ML20077R918
| ML20077R918 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 01/17/1995 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20077R917 | List: |
| References | |
| NUDOCS 9501230196 | |
| Download: ML20077R918 (45) | |
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Table 7 Piping Design Acceptance Criteria (1)
Tier 2 Tables and Commitment Tier 2 Sectionst2) Figurest2)(3)
Intersystem LOCA 3.9.3.1 Pipe Support Jurisdictional Boundaries 3.9.3.4.1 Pipe Support Baseplate and Anchor Bolt Design 3.9.3.4 ;
l Use of Energy Absorbers and Limit Stops 3.7.3.3.1.7, Table 1.8-20 l 3.9.3.4.1(6)(a)
Use of Snubbers 3.9.3.4.1 Decoupled Branch Pipe - Displacement Criteria 3.7.3.3.1.4 Seismic Self-Weight Excitation 3.7.3.3.4 Supplementary Steel 3.9.3.4 Friction Forces 3.7.3.3.4 Gaps Between Pipe and Supports 3.7.3.3.4 Instrumentation Line Support Criteria 3.7.3.8.1.9, 3.9.3.4.1 Pipe Defection Limits 3.9.3.4.1 Pipe-Mounted Equipment Allowable Loads 3.9.3.1.21 As Built Piping Verification 3.9.3.1.20,3.9.8 (Ref. 3.9-10)
Pipe Interferences 3.9.3.1.22 Postulated Break and Crack Location and 3.6.2.1.4.1 through Configuration 3.6.2.1.4.5, 3.6.2.1.5.2, 3.6.2.1.5.3 l Dynamic Analysis for Postulated Break 3.6.2.3.1, 3.6.2.3.2 Notes:
(1) See Tier 2, Subsection 3.9.1.7. 'Ihe change restriction noted in this subsection applies to the delineated commitments only in their application to piping design.
(2) The applicable portions of these ections. tables and figures are italicized on the sections, tables and figures themselves. See Note (1) of Table 1.
(3) Tables 1.8-21 and 3.2-3 are applicable to a commitment invohing the ASME Code,Section III, and Tables 5.2-1 and 1.8-21 are applicable to a commitment involving the ASME Code. Section Ill. Code Cases.
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Table 8 Fuel System: Design Criteria and First Cycle Design and Methods (1)
Tier 2 Tables Commitment Tier 2 Sections (2) and Figurestz) Expiration (3)
Fuel System Design 4.2, 4.2.2.1, 4.2.5 First Full (Reference 4.2-1) Power l Fuel Assembly Design 4.2.2, 4.2.5 Figures 4.2-1, First Full (Reference 4.2-1) 4.2-2 Power Nuclear Design 4.3.2.1 Figure 4.3-1 First Full Power Fuel Evaluation Methods and Results 4.2.3 Table 1.8-19 First Full (References 4.2-2, Power 4.2-3)
Equilibrium Cycle and Control Rod Patterns 4A.1, 4A.2, 4A.3 First Full Power Fuel Licensing Acceptance Criteria App. 4B Table 1.8-19, None Table 1.8-21 Control Rod Licensing Acceptance Criteria App. 4C None Notes:
(1) See Tier 2, Section 4.2.
(2) The applicable portions of these sections, tables and figures are italicized on the sections, tables.
and figures themsches. See Note (1) of Table 1.
(3) The iequirement for prior NRC Staff approval expires as noted.
Table 9 instrument Setpoint Methodology (1)
Commitment Tier 2 Sectionst2) Tier 2 Tablestz)
Instrument Setpoint Methodology 7.1.2.10.9, 1.8-20 7.2.2.2.1(6),
7.3.2.1.2(3)(f),
7.3.4.(Ref. 7.3 2),
7.4.2.3.2(3)(f),
7.6.2.1.2(3)(f),
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Notes:
(1) See Tier 2. Subsection 7.1.2.10.9.
(2) The applicable portions of these sections and tables are itahcized on the sections and tables themsches. See Note (1) of Table 1.
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ABWR oesion comresoecamentmer2 i V( 4 i load response is then determined by either the time-history method or the response-spectrum method. The load on the reactor internals due to faulted i event SSE are obtained from this analysis. l The above loads are considered in combination as defined in Table 3.9-2. The SRV, LOCA (SBL, IBL or LBL) and SSE loads defined in Table 3.9-2 are all assumed to act in the same direction. The peak colinear responses of the reactorinternals to each of these loads are added by the square-root-of-the-summf-the-squares (SRSS) method. The resultant stresses in the reactor internal structures are directly added with stress resulting from the static and steady-state loads in the faulted load combination, including the stress due to peak reactor internal pressure differential during the LOCA.
The reactor internals satisfy the stress deformation and fatigue limits as defined in Subsection 3.9.5.3.
3.9.2.6 Correlations of Reactor Internals Vibration Tests with the Analytical Results Prior to initiation of the instrumented vibration measurement prognm for the prototype plant, extensive dynamic analyses of the reactor and internals are perforTned. ,
The results of these analyses are used to generate the allowable vibration levels during the vibration test. The vibration data obtained during the test will be analyzed in detail.
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Such comparisons provide the analysts with added insight into the dynamic behavior of the reactor intemals. The additional knov ledge gained from previous vibration tests has been utilized in the generation of the dynamic models for seismic and LOCA analyses for this plant. The models used for this plant are similar to those used for the vibrati m analysis of earlier prototype BWR plants.
3.9.3 ASME Code Class 1,2, and 3 Components, Component Supports, and Core Support Structures 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits This section delineates the criteria for selection and definition of design limits and loading combinatiors associated with normal operation, postulated accidents, and specified seismic and other Reactor Building vibration (RBV) events for the design of safety-related ASME Code components (except containment components, which are discussed in Section 3.8).
This section discusses the ASME Class 1,2, and 3 equipment and associated pressure-retaining parts and identifies the applicable loadings, calculation methods, calculated stresses, and allowable stresses. A discussion of major equipment is included on a
(
b) component-by-component basis to provide examples. Design transients and dynamic loading for ASME Class 1,2, and 3 equipment are covered in Subsection 3.9.1.1.
MechanicalSystems and Components 3.9-27 1
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Seismic-related loads and dynamic analyses are discussed in Section 3.7. The suppression pool-related RBVlaads are described in Appendix 3B. Table 3.9-2 presents the combinations of dynamic events to be considered for the design and analysis of all ABWR ASME Code Class 1,2, and 3 components, component supports, core support structures and equipment. Specific loading combinations considered for evaluation of each specific equipment are derived from Table 3.9-2 and are contained in the design specifications and/or design reports of the respective equipment.
Piping loads due to the thermal expansion of the piping and thermal anchor movements at supports are included in the piping load combinations. All operating modes are evaluated and the maximum moment ranges are included in the fatigue evaluation. [ Piping systems with maximum operating temperatures ofless than or equal to 65 *C are not required to be analyzedfor thermal expansion loading.}*
[Loupessurepiping systems that interface uith the reactor coolant pressure boundary will be designed with a pipe wall thickness calculatedfor a pressure equalto 0.4 times the reactor coolant system pressure but no less than that ofa schedule 40 pipe.}" See Appendix SM for additional information on intersystem LOCA.
Thermal stratification of fluids in a piping system is one of the specific operating conditions included in the loads and load combinations contained in the piping design specifications and design reports. It is known that stratification can occur its the feedwater piping during plant startup and when the plant is in hot standby ccnditions following scram (Subsection 3.9.2.1.3). (If during design orstartup, evidence ofthermal stratification is detected in any otherpiping system, then strattpcation willbe evaluated. Ifit cannot be shown that the stresses in thepipe are low and that movement due to bowing is acceptable, then stratifcation will be treated as a design load. In general, if temperature differences between the top and bottom of the pipe are less than 27*C, it may be assumed that design spectfication and stress reports need not be revised to include stratifcation.1 hepiping design reports shall be inspected to con)irm that thepipingsystems have been designedforthermalstratifcation in accordance with the requirements of this paragraph.} '
Under thermally stratified flow conditions, it has been observed that a relatively thin dynamic interface region exists, which oscillates in a wave pattern. This results in undulation in the hot-to-cold interface region which produces thermal striping on the inside of the pipe wall. Thermal striping stresses are the result of differences between the pipe inside surface temperatures which vary with time due to the interface oscillation and the average through-wall temperatures. The results of the feedwater piping thermal striping stress analysis confirm that the feedwater therrr.M nriping fatigue usage is minimal per the ASME Code,Section III, fatigue evalusion requirement; therefor e, thermal striping fatigue effects are negligible.
- See Subsection S.9.1.7.
O 3.9-28 MechanicalSystems and Components
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[N, A 91.5m depth is, thus, a m onable upper bound and, therefore, is selected to be the deep soil deposit case for this appendix. Between the shallow and deep soil cases, an intermediate depth is chosen to be at 61m.
In summary, the variations ofsoil deposit thickness are accounted for by considering the following four representative soil deposit depths. ,
Minimum (embedment depth) 25.7m -
Shallow soil deposit 45.7m Intermediate soil deposit 61.0m Deep soil deposit 91.5m 3A.3.2 Sois Profile and Properties The range of soil profiles considered in this appendix is based on the velocity profiles ,
used in GESSAR (References SA-1 and SA-2). A total of six velocity profiles are selected and shown in Figure SA-3. These velocity profiles are designated with the abbreviations:
UB, VPS, VP4, VP5, VP7 and an upper bound case with rigid soil properties (R cases).
The profile UB represents a soil profile of which the shear wave velocity at a depth z below the ground surface is obtained using the modulus parameter K2 max and the ,
following equations.
l G,,,(z) = 213.8K2maxdUm (z) (SA-1) r V, = gG,,x/p (SA-2) t where G max = maximum shear modulus, kPa ;
om = effective mean pressure (kPa) at depth (z); it is assumed equal to 0.7 times the effective overburden pressure, which corresponds to the use of an at rest coefficient oflateral pressure equal to 0.55.
Kemix = modulus parameter p = mass density A
( V, = shear wave velocity Seismic Coll Structure interaction Analysis 3A-3
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This soil profile is issumed to consist of seven horizontal layers. The K 2 max value, total unit weight, and Poisson's ratio for each layer are as shown in Table SA-1.
Note that for submeiged layers the Poisson's ratio is to be adjusted such that the minimum 1 -wave velocity ofwater is retained. The values of average shear wave velocity, shown in Figure SA-3 for UB, are computed using Equations SA-1 and SA-2 at the mid-depth of each layer for the ground water level at a depth of 0.61rn below grade.
The profiles UB, VP3 through VP5 are selected based on three generalized soil zones shown in Figure SA-4: a soil zone (sands, silts, clays, and gravely soils), a transition zone, and a soft rock and well-cemented soil zone. Velocity profile UB represents an average profile of'he soil zone;VP3 and VP4 hound the transition zone and VP5 represents an average profile for well<emented soi! zone. Those velocity profiles are smooth cunes representative of the average variation of shear modulus with depth that can be expected within each of the soil zones.
The prefile VP7 represents a hard rock site with a uniform shear wave velocity of 1524 m/s. The profile for R cases represents an upper bound profile with rigid soil properties (uniform shear wave velocity of 6096 m/s).
The lower bound shear wave velocity for the top 9m of soil among all profiles considered is 303 m/s. The upper bound velocity is 6096 m/s. This constitutes a wide range of potential site conditions that are suitable for nuclear power piants.
The general soil layer properties (layer thickness, total unit weight, and Poisson's ratio) defined for the UB profile are also adopted for all other profiles except for VP7 and R.
The profiles VP7 and R are considered to be uniform elastic half-space with a constant Poisson's ration of 0.3. The unit weight density for VP7 profile is the same as UB profile.
The unit weight density for R profiles is 2.2 t/m3. The base rock in all soil profiles is s
modeled using density 2.2 t/m , shear wave velocity of 1594 m/s. Poisson's ratio of 0.3, and material damping of 1E The average shear wave velocities in layers for all soil profiles are tabulated in Table 3A-2.
The shear modulus and material damping of soil are strain dependent. Figure SA-5 shows the variation ofshear aodulus and damping ratio with shear strain forvarious soil profiles considered. The soil cun'es shown correspond to average curves of Reference SA-6. On the basis of the recommendations made in Reference SA-3 the soil material damping of a hysteretic nature is limited to a maximum of 15% of critical. In addition to use of average soil curves, a parametric study was performed in which the upper bound shear modulus soil degradation curve of Reference SA-6 is used. This cune is the same as the shear modulus degradation curve reported in Reference SA-7 for sands. The results of this case tre presented in Section 3A.9.5. Variation of shear rnodulus and damping for rock profiles (VP5 and VP7) are showa in Figure SA-6. The shear moduhts reduction factors and damping ratios at various strain levels are shown 3A4 Seisnsic SOU Structure Interaction Analysis
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V Table 3H.1-12 shows the results for Load Combination 15, LBL (30 min) + SSE, which consists of D + L + P, + CO + T,+SSE + F, + Y. l l
Table SH.1-13 shows the load combination results corresponding to Load Combination !
15a,15b, IBL/SBL (6 hrs) + SSE, and consists of D + L + P, + CO + SRV + T, + SSE.
In these tables, the forces and moments are divided into three groups, for the convenience ofinput into the CECAP program for the rebar stress evaluation. The first row represents the forces and moments caused byThermal Load (T3 ). The second row lists the forces and moments due to loads 1 through 8 listed in Subsection 3H.1.5.3. The ;
last row lists the resulting forces and moments due to seismic loads (SRSS).
3H.1.5.5 Structural Design The evaluation is based on the loads, load factors and load combinations indicated in Subsection 3H.1.4. !
Figure 3H.1-21 shows the location of the sections which were selected for evaluation.
Bechtel computer program 'CECAP' was used for the evaluation of stresses in rebar and concrete and strains in the liner plate. The input to CECAP consists of rebar ratios, i material properties, and element geometry at th e section under consideration together j with the forces and rnoments from the STARD*lNE analysis, which are shown in Tables 3H.1-10 through 3H.1 13. Table SH.1-1 i lists the rebar ratios used in the evaluation. At each section, in general, three ehments were analyzed at azimuth 180*,
225* and 270 . Table SH.1-15 through SH.1-18 show the rebar and concrete stresses at these sections for the representative elements.
Figure SH.1-22 shows flow chart for the structural analysis and d.: sign.
Figures SH.1-28 through 3H.137 present the design drawings used for the evaluation of the containment and the Reactor Building Structural design.
3H.1.5.5.1 Containment Structure 3H.1.5.G.1.1 Containment Wall Section , I through 6 shown in Figure 3H.1-21 are considered critical sections for the containment wall. hiaximum stress was found to be 3.72x105 kPa in the meridional rebar at Section 1 near the bottom of the RCCV wall due to load combinations .15a and 15b, as shown in Table SH.1-18. The maximum stress in the circumferential rebar was found to be 3.53x10 5kPa which also occurs at Section 1. Table 3H.1 19 shows liner plate strains. The liner maximum stmin was found to be 0.00197 at Section 1, which is within allowable limits given in Table CC-3720-1, AShfE Code Section III, Division 2.
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3H.1.5.5.1.3 Containment Foundation Mat Sections 10 to 14 were evaluated for the basemat within the Containment Walls and Section 18, for outside the Containment. The sections are shown in Figure SH.1-21. The maximum rebar stress was calculated as 3.18x105 kPa at Section 12just outside the RPV Pedestal and is shown in Table SH.1-18. The liner plate maximum strain was found to be 0.000439 at Section 14 as shown in Table SH.1-19.
3H.1.5.5.2 Containment Internal Structures 3H.1.5.5.2.1 Diaphragm Floor Sections 15,16, and 17, were selected for evaluation of the diaphragm floor. The sections are shown in Figure 3H.1-21. The results of the analysis are shown in Tables 3H.1-15 to SH.1-19. The m:c:imum stress in the radial rebar was found to be 2.45x105k"a at Section 15 shown in Table 3H.1-18, whereas the maximum stress in the l circumferential rebar was found to be 9.9x10 4kPa at Section 17, as shown in Table SH.1-17. The maximum strain in the liner plate was found to be 0.000848 cm/cm (compressive) as shown in Table 3H.1-19.
3H.1.5.5.2.2 Reactor Pedestal Sections 19,20 and 21 were selected for evaluation of the pedestal. These are shown in Figure 3H.1-21. The forces and moments for the load combinations are shown in Tables SH.1-10 to SH.1-13. The results of the analysis are shown in Table SH.1-20. The maximum membrane stress in cylindncal steel plate was found to be 2.07x105 kPa and the maximum shear stress in the stiffener plates was found to be 1.84x105kPa, which are within the allowable stress limits.
3H.1.5.5.3 Reactor Building Sections 22 through 34 were analyzed for the Reactor Building outside the containment. The sections are shown in Figure 3H.1-21. Sections 22 to 24 were selected for the R/B Outside Viall, Sections 25 to 29 for the spent fuel pool walls and floor and Sections 30 to 34 for the R/B Slabs.
3H.1.5.5.3.1 R/B Outside Walls These walls resist the lateral soil pressure besides the forces and moments from the other loads. The design lateral soil pressures are shown in Figure 3H.1-11. Out of-plane 3H.1-16 Design Details and haluation Results of Seistnic CategoryI Structurer
1 t i
,v -
1 Riv. 2 l ABWR oesion comresoecamenvrier2 i
,rs Table 15.6-5 Steamline Break Accident Parameters i Data and assumptions used to estimate source terms. l A. Power Level 4005 MWt
- 8. Fuel damage none C. Reactor coolant activity Subsection 15.6.4.5 D. Steam mass released 12,870 kg E. Water mass released 21,953 kg Il Data and assumptions used to estimate activity released A. MSIV closure time (break time 5.5 s l until fully closed)
- 8. Maximum release time 2h lil Dispersion and Dose Data A. Meteorology Table 15.6-7
- 8. Boundary and LPZ distances Table 15.6-7
, C. Method of Dose Calculation Reference 15.6-2 D. Dose conversion Assumptions Reference 15.6-2, RG 1.109, and ICRP 30 E. Activity inventory / release Table 15.6-6
, F. Dose Evaluations Table 15.6-7 t
h O
b Decrease in Reactor Coolant Inventory 15.6-27
s ..
Rev.O ABWR oesion controloocanonener:
O Table 15.6-6 Main Steamline Break Accident Activity Released to Environment (megabecqueral)
Isotope Case 1 Case 2 1-131 7.29E+04 1.46E+06 l-112 7.10E+05 1.42E+07 l133 5.00E+05 9.99E+06 l-134 1.40E+06 2.79E+07 l-13b 7.29E+05 1.46E+07 Total Halogens 3.41E+06 ' 6.81 E+07 KR-83M 4.07E4 02 2.44E+03 KR-85M 7.18E+02 4.29E+03 KR-85 2.26E+00 1.36E+01 KR 87 2.44E+03 1.47E404 KR-68 2.46E+03 1.48E+04 KR-89 9.8PE+03 5.92E+04 KR-90 2.55E+03 1.55E+04 XE-131M 1.76E+00 1.'s6E+01 XE-133M 3.39E+01 2.04E402
- XE 133 9.47E+02 5.70E+03 XE-135M 2.89E+03 1.74E+04 XE-135 2.70E +03 1.62,1+04 XE-137 1.23E+04 7.40E+04 l XE-138 P.44E+03 5.66E+ 04 XE-139 4.33E+03 2.59504 Total Noble Gases 5.11E+04 3.07E4 05 l .
l l
l 9
15.6-28 Decrease in Reactor Coolant Inventory l