ML20151P093

Rev 2 to Mark I Long-Term Program Plant-Unique Analysis, Peach Bottom Units 2 & 3

ML20151P093
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1985
From:	BECHTEL GROUP, INC.
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References
NUDOCS 8601030299
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PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 MARK I LONG-TERM PROGRAM PLANT UNIQUE ANALYSIS DOCKET NUMBERS: 50-277 AND 50-278 PREPARED FOR PHILADELPHIA ELECTRIC COMPANY WESTERN POWER DIVISION SAN FRANCISCO, CALIFORNIA REVISION 2 l DECEMBER 1985 ReA*28828a!8677 P PDR

_ - - - . _ _ . _ _ _ . _ . _ _ _ _ _ . . _ - . _ _ . _ _ _ . . _ _ _ . . _ , _ . . ~ . _ _ _ _ _ _ _ _ , . _ . _ _ . - - _ _ _ _ _ _ _ _ . .

. . e on the ASME Code, and to develop a Plan't Unique Analysis Application Guide (PUAAG). The generic program also investigated ways to mitigate the loads.

(b) The: plant unique program 'was to identify areas that needed load mitigation and/or structural modifications, to design and install the necessary hardware, and to demonstrate by plant unique analysis (PUA) that'the~ modified structure met the specified acceptance criteria and the intended original margin of safety had been restored.

The : generic program was 'a major effort to define the - loads. Various subscale and lin-plant tests were performed, scores 'of reports were generated, and a full-scale test facility was constructed which consisted of one complete torus bay, the equivalent of 1/16th of the Peach Bottom toru.5. All this was done in an effort to formulate methods of defining the magnitude of the dynamic pool loads that occur during LOCA and SRV events.

In addition, parallel analytical efforts were undertaken. -

The generic program was substantially completed ~ in 1980 and the plant unique loads were defined. Analyses of the PBAPS torus revealed the need for mitigation of the loads and modification - of the structural elements.

T-Quenchers and vent header deflectors were' designed for load mitigation.

All the ne'cessary hardware installation ~and modifications to the Unit 3 torus were completed in the fall of 1981 and to the Unit 2 torus in the spring of 1982. Modifications to torus attached pipes and supports in both units are' in progress. Major modifications were:

, (a) Installation of quenchers and 8-inch vacuum breakers on the SRV lines (b) Tie down of the torus (c) Installation of torus shell stiffeners (d) Installation of torus nozzle stiffeners i

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(e) Installation of deflectors below the vent header in the non-vent bay (f) Reinforcing of vent header-downcomer intersections

' (g) Addition of new supports and modification of existing supports to the torus attached pipes 4

(h) Strengthening of torus internal structures and their. supports.

Plant unique analyses were performed in accordance with NUREG-0661

. guidelines (Reference 10) for the specified loads. The IOCA loads are defined in the Mark I Load Definition Report (T.DR) (Reference 11). An extensive in-plant SRV discharge test program was ccnducted to define the SRV discharge loads. The results _ of the plant unique analyses were evaluated and compared with structural acceptance criteria.

This report contains the results of the evaluation for Peach Bottom Units 2 and 3. Section 2 describes the physical and stnictural characteristics of all the structures and piping considered. Section 3 describes the phenomena of LOCA and SRV discharge. Structural acceptance criteria are summarized in Section 4. Applied loads and load combinations given in the

> structural acceptance criteria are discussed in Section 5. The structural evaluation of the torus is given in Section 6. Sections 7 and 8 contain the results ' of the evaluation of the vent system and torus internals respectively. The last chapter summarizes the analyses. Modifications to torus attached piping supports in both units are underway. The results of the plant unique analysis' for the torus attached piping are contained in addendum No. 1 to this PUA report.

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l PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 ADDENDUM NO.1 TO THE MARK I LONG-TERM PROGRAM PLANT UNIQUE ANALYSIS DOCKET NUMBERS: 50-277 AND 50-278 PREPARED FOR PHILADELPHIA ELECTRIC COMPANY WESTERN POWER DIVISION

@: SAN FRANCISCO, CALIFORNIA REVISION 0 DECEMBER 1985

1 TABLE OF CONTENTS SECTION PAGE

1.0 INTRODUCTION

l-1 2.0 DESIGN LOADS 2-1 3.0 TORUS ATTACHED PIPING 3-1 4.0 STRUCTURAL STEEL MODIFICATIONS 4-1 FOR LOADS FROM TORUS ATTACHED PIPING HANGERS 5.0 SRV DISCHARGE PIPING 5-1 6.0 TORUS PENETRATIONS 6-1 7.0

SUMMARY

OF ANALYSIS 7-1

8.0 REFERENCES

8-1 l

PBAPSPF2/7

1.0 INTRODUCTION

1.1 A report describing the Mark I Long Term Program Plant Unique Analysis for Peach Bottom Atomic Power Station Units 2 and 3 (Reference 1) was submitted to the U.S.

Nuclear Regulatory Commission (NRC) in April 1982 and revised in August 1983. This report described the structural acceptance criteria and the results of plant unique analyses for the structural evaluation of the torus shell, the vent system and torus internal structures.

1.2 This report is an addendum to the Reference 1 report and contains the results of the plant unique analyses for the Safety Relief Valve (SRV) discharge lines (in the d ryw el l and the wetwell) and the external torus attached piping under the Mark I Long Term Program. Section 2 describes all the loadings that were considered for the evaluation of torus attached piping and SRV discharge lines. Section 3 describes the evaluation of the torus attached piping including the design loads, analysis, acceptance criteria, and f atigue evaluation necessary to meet the original intended safety margins. Section 4 describes the evaluation of the structural steel which carries the additional reaction loads resulting from the hydrodynamic loads on the piping supports. The evaluation of the SRV discharge lines, both inside the drywell and the wetwell, is described in Section 5. Section 6 describes the analysis of torus penetrations for hydro-dynamic loads and Section 7 summarizes the analyses.

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2.0 DESIGN LOADS

2.1 INTRODUCTION

This section describes the loads used to analyze the external torus attached piping and the SRV discharge lines, including normal operating, seismic and hydrodynamic loads. Hydro-dynamic loads are further categorized as either Loss of Coolant Acci dent (LOCA) related or due to SRV discharge phenomenon.

2.2 TORUS ATTACHED PIPING LOADS The loads used to analyze the external torus attached piping are described below.

2.2.1 Thermal Loads: Thermal analysis was performed for thermal modes based on temperatures described in references 2, 12, 13, 14 and 15. Piping anchor movements for the thermal analyses included equipment nozzle displacements, torus penetration displacements, and branch connection displacements for small pipe connected to large pipe. These displacements were due to normal operating and LOCA conditions. The torus penetration displacements also take into consid-eration the movement of the torus shell at the maximum suppression pool temperature which was obtained f rom the torus shell analysis described in reference 1.

An analysis of the suppression pool temperature response to SRV discharge was performed using proprietary General Electric models. Seven events were analyzed which bound those required by the NRC. The analysis concluded the NRC local of 200*F near the spargers (pool temperature limitreference 15) was not exceeded for an events.

A suppression pool temperature monitoring system has been designed and installed to meet the requiremerits of NUREG-0661. This system meets the acceptance criteria in a variety of ways including separate, redundant sensors at 13 locations, room temperature indicators, alarms to assist operators and seismically qualified sensors.

2.2.2 Dead Weight: The dead weight analysis was performed using the actual as-built load scttings for the spring hanger supports, which were applied as preloads on the corresponding support components in the piping model.

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2.2.~ Piye pressure: Internal pipe pressure loads were con-

' d e .a c - as specified by the original design criteria.

2.2.4 Se i smi ; Loads _: Response Spectrum analyses were per-formed for both the Operating Basis Earthquake (0BE) cred the Safe Shutdown Earthquake (SSE), using spectra specified by the original design criteria. The analysis method is consistent with Reference 2.

2.2.5 Pool Swell Loads: Dynamic torus shell motions at piping penetrations under the pool swell loading were obtained from the torus shell analysis for the pool 4 swell loads described in Reference 1. Torus shell

' motions appeied to the piping models at anchor points are in the form of six components of displacement time histories - three translations and three rota-tions. In addition, due consideration was given to the imposed displacements at torus penetrations due

" to torus pressurization.

J' 1: 2.2.6 Condensation Oscillation Loads: Dynamic torus shell motions at penetrations under condensation oscillations were applied at piping anchors in the form of harmonic j

displacements, three translations and three rotations.

4 These displacement components were obtained from i harmonic torus shell analysis described in Reference 1.

Harmonic displacements are considered within 1 Hz bands, at torus natural frequencies if any, up to 30 Hz.

. 2.2.7 Chugging Loads: Chugging loads are similar to conden-sation oscillation loads. The harmonic amplitudes were obtained from the torus shell analysis for chugging loads (Reference 1). In this case, loading frequencies up to 50 Hz were considered for the piping

i. evaluation.

2.2.8 SRV Discharge Loads: SRV discharge loads for the piping evaluation are in the form of displacement time histories at the piping ar.chor point (torus penetra-L tion), which were obtained from torus shell analysis (Reference 1). SRV discharge loads are considered

! f or two governing load cases, one for the single valve actuation under Design Basis Accident (DBA) LOCA and the other for the multiple valve actuation (Six adja-cent valves) under Intermediate Break Accident (IBA)/ ,

Small Break Accident (SBA) LOCA.

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2.3 SRV DISCHARGE LINE INSIDE THE DRYWELL LOAUS This section describes the loads considered for evaluating the SRV discharge lines located in the drywell, extending from the outlet flange connection on the SRV to the main vent penetration anchor.

2.3.1 Thermal Loads: Thermal loads in the form of appropriate Reactor Pressure Vessel (RPV) thermal nozzle displace-ments with main steam line in the hot condition at a design temperature of 585*F. The temperature of the SRV pipe would be governed by the operating mode under consideration.

2.3.2 Dead Weight: These loads were based on static dead wei ght analysis of the SRV piping model.

2.3.3 Pipe Pressure: Internal pipe pressure loads were considered based on the original design criteria.

2.3.4 Seismic Loads: Response Spectrum analyses were performed for OBE and SSE using spectra specified oy the original design criteria. The analysis method is cor,;istent with Reference 2.

2.3.5 SRV Discharge Loads 2.3.5.1 Air In the Drywell Cases The pressure transient- following the opening of an SRV results in a significant transient thrust loads on the SRV pipi ng. For all load cases Al.1 through C3.2 the derelopment of this thrust loading was done in accordance with the Mark I Containment Program - Load definition report (Reference 3). The bounding SRV load case f or the drywell porti-on of the SRV piping was determined to be load case C3.1 (multiple valve subsequent actuation under normal operating conditions).

2.3.5.2 Steam in the drywell case Load case C3.3 is the condition in which the drywell is full of steam and the subsequent SRV actuation occurs. The Mark I criteria provides a recommended value for the inter-

! facial heat transfer coefficient to describe the rate of condensation at the steam / water L interface. A large heat transfer coefficient was specified in an effort to introduce conservatism into the calculation since no 2-3 PBAPSPF2/5-5

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test data is available for reflood when steam is present in the drywell. (Fullscale tests were undertaken under normal plant operating conditions.) As condensation occurs during reflood, the temperature of the water at the interface increases. Since the RVRIZ code does not keep track of the temperature change, it is possible to exercise the code so that more steam is apparently condensed than can actually be supplied to the process. This results in the calculation of conservative reflood heights.

An analysis was perf ormed to estimate the amount of steam condensed during reflood and the resulting rise in water temperature. When sufficient water is raised to the saturation temperature, the steam and cold water will be effectively insulated from each other and condensation will cease. It has been calcu-lated that a 4 foot thermal layer will be formed by the time the water level in the SRVDL reaches 8 feet. This layer will ef fectively halt further interfacial steam condensation during Load Case C3.3.

Therefore, the heat transfer coef ficient recommended in the Mark I Criteria was used during quencher refill and initial flooding of the SRVDL. Thereafter, the heat transfer rate is set to zero.

The assumptions used in the calculation of the 8 foot cutoff of condensation are still based on conservative estimates. Hence, calculations of reflood during Load Case C3.3 will still be conservative.

2.3.6 Vent System Motion: SRV lines inside the drywell are not subjected to direct loading f rom the hydrodynamic phenomena in the suppression pool. However, they are likely to be subjected to these loads indirectly through the dynamic motions of the vent system. The motion of the main vent at SRV/ main vent intersection was evaluated and it was judged that the vent system motions L would not have any significant effects on the SRV lines.

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2.4 SRV DISCHARGE LINE IN THE WETWELL LOADS This Section describes the loads considered for evaluating the SRV discharge lines located in the wetwell, extending l from the main vent penetration anchor to the inlet of T-Quencher. Therefore, a portion of SRV discharge line is submerged.

2.4.1 Thermal Loads: The thermal loads for the wetwell SRV piping were based on the line temperature of 400 F.

2.4.2 Dead Weight: These loads were obtained in a manner similar to SRV piping in the drywell.

2.4.3 Sei smi c Loads : Seismic loads were obtained in a manner similar to SRV piping in the drywell.

2.4.4 SRV Actuation Loads: These loads also were obtained simultaneouly with those for SRV piping in the drywell, as described above.

2.4.5 Vent System Motion: Submerged portion of SRV line in the wetwell is subject to the loading from hydro-dynamic phenomena in the suppression pool. They are also likely to b'e subjected to these loads indirectly through dynamic motion of the vent system. However, it was judged that this motion would have insignificant effect on the.wetwell SRV piping.

2.4.6 Pool Swell Related Loads: Pool swell related loads on the wetwell SRV piping include pool swell impact, drag and fallback loads the for the portion of the SRV line lo.ated above the suppression pool. Also, the submerged portion of the SRV line would be subjected to drag loads induced by LOCA water jets or LOCA air bubbles. These loads are defined according to the guidelines of Reference 3 and meet the criteria of NUREG-0661 (Reference 4).

2.4.7 Condensation Oscillations and Chugging Drag Loads:

l These drag loads are harmonic in nature and include l fluid structu re interaction ef f ects. Development of these loads, which act on the submerged portion of the SRV piping during condensation oscillation and chugging phases of a LOCA, is in accordance with guidelines of Reference 3 and meet NUREG-0661 (Reference 4) criteria.

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l 2.4.8 T-Quencher Air Bubble Drag Loads: The submerged portions of the SRV discharge piping and supports are subjected to standard and acceleration drag loads caused by the oscillations of the air-steam bubbles expelled from the T-Quenchers following an SRV actuation. These loads were obtained according to the guidelines of Reference 3 and meet the criteria of NUREG-0661 (Reference 4).

2.4.9 Torus Motion: SRV discharge lines are not affected by torus motions because T-Quencher supports or t

other SRV supports are directly supported on the ring _ girders, and the ring girder with the saddle plate and columns are assumed to be rigid for load transfer purposes.

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3.0 TORUS ATTACHED PIPING l

3.1 INTRODUCTION

l This section describes the results of the evaluation perf ormed on the PBAPS piping systems externally attached to the torus.

b bore All of the piping (ODlarge'n.

< 4 i ore piping (0D>

) systems 4 inches) attached andare to torus small considered as essential under all load combinations. Components included j in the evaluation are the piping (large and small piping

' attached to the torus and small pipe branch lines attached to the large piping) and piping supports, valves, flanges, equipment nozzles, and equipment anchorages. For each component evaluated, this section provides a description of design load combinations, acceptance criteria, and results. A summary

of the components evaluated is contained in Table 3.4.1.

Analytical methodology is provided in the following sections and Reference 11.

3.2 DESIGN LOAD AND COMBINATION 3.2.1 A total of 9 design loads were considered in the analyses and tabulated in Table 3.2.1. Thermal modes were considered according to the FSAR (Reference 2) as modified by the PULD (Reference 5) PBAPS Technical Specifications (Reference 12) and G.E. flow diagrams (Reference 13) and References 14 and 15. Dead Weight, pipe pressure, and seismic loads were based on the original design criteria described in the FSAR. Hydro-dynamic loads (SRY, PS, CO, CH) considered in the analysis are described in sections (4) & (5) of Reference 3. The PUAAG (Reference 6) Table 5.2 shows the 27 design load combinations for the torus attached piping. The external torus attached piping does not experience any direct hydrodynamic loads during a LOCA or SRV discharge. These loads are applied indirectly through the motion of the torus shell penetration.

! Therefore the bounding load combinations for performing the code evaluation can be reduced to those shown in Tables 3.2.2, 3.2.3, and 3.2.4.

3.2.2 For load combinations containing more than one dynamic load case, the two largest load cases were combined by SRSS and additional dynamic load cases were added by absolute summation to the result. An al+,ernati ve accep-

! tance method was to combine all three dynamic load cases by SRSS.

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3.3 PIPING ANALYSIS 1

3.3.1 All external piping attached to the torus is listed in Table 2-2 of Reference (1) and Table 3.4.2.

3.3.2 Each piping model external to the torus was modeled from the torus penetration nozzle to an anchor or a point at which the load effects had attenuated below 10% of the design allowables. Table 3.4.3 summarizes all the branch piping connected to the large bore piping.

3.3.3 All torus attached piping was evaluated by performance of uncoupled analysis in which the displacements resulting from the torus shell motion were input into the torus. Computer program ME101 was used to evaluate all the load cases shown in Table 3.2.1. For large bore piping analysis, the type of analysis method used for each load case is given in Table 3.2.1.

3.3.4 In SRV and PS. load cases, displacement time history analyses were performed using the modal superposition method. The displacement inputs consisted of 6 compo-nents (3 translations and 3 rotations) and applied at the torus nozzle interface connection to the piping.

In these modal time history analyses, modes up to 50 Hz were considered sufficient to reflect the piping response after the research based on conversion of time domain to frequency domain through Power Spectral Density. Time step used was 0.002 seconds and the end time was 2.4 seconds.

3.3.5 For CO and CH load cases, steady state harmonic analyses were performed as follows:

1. Input frequencies were considered from 1 Hz to 30 Hz in C0 (1 Hz to 50 Hz in CH) -

2. 6 components of input displacements (3 translation and 3 rotations) were applied at each torus nozzle in the form of sinusoidal functions with appropriate amplitude and phase angle for every f requency.

3. 30 individual analyses in C0 (50 in CH) were run, each corresponding to one individual frequency.

4. In the C0 load case the results of each of the 30 modes were compared for each component. The 4 largest results were summed absolutely, and the 26 remaining results were combined by SRSS and added to the 4 largest results. In the CH load case, the results of each of the 50 modes were combined by absolute summation.

3-2 PBAPSF2/1-2

i In harmonic analysis, the moduli (vector sums of real and imaginary parts) were considered for the resulting stresses, support reaction loads and displacements, and forr.e and moment components in the computer analysis output.

All hydrodynamic piping analyses were performed using damping values consistent with the guidelines in NRC Regulatory Guide 1.61 (Reference 7).. The damping values (percent of critical damping) used in the analyses are shown in Table 3.3.1.

3.4 ACCEPTANCE CRITERIA 3.4.1 The Code used to evaluate the acceptability of the design of the existing piping systems was the ASME 1977 BPV Code Section III summer 1977 addenda as recommended by Reference 5. The original Code used for the initial design of the piping systems was the ANSI B31.1 Code 1967 Ed. which is still valid for the static and seismic analyses. However, an evaluation was performed which indicated that the 1973 edition of B31.1 was equal to or more conservative than the 1967 edition:

therefore, the static and seismic calculations were evaluated to ANSI B31.1 Code 1973 Ed. including the summer 1973 addenda. ANSI B31.1-1977 Ed. was used for stress iiitensification f actors. The maximum pipe stresses, as defined by equations 8 through 11 in Sub-section NC-3650 of the Code, resulting from the bounding load combinations in Tables 3.2.2, 3.2.3, and 3.2.4 were all shown to be within the corresponding Code allowables.

Existing pipe supports were evaluated to the ANSI B31.1 1973 through Summer 1973 addenda. New pipe supports or modifications to existing supports were designed to ASME Section III subsection NF, 1977 edition through Summer 1977 addenda. Structural components were evaluated to Reference 8.

3.4.2 All the components, flanges, valves, equipment nozzles and equipment anchorages were analyzed or' mechanically supported to meet stress and operability requirements for all load conditions specified by the PUAAG and as noted herein. Flanges were evaluated, based on NC-3658.2 and NC-3658.3 of Section III 1979 summer Addenda.

Valves, equipment nozzles and equipment anchorages were evaluated based on the original criteria or detailed analysis was performed.

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3.5 FATIGUE EVALUATION 3.5.1 The fatigue evaluations for torus attached and SRV discharge piping are incleded in the-joint efforts of the Mark I Owners Group. A generic response (Reference 9) has been presented to the NRC by the Mark I A/E task group. The results show that all piping systems included in the Mark I program have cumulative usage f actors less than 0.5, most being below 0.3.

3.5.2 The SRV discharge lines and torus attached piping are analyzed in accordance with ASME Section III Class 2/3 rules. The design loads and the load combinations are established based on NUREG-0661. The equivalent maximum stress cycle factors are used to determine the equivalent full stress cycles for each loading. This procedure is similar to the approach used for Mark II plants and uses an exponent consistent with that specified for ASME Class 2 piping cyclic load analysis, NC-3611.2. Equation (11) f rom NC-3652.3 is used to evaluate the' stress range for each load combination (including dynamic loads). The usage factors are determined using the Markl fatigue curve which is consistent with Class 2/3 rules.

3.5.3 For Peach Bottom Units 2 & 3, the calculated stresses for each Mark I load were reviewed for each piping system. The three most critical components (2 for torus attached piping, 1 for SRV discharge lines) were selected for fatigue evaluations. The maximum usage f actor for these 3 components considering 2 event sequences (N0C + DBA, NOC + IBA/SBA) is 0.202.

(N0C = Normal Operating Conditions).

3.6 MARK-I ASSESSMENT RESULTS 3.6.1 The objective of the Mark I containment Long Term Program (LTP) for PBAPS is the restoration of originally intended design safety margins for the new suppression pool hydrodynamic loads. All required analyses have been performed and modifications have been or will be made, where appropriatc, to ensure that this objective is met.

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TABLE - 3.2.1 Design Load and Analysis Method NO LOAD CASE ANALYSIS METHOD 1 Thermal Static 2 Dead Weight Static 3 Pipe Pressure Static 4 Operational Basis Earthquake (0BE) Response Spectrum 5 Safe Shutdown Earthquake (SSE) Response Spectrum (l) 6 Safety Relief Valve discharge (SRV) Time History (2) 7 Pool Swell (PS) Time History (2) 8 Condensation Oscillation (CO) Harmonic (2) 9 Chugging (CH) Harmonic (2)

Note: (1) SSE = 2.4 (0BE) .

(2) In small bore piping evaluation, equivalent static analyses were performed.

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TABLE 3.2.2 Piping Stress Combinations Mark I .

Equation S-1. ASME III Eq. 8 (A) P+W i Sh -

S-2. ASME III Eq. 9 (B) P+W + SRVall i 1.2 Sh S-3. ASME III Eq. 9 (B) P+W + OBE i 1.2 Sh S-4. ASME III Eq. 9 (C) P+W + OBE 2 + SRV all 2 1 1.8 Sh S-5. ASME III Eq. 9 (C) P+W + SRV all + CH2 1 1.8 Sh The following three equations refer to ASME III Eq 9 (D).

S-6 Eq 9(FB) ='P+W+ D2 +E 2 + F 12.4 Sh where D&E are the two largest values of: SSE, SRVone and PS. F is the smallest value of: SSE, SRV o and PS. (See Section 3.2.23 S-7 Eq 9(FA) = P+W+ A2 +B 2 + C 1 2.4 Sh where A&B are the two largest values of: SSE, SRVall and CH. C is the smaTlest value of: SSE, SRVal and CH. (See Section 3.2.2 S-8 Eq 9(FC) = P+W + SSE2+C0 2 1 2.4 Sh (See Note 6 Table 3.2.4)

.S 9 c o n d a ry Stress Equation Thermal expansion stresses and seismic anchor movement stresses due to 0BE shall be considered as required by ASME Section III, subsection NC-3652.3 (Eq.10) or (Eq.ll).

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TABLE 3.2.3 PIPE SUPPORT, VALVE AND EQUIPMENT N0ZZLE LOADING COMBINATIONS PBAPS UNITS 2 AND 3 DEFINITIONS:

Normal = ASME Level A Loading Upset = ASME Level B Loading

. Faulted = ASME Level D Loading THRM - As defined on Table C lWT + THRMl = max, of (WT) or (WT + THRM)

• LOAD COMBINATIONS:

Normal:

S-1: lWT + THRMl Up. set: Maximum of the following: -

S-2: lWT + THRMl + OBE S-3: lWT + THRMl + SRVall Faulted: Maximum of the following:

S-4: (Part 1): Place the 2 largest of the 3 terms SSE, SRV one and PS under the radical and solve the appropriate equation: (Also see Section 3.2.2)

A) lWT + THRM l +/SSE 2 + 3RV2 one + PS j, or B) lWT + THRM l+/SSE 2 + p3]2 + SRV one' or C) lWT + THRM l +/SRV2 one +E3 1

+ SSE,

• - Unless noted otherwise in analysis.

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TABLE 3.2.3 PIPE SUPPORT, VALVE AND EQUIPMENT N0ZZLE LOADING COMBINATIONS (Part 2): lWT + THRM l + /SSE 2+C 2 where C =

max of C0 or CH (Part 3): Place the 2 largest of the 3 terms SSE, SRVall and CH under the radical and solve the appropriate equation:

(See also Section 3.2.2)

A) lWT + THRM l + /'SSE 2 + SRV2 all +CH, B) lWT + THRM l+/SSE2 + CH 2 + SRV all*

4 or C) lWT + THRM l+/SRV2 all + CH2 + SSE, (Part 4): lWT + THRM l+ SSE NOTE: 1) SSE = 2.4 x OBE SRVone

= 2 x SRV SRVall = 4 x SRV PS1

= 1.1 x PS loads 3-8 PBAPSPF2/1-8

TABLE 3.2.4 .

CORRELATION OF STRESS, PIPE SUfPORT AND NUREG-0661 LOADING COMBINATIONS ,

Stress (1) Pip.e Support (2) Combinations (5) Enveloped (5) '

Equation Equation Utilized C o mb i n a t i o n s_. y S-1 S -1 -

S-2 S-3 1 S-3 S-2 '

S-4 -

2.

S-5 - 11 4,5,10 S-6 S-4 25 _

16,18,19,22,24 S-7 S-4 15 (4) 3,6,7,8,9,12, 13,14,2i S-8 S-4 26/27 (3,'.6) 17,20,23 ,

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j. NOTES: .

(1) Reference - Table 3.2.2 (2) Reference - Table 3.2.3 (3) SSE used in combination with C0 for conservatism ,

(4) DBA (CH) envelopes IBA (C0 & CH) and SBA (CH)

(5) NUREG-0661 Fi gu re 4.3-2 (Ta bl e 5.1 )

(PUAAG Table 5.2)

(6)

Combination with CH is enveloped by equation S-7. -

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!r PERCENT OF CRITICAL DAMPING FROM REG. GUIDE 1.61 Line Size __

Load SRV* PS, CO, CH Pipe 0.D.> 12 inches 2 3 Pipe 0.0.< 12 inches 1 .

2 s

Note:

• SRV for Level,C'and D could be increased 1%. However, PBAPS conservatii'ely used lower damping as shown in this table. -

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Table 3.4.1 Summary of Components Evaluted Unit 2 STRESS PENETRATION NO. OF FLANGES VALVES l EQUIPMENT NUZZLES ISO N0. N0ZZLE NO. EVALUATED EVALUATED l EVALUATED 26B (VAC Breaker)

P-9-1 N205A 1 A02502B __

N205B KD2502A,A02521B,A02521A, N219 A02511, A02512, A02515 A02519,A02505, A02520 (Non Mark I Impact) l P-10-1 N226A - M017 20P35, RHR Pump Suct.

N226B M015D 2BP35, RHR Pump Suct.

M0130 M013B M015B 1

P-10-2 N226C - M015C 2AP35, RHR Pump Suct.

N226D M015A 2CP35, RHR Pump Suct.

. M013C M013A P-10-3 N210A 3 M0174 2BE24, RHR Heat Exch.

N211 A M0176 Outlet M025B 2BE24, RHR Heat Exch.

M026B Service Water Inlet M031B M0154B M034B

, M038B -

M039B 1

P-10-4 N210B 1 M033 2CE24, RHR Heat Exch.

N2118 M026A Outlet M039A M025A M031A M0154A M020 M038A M034A CV-43 (GE) 3-11 PBAPSPF2/4-7

TABLE 3.4.1 Summary of Components Evaluated Unit 2 STRESS PENETRATION NO. OF FLANGES VALVES EQUIPMENT N0ZZLES ISO N0. N0ZZLE NO. EVALUATED EVALUATED EVALUATED P-13-1 N212 2 - 20S38, RCIC Turbine Exhaust 1

P-13-2 N225 2 M041 20P36, RCIC Pump Suct.

M039 [20P142 WCU Pump Suct]

M070 (Non-Q)

M071 M018 P-14-1 N228C - M07B 20P37, Core Spray Pump Suction N2280 M070 2BP37, Core Spray Pump Suction j

P-14-2 N228A - -

M07A 2CP37, Core Spray Pump Suction N2288 M07C 2AP37, Core Spray Pump Suction P-14-3 N224 3 M012A 2CP37, Core Spray Pump Disch.

M0llA 2AP37, Core Spray Pump I Disch.

MOSA M05C -

M026A P-14-4 N229A 3 MOSB 2DP37, Core Spray Pump Disch.

N234 M05D 2BP37, Core Spray Pump Disch.

M0llB M012B M026B P-23 N214 3 - 20S37, Turbine Exhaust P-23-2 N227 2 M017 20P33, HPCI Pump Suct.

M057 M058 3-12 PBAPSPF2/4-8

TABLE 3.4.1 Summary of Components Evaluated Unit 2 STRESS PENETRATION .NO. OF FLANGES VALVES EQUIPMENT N0ZZLES ISO NO. N0ZZLE NO. EVALUATED EVALUATED EVALUATED P-23-3 N233 2 M020 20P36, RCIC Pump Disch.!

M030 M021 M031 P-23-4 N216 2 M025 20P38 HPCI Pump Disch.

i I

3-13 PBAPSPF2/4-9 4

- - - - - - , . - . , , ,.-.__._.._,_,,,,,,..,,r,, _ _ , , , _ , , _ , , , _ . , _

Table 3.4.1 (Cont 'd)

Summary of Components Evaluated Unit 3 STRESS PENETRATION NO. OF FLANGES VALVES EQUIPMENT NUZZLES ISO N0. N0ZZLE N0. EVALUATED EVALUATED EVALUATED .

26B (VAC Breaker)

P-9-2 N205A - A03502B N205B A03502A,A03521B,A03521A,__

A03520, A03519, A03505 N219 A03511, A03512

._(Non Mark I Impact) -

P-10-7 N226C - M013A 3AP35, RHR Pump Suct.

M015A 1

P-10-8 N226A -

M013B 3DP35, RHR Pump Suct.

M013D 3BP35, RHR Pump Suct.

N226B .M017 l M015B M015D l

P-10-9 N210A -

M033 3BE24, RHR Heat N211A M031B Exchanger Outlet M020 M026B M039B '

M0154B M026B M034B M038B CV43 (GE)

P-10-10 N226D -

M013C 3CP35, RHR Pump Suct.

M015C P-10-12 N210B 3 M0154A 3CE24, RHR Heat Exch.

N211B M025A Outlet M026A 3CE24, RHR Heat Exch.

M031 A Service Water Inlet M0174 M0176 M034A M038A M039A 3-14 PBAPSPF2/4-1 e

Summary of Components Evaluated Unit 3 STRESS PENETRATION N0. OF FLANGES VALVES EQUIPMENT N0ZZLES N0ZZLE NO. EVALUATED EVALUATED EVALUATED ISO NO.

P-13-3 N212 2 - 30S38, RCIC Turbine Exhaust

\ l P-13-4 N225 3 M070 30P36, RCIC Pump Suct.

M071 [30P142, WCU Pump Suct]

M041 (Non-Q)

M039 M018

\

l P-14-5 N228C -

M07A 3AP37, Core Spray Pump '

Suction N2280 M07C 3CP37, Core Spray Pump Suction P-14-6 N228A 1 M07D 3DP37, Core Spray Pump Suction N228B M07B 3BP37, Core Spray Pump Suction P-14-7 N234B 4 M05C 3CP37, Core Spray Pump Disch.

N236B MOSA 3AP37, Core Spray Pump

- M012A Disch.

M0llA M026A 1

1 P-14-8 N234A 4 M05B 30P37, Coce Spray Pump Disch.

N236A M050 3BP37, Core Spray Pump M0llB Disch.

M012B M0268 1

P-23-5 N214 3 - 30S37, HPCI Turbine Exhaust

-l P-23-6 N227 2 M057 30P33, HPCI P=p Suct.

M058 M017 l

l 3-15 PBAPSPF2/4-2

Summary of Components Evaluated Unit 3 STRESS PENETRATION NO. OF FLANGES VALVES I EQUIPMENT N0ZZLES ISO NO. N0ZZLE NO. EVALUATED EVALUATED  ! EVALUATED P-23-7 N235 2 M030 30P36, RCIC Pump Disch.

M020 M021 M0503 P-23-8 N216 2 M025 30P38 HPCI Pump Disch.

3 3-16 PBAPSPF2/4-3

Table 3.4.2 Peach Bottom Mark I - Units 2 and 3 Summary of External Large Bore Piping Attached to Torus UNIT 2 UNIT 3 TORUS TORUS STRESS PENETRATION STRESS PENETRATION ISO NO. N0ZZLE N0. ISO NO. N0ZZLE NO.

' N205A P-9-1 N205A i-9-2 N205B N205B N219 N219 P-10-1 N226A, N226B P-10-7 N226C P-10-2 N226C, N226D P-10-8 N226A, N226B P-10-3 N210A, N211 A P-10-9 N210A, N211A P-10-4 N2108, N211B P-10-10 P2260 l P2108, N211B P-13-1 N212 P-10-12 P-13-2(l) N225 P-13-3 N212 P-23-2 N227 P-13-4 N225 P-14-1 N228C, N2280 P-14-5 N228C, N2280 P-14-2 N228A, N228B P-14-6 N228A, N228B P-14-3 i N224 P-14-7 L N2348,N236B(2)

P-14-4 N229(2),N234 P-14-8 N234A, N236A( )

P-23-1 N214 P-23-5 N214 P-23-3 N233(2) P-23-6 N227 P-23-4 N216(2) i P-23-7 N235(2)

P-23-8 N216(2)

Note: (1) P-13-2 and P-23-2 are analyzed as one piping system (2) Actual attached piping is 4" nominal diameter but analyzed by using large bore piping approach.

3-17 PBAPSPF2/4-4

Table 3.4.2 (Cont 'd)

Peach Bottom Mark I - Units 2 and 3 Summary of External Small Bore Piping Attached to Torus

~~

Stress Torus Isometric Drawing Line Nozzle Unit 2 Unit 3 Size System Identification N203 SP-9-21 SP-9-22 1" 0xygen Analyzer N206A SP-23-23 SP-23-26 2" Level and Pressure Instrument B SP-23-23 SP-23-26 2" Level and Pressure Instrument N209A No supports 1" Air and Water Temperature to D No supports 1" Air and Water Temperature N217B SP-23-25 SP-23-24 3" RCIC Turbine Exhaust Vacuum Relief N218A FSK-M438-ll4 SP-35-22 1" Instrument Air and Oxygen B SP-9-21 SP-9-22 1" Instrument Air and Oxygen C SP-35-21 FSK-M3035-58 1" Instrument A1: and Oxygen N221 SP-13-21 SP-13-22 2" RCIC Vacuum Pump Discharge N223 SP-23-21 SP-23-22 2" Condensate from HPCI Turbine Drain N230 SP-13-21 SP-13-22 2" RCIC Pump Recirculation N215 FSK-M438-133 --- 2" Level and Pressure Instrumenta-tion N250 ---

FSK-M3023-42 2" Level and Pressure Instrumenta-tion N213A FSK-M438-128 FSK-M3023-43 2" Construction Drains With

-41 Torus Water Instrumentation 3-18 PBAPSPF2/4-5

Table 3.4.3 Peach Bottom Mark I - Unit 2 Summary of Small Bore Branch Lines Connected to Large Bore Pipe SMALL PIPE LARGE PIPE SMALL PIPE LARGE PIPE SMALL PIPE LARGE PIPE 150. NO. ISO. NO. 150. NO. ISO. NO. ISO. N0. ISO. NO.

FSK-M-428-8 P-9-1 FSK-M-438-54 P-10-4 FSK-M-775-14 P-14-3 FSK-M-428-40 FSK-M-448-28 FSK-M-438-43 FSK-M-428-17 FSK-M-443-19 FS K-M-7 75-13 P-14-4 FSK-M-439-40 FSK-M-448-9 FSK-M-438-141 FSK-M-438-20 P-10-1 FSK-M-439-39 FSK-M-438-42 FSK-M-430-122 FSK-M-439-52 FSK-M-438-86 FSK-M-438-104 FSK-M-443-18 FSK-M-435-43 F SK-M-448-24 FSK-M-775-27 FSK-M-775-17 FSK-M-430-123 FSK-M-448-23 FSK-M-775-16 FSK-M-448-16 FSK-M-438-106 FS K -M-775-1 FSK-M-438-26 FSK-M-438-51 FSK-M-775-2 FSK-M-448-7 FSK-M-443-16 FSK-M-438-136 FSK "-438-78 FSK-M-775-26 FSK-M-775-25 FSK-M-438-14 FSK-M-448-10 FSK-M-775-15 FSK-M-438-9 10GB-4" P-10-5 FSK-M-438-85 FSK-M-448-17 10GB-4" P-10-6 FSK-M-438-38 FSK-M-448-38 FSK-M-430-181 P-13-1 FSK-M-430-li7 P-23-1 FSK-M-443-9 FSK-M-438-91 FSK-M-430-183 FSK-M-438-77 P-10-2 FSK-M-430-179 FSK-M-430-115 FSK-M-438-19 FSK-M-430-180 FSK-M-772-9 FSK-M-438-101 FSK-M-430-158 FSK-M-772-7 FSK-M-438-102 FS K-M-772-14 F S K -M-772-8 FSK-M-448-8 FSK-M-430-98 FSK-M-430-159 FSK-M-448-17 FSK-M-430-146 FSK-M-430-178 FSK-M-438-33 FSK-M-430-153 FSK-M-430-188 FSK-M 448-16 FSK-M-430-104 P-13-2 FSK-M-430-97 FSK-M-438-15 FSK-M-430-157 FSK-M-438-138 FSK-M-438-30 FSK-M-430-102 FSK-M-430-106 P-23-2 FSK-M-438-103 FSK-M-430-152 FSK-M-438-126 FSK-M-438-49 P-10-3 FSK-M-430-175 FSK-M-430-135 FSK-M-443-8 FSK-M-775-21 P-14-1 F5K-M 430-100 P-23-3 FSK-M-443-10 FSK-M-438-39 FSK-M-430-137A FSK-M-448-33 FSK-M-775-19 FSK-M-772-13 FSK-M-448-39 FSK-M-438-41 FSK-M-430-182 FSK-M-438-53 FSK-M-438-92 FSK M-430-194 FSK-M-438-50 FS K-M-7 7 5-18 P-14-2 FSK-M-430-171 FSK-M-448-25 FSK-M-438-46 FSK-M-430-172 FSK-M-775-28 FSK-M-438-92A FSK-M-438-124 P-23-4 FSK-M-448-ll FSK-M-438-44 FSK-M-438-125 FSK-M-448-1 FSK-M-775-20 FSK-M-430-134 FSK-M-448-25 FSK-M-438-83 P-14-3 FSK-M-438-127 FSK-M-448-6 FSK-M-775-4 FSK-M-438-192 FSK-M-438-105 FSK-M-438-140 FSK-M-438-110 FSK-M-448-22 FSK-M-775-3 FSK-M-439-39 FSK-M-775-22 FSK-M-439-50 FSK-M-438-45 FSK-M-448-2 P-10-4 FSK-M-438-137 FSK-M-448-G FSK-M-775-24 FSK-M-448-12 FSK-M-448-34 3-19 PBAPSPF2/4-6

Table 3.4.3 (Cont'd)

Peach Bottom Mark I - Unit 3 Summary of Small Bore Branch Lines Connected to Large Bore Pipe LARGE PIPE SMALL PIPE LARGE PIPE SMALL PIPE LARGE PIPE SMALL PIPE ISO. N0. ISO. N0. '

ISO. N0. ISO. NO. ISO. NO. ISO. NO.

FSK-M-3009-1 P-9-2 FSK-M-3010-28 P-10-12 FSK-M-3014-3 P-14-7 FSK-M-3052-3 FSK-M-3010-92 FSK-M-3014-53 FSK-M-3009-6 FSK-M-3010-112 FSK-M-3014-43 FSK-M-3010-38 P-10-7 FSK-M-3052-2 FSK-M-3014-38 FSK-M-3010-45 FSK-M-3010-13 FSK-M-3014-56 FSK-M-3010-49 FSK-M-3010-90 FSK-M-3014-17 FSK-M-3010-59 FSK-M-3010-88 FSK-M-3014-7 FSK-M-3010-74 P-10-8 FSK-M-3010-22 FSK-M-3014-33 P-14-8 FSK-M-3010-46 FSK-M-3032-10 FSK-M-3014-14 FSK-M-3010-19 FSK-M-3010-5 FSK-M-3014-39 FSK-M-3010-18 FSK-M-3010-10 FSK-M-3014-49 FSK-M-3010-ll3 FSK-M-3010-6 FSK-M-3014-51 FSK-M-3010-62 FSK-M-3010-108 FS K-M-3014-ll FSK-M-3010-52 FSK-M-3013-40 P-13-3 FSK-M-3014-27 FSK-M-3010-47 SP-23-24 FSK-M-3014-9 FSK-M-3010-31 FSK-M-3013-18 FSK-M-3014-52 FSK-M-3010-53 FSK-M-3013-34 FSK-M-3014-55 FSK-M-3.010-32 FSK-M-3013-10 FSK M-3014-28 FSK-M-3010-61 FSK-M-3013-14 FSK-M-3014-29 FSK-M-3010-57 FSK-M-3013-35 F S K-M-3023-32 P-23-5 F SK-M-3010-27 P-10-9 FSK-M-3013-ll FSK-M-3023-25 FSK-M-3052-2 FSK-M-3013-27 P-13-4 FSK-M-3023-49 FSK-M-3010-16 FSK-M-3013-29 FSK-M-3023-36 FSK-M-3010-95 FSK-M-3013-42 FSK-M-3023-58 F SK-M-3010-91 FSK-M-3014-45 P-14-5 FSK-M-3023-12 FSK-M-3010-15 FSK-M-3014-4 FSK-M-3023-24 FSK-M-3010-12 FSK-M-3014-47 FSK-M-3023-22 P-23-6 FSK-N-3010-25 FSK-M-3014-6 FSK-M-3023-13 FSK-M-3010-29 FSK-M-3014-46 P-14-6 FSK-M-3023-54 FSK-M-3010-89 FSK-M-3014-9 FSK-M-3013-2 P-23-7 FSK-M-3010-8 FSK-M-3014-ll FSK-M-3013-32 FSK-M-3010-93 FSK-M-3014-48 FSK-M-3013-43 FSK-M-3010-14 FSK-M-3027-46 FSK-M-3013-24 "

FSK-M-3010-104 FSK-M-3014-16 P-14-7 FSK-M-3013-25 FSK-M-3010-23 FSK-M-3014-37 FSK-M-3023-31 P-23-8 FSK-M-3010-87 FSK-M-3014-40 FSK -M-3023-51 FSK-M-3010-99 FSK-M-3014-41 FSK-M-3010-9 FSK-M-3014-6 FSK-M-3010-35 P-10-10 FSK-M-3014-26 FSK-M-3010-43 FSK-M-3014-25 FSK-M-3010-50 FSK-M-3014-4 FSK-M-3010-60 10GB-4" P-10-l l 10GB-4" P-10-13 3-20 PBAPSPF2/4-/

l

4 4.0 STRUCTURAL STEEL MODIFICATIONS FOR LOADS FROM TORUS ATTACHED PIPING HANGERS

4.1 INTRODUCTION

4.1.1 This section describes the structural steel mooifica-tions required for the loads from torus attached piping supports. Pipe hangers for 15 large piping systems of Unit 2 and 16 large piping systems of Unit 3 were reviewed. It was determined that s t ru ct u -

ral steel modifications were required for loads from 9 systems of Unit 2 and 8 systems of Unit 3. Pipe hangers for remaining systems were either supported on concrete slabs or walls without inducing additional load on structural steel beams or the additional loads induced were insignificant. Loads from branch lines and small pipes were not significant enough to require any structural steel modifications.

4.2 LOAD COMBINATIONS AND ALLOWABLE STRESSES 4.2.1 The existing structural steel is designed f or load combinations and allowable stresses as given in Reference (2). The same criteria is used for the design of new beams. The ultimate strength design method was used to evaluate concrete slabs supporting the hangers. Allowable loads of expansion anchors are determined according to manuf acturer's recommenda-tions with safety factors per NRC IE Bulletin 79-02.

4.3

SUMMARY

OF DESIGN, ANALYSIS AND MODIFICATIONS 4.3.1 The pipe hangers attached to structural steel beams are basically spring hangers, ri gid hangers or anchors.

Depending upon the end connections, these hangers cause vertical, lateral and axial loads and biaxial bending and torsional loading on the structural steel beam. Modifications we e designed when the combined stresses from these new loads and existing dead and live loads exceeded the allowable stresses.

4.3.2 The modifications were designed on a case by case basis considering the speci fic f raming, the new loads, existing field conditions and space availability.

4.4 CONCLUSION

S The existing structural steel beams were capable of taking additional loads in most cases. For other cases, the supporting s', eel beams closer to the nozzles experienced the heavier loads and modifications were designed to take the new loads.

4-1 PB AP SPF 2/1 -12

5.0 SRV DISCHARGE PIPING

5.1 INTRODUCTION

This section describes the methods of analyses, load cases considered and design data utilized for the stress evaluation of Peach Bottom Units 2 and 3 safety relief valve (SRV) discharge piping systems. Eleven SRV discharge lines are located in each of the two units. Each SRV discharge line was given a unique line designation label wita label designa-tions running sequentially from 71A to 71L with the exception of 71I which was purposely omitted. The SRV discharge lines extend from the outlet flange connection on the safety / relief valves located in the drywell, through the vent-pipe penetra-tion to their connections on the T-Quenchers in the wetwell.

5.2 DESIGN INFORMATION The SRV discharge systems are classified as essential Class 3 piping systems. Design allowables and load combinations for the stress evaluation of the SRV discharge lines were taken from the ASME Boiler and Pressure Vessel Code Section III, Subsection ND-3600 and the Mark I Structural Acceptance Criteria in the PUAAG, respectively. Table 5.1 lists all the applicable design load combinations for the SRV discharge lines. The 27 combinations were consolidated into the bounding load combinations shown in Tables 5.4 and 5.6 for drywell and wetwell piping respectively.

5.2.1 Piping Stress Allowables The SRV discharge line piping stress allowables Sh and Sc were taken from ASME Boiler and Pressure l Vessel Code Section III, Appendix I.

Footnotes 3 and 4 in Table 5.1 allow for Service Levels C and D stress allowable values respectively to be used in lieu of Service Level B stress allow-ables as long as other fatigue and operability criteria are met. One exception to this criterion is as follows; load combination 3 f rom Table 5.1 was considered as a faulted load case. Service Level D limits were used for this load combination.

5.2.? Stress Intensification Factors The piping and component stress intensification factors were taken from ASME Boiler and Pressure Vessel Code Section III, Sub-article ND-3673.2(b).

5.2.3 Applicable Edition of Code The SRV discharge piping systems were analyzed to ASME Boiler and Pressure Vessel Code Section III 1977 Edition through the Summer Addendum of 1977.

5-1 PBAPSPF2/1-13

5.2.4 Structural Damping values The f.tructural damping values utilized in the dynamic transient analyses of the SRV discharge lines were based on the damping coefficients given in NRC Regu-l at ory Guide 1.61. Structural damping values utilized in the seismic analysis are based on the FSAR.

l

5.3 DRYWELL SRV DISCHARGE LINES Eleven SRV discharge lines are located in the drywell for each of Peach Bottom Units 2 and 3, extending from the outlet flange connection on the safety / relief valve to the vent-pipe pene-tration anchor. One 8" vacuum breaker valve is located on each SRV discharge line. Each of the 11 SRV discharge lines in each Unit is routed and supported differently but the routings and support configurations in one unit are nearly identical to those in the second unit. The 11 SRV discharge lines for Unit 3 were individually analyzed with their results applicable and utilized for their corresponding lines in Unit 2. The following subsections describe the analyses performed on the SRV discharge piping in the drywell for Unit 3.

5.3.1 Piping Models i

The computer program used to analyze the SRV discharge lines in the drywell was ME101. Below is a descrip-tion of the contents in.each computer model.

5.3.1.1 Weight, Thermal' Expansion and Seismic Inertia Model Four separate computer models were used for the weight, thermal expansion and seismic inertia (0BE and SSE) load case stress evaluation of the 11 SRV d-is' charge lines in the drywell . Each computer model consisted l of a main steam line and their associated

! attached SRV lines (see Table 5.2 for the l contents of each computer model). In each ,

! computer model, the main steam line was

! modeled from the Reactor Pressure Vessel (RPV) nozzle to its flued head anchor located down-j stream of the isolation valve with all attached SRV discharge lines included extending from their branch connection on the main steam line to their vent-pipe penetration anchor.

All attached spring hangers (in weight analyses only) and anchors (located at the RPV nozzles, flued heads and vent-pipe penetrations ) were 5-2 PB P AS PF 2 /1 -14

i modeled rigid. Stiffnesses were calculated and utilized for all snubbers in the computer models. Concentrated masses were lumped at valve C.G. locations. Additional masses representing contributions from restraint-hardware (spring hangers and snubbers) were modeled on the piping system at their respec-tive locations.

5.3.1.2 SRV Discharge Loads Model i

Each SRV discharge line in the drywell was individually analyzed from the branch location l on the main steam line up to and including the vent-pipe penetration anchor. The branch pipe connection on the main steam line was modeled as an anchor with equivalent stiff-nesses calculated and input to represent the stiffness effect of the main steam line. The vent-pipe penetration anchor was modeled r i g i d .-

Stiffnesses were calculated and utilized for all snubbers in each computer model. Masses were lumped at valve C.G. locations as well as the imposition of additional masses in the piping systems due to restraint hardware (spring hangers and snubbers). Input SRV discharge loads were applied to their appro-priate piping segments.

5.3.2 Load Case Analysis Methods Listed below is a summary of the analysis methods utilized for each load case analyzed for the SRV discharge lines in the drywell .

5.3.2.1 Weight A static dead weight analysis was perfomed on each of the four computer models.

5.3.2.2 Thermal Expansion l

! A static analysis was performed on each of the four computer models. In each computer model, appropriate RPV thermal nozzle displacements were applied with the main steam line analyzed l

in the hot condition 3t a desi gn temperature

of 585*F. Attached'SRV lines were analyzed either hot (at a design temperature of 500*F)

! or cold (at 70*F) based on postulated modes of SRV actuation (see Table 5.3 for a list of '

thermal analysis conditions of in each computer' model).

l 5-3 P B A P S PF 2/1 -15

5.3.2.3 Seismic Inertia A modal superposition response spectrum analy-sis was performed on each of the four computer models. The OBE and SSE load cases were analyzed with app.ropriate response spectrum curves applied in the XY and YZ global direc-tions. The minimum cut-off frequency used in any computer model analysis was 20 Hz.

Results are based on the maximum of XY or YZ earthquake.

5.3.2.4 SR V Actuation A modal superposition force-time history analy-sis was performed for each SRV discharge piping model. App ropri ate f orce-time hi story input loads were applied to every piping seg-ment starting from the first piping segment downstream of the safety / relief valve up to and including the pipe segment leading into the

• ent-pipe anchor. A damping ratio of 1% of critical was used. Each computer model was analyzed to a minimum specified cut-off f requency of 200 Hz, a time increment of 0.0006 sec. and a time duration of 0.40 sec.

5.3.3 Results Table 5.4 lists the bounding load combinations con-sidered, extracted from Table 5.1, for the SRV. discharge lines in the drywell. All 11 SRV discharge lines in the drywell complied with Table 5.4 stress criteria and load combinations. Maximum stresses generated for each load case and pipe size were added by absolute summations. SRSS combination is acceptable and was used for support evaluation on a case by case basis to remove conservatism. Fatigue analysis was consi-dered as stated in Section 3.5.

5.4 WETWELL SRV DISCHARGE LINES Eleven SRV discharge lines are located in the wetwell for each of Peach Bottom Units 2 and 3, extending from the vent-pipe penetration anchor to their connection on the T-Quenchers.

Each unit has three unique routings and support configurations representing the 11 SRV discharge lines. The three unique configurations f or one unit are identical to those of the second unit. The three unique SRV discharge line configurations for Unit 2 were individually analyzed with their results applicaule and utilized for their corresponding lines in Unit 3. The following subsections summarize the analyses performed on the SRV discharge piping in the wetwell for Unit 2.

5-4 P B A P S PF 2 /1 -16

v 5.4.1 Piping Models

~

Three separate computer models, utilizing computer program TMRSAP of Teledyne Engineering Services, w2re

, used f or the weight, thermal expansion, seismic inertia (0BE), SRV discharge and pool swell load. case stress evaluation of the 11 SRV discharge lines in

! the wetwell. Each computer model represented one of l- the three unique line and support configurations (See Table 5.5 for the contents of each computer model).

In each computer model, the SRV discharge lines were

modeled from the vent-pipe penetration anchor to the I

connection on the'T-Quencher. Stiffnesses were input at the vent-pipe anchor, intermediate restraints and the T-Quencher connection. Anchor and rigid support

stiffnesses were modeled in as they were deemed required because of the relative rigidity of the j piping in the wetwell.

5.4.2 Load Case Analysis Methods

Listed below is a summary of the analysis methods

utilized for each load case analyzed for the SRV discharge lines in the wetwell.

5.4.2.1 Weight A static dead weight analysis was performed j on each of the three computer models.

i 5.4.2.2 Thermal Expansien A static analysis was performed on each of the three computer models. Each model was analyzed i in the hot (400*F) condition.

4

5.4.2.3 Seismic Inertia t

A modal superposition r;sponse spectrum analysis was performed in each of the three computer models. The first mode of vibration of the piping systems was found to be above 37 cycles.

An OBE load case analysis was performed on each computer model to the first 20 modes of vibration with the' input response spectra accel-

! eration values applied in the rigid range above 20 Hz. The results of the OBE computer models were multiplied by 2.4 to obtain SSE results.

5-5 PB APS PF2 /1 -17 a

?

j.-

5.4.2.4 SRV Actuation A modal superposition force-time history analysis was performed for each SRV discharge piping model. Appropriate f orce-time history input I

loads were applied in each model. A damping ratio of 1% of critical was specified. Each computer model was analyzed to a specified c'ut-off frequency of 300 cps with time incre-  !

ments set of 0.001 second and with a time duration of 0.560 seconds.

I 5.4.2.5 LOCA Pool Swell Impact and Drag Loads l

A modal superposition force-time history analy-sis was performed on each SRV discharge piping model. Appropriate LOCA pool swell impact and drag input loads were applied simultaneously along the entire piping system in each piping ,

model. A damping ratio of 1% of critical was used. Each' computer model was analyzed to a specified cut-off frequency of 300 cps with time increments set a 0.001 second and with a time duration of 0.285 seconds.

5.4.3- Results Table 5.6 lists the bounding load combinations consi-dered, extracted from Table 5.1, for the SRV discharge ,

lines in the wetwell. All SRV discharge lines in the wetwell complied with Table 5.6 stress criteria and load combinations. Maximum stresses generated for each load case and pipe size were added by absolute summation. Fatigue analysis was considered as stated in Section 3.5.

S 5-6 PBAPSPF2/1-18

TABLE 6-1 CLASS 2 AND 3 PIPING SYSTEMS SBA SBA + EQ SBA+SRV SBA + SRV + EQ S SRV IBA IBA + E0 IBA+SRV IBA + SRV + EQ DBA DBA + EQ DBA+SRV DBA + EQ + SRV EVENT

& C0 CO PS CO C0 COMBINATIONS R CO ,CH CH C0 CH (1) CH PS CO .CH PS CH PS CO CH V EQ CH 0 S 0 S 0 S 0 S 0 S O S 0 S 0 S TYPE OF EARTHQUAKE O S 4 5 6 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 COMBINATION NUMBER 1 2 3 7 LOADS X X X X X X X X X X X X X X X X X X X X X Normal (2) N X X X X X X X X X X X X X X X X X X Earthquake X X X X X X EQ X X X X X X X X X X SRV Discharge SRV X X X X X X X X X X X X X X X X X X X X X X X X X X X Thrrmal Ta X X X X X X X X X X X X X X X X X X X X X X X X X X X X X Pip 7 Pressure Pa X X X LOCA Pool Pp3 '

X X X X X X Swall LOCA Condensation X X X X X X Oscillation Pen X X X X X X X X X X X X X LOCA Chugging Pru X X X STRUCTURAL ELEMENT RDW Ess ntial Piping Systems B B B B B B B B B B B B B B B B B 10 B B B B B B With IBA/DBA B B B B (4)

(3) (3) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4) (4)

B B B B B B B B B B B B - -

With SBA 11 (3) (3) (4) (4) (4 )' (4) (3) (3) (4) (4) (4) (4)

Non-nssential Piping Systems D D D D D D D D D D D D D D D D D D D D With IBA/DBA 12 B C D D D D D (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5) (5)

(5) (5) (5) (5)

D D D D D D D D D With SBA 13 C C D - - - - - - - - - - - -

(5) (5) (5) (E) (5) (5) (5) (5) (5) (5) (5) (5)

This Table corresponds to Table 5-2 of Reference 6 (PUAAG) 5-7 PBAPSPF2/1-19

The following notes apply to Table 5-1:

(1) Where drywell-to-wetwell-p ressu re di f f erenti al is normally utilized as a load mitigator, an additional evaluation shall be performed without SRV loadings but assuming the loss of the pressure differential.

Service Level D Limits shall apply for all structural elements of the piping system for this evaluation.

The analysis need only be accomplished to the extent that integrity up to and including the first pressure boundary isolation valve is demonstrated, including operability of that valve. If the normal plant operating condition does not employ a drywell-to-wetwell pressure differential, the listed Service Level assignments shall be applicable.

(2) Normal loads (N) consist of dead weight loads (D).

(3) As an alternative, the 1.2S h limit ia Equation 9 of NC-3652.2 may be replaced by 1.8Sh provided that all other limits are satisfied. Fatigue requirements are applicable to all columns with the exception of 16, 18, 19, 22, 24 and 25.

(4) Fr otnote 3 applies, except that instead of using 1.8S h in Equation 9 of NC-3652.2, 2.4Sh may be used.

(5) Equation 10 of NC or NO-3650 shall be satisfied, except that fatigue requirements are not applicable.

to columns 16, 18, 19, 22, 24 and 25, since pool swell loadings occur only once. In addition, if operability of an active component is required to ensure containment integrity, operability of that component must be demonstrated.

~

O 5-8 PBAPSPF2/1-20

Table 5.2 Computer Model Contents for Weight, Thermal Expansion and Seismic Inertia for SRV Discharge Lines in the Drywell Computer Model Contents Number Main Steam Line i SRV Discharge Line 1 A 71A 71B 71C 710

-2 B 71E 71F 71G 3 C 71H 71J 4 0 71K 71L e

h r

5-9 PBAPSPF2/3-1

..__..,.,,_m.,,-,,,_, ._,..r .__..,r_,__,,,._ , .,... .__. _._ ,,7- , _ - - _ . , _ _ _ _ _ . - . , . , , _ _ . . . _ . - . , . _ _ _ . . , . . , , _ _ , , , . . ,

4 TABLE 5.3 Postulated Thermal Expansion Load Cases for SRV Discharge Lines in the Drywell Thermal Expansion Load Cases i 1 2 3 4 5 6 7 8 9 10 Main Steam Line A H H H H l

SRV Line 71A C H H C SRV Line 71B C H C H i

H H H H H Main Steam Line B H H H H H H H H

!' SRV Line 71C C H C C C C C H C C H SRV Line 710 C H C C H H I SRV Line 71E C H H H. H H C C C C SRV Line 71F C H C H H C .C C H C l

Main Steam Line C H H H H H SRV Line 71G C H C C H SRV Line 71H C H H H C -

l SRV Line 71J C H C H H Main Steam Line D H H H H SRV Line 71K C H H C SRV Line 71L C H C H Note: H = Line " Hot" at 500*F (C'ywell), 400*F (Wetwell)

C = Line " Cold" at ambient tempera tu re (70*F) 5-10 PBAPSPF2/3-2

Table 5.4 Bounding Load Combinations for SRV Discharge Lines in the -Drywell l SRV

+

Event Combination N0C EQ SRV EQ f  !

Typt of Earthquake 0 0 S 2 3 Combination Number 1 Loads:

Dead Weight X X X X X X X Earthquake X SRV Discharge X X X Thermal Expansion X X X X X Pipe Pressure X X X X X Service Level A B B C D(2)

Notes:

(1) Extracted from Table 5.1

-(2) Exception f rom Table 5.1 - Loaa Combination #3 is considered as a faulted condition. Sevice Level D limits are met for this load combination.

5-11 PBAPSPF2/3-3

l Table 5.5 Applicability of SRV Discharge Wetwell Piping Models l

Computer Model Number Applicable SRV Lines 1 71A 71F 71G 2 718 710 71H 71J 3 71C 71E 71K -

71L i

I l

I l

5-12 PBAPSPF2/3-4

Table 5.6 Bounding-Load Combinations for SRV Discharge Lines in the Wetwell l

l SRV DBA+E0+SRV

+

i Event Combination NOC EQ SRV E.Q. PS Type of Earthquake 0 0 S l

Combination Number 1 2 25 Loads:

Dead Weight X X X X. X Earthquake X X X SRV Discharge X X X Thermal Expansion X X X X X Pipe Pressure X X X X X LOCA Pool Swell X Service Level A B B C D Notes:

(1) Extracted from Table 5.1 -

5-13 PBAPSPF2/3-5

6.0 TORUS PENETRATIONS

6.1 INTRODUCTION

Vibration of the torus attached pipes impose reaction loads at the junction of the torus and the pipes, i.e. the torus penetrations. In anticipation of stresses exceeding the' allowables, all the nozzles six inches and above in diameter were reinforced with circular and radial stiffeners as shown in Figure 6-4 of the plant unique analysis report (reference 1).

This was done in both of the units, since the torus penetra-tions are very similar. This section describes the evaluation of the torus penetrations. Methods of stress analysis and fatigue analysis are summarized.

6.2 LOAD COMBINATIONS AND STRESS ALLOWABLES 6.2.1 The torus penetrations are part of the metal contain-ment and thus are classified as MC components by the ASME Code. The load combinations and stress allowables applicable to the torus apply to the penetrations also.

6.2.2 Section 5.0 of the plant unique analysis report gi ves the summary of these governing load combinations and the allowable stresses. For fatigue analysis the number of cycles used are given in reference 9.

Conservatively, 800 governing SRV actuations were assumed for fatigue evaluation.

l 6.3 METHOD OF STRESS ANALYSIS 6.3.1 Pipe reaction loads from torus external piping were obtained from the analyses of torus attached piping described in Section 3.0 of this addendum. Reaction loads on the torus penetration from torus internal atachments were obtained from the analysis nf torus internal structures described in Section 8 of Refe-rence 1. The coupling effects between the torus and the attached pipes were neglected, and this results in conservative analysis of the torus attached pipes and thus conservative loads on the torus penetrations.

For the governing load combinations, the seismic and i hydrodynamic loads were combined by the SRSS method.

From these reaction loads, an in-house computer program, ME210, based on the methodology given in the Welding Research Council Bulletin 107 (reference 10) was used to calculate the stresses at four locations, 90' apart, of each given penetration.

6.3.2 Stresses at the penetrations due to dead load and quasistatic pressure on the torus are obtained from the torus analyses described in the plant unique analysis.

6-1 PBAPSPF2/1-21

6.3.3 The total stresses were calculated by adding the stresses due to pipe reaction. loads from torus external piping, stresses due to reaction loads from torus internal attachments, and the stresses due to dead load and quasistatic pressure on the torus. The maximum primary stress intensities and the range of maximum primary plus secondary stress intensities were calculated at all penetrations. These stress intensitjes were compared with the allowables.

6.4 METHOD OF FATIGUE ANALYSIS The f atigue analysis was carried out for the highly stressed penetrations. The dynamic loads considered were seismic loads, safety relief valve discharge loads, condensation oscillation, and chugging. For each of the governing load combinations given in reference 9, the alternating stresses were with the following correction factors:

1. Stress concentration factors, 2.0

11. Correction factor for modulus of elasticity, 1.075 iii. Factor to correct to alternating stresses, 0.50 Allowable number of cycles were obtained for each of the loads f rom Figure I-9.1 of the ASME Code. Usage factors for each load combination were calculated as a ratio between the stress cycles f rom reference 9 and the allowable number of cycles obtained above. Cumulative usage f actor is the sum of all the individual usage factors.

6.5 RESULTS For the governing load combinations, the calculated stresses and cumulative fatigue usage factors stay within their corresponding allowable values.

1 6-2 PBAPSPF2/1-22 7- ,- .- --.-__ --., , . -. , - - _ - -

i 7.0

SUMMARY

OF ANALYSIS t

A plant unique stress analysis was performed for the MSRV discharge  !

lines (in the drywell and the wetwell) and the torus attached piping for the Peach Bottom Atomic Power Station Units 2 and 3. The piping analysis was performed for all applicable LOCA, SRV discharge  !

and normal loads. The supporting system, valves, equipment nozzles and equipment anchorages were evaluated and have been or will be modified as necesary to accommodate these loads. j When all required modifications are completed, the stresses in the  !'

piping systems and pipe supports will meet code allowables, valve accelerations will satisfy allowables, equipment nozzle reactions will meet allowables and operability criteria and equipment anchorages will meet design allowables. ,

Based.on the results.of this plant unique analysis, it is concluded that the SRV and torus attached piping systems and related compo- L nents will meet all code allowables and design criteria (when modif-ications are completed) and, therefore, meet the original intended -

margin of safety.

• ' ' ^ ^ ' * '* , * -

p g ...- - . , . _

t b

7-1 PBAPSPF2/1-23

8.0 REFERENCES

1. Peach Bottom Atomic Power Station Units 2 and 3, Mark I Long

~ Term Program - Plant Unique Analysis Report: Revision 1,

. August 1983.

2. Peach Bottom Atomic Power Station Units 2 and 3, Final Safety Analysis Report.

3. General Electric Company, " Mark I Containment Program-Load Definition Report", Revision 2, NE00-21888, November 1981.

4. .U.S. Nuclear Regulatory Commission, " Safety Evaluation Report, Mark I Containment Long Term Program, NUREG-0661, July 1980.

5. General Electric Co. Mark I Containment Program, ". Plant Unique Load Definition Peach Bottom Atomic Power Station Units 2 and 3", NED0-24577, Rev. 1, January 1981.

6. . PUAAG - Mark I Containment Program, Plant Unique Analysis

~

Application Guide, NED0-24583-1, October 1979 by General Electric. . .

7. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.61,

" Damping Values for Seismic Design of Nuclear Power Plants",

October 1973.

8. American Institute of Steel Construction Design Manual, 6th & 8th Edition.

9. Mark I' Containment Program, " Augmented Class 2/3 Fatigue evaluation method and results for typical torus attached and SRV piping systems" Report No. MPR-751, November 1982.

10. Revision: to " Welding Research Council Bulletin 107", March 1979.

11. " Design Criteria for Evaluation of Torus kttached Piping and Associated Components Due to the Effects of Mark I Loads'"

Bechtel Document l i l 8 7-P 317 Q.

12. Appendix A to Operating License DPR-44 and DPR-56 Technical Specifications and Bases for Peach Bottom Atomic Power Station, .

Units 2 and.3, Section 3.7.A.  !

13. General Electric Flow Diagram: '

RHR : 6280-M-1-DD-9-4 RCIC: 6280-M-1-DD-4-4 ,

Containment Inerting: 6280-M-98-28-1 Core Spray: 6280-M-1-DD-7-5 HPCI: 6280-M-1-DD-6-4 l 1

8-1 PBAPSPF2/1-24

14. Specification for Piping Materials, Instrument Piping Standards and Valve Qualifications for the Peach Bottom Atomic Power Station Units 2 and 3. Bechtel Specification '

6280-M-300, Rev. 11 with Addition No. 1.

15. General Electric Co., " Peach Bottom Atomic Power Station Units 2 and 3, Suppression Pool Temperature Response",

Report NEDC-24380-P, December 1981.

l-I e

j >

. 8-2 ,

PBAPSPF2/1-25 l i

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