ML20080P789
| ML20080P789 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/31/1984 |
| From: | Edwards N, Quinn R, Yin Y NUTECH ENGINEERS, INC. |
| To: | |
| Shared Package | |
| ML20080P730 | List: |
| References | |
| BPC-01-300-4, BPC-01-300-4-V04-R00, BPC-1-300-4, BPC-1-300-4-V4-R, NUDOCS 8402230116 | |
| Download: ML20080P789 (35) | |
Text
__
BPC-01-300-4 Revision 0 January 1984 HOPE CREEK GENERATING STATION PLANT UNIQUE ANALYSIS REPORT VOLUME 4 INTERNAL STRUCTURES ANALYSIS Prepared for:
Public Service Electric and Gas Company Prepared by:
NUTECH Engineers, Inc.
San Jose, California Prepared by: Reviewed by:
R. D. Quinn, P.E. Y. C. Yiu, P.E.
Senior Engineer Group Leader Approved by: Issued by:
N. W. Edwards, P.E. R. A. Lehnert, P.E.
President Project Manager A
~'
8402230116 840210 PDRADOCK05000g g{
REVISION CONTROL SHEET Hope Creek Generating TITLE: Station DOCUMENT FILE NUMBER: BPC-01-300-4 Plant Unique Analysis Revision 0 I I Report, Volum 4 d
INi;i ALS L. R. Hussar / Engineer A.
$. he Imandoust/ Consultant I
.c/d A . I.
INITIALS MA&
R. D. Quinn/ Senior Engineer R od INITI At.S C/% s. LJ. csu)
INITIALS C. S. Wu/ Consultant I
' D isiTIKLS Y. C. Yiu'/ Group Leader AFFECTED OOC PREPARED ACCURACY CRITERIA REMARKS PAGE(S) REV BY/DATE CNECK SY / OATE CHECK SY / DATE ii 0 26k2fZG'0+ LSI-(l-1h-54 lI-Ni@
q fii iv 0
0 v 0 vi 0 vii 0 viii 0 ix 0 4-1.1 0 4-1.2 0 4-1.3 0 4-2.1 0 4-2.2 0 4-2.3 0 4-2.4 0 4-2.5 0 4-2.6 0
- -2.7 0 I b2.8 0 4 -2. 9 0 'r y V 4-2.10 0 gpg/,.c.84 (_f// /-16.f4- [l-%'ft PAGE 1
OF 2
, CEP 31.1.1
REVISION CONTROL SHEET Hops Crock Generating (CONTINUATION)
TITLE: Station DOCUMENT FILE NUMBER:BPC-01-300-4 Plant Unique Analysis Revision 0 Report, Volume 4 AFFECTED DOC PREPARED ACCURACY CRITERI A REMARKS PAGE(S) REV SY / DATE CHECK SY / OATE CHECK SY / OATE 4-2.11 0 8M/3 fG-84 [ /2/[ I-14-N %[I-36-$f 4-2.12 0 4-2.13 'O 4-2.14 0 4-2,15 0 4-2.16 0 4-2.17 0 T y 4-2.18 0 AM/l-2' ~8 + LM /-26-F+
4-2.19 0f,Ilz(pl84 CSU /-16-5% 4-2.20 0 AM/ o - zG- 04 (_JU{ l-26-F+ y 4-3.1 0 /20Gl'-2"~04 Lgl[t-zs-g4 QlI-t4-tq I l I l j PAGE OF CES 20*2
ABSTRACT The ' pr ima ry containment for the Hope Creek Generating Station , was de s igned , erected, pressure-tested, and N-stamped in accordance with the ASME Boiler and Pressure Ve ssel Code, Section III, 1974 Edition with addenda up to and including Winter 1974. These activities we re pe rf ormed for the Public Service Electric and Gas Company (PSE&G) by the Pitt sburgh-De s Moines Steel Company. Since then, new requirements which af fect - the des ign and operation of the primary containment system have been established. These requirements are defined in the Nuclear Regulatory Commission's (NRC) Safety Evaluation Report, NUREG-0661. The NUREG-0661 requirements define revised contain-ment ' des ign loads postulated to occur during a loss-of-coolant
- , accident or a safety-relief valve discharge event which are to be evaluated. In addition, NUREG-0661 requires that an assessment of the effects that these postulated events have on the operation of the containment system be performed.
n U This plant unique analysis report ( PUAR) documents the efforts undertaken to address and resolve each of the applicable NUREG-0661 requireme nts for Hope Creek. It demonstrates,. in accordance ' with NUREG-0661' acceptance criteria, that the design of' the primary containment system is adequate and that original i . de s ign safety margins have been restored. The Hope Creek PUAR is composed of the following six volumes : 1 o Volume 1 - GENERAL CRITERIA AND LOADS METHODOLOGY i o Volume 2 - SUPPRESSION CHAMBER ANALYSIS o Volume 3 - VENT SYSTEM ANALYSIS o Volume 4 - INTERNAL STRUCTURES ANALYSIS o Volume 5 - SAFETY RELIEF VALVE DISCHARGE PIPING ANALYS IS o Volume 6 - TORUS ATTACHED PIPING AND SUPPRESSION CHAMBER PENETRATI(.N ANALYSES o B PC-01-3 00 -4 Revision 0 4-ii nutg_qh
Major portions of all volumes of this report have been prepared by NUTECH Engineers, Incorporated (NUTECH), acting as a consultant responsible to the Public Service Electric and Gas Company. Selected sections of Volumes 5 and 6 have been prepared by the Bechtel Powe r Corporation acting as an agent responsible to the Public Service Electric and Gas Company. This volume, Volume 4, documents the evaluation of the internal structures. NOTE: Identification of the volume number precedes each page, section, subsection, table, and figure number. O BPC-01-300-4 9 Revision 0 4-iii nut _ech
TABLE OF CONTENTS Page ABSTRACT 4-il LIST OF ACRONYMS 4-V LIST OF TABLES 4-Vili LIST OF FIGURES 4-ix 4-
1.0 INTRODUCTION
4-1.1 4-1.1 Scope of Analysis 4-1.2 4-2.0 INTERNAL STRUCTURES ANALYSIS 4-2.1 4-2.1 Component Description 4-2.2 4-2,1.1 Catwalk 4-2.3 4-2.1.2 Monorail 4-2.8 4-2.2 Loads and Load Combinations 4-2.11 4-2.2.1 Loads 4-2.12 4-2.2.2 Load Combinations 4-2.14 4-2.3 Analysis Acceptance Criteria 4-2.16 4-2.4 Method of Analysis 4-2.17 4-2.5 ' Analysis Results 4-2.18 4-2.5.1 Conclusions 4-2.20 4-3.0 LIST OF REFERENCES 4-3.1 i ? O BPC-01-300-4 Revision 0 4-iv nut
j LIST OF ACRONYMS ACI American Concrete Institute ADS Automatic Depressurization System AISC American Institute of Steel Construction ASME American Society of Mechanical Engineers ATWS Anticipated Transients Without Scram BDC Bottom Dead Center BWR Boiling Water Reactor CDF Cumulative Distribution Function CO Condensation Oscillation DBA Design Basis Accident DC Downcomer DLF Dynamic Load Factor ECCS Emergency Core Cooling System PSAR Final Safety Analysis Report FSI Fluid-Structure Interaction FSTP Full-Scale Test Facility HNWL High Normal Water Level HPCI High Pressure Coolant Injection IBA Intermediate Break Accident I&C Instrumentation and Control ID Inside Diameter IR Inside Radius LDR Load Definition Report LOCA Loss-of-Coolant Accident BPC-01-300-4 9 Revision 0 4-v nutggh
LIST OF ACRONYMS (Continued) LPCI Low Pressure Coolant Injection LTP Long-Term Program MC Midcylinder MCF Modal Correction Factor MJ Mitered Joint MVA Multiple Valve Actuation NEP Non-Exceedance Probability NOC Normal Operating Conditions NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NVB Non-Vent Line Bay
/ } OBE Operating Basis Earthquake OD Outside Diameter PSD Power Spectral Density PSE&G Public Service Electric and Gas Company PUA Plant Unique Analysis PUAAG Plant Unique Analysis Application Guide PUAR Plant Unique Analysis Report PULD Plant Unique Load Definition
, OSTF Quarter-Scale Test Facility RCIC Raector Core Isolation Cooling RHR Residual Heat Removal RPV Reactor Pressure Vessel
<~s RSEL Resultant Static-Equivalent Load (G
BPC-01-300-4 Revision 0 4-vi nute_cb
l LIST OF ACRONYMS (Concluded) SBA Small Break Accident SBP Small Bore Piping SER Safety Evaluation Report SORV Stuck-Open Safety Relief Valve SRSS Square Root of the Sum of the Squares SRV Safety Relief Valve SRVDL Safety Relief Valve Discharge Line SSE Safe Shutdown Earthquake STP Short-Term Program SVA Single Valve Actuation TAP Torus Attached Piping VB Vent Line Bay VH Vent Header VL Vent Line VPP Vent Pipe Penetration ZPA Zero Period Acceleration BPC-01-300-4 G Revision 0 4-vii nutggh
LIST OF TABLES Number Titl'e Page 4-2.2-1 Internal Structures Component Loading 4-2.13 Identification 4-2.2-2 Controlling Internal Structures Load 4-2.15 Combinations 4-2.5-1 Internal Structures Stresses for control- 4-2.19 ling Load Combinations O l d BPC-01-300-4 "eerision 0 4-viii flu
LIST OF FIGURES Number Titic Page 4-2.1-1 Plan View of Catwalk Segment 4-2.5 4-2.1-2 Catwalk Support De tails 4-2.6 4-2.1-3 Catwalk Support Details at Vacuum Breaker 4-2.7 Platform 4-2.1-4 Plan View of Suppression Chamber Monorail 4-2.9 4-2.1-5 Monorail Support Details 4-2.10 0 O BPC-01-300-4 Revision 0 4-ix nutp_qh
4-
1.0 INTRODUCTION
In conjunction with Volume 1 of the Plant Unique Analysis Report (PUAR), this volume documents the efforts undertaken to address the requirements defined in NUREG-0661 which affect the Hope Creek internal structures. The internal structures PUAR is organized as follows: o INTRODUCTION Scope of Analysis o INTERNAL STRUCTURES ANALYSIS Component Description Loads and Load Combinations Analysis Acceptance Criteria Method of Analysis Analysis Results and Conclusions The INTRODUCTION section contains a general overview discussion of the internal structures evaluation. The INTERNAL STRUCTURES ANALYSIS section discusses the specific components, loads, criteria, methods, and results associated with the evaluation. A summary of the conclusions derived from the internal structures evaluation is also included.
/"N b
BPC-01-300-4 Revision 0 4-1,1 nutggh
4-1.1 Scope of Analysis The general criteria presented in Volume 1 are used as O the basis for the Hope Creek internal structures evaluations described in this volume. The internal structures evaluated include the catwalk and monorail. These structures are not required for the safe operation of the primary containment system during accident conditions. The internal structures are evaluated for the effects of LOCA related loads, as defined by the NRC's Safety Evaluation Report NUREG-0661 (Reference 1) and the Mark I Containment Program Load Definition Report (LDR) (Reference 2). The LOCA loads used in this evaluation are formulated using the procedures discussed in Volume 1 of this report. The evaluation includes structural analyses of the internal structures to ensure that these structures do not fail and result in damage to safety related components. The results of the structural evaluation for each load are used to evaluate load combinations for the internal structures, in accordance with the Mark I Containment O BPC-01-300- 4 Revision 0 4-1.2 nut _ech
Program Plant Unique Analysis Application Guide (PUAAG) (Reference 3). The evaluation results are conserva-tively compared with the acceptance limits specified by , the applicable sections of the American Society of Mechanical Engineers (ASME) Code (Reference 4) to ensure that failure will not occur. 4 i l l l l BPC-01-300-4 Revision 0 4-1.3 n
4-2.0 INTERNAL STRUCTURES ANALYSIS a An evaluation of each of the NUREG-0661 requirements which affect the design adequacy of the Hope Creek internal structures is presented in the sections which follow. The criteria used in the evaluation are con-tained in Volume 1 of this report. The component parts of the internal structures which are examined are described in Section 4-2.1. The loads and load combinations which are evaluated are described and presented in Section 4-2.2. The analysis method-ology used to evaluate the effects of these loads and load combinations is discussed in Section 4-2.4. The 's acceptance limits to which the analysis results are compared are discussed and presented in Section 4-2.3. The analysis results and the corresponding design margins are presented in Section 4-2.5. I % ,I BPC-01-300-4 Revision 0 4-2.1 nutsch
l l 4-2.1 Component Descrip e. ion The internal structures which are evaluated include the O catwalk and the monorail, which are described in Sections 4-2.1.1 and 4-2.1.2, respectively. O s O BPC-01-300-4 Revision 0 4-2.2 nutggh
4-2.1.1 Catwalk O
- U The catwalk is a platform-type structure approximately 3 feet wide, which extends around the full circumfer-ence of the suppression chamber. The catwalk is located in the upper outside quadrant of each suppres-sion chamber. mitered cylinder.
The catwalk frame consists of two W8 x 21 stringers which span between S8 x 18.4 support beams which are located at the mitered joint and at two intermediate locations between mitered joints. The mitered joint and intermediate supports consist of two vertical hangers and one horizontal support strut. The catwalk
'E and catwalk support configurations are shown in Figures 4-2.1-1 and 4-2.1-2.
A vacuum breaker platform approximately 3' wide by 9' long is located adjacent to the catwalk at midcylinder of.each vent line bay. The vacuum breaker platform is supported by two S8 x 18.4 platform support members. The vacuum breaker platform and support configurations are shown in Figures 4-2.1-1 and 4-2.1-3. A BPC-01-300-4 Revision 0 4-2.3 nutggh
The catwalk and vacuum breaker platform are comprised ' of grating which is bolted to the catwalk frame. The catwalk frame is braced against lateral loads by the horizontal support struts. O O BPC-01-300-4 Revision 0 4-2*4 gh
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- s BPC-01-300-4 Revision 0 4-2.7
1 4-2.1.2 Monorail The monorail consists of S10 x 25.4 beam sections curved to a 60'-7" radius and connected end-to-end to form a continuous monorail beam around the circumfer-ence of the suppression chamber. The monorail is located in the upper ou tside quadrant of each suppression chamber mitered cylinder. The monorail configuration is shown in Figure 4-2.1-4. The monorail beam is supported at each mitered joint and at one intermediate location between mitered joints. The intermediate , supports are located at or near midcylinder. The monorail supports consist of a vertical support O member and a horizontal support member. The horizontal support members are constructed from a 2-1/8" x l-1/2" bar sandwiched by two 2-1/2" x 2-1/2" x 5/16" thick angles. The bottom legs of the support angles are removed from the intermediate hori zon tal support members. The vertical suppo rt members are constructed from a combined section of a 2" x 1/2" solid bar and a 1-1/2" x l-1/2" solid bar. De tails of the monorail supports are shown in Figure 4-2.1-5. B PC-01-3 00-4 O Revision 0 4-2.8
, nut _ech_
O e-SUPPatESSION M ONCR AIL CMAMBER BEAM SHELL I_ h N . 2., _ ee kt. # N I CENTERLIN E DRESSCN f M CNORAIL
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i t 4-2.2 Loads and Load Combinations The loads for which the Hope Creek internal structures are evaluated are defined in NUREG-0661 on a generic basis for all Mark I plants. The methodology used to develop plant unique loads, for each applicable load defined in NUREG-0661, is discussed in Section 1-4.0. The results of applying the methodology to develop specific values for each of the controlling loads are discussed and presented in Section 4-2.2.1. The controlling load combinations which affect the internal structures are formulated by usicg the event combinations and event sequencing defined in NUREG-0661 and discussed in Sections 1-3.2 and 1-4.3. The controlling load combinations are discussed and presented in Section 4-2.2.2. s O BPC-01-300-4 Revision 0 4-2.11 11U
4-2.2.1 Loads The loads acting on the internal structures are categorized as follows:
- 1. Dead Weight Loads la. Dead Weight of Steel
- 2. Seismic Loads 2a. OBE Loads
- 26. SSE Loads
- 3. Pool Swell Loads 3a. Pool Swell Impact and Drag Loads 3b. Froth Impingement and Fallback Loads 3c. Pool Fallback Loads
- 4. Containment Interaction Loads 4a. Containment Structure Motions Table 4-2.2-1 shows the specific internal structures which arc affected by each of the above loads. The methodology used to develop values for each of these loadings is discussed in Section 1-4.0. The resulting magnitudes and characteristics of each loading are similar to those described in Volume 3 of this report.
O BPC-01-300-4 Revision 0 4-2,12 nutggh
O Table 4-2.2-1 INTERNAL STRUCTURES COMPONENT LOADING IDENTIFICATION e Volume 4 Load d ,d Designation PUAR Volume e m Section 3 o Category ase Refennce y C Number b , Dead Weight la 1-3.1 X X 2a 1-3.1 X X Seismic 2b l-3.1 X X
/ 3a 1-4.1.4.2 X
< l Pool Swell 3b l-4.1.4.3 X X Loads 1 3c l-4.1.4.4 X Containment h 2-2.2 X X Interaction h N D BPC-01-300-4 Revision 0 4-2.13
- - - - - - - - - - - l
i 4-2.2.2 Load Combinations The loadings which affect each of the internal struc-tures are presented in Section 4-2.2.1. The general NUREG-0661 criteria for grouping these loads into event combinations are discussed in Section 1-3.2. Since the internal structures are located above the suppression pool, the event combinations which produce controlling stresses are those which contain pool swell loads. These include the DBA 18 and DBA 25 combinations as shown in Table 4-2.2-2. The catwalk and monorail are therefore evaluated for the DBA 18 and DBA 25 event combinations. O BPC-01-300-4 O Revision 0 4-2.14 , nuteSh i
O Table 4-2.2-2 CONTROLLING INTERNAL STRUCTURES IDAD COMBINATIONS Section Event DBA 4-2.2.1 Load NUREG-06C1 Designation Combination 18 25 Number
- 1) Dead Weight la la OBE 2a
< 2) Seismic SSE 2b
- 3) Pool Swell Loads 3a- c 3a-3c
- 7) Containment Interaction 7a 7a Service he# B( } C Level Internal Structures g E E Note:
- 1. Evaluation of secondary stress range and fatigue not required.
4 O BPC-01-300-4 Revision 0 4-2.15 g{
4-2.3 Analysis Acceptance Criteria l The service level assignments for the internal struc-9 tures and the suppression chamber shell at attachment points to internal structures are shown in Table 4-2.2-2. The table shows that all internal structures are designated as Service Level E components, and as such, are not required to meet ASME Code acceptance limits. In order to employ a consistent set of design 5.riteria which ensures that failure will not occur, the internal structures are conservatively evaluated for the Service Level D scceptance limits contained in the ASME Code. The suppression chamber shell near attachment points to internal structures is evaluated in accordance with the requirements for Class MC components contained in the ASME Code. The correspond-ing allowable stresses for the internal structures are presented in Section 4-2.5 for the DBA 18 and DBA 25 combinations. BPC-01-300-4 O Revision 0 4-2.16 nut _ech l
4-2.4 Method of Analysis _ O The loadings for which the internal structures are evaluated are identified in Section 4-2.2.1. The analysis of the catwalk is performed using manual calculations to evaluate the stringers, hangers, and . associated catwalk components. Equivalent static analyses are performed for pool swell impact loads, froth impingement loads, seismic loads, and containment interaction loads. The reaction loads in the catwalk hangers are used to evaluate local stresses in the suppression chamber shell. The analysis of the monorail is performed using a beam model which includes the monorail beam and the monorail supports. Equivalent static analyses are performed for all monorail loadings. The reaction loads in the monorail supports are used to evaluate local stresses in the suppression chamber shell. o BPC-01-300-4 Revision 0 4-2.17 O
4-2.5 Annlyeis R1sults The geonie t ry , loads, load combinations, acceptance criteria, and analysis methods used in the evaluation of the internal structures are presented in the preceding sections. The resulting maximum stresses for the catwalk and monorail are shown in Table 4-2.5-1. The maximum suppression chamber shell and pad plate to shell we ld stresses due to catwa lk and monorail reaction loads are also reported in Table 4-2.5-1. As is evident from this table, the calculated stresses for these components are less than the corresponding allowable stresses. O B PC 3 00 -4 Revision 0 4-2.18 nutggh
Table 4-2.5-1
/D INTERNAL STRUCTURES STRESSES FOR CONTROLLING LOAD COMBINATIONS V
Load Combination Stresses (ksi) Material Stress DBA 25 Item Material Properties g, (ksu A11C"8D1* Cale. Calc. Stress Stress AllCW. (Service Level El Compressive 1.01 21.58
, . ,,,,, (1)
SA- Weak Axis 12.58 50.80 0.94 St i er Bending Su = 58.0 Strong Axis 29.00 44.69 Bending 21.70 yf"
- Sy = 33.85 Compressive 5.10 g t
8u = 58.0 Bendina 23.20 40.63 0.88 Catwalk S = 33.86 Compressive 0.62 6.27 (1) Platform Y Su port SA-36 0.56 S = 58.0 Bending 18.88 40.63 Tensile 1.21 40.63 (1) Monorail S = 33.86 Beam SA-36 ed er gm Su = 58.0 ( } Strong Axis Bending 5.24 30.96
,% )'
Monorail S = 33.86 Compressive 4.85 22.57 (1) vertical SA-36 Y 0.65 Support S = 58.0 Bending 16.96 40.63 III Monorail S y " 33.86 Compressive 2.96 8.61 Horizontal SA-36 0,73 Support Su = 58.0 Bending 24.03 40.64 Suppresstor sg.516 (3) Chamber S = 19.3 Local Primary 20.73 28.95(2) 0.72 Shell Gr. 70 mc M e rane Pad m SA-5 g 2.54 15.01(2) mc
= 19.3 Prl. tar'/ 0.17 Weld Notes:
- 1. Values shown obtained from beam interaction equation.
- 2. Service Level B allowable conservatively used.
- 3. Stress shown includes combined effects of internal structure reaction leads and general shell stresses obtained from the suppression chamber analysis documented in Volume 2.
, m) BPC-01-300-4 Revision 0 4-2.19 nutech
4-2.5.1 Conclusions The values of the loads used to evaluate the internal O structures are conservative estimates of the loads postulated to occur during an actual LOCA event. The event combinations for which the internal structures are evaluated envelop the actual events expected to occur during a LOCA event. Th.a acceptance limits to which the evaluation results are compared are more restrictive than those required by NUREG-0661. Use of these acceptcnce limits ensures that the internal structure components will not fail and cause damage to safety-related components. O ,. As is evident from the analysis results presented, stresses in the internal structure ecmponents are within these conservative acceptance limits The intent of the NUREG-0661 criteria as it relates to the design adequacy of the Hope Creek internal structures is therefore considered to be met. s BPC-01-300-4 O Revision 0 4-2.20 nutpg 1
4-3.0 LIST OF REFERENCES
- 1. " Mark I Containment Long-Term Program," Safety Evaluation Report, Nuclear Regulatory Commission, NUREG-0661, July 1980.
- 2. " Mark I Containment Program Load Definition Report," General Electric Company, NEDO-21888, Revision 2, December 1981.
- 3. " Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide,"
Task Number 3.1.3, General Electric Company, NEDO-24583-1, October 1979.
- 4. ASME Boiler and Pressure Vessel Code, Section III, Division 1, 1977 Edition with Addenda up to and including Summer 1977.
. (A i l 1 3 (V BPC-01-300-4 hevision 6 4-3.1 nutg,qh '
- - -. .. .- - -}}