ML20214V192

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Evaluation of RCS for Elimination of Primary Component Support Snubbers
ML20214V192
Person / Time
Site: Beaver Valley
Issue date: 05/31/1987
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DUQUESNE LIGHT CO.
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ML19292H353 List:
References
NUDOCS 8706120013
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ATTACHMENT C DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION - UNIT NO. 1 EVALUATION OF REACTOR COOLANT SYSTEM FOR ELIMINATION OF PRIMARY COMPONENT SUPPORT SNUBBERS May 1987 i

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TABLE OF CONTENTS Page I. INTRODUCTION 1 II. BACKGROUND 1 III. ANALYSIS 3 A. Mathematical Models 3 B. Loading Conditions 3 C. Codes and Standards 4 D. Computer Programs 4 IV. RESULTS AND DISCUSSIONS 5 A. Stress in Reactor Coolant Loop Piping 5 B. Fracture Mechanics Evaluation 5 C. Component Support Evaluation 6 V. CONSERVATISMS 6 VI. INDEPENDENT VERIFICATION 7 VII. QUALITY ASSURANCE 8 VIII. ENHANCEMENT OF RELIABILITY 8 IX. CONCLUSIONS 9 X. REFERENCES 9 LIST OF TABLES Table 1: Reactor Coolant Loop Piping Stresses Table 2: Reactor Coolant Loop Natural Frequencies and Modes Table 3: Factors of Safety for Component Supports LIST OF FIGURES Figure 1: Steam Generator and RC Pump Supports Figure 2: Steam Generator and RC Pump Lower Supports Figure 3: Stone & Webster Component / Support Model 4

Figure 4: Stone & Webster Piping Model Figure 5: Westinghouse Model Figure 6: Stone & Webster / Westinghouse Interface 8218A-1628301-B1 1

I. INTRODUCTION This report is submitted in support of Duquesne Light Company's request for a licensing amendment for Beaver Valley Power Station Unit 1. The amendment would permit the elimination of several large bore snubbers from the primary coolant loop component supports (Figures 1 and 2) as a result of excluding the dynamic effects of postulated primary loop ruptures. These changes to the design basis are in compliance with General Design Criteria 4 (GDC-4) and are intended to improve support reliability and reduce occupational exposure to radiation.

Two snubbers at each steam generator upper support will be retained and the other two, at locations of small thermal movement, will be replaced with rigid struts. All snubbers at the coolant loop elevation will be eliminated from the steam generator and reactor coolant pump supports. Thus, a total of six large bore snubbers will be retained, twenty-four will be eliminated and six replaced with rigid struts.

The technical basis for the licensing amendment is the use of

" leak-before-break" technology for excluding from the design basis the dynamic effects of postulated pipe ruptures in primary coolant piping as allowed by the current GDC-4 rule. Westinghouse topical reports (WCAP 11317/11318, Reference 9) documenting the fracture mechanics analyses are submitted as separate attachments.

The purpose of this report is to demonstrate that the reactor coolant piping, components, and supports with the modified support config-uration are able to withstand all remaining loads, including those due to the design basis earthquake, with an acceptable margin of safety. Specifically:

1. The maximum stresses in the primary loop piping are within allowables.
2. The reactor coolant system components and supports continue to have acceptable margins of safety.

II. BACKGROUND The primary loop piping of Pressurized Water Reactors (PWRs) is highly reliable, and for Westinghouse plants (including Beaver Valley Unit 1) there is no history of cracking failure. The Westinghouse Reartor Coolant System (RCS) primary loop has an operating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g., intergranular stress corrosion cracking), water hammer, or fatigue (for both low and high cycle). This operating history totals over 450 reactor years, including five plants each having 16 years of operation and fifteen other plants with over 11 years of operation.

l The application of the leak-before-break approach (References 9 and 12) to prevent ruptures of the primary coolant loop piping 8218-1628301-B1 1

eliminates the requirement to design for the extreme loads associated with these previously postulated pipe rupture events. This provides the opportunity to eliminate selected primary component support snubbers which principally carry pipe rupture loads.

Large bore snubbers, being active components, required periodic removal for functional testing and implementation of a seal service life program. Removal / inspection activities of large bore snubbers have exposed maintenance personnel to high radiation because the snubbers are located in the reactor containment cubicles. The deletion of these snubbers will eliminate this source of occupational exposure and facilitate maintenance and in-service inspections of piping and components by reducing plant congestion.

Support system reliability is also increased with the removal of these active elements. Inadvertent lockup, bleed rate variance, and hydraulic fluid leakage are possible large bore snubber problems that are eliminated.

The function of each primary loop support snubber has been reviewed to determine those which may be eliminated. The object of this review was to maximize snubber elimination allowed by the exclusion of dynamic effects of main reactor coolant loop breaks while having minimal effect on the design margins for other loads. Except for the application of " leak-before-break" for elimination of dynamic effects of main reactor coolant loop breaks, the use of seismic loads based on 2 percent damping, and the combination of the remaining pipe rupture loads with seismic loads using a square root of the sum of the squares (SRSS) combination, other licensing basis requirements are maintained.

The large bore (12 in. I.D.) snubbers on the Beaver Valley steam generator (SG) and reactor coolant pump (RCP) supports which may be eliminated by use of " leak-before-break" criteria are indicated in Figures 1 and 2:

  • Both snubbers (Nos. 5 and 6) acting parallel to the hot leg in each of the three lower SG supports.
  • Both snubbers (Nos. 7 and 8) acting parallel to the cold leg in each of the three RCP supports.
  • Four snubbers (Nos. 9, 10, 11, and 12) acting between the RCP support and the SG lower support in each cubicle.

Also, during this modification, the two snubbers (Nos. 2 and 4) located at each SG upper support and acting in a direction perpen-dicular to the hot leg will be replaced by rigid struts. The existing snubbers at these locations have minimal axial movement during normal operation. It is expected that these snubbers can be replaced by struts even under the current licensing basis. These struts will serve as lateral supports and permit normal steam 8218-1628301-B1 2

generator movement in the direction radial from the reactor during -

thermal movement of the reactor coolant loop.

The snubbers to be retained, as noted in Figures 1 and 2, are two snubbers (Nos. I and 3) located at the SG upper support and acting in ,

a direction parallel to the hot leg.  ; ,

Loading evaluations performed with the revised support configuration establish that the piping, components, and supports are stressed within UFSAR acceptable limits.

III. ANALYSIS A. Mathematical Models Two independent analyses of the primary RCS loop were performed for this work. Stone & Webster Engineering Corporation (SWEC) performed analyses using the model of Figure 3 to obtain com-ponent support loads. SWEC used the model of Figure 4 for analysis of the primary loop piping. An independent verifica-tion of this work was provided by Westinghouse Electric Corporation using the analytic model shown in Figure 5. SWEC had original design responsibility for this work. All analytical models were revisions to existing models and incorporated changes due to the proposed modifications to the supports.

B. Loading Conditions The RCS with the revised support configuration was analyzed for the following loading conditions:

  • Deadweight,
  • Internal pressure, q
  • Thermal expansion, l
  • Seismic events (OBE and DBE), and 1

i These loading conditions cover all UFSAR specified loadings

! other than the dynamic effects of postulated pipe ruptures in

! the main reactor coolant piping.

! No other hydraulic transient loading was considered as j significant.

For seismic analysis, peak spread amplified response spectra (ARS) for 2 percent equipment damping (OBE and DBE) were used.

l This damping is well within that permitted by ASME Code Case i N-411. Responses to the three directions of earthquake loading

{ were evaluated in accordance with the UFSAR by combining the j vertical earthquake absolutely with the higher horizontal l

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'f response for component supports. The piping analysis used three earthquake directions, one vertical and two horizontal, and combined the results by square root sum of the squares (SRSS).

The combination of closely spaced modes conformed to the "10 percent method" of NRC Regulatory Guide 1.92, Rev. 1 (Reference 5).

The SRSS method was used for combining pipe rupture and DBE loads.

C. Codes and Standards The following Codes and Standards were utilized in the analysis:

  • Power piping, USAS B31.1-1967 (Reference 6) . This is the original code of record to which the plant was constructed.

Beaver Valley Power Station Unit 1, Duquesne Light Company.

Allowable stresses currently documented were used for requalification.

/

  • ASME Boiler and Pressure Vessel Code,Section III (see UFSAR Table B.3-5 for specific Code and Addenda), Nuclear Power Plant Components (Reference 7), was used for the 4

design and construction of the equipment.

D. Computer Programs The Westinghouse analysis used the WESTDYN computer code (Reference 2). The WESTDYN computer code has been utilized on numerous Westinghouse plants and was revien d and found acceptable by the NRC in 1974. i Component sulport stiffness

, matrices were supplied by SWEC and are identical to matrices used in earlier analyses, except for stiffnesses representing the eliminated snubbers.

The SWEC analysis used the STARDYNE computer code (Reference 3).

STARDYNE is a public domain computer program and is recognized as a Category I computer program suitable for nuclear work. The following modules of STARDYNE, Version 3, Level H, were used:

  • STAR (Static and Modal Extraction)
  • DYNRE4 (Seismic Response Spectrum)

-* DYNRE6 (Time History Transient Analysis) - only used for evaluating pipe rupture loadings.

This program is monitored with SWEC's Qualit;y Control procedures with respect to any program errors which are encountered through industry usage.

Static analysis of the steam generator support frame used the STRUDL-SW program (Reference 4). This program is also maintained and monitored by SWEC's Quality Control procedures.

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8 IV. RESULTS AND D,ISCUSSIONS A. - Stress fn Reactor Coolant Loop Piping Table 1 provides - the level of stress in the piping and the code allowable stresses ._ (Reference 6). A comparison is also shown- l between the maximum stress in the reactor coolant loop piping t for the existing and proposed support configurations for con-t trolling-_ load' combinations. The results clearly show the stresses'in the piping are well within allowables.

i- B. Fracture Mechanics Evaluation

' Forces and moments developed at several locations in the reactor coolant loop piping for different loading conditions and the proposed support confi'uray,iom g were used in a fracture mechanics evaluation of thoses locaticrx i

i Westinghouse toMcal reports, WCAP-11317/11318. (Reference 9),

' y '. which are included as a separate attachment, justify the

, elimination of; RCS primary' loop pipe breaks for Beaver Valley Unit 1 as fo!.lowi -

1. tStress corrosion cracking. is precluded by use of fracture-

', ' resistant materials in the piping system and controls on

~icactor coolant chemistry, temperature, and flow . during 1 normal operation.

s 2. Water hanner should not occur in the RCS piping because of 5

system, design, testing and operational- considerations.

3 If 3. The effects of low and high cycle fatigue on the integrity 4* of the' primary piping are negligible.

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,, '4. A large margin exists between - the leak rate of the re-ference flaw and the criteria of Reg. Guide 1.45.

5} Ample margin exists between the reference flaw chosen for leak detectability and the " critical" flaw.

6. Ample mM11t. exists in the material properties used to demor -ir te end-ef-service life (relative to aging) sta'il .tl * - 1e reference flaw.

i " The reference flaw will be stable throughout reactor life i becabis 'of the ample margin in items 4, 5, and 6 above and will leak at a detectable rate which will assure a safe plant shut-down.' A detailed review of leakage detection capabilities at Beaver Valley Unit I are included as Attachment D.

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Based on the above, it is concluded that postulation of primary

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loop pipe breaks'need not be considered.in the structural design basis of Be' aver. Valley Unit 1.

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C. Component Support Evaluation Using the analytical model of Figure 3, SWEC has evaluated the proposed support system configuration. As indicated by Table 2, the frequencies of most vibrational modes are . virtually unchanged. The loads on primary components and supports continue to be low.

The postulated terminal end and intermediate breaks in the pressurizer surge, residual heat removal, accumulator, main steam, and feedwater lines were reviewed by SWEC to determine those breaks which would cause the most severe loadings on the revised'- support configuration with snubbers removed. Time history forcing functions were applied to the analytical model.

of Figure 3, representing these potentially limiting breaks, to obtain maximum loads with the revised support configuration.

These loads were combined by SRSS with seismic DBE loads and then summed with deadweight and pressure loads. In all evalu-ated cases, the support loads are within UFSAR and Code allowables.

Table 3 provides Factors of Safety for the existing and proposed support configurations for various load combinations, where:

Allowable Load Factor of-Safety = Calculated Combined Load For the - component interfaces, the Allowable Load is taken from.

the Westinghouse Equipment Specification and, therefore, contains additional conservatism.

The snubber and strut loads at the upper steam generator supports are governed by postulated ruptures of ' the main steam lines. The proposed change to eliminate the lower snubbers has a negligible effect on the loads-on the remaining upper steam generator snubbers and struts. The struts carry small thermal loads because they are oriented perpendicular to the hot leg and the steam generator has minimal thermal movement in this direction.

V. CONSERVATISMS Recently, the ACRS (Reference 11) has expressed a concern about

-modifications to primary component supports stemning from the app 4-ication of the recent revision to GDC-4 (Reference 12). This concern questions the adequacy of ASME Code allowables to provide sufficient margin against indirect pipe rupture from support failures postulated in probabilistic studies. NRC funded research has

, concluded that the probability of indirectly-induced double ended guillotine break (DEGB) in RCS piping due to earthquake is very small for. Westinghouse reactors.

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There 'are several aspects of this submittal for primary component

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support snubber reduction which indicate that this is not a concern for Beaver Valley Unit 1.

The primary support designs predate Subsection NF of the' ASME Code.

Although fabrication and inspection closely paralleled the requirements of the ASME Code, the allowable stresses were limited to-90 percent of the yield strength at the plant faulted temperature level. The more liberal ASME Service Level D allowable stresses of up to 120 percent of yield were not used. In addition, comparison of stresses or loads based on elastic limits are very conservative and are not true indicators of failure loads.

The ACRS is concerned about seismic risk. In these evaluations, a damping value of 2 percent was used for OBE and DBE. Thus, the reported safety factors inherently contain more conservatism than the probabilistic work (Ref. 8) based on damping greater than 2 percent.

Also, pipe ruptures in primary loop branch lines and in the main steam and feedwater lines were considered in conjunction with the

~ seismic event. These postulated ruptures were major contributors to the computed stresses. As shown in Table 3, the factors of safety, i based on the 90 percent yield criteria, are substantial (> 2.1) for

! the load combination consisting of seismic DBE, deadwight, thermal, and pressure. Such large margins should mitigate concerns about seismic risk.

Many of-the snubbers proposed.for elimination are parallel to the hot and cold leg piping. These snubbers function almost exclusively to protect against primary loop ruptures. Their stiffness and strength is small compared to the adjacent . piping. Also, the snubber reduction proposal calls for the substitution of rigid struts of comparable strength and stiffness in place of the snubbers being eliminated at the top of each steam generator. These changes do not have a substantial effect on the existing design margin.

The purpose of this licensing amendment for changes to the primary component supports is limited to the elimination of as many snubbers as feasible. Considering the operating history of. snubbers, and the high factors of safety, the safety benefits of the proposed revisions outweigh the concerns based on the probabilistic risk studies.

VI. INDEPENDENT VERIFICATION As discussed previously, two essentially independent analyses were performed by SWEC and Westinghouse. Both completely modeled a single primary loop. The results of both analyses at support-to-component interface points were reviewed and found to be in close agreement.

RCS natural frequencies / modes and pipe stresses listed in Table 1 and 2 show good agreement in results obtained by Westinghouse and SWEC.

The interfaces between Westinghouse and SWEC for this work have been carefully monitored. Interface details are provided in Figure 5.

8218-1628301-B1 7

The primary equipment, piping, and supports are shop fabricated to strict tolerances. The components and supports were then accurately positioned in accordance with detailed installation specifications and, finally, the loop piping was installed between the components.

The RCS primary piping is supported only by the RCS primary equip-ment. Therefore, the original as-designed geometry adequately reflects the as-built condition.

The computer models prepared by Westinghouse and SWEC were based on the original models. As part of the current reanalysis, these models were reviewed and modified to reflect the proposed snubber elimina-tion and use of rigid struts. Personnel involved in these reviews differed from those involved in generating the original models. The calculations, including modifications to the analytical models, were independently reviewed.

The support configuration is essentially the same as the original configuration, except for the removal of snubbers. The reactor coolant pump and lower steam generator support stiffness matrices, without snubbers, are essentially the same as previously used for the deadweight and thermal cases (i.e., with snubbers inactive). The stiffness matrix for the steam generator upper support reflects substituting two struts for two support snubbers.

VII. QUALITY ASSURANCE Except for elimination of dynamic effects due to pipe rupture of the primary RCS piping, the SRSS combination of other pipe rupture loads with seismic loads, and the use of 2 percent equipment damping in dynamic analyses, the analyses are in conformance with the existing licensing basis (Reference 1), both with respect to design criteria and the control of the engineering process. The work has been independently reviewed as safety-related calculations and meets 10CFR50, Appendix B, Quality Assurance requirements. The results of the analyses are maintained in Project Document Control.

VIII. ENHANCEMENT OF RELIABILITY NUREG/CR-3718, " Reliability Analysis of Stiff versus Flexible Piping

- Status Report" (Reference 10) established that piping designs using snubbers as support devices may not exhibit the intended reliability because the snubbers may fail to perform the desired function.

Inadvertent lockup, bleed rate variance, and hydraulic fluid leakage are a few of the many problems experienced by the nuclear industry with regard to large bore snubbers. It was further demonstrated in NUREG/CR-3718 that certain piping systems with snubbers removed actually exhibit higher reliability than do those of the original design. Certain large bore snubbers proposed for elimination here are parallel to the cold, hot, and crossover legs of the reactor coolant loop piping. Inadvertent lockup of these could induce high thermal stresses during normal plant operation. The elimination of these snubbers, therefore, enhances reliability.

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IX. ' CONCLUSIONS Based on the results of loading evaluations of the reactor coolant system with the proposed support configuration the following conclusions are made:

  • Piping, components, and supports are stressed within UFSAR allowable limits.
  • Adequate safety margins exist .with respect to strength and fatigue, and structural integrity will be maintained during seismic events.

X. REFERENCES

1. Updated Final Safety Analysis Report (UFSAR), Beaver Valley Power Station, Unit 1, Duquesne Light Company.
2. WESTDYN, Westinghouse Electric Corporation.
3. STARDYNE, Version 3, Level H, System Development Corporation, February 1, 1984.
4. STRUDL-SW, Version 3, Level 3, Massachusetts Institute of Technology / Stone & Webster Engineering Corporation.
5. Regulatory Guide 1.92, Rev. 1, Combining Modal Response and Spatial Components in Seismic Response Analysis, U.S. Nuclear Regulatory Commission, February 1976.
6. USAS B31.1, Power Piping, American Society of Mechanical Engineers, 1967.
7. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Plant Components, American Society of Mechanical Engineers.
8. NUREG/CR-2189, Vol. 8, Porbability of Pipe Fracture in the Primary Coolant ' Loop of a PWR Plant, prepared for the U.S. Nuclear Regulatory Commission by Lawrence Livermore Laboratory, September, 1981.
9. WCAP-11317/11318, Technical Bases for Eliminating Large Primary Loop Pipe Rupture As A Structural Design Basis for Beaver Valley Unit 1, Westinghouse Electric Corporation, 1987.
10. Lu, S.C. and Chou, C.K. Reliability Analysis of Stiff vs. t l Flexible Piping, NUREG/CR-3718, Lawrence Livermore National l

Laboratory, Livermore, California, 1984.

11. Ward, D.A., Letter to V. Stello, ACRS Comments on the 1: Interpretation of 10CFR Part 50, General Design Criterion 4,

, " Environmental and Missile Design Bases", December 17, 1986.

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12. 10CFR Part 50, Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures, Federal Register, Vol. 51, No. 70, April 11, 1986, p. 12502.

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TABLE I REACTOR COOLANT LOOP PIPING STRESSES ANSI Existing Proposed Code Factor of Safety Code Configuration Configuration Allowable Existing Con- Proposed Con-Equation Leg-Pt.-Stress (KSI) Leg-Pt.-Stress (KSI) Stress (KSI) figuration figuration 102.3.2(d) HL 60 6.0 HL 10 7.0 (6.5) 14.8 2.5 2.1 (2.3)

XL 205 5.8 XL 205 6.4 (6.0) 15.2 2.6 2.4 (2.5)

CL 300 6.2 CL 300 6.5 (6.4) 15.2 2.5 2.3 (2.4) 102.3.3(a) HL 65 7.5 HL 10 7.4 (7.5) 17.8 2.4 2.4 (2.4)

Normal & XL 200 7.8 XL 200 11.5 (9.9) 18.2 2.3 1.6 (1.8)

Upset CL 300 6.5 CL 365 8.5 (8.6) 18.2 2.8 2.1 (2.1) 102.3.3(a) IIL 65 7.8 HL 10 7.7 (8.1) 26.7 3.4 3.5 (3.3)

Faulted XL 200 8.2 XL 200 13.9 (11.7) 27.3 3.3 2.0 (2.3)

CL 300 6.7 CL 365 9.6 (9.7) 27.3 4.1 2.8 (2.8) 102.3.3(c) HL 60 6.0 HL 10 6.4 (9.6) 25.6 4.3 4.0 (2.6)

XL 200 4.5 XL 230 5.9 (5.6) 25.7 5.7 4.4 (4.5)

CL 300 3.2 CL .300 2.5 (3.2) 25.7 8.0 10.3 (8.0)

NOTES:

1. HL - llot leg
2. XL - Crossover leg
3. CL - Cold leg
4. See Figure 4 for Node Numbers
5. Westinghouse values are shown in parenthesis, though not necessarily taken at the same node point.

8218B-1628301-B1

TABLE 2 RCS NATURAL FREQUENCIES AND MODES Frequency (Hz)

Existing Proposed Configuration Configuration Description of Mode

-b- 4.451 RCP translation and rocking in plane normal (4.512) to Cold Leg.

6.60 6.66 SG rocking in plane normal to Hot Leg; (6.41) RCP rocking in plane normal to Cold Leg 6.78 6.77 SG rocking in plane of Hot Leg; (6.49) RCP motions small 10.10 -b- RCP translation and rocking, pulling lower SG via connecting snubbers.

-b- 13.15 RCP rocking in plane normal to Cold Leg; SG (14.05) lower mass swinging normal to Hot Leg.

14.52 14.23 RCP rocking parallel to Cold Leg (12.22) SG motions small 16.00 15.94 SG bouncing coupled with horizontal motions; (15.51) RCP rocking normal to Cold Leg NOTES:

(a) Frequencies in parentheses are from Westinghouse, others are by Stone & Webster (b) No corresponding mode (c) First 3 modes contribute over 90 percent of the system response 8256-1623801-B1 1

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TABLE 3 FACTORS OF SAFETY FOR COMPONENT SUPPORTS Minimum Factor of Safety Allowable Existing Proposed Component Load (a) Configuration Configuration l Load Case I l-1 2 3 l 2 3 l Steam Generator Shell Bending 400,000 1.2 2.0 13.9 1.8 5.7 Steam Generator Upper Support Resultant Lateral Load 8500 2.8 2.8 15.2 2.8 15.2 Restraints 3.to Hot Leg (2 Snubbers for Existing 2 Struts for Proposed) 1747/1508 1.9 1.9 13.5 1.4 6.3 Anchorage to Building JL to Hot Leg (b) 1746 1.9 1.9 13.5 1.6 7.2 Snubbers // to Hot Leg 1747 1.6 1.6 11.3 1.6 11.0 Anchorage to Building //

to Hot Leg (b) 1746 1.6 1.6 11.3 1.6 11.0 Steam Generator Lower Support Most Critical Member: (c) 1.04 1.06 2.7 1.1 2.5 Anchorage to Building (d) 1.3 1.4 6.2 1.3 2.8 Steam Generator Feet Vertical Force 3075 2.3 2.7 6.7 2.6 6.0 Tangential Force 3545 2.8 9.2 46.6 9.0 19.1 Reactor Coolant Pump Support Most Critical Member (c) 1.02 1.9 3.2 1.4 2.1 i Anchorage to Building (e) 5953 1.2 2.1 3.7 2.1 3.1 Reactor Coolant Pump Feet Vertical Force 3400 1.3 2.9 5.8 2.6 3.7 Tangential Force 2605 1.7 14.3 35.6 9.3 21.1 1

8218C-1628301-B1 1

TABLE 3 (CONT.)

~ LOAD CASES:

1. LARGE LOCA CONDITION: Loads due to pressure, deadweight, thermal, and Square Root of the Sum' of the Squares (SRSS) of Design Basis Earthquake (DBE) and the worst effect of postulated Reactor Coolant Loop Break, Main : Steam line break, Main Feedwater line break, or Reactor Coolant Loop branch line (RHR, Pressurizer Surge, Accumulator) break.
2. REMAINING PIPE BREAK CONDITION: Loads due to pressure, deadweight, thermal and SRSS of DBE and the worst effect of Main Steam line break, Main Feedwater ' line break, or Reactor Coolant Loop branch line (RHR, Pressurizer Surge, Accumulator) break.
3. DBE CONDITION: Loads due to pressure, deadweight, thermal and DBE.

NOTES (a)- All allowable loads are in kips except steam generator shell bending which is in kip-inches.

(b) Allowable and _ applied loads vary from anchorage to anchorage. The allowable load - and factor of safety shown apply to the most critical location.

(c) Allowable' load is based on allowabic principal stress in the member (90 percent of minimum yield stress at temperature).

(d) Allowable and applied loads vary from anchorage to anchorage. The factors of safety shown apply to the most critical anchorage location.

Allowable tension: _ 1890 kips; allowable compression: 3286 kips; allowable shear: 2380 kips (Node 71) or 700 kips (Node 82); see Figure No. 2.

(e) Applied load varies from anchorage to anchorage. The combined allowable load for three anchorages equals 5953 kips, and factors of safety generally are presented for combined embedment loads as the worse case. Each of the individual embedment loads has a higher factor of safety, except for Proposed Configuration, Load Case 3, in which the minimum F.S. is for an individual embedment compressive load (allowable equals'3110 kips).

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