 F respectively.
No damage of piping components or supports was found after inspection of all p.iping and supports.
4.
High radiation levels exist in the pressurizer cavity.
A 5.
The pressurizer cavity is very congested and there is a lack of support anchoring space.
Page 13 Question 9 Provide the maximum predicted backpressure for the safety valves and the PORVs.
RESPONSE
The maximum predicted backpressure on the safety valves is approximately 950 psia, which occurs as a short l
duration (<40 millisecond) spike immediately af ter the loop seal enters the valve discharge piping.
This pressure decays to 400 psia as the loop seal is accelerated downstream. The flow rate through the Crosby 6M6 safety valves has been shown by test to be relatively insensitive to backpressure.
The PORVs were modeled in the analysis as adjustable i
orifices.
The flow rate through the PORV is, therefore, a direct function of backpressure except when flow through the PORV is choked.
The analysis performed to address the transient mechanical loads on the discharge piping contained models and assumptions selected to provide conservatism in the size of tha loads. The assumptions are consistent with those which would have been selected had a conservative prediction of backpressure been the objective of the analysis.
Question 10 Question 10(b) requested information on how the
~
Thermal-hydraulic code EDSFLOW was verified.
The licensee did not respond. This information is needed to complete the review.
RESPONSE
See response to Question la.
0
TABLE 2.1
. Load Combinations and Stress Criteria for Supports, Restraints, and Anchors Unit 1 & 2 CONDITION LOAD COMBINATION CRITERIA
- 1. Normal Thermal Displacement AISC Normal Pressure (As Applicable)
Allowable Stress Weight
.2. Upeet Thermal Displacement AISC Normal Pressure (As Applicable)
Allowable Stress Weight with 1/3 increase for OBE (Inertia)-
seismic OBE (Displacement)
SRV Thrust
- 3. Faulted Thermal Displacement Yield Stresses At Pressure (As Applicable)
Operating Temperature Weight i
SSE (Inertia)
SSE (Displacement)
Pipe Rupture or SRV Thrust i
l
l Table 3.1 Un3t 2 (NOTE 3)
UPSTREAM (CLASS 1)
Maximum Stresses Ccmponent Joint Stress Allowable-Condition Description Name Results Stress Ratio (psi)
(psi)
Design 3/4 x 6" Branch 64L 22,734 24,120 0.94 connection I
Emergency 6" Elbow 928 31,856 32,004 0.99 (See Note 1)
Faulted 6" Elbow 102 38,019 48,240 0.79 Tapered Equation 10 6" Transition 74B 135,080 47,940 2.82 (See Note 2)
Joint Equation 12 6" Elbow 117 36,754 47,940 0.77 Tapered Equation 13 6" Transition 92B 48,086 60,000 0.80 Joint Usage 6" x 1 1/2" 111 u=0.85 u=1.0 0.85 Sweepolet NOTES 1.
The revised blowdown event was incorporated into emergency condition.
2.
Stress ratio greater than 1 is acceptable since Equations 12 and 13 are satisfied.
3.
The Unit 1 revised analysis results are enveloped by the j
Unit 2 analysis.
]
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Table 3.2 Unit 2 (NOTE 1)
Downstream (ANSI B31.1.0)
Maximum Stresses Component Joint Stress Allowable Stress Condition Description Name Results(psi)
Stress Ratio Upset 12"x6" Red Tee 17 16349 19080 0.857 Emergency 12" Tee 16 25161 28620 0.879 Faulted 12" Tee 16 34426 38160 0.902 NOTE:
(1) The Unit 1 revised analysis results are enveloped by the Unit 2 analysis.
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REVISED RESPONSE TO QUESTIONS TRANSMITTED BY E. G. RDENSAM'S LETTER OF MARCH 11, 1985 Question 11 Your submittals (Refs. 2, 3, and 4) state that a structural analysis of the safety pORV valve piping system has been conducted, but does not present details of the analysis.
To allow for a complete evaluation of the methods used and results obtained from the structural analysis, please provide:
(a) A detailed description of the methods used to perform the analysis.
Identify the computer programs used for the analysis and how these programs were verified.
(b) A description of the method used to apply the fluid forces to the structural model. Since the forces acting on a typical pipe segment are composed of a net, or " wave", force and opposing " blowdown" forces, describe the methods for handling both types of forces.
(c) A description of methods used to model supports, the pressurizer and relief tank connections, and the safety valve bonnet assemblies and pORV actuator.
(d) An identification of the load combinations performed in the analysis together with the allowable stress limits.
Differentiate between load combinations used in the piping upstream and downstream of the valve. Explain the mathematical methods used to perform the load combinations, and identify the governing codes and f
standards used to determine piping and support adequacy.
(e) An evaluation of the results of the structural analysis, including identification of overstressed locations and a description of modifications if any.
(f) A sketch of the structural model showing lumped mass locations, pipe sizes, and application points of fluid forces.
i l
(g) Identify where a copy of the EDS Nuclear, Inc. struc-l I
tural analysis report may be obtained.
Response
(a) McGuire Unit 2 The dynamic structural rerponse for S/RV discharge events was evaluated by a step-by-step direct l
integration force time history analysis using l
l SUPERPIPE.
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page 2 (a) McGuire Unit 2 (continued)
The modeled stiffness included the effect of bending, shear, axial and torsional deformations, as well as any changes due to the effects of internal pressure on curved members as required by NB-3687 of the ASME Section III Code. The masses of the pipe, piping components, valves, pipe contents and insulation were considered in forming the mass matrix.
The forcing functions from the thermal-hydraulic analysis were applied at elbows, reducers, and tees.
For the analysis, a damping value of one percent of critical damping was used. This is based on the recommendations of Regulatory Guide 1.61.
The modal damping matrix for direct integration analysis assigned arbitrarily for each mode is given as:
[c] = c([M] + p [K]
in which
[M] = mass matrix
[K] = stiffness matrix, which was calculated by considering all rigid supports and shock suppressors (snubbers) in the piping system o(, p = factors which control the amount of damping
( = 4Tf f If A -f h }
y2 22 2l f 2,f 2 p
k*b
-b l1 22 7(f -f*)
2 1
2 where f'f
= 1 wer and upper response frequencies of the 1
1 2
piping system respectively hy,k = corresponding damping ratios to frequencies f, f ' reSPectively y
2 The one percent critical damping ratio was assigned to the corresponding frequency at 10 Hz and 100 Hz respectively for f and f '
2
page 3 (a) McGuire Unit 2 (continued)
The integration time utep for the time history analysis was selected to be fine enough to include the structural response to the highest frequency noted in the load histories. A time step of 0.002 seconds was used, which is considered to be accurate for the evaluation of structural response for a duration of 1.0 second.
For the linear dynamic structural analysis, support stiffness for rigid supports or shock suppressors was assumed to be infinite.
McGuire Unit 1 The analysis methods for Unit 1 are the same as for Unit 2 except that 2% damping was used. Damping of 2%
is the value stated in Regulatory Guide 1.61 for large 1
piping (OBE) and for small piping (SSE). The majority of the pipe in this model is 12 inch.
(b) McGuire Unit 1 and 2 The general method of force history generation was to develop the total forces in the axial direction at opposing components such as bends or tees according to the following equation:
F=F
+F g
CS where F is the wave force (due to the fluid g
acceleration) and F is the control surface force in the direction of F.
In general, the wave force component tends to accelerate the entire piping segment whereas the control surface force also tends to differentially expand or contract the segment. The extent of such axial deformation is dependent upon the piping properties and the magnitude of the control surface term, but is generally small for gaseous discharge problems in S/RV piping. Total forces calculated in this fashion are intended for application at the tangent points of each bend.
Two total forces are, therefore, calculated at each direction change, or equivalently, two opposing total forces are calculated for each straight segment of piping. Where the piping has two bends back-to-back or the distance of a straight piping segment is relatively small, only a single net force is calculated and F is represented by just the wave force F.
g
Page 4 (c) McGuire Unit 2 Rigid supports or shock suppressors were modeled as infinitely rigid.
The pressurizer is modeled as a beam with the nozzles connected to the pressurizer centroid using rigid members.
The relief tank connection is modeled as an anchor since the pipe penetrates the tank and is supported at two additional locations inside of the tank.
I Safety valve bonnets and PORV actuators are modeled as rigid members with the applicable mass lumped at the bonnet or actuator centroid.
McGuire Unit 1 McGuire Unit 1 pressurizer, relief tank connection, safety valve bonnets and PORV actuators are modeled the i
I same as for Unit 2.
The supports are modeled the same as Unit 2.
(d) McGuire Units 1 and 2 Load combinations and allowable stresses are identical to the original analysis (Ref. FSAR Chapter 3) with the exception of the blowdown loading being included as an emergency loading.
1 GOVERNING PIPING CODES & STANDARDS 1.
ASME Boiler and Pressure Vessel Code,Section III,
=
1971 including Swmmer and Winter 1971 Addenda.
2.
ANSI Code for POWER Piping B31.1.0, 1967.
PIPE SUPPORT CODES AND STANDARDS (Ref. McGuire FSAR Section 3.9)
(e) McGuire Unit 2 Analysis results show that piping design is acceptable in accordance with NB-3600 of ASME Section III code.
Pipe supports for McGuire Unit 2 were redesigned for the increased loads from the blowdown event to satisfy design criteria requiring modification of approximately 9 supports.
Q-e I-page 5 McGuire Unit 1 Analysis results show that piping design is acceptable in accordance with NB-3600 of ASME Section III code.
For details, see the repsonse'to Question.3 of Supplemental Request for Additional Information 1
Regarding NUREG-0737, Item II.D.1, " Performance Testing
~ of Relief and. Safety Valves" McGuire Nuclear Station, Units 1 and 2, May 27, 1987.
(f) The structural model is shown in Attachment B.
(g) The EDS Nuclear, Inc. structural analysis report is available at DPCo/ Design Engineering Department.
l I
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