Text

Evaluation of RCS Components & Component Supports for Optimized Reactor Coolant Pump Support Configuration Loadings
	ML20137E719
Person / Time
Site:	Crystal River
Issue date:	01/13/1986
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML20137E698	List: 3F0186-12, Forwards Addl Info Re Request for Exemption from Portion of GDC 4,consisting of Responses to NRC Concerns Raised During 851031 Meeting,Rept on Reactor Coolant Pump Support Configuration Loadings & Benchmarking Info; ML20137E719;
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ATTACHMENT B EVALUATION

. OF REACTOR COOLANT SYSTEM COMPONENTS AND COMPONCNT SUPPORTS FOR OPTIMIZED REACTOR COOLANT PUMP SUPPOFT CONFIGURATION LOADINGS CRYSTAL RIVER 3 GENERATING PLANT Prepared For FLORIDA POWER CORPORATION By BABCOCK & WILCOX COMPANY Lynchburg, Virginia

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o; c Table of Contents Section 'pescription Pagg 1.0 ~ -Introduction 2 2.0 RCS Pump Support Optimization Methodology 6 3.0 Reactor Coolant Loop Analysis Methodology 8 4.0 References 18 5.0 RCS Components Stress Results Summary 19 5.1 Reactor Vessel 23 5.2 Steam Generator 32 5.3 Reactor Coolant Piping 39 5.4 Reactor Coolant Pumps 45 6.0 Leak-Before-Break Evaluation 49 7.0 Evaluation of Attached Piping 51 8.0 Discussion of Branch Piping Guillotine Effects 54 9.0 Discussion of Additional Loadings Considered 57 10.0 RCS Component Support Seismic Support 58 Capacity Factors Appendix A 60 Ficure Description 1 FPC Loop Math Model 2 FPC Loop Math Model 3 FPC Loop Math Model - RC Piping Elevation 4 FPC Seismic Model-Base Mat and Primary Piping Wall Structure

-5 FPC D-Ring 1 and D-Ring-2 6 FPC Isometric of D-Ring-3 B- 1 l

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1.0 INTRODUCTION

Florida Power Corporation (FPC), Babcock & Wilcox (B&W) and the NRC staff have discussed on several occasions.

-the applicability.of advanced fracture mechanics techni-ques in demonstrating that the evaluation of the effe; cts df-

. postulated pipe breaks in the Reactor Coolant System (RCS),

main loop piping is not required. Furthermore, in Reference L(1), the NRC'specifically indicated that it is permissible to utilize advanced fracture mechanics techniques as an alternate piping qualification method in lieu of postulating main loop pipe breaks.- FPC, based on these events, has proposed to utilize those techniques at the Crystal River 3 (CR-3) Generating ' Station to eliminate postulated RCS main loop pipe. breaks and consequently their inherent mechanical and structural load effects. FPC culminated this proposal by submitting a request for partial exemption from General Design Criteria 4 (GDC-4) in Reference 2.

Specifically in Reference (2), FPC requested a partial exemption from those portions of GDC-4 which require protection of structures, systems, and components against certain dynamic (including mechanical and structural loading) effects associated with postulated RCS main loop pipe breaks. This exemption pertains to all postulated breaks specified in the CR-3 Reactor Coolant main loop piping. FPC has not requested exemption from GDC-4 for other postulated breaks. Additionally, the request does not affect the CR-3 Nuclear Generating Station design basis for environmental, containment, equipment qualification or ECCS analysis.

The following information is provided in this document as additional justification for the subject exemption:

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1. In Reference (3), Babcock and Wilcox submitted for NRC staff' review a fracture mechanics analysis to validate the " Leak-Before-Break" (LBB) failure scenario

'for their Nuclear Steam System (NSS) designs. Staff

. review of this submittal is complete. Approval of the report was documented in Reference 6.

Reference (3) demonstrates that for the NSS RCS main loop piping:

a) A substantial ~ sized flaw in the piping would not grow through the wall nor significantly extend in length during the plant design lifetime.

b) If a flaw were to grow through the wall of the pipe, it-still hac a large margin against reaching the critical crack length. This margin is exemplified by the fact that the postulated crack could be cycled many times from (1) an unloaded condition to (2) a 10 gpm leak condition and (3) back to an unloaded condition without growing even .

near the critical crack length.

c) -A very long through wall crack (many times longer than a leak detectable longitudinal or circumfer -

ential crack length) would remain stable under normal operation plus SSE loadings.

This demonstration provides sufficient justification for elimination of large postulated breaks from the design basis for the CR-3 NSS RCS main loop piping.

2. The ACRS in Reference 4 has approved the application of the aforementioned fracture mechanics techniques to the analysis of asymmetric blowdown loads. Reference (4) states "That is, there is no known mechanism in PWR primary piping material for developing a large break B-3

p

,= e without going through an extended period during which

'the crack would leak copiously."

3.- Section 5.0 of this report presents an assessment of the new optimized RC pump support configuration-

-loadings with respect to the RCS components (i.e. Reactor Vessel, Steam Generator, RC Primary Piping and RC Pumps). The assessment presents new calculated ' stresses for the affected RCS components.

Stress values are tabulated showing along with the applicable code allowable stresses the change with respect to existing-component stresses.

14 . Section 6.0 presents an assessment of the new optimized RC pump support configuration loadings with respect to the . existing loadings evaluated in the B&W Owners Group LBB Topical Report, Reference 3.

5. Section _7.0 presents a discussion of the new optimized RC pump support configuration effects on the RC loop attached piping. This section presents an assessment of.the various amplified response spectra at piping attachment points with respect to the design spectra used in the existing analysis of the attached piping.

6. Section 8.0 presents a discussion of loadings on the RC piping and RC pump supports due to postulated

! guillotine breaks of piping attached to the RC main loop piping.

! 7. Section 9.0 presents a discussion of other loadings on 7

the RC main loop piping considered in this evaluation.

8. Section 10.0 presents an assessment of the RCS Compon-t ent's Support SSE seismic margins using loadings from i B-4 b

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the new optimized RC pump support configuration loading analysis. The new support configuration margins are calculated using techniques employed in the Lawrence Livermore National Laboratory (LLNL) "Proba-bility of Pipe Failure Evaluation" of Babcock,& Wilcox

- (B&W) plants, Reference 5. The new and existing margins are listed clong with the minimum margin of any B&W plant' evaluated in their report.

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, . 4 L2.0: RCS PUMP SUPPORT OPTIMIZATION METHODOLOGY The following is a listing of the important aspects'evalu-

-ated; and the results which led to the final: optimized RC ~

! Pump-Support Configuration for Crystal River 3 Generating Station.._The new-support configuration was based on the assumption that, in.the future,-certain effects associated 7 , with postulated _RC' main loop pipe breaks can be ignored in any. safety evaluation of pumps, primary piping, nozzles and major components and component supports (due to the NRC acceptance.of leak-before-break technology).

1.: .A study of.the RC Pump Supports was performed to determine:if existing snubbers could be removed or replaced with rigid. struts and smaller, more reliable snubbers. For this study, a one half primary loop mathematical model was developed for the Reactor Coolant System.

2. The results of two-studies performed for the CR-3 IN: pump supports showed that the existing 32 large

~ bore snubbers (1000-2000 kips)1can be replaced by 4

-rigid struts and 4 smaller snubbers (400 kips).

3. The study mathematical model (developed in 2 above) of the-RC primary loop was refined and all inputs to the analysis model were verified by use of the latest design information available. The model was expanded to perform a complete evaluation of the seismic, i deadweight and-thermal loadings for CR-3.

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4. Deadweight, thermal and OBE seismic loadings were generated for the optimized RC pump support configuration. Loadings for SSE were determined by B6

-.. . w factoring the OBE loadings as discussed in the present FSAR.

5. ' The reactor vessel, steam generator, reactor coolant

- pumps and reactor coolant piping were re-evaluated using the revised RC main loop loadings. ,

6. All calculations performed to justify the revised RC pump support arrangement were verified and reviewed

- and are intended as supplements to the existing RC Component Stress Reports.

7. The new RC main loop piping loads were evaluated with respect to the B&W Owners Group LBB Topical Report enveloping loads and piping safety factors. The new RC piping loadings were shown to be enveloped by the generic LBB topical report safety. factors.

8. The RCS Component's Supports SSE seismic capacity factors previously ct culated by LLNL in Reference 5 yere recalculated using the revised RC loop piping loadings.

In addition to the analysis performed above, responses have been made to specific NRC questions. These responses along with discussions of methods used and documentation of their acceptability have been presented and submitted.

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'8 ay 3.0. REACTOR COOLANT LOOP ANALYSIS METHODOLOGY

-The;RCS analysis mathematical model is comprised of beam type 1 elements, which simulate a required physical property

- of the z structure in the model. The elements representing the vessel'or pipe are located along the centerline of.the

component being modeled. The structural analysis computer code used is STALUM: (B&W proprietary code).

3.1 Modelina of Pioina Modeling of the primary piping is simplified by STALUM's capability of accepting the- outer radius and wall thickness to describe a pipe cross section. Each pipe element is

-located in the model by reference to the joints at each of

'its ends,'which are located'by their Cartesian coordinates.

Elbows are modeled by declaring the element curved and referring to a joint lying at the elbow's center of curva-ture.- Flexibility calculations neglect cladding in the pipe cross sections. The weights of cladding, insulation, and fluid contents are considered.in the lump mass calculations.

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' 3'. 2 Modelina of vessels When possible, the RCS vessels are modeled using hollow cylindrical cross sections requiring as input the outer-radius and wall thickness. However, the reactor vessel is i modeled using general beam cross sections in which all the

required-section properties are input. Symmetry considera-

'tions ~ require that only. half of the reactor ver _ vl proper-ties appear in t;.a mathematical model together with the

' correct symmetry boundary conditions.

The steam generator shell is input as a hollow cylindrical cross section where all section changes are considered. No l stiffening effect resulting from the tubes and intermediate tube support plates'is considered. For the mass calcula-B8 l

+ ,. y ., -.- e-m_y y . --,. ..,_..m,_~ , .. -,_. ,_m_,.y,,__c-mm.--_,-,.m,_,_,,_._,, ,,, , _ - - , , . . , , , . _ . ._ _ - , - . - , - - -

[ .tions of.the steam generator, however, all parts including weight of the contained water _ (considering the secondary side flooded) and insulation, are taken into account.

The upper and-lower heads of the steam generator were modeled by beam elements. A' separate finite element model

!- of the heads was subjected to unit loads. The d:flections

-from these specified loads are calculated and the equivalent beam cross section properties are obtained.

- 3. 3 Modelina of Reactor Coclant Pumos and Motors p .The pump casing, motor stand, and motor are modeled using-beam elements.. The stiffness values for axial and bending forces and bending moments in two perpendicular directions

for each of these simulated beams are calculated. Using these stiffness values, properties such as cross sectional area, bending moments of inertia, torsional moments of inertia, and shear deflection factors are calculated for the-equivalent beams.

Simulated radial elements are used to represent _the correct positioning of the supports on the pump. These simulated elements serve to transmit the forces occurring at the lugs to the centerline pump model.

[ 3.4- Modelina of Reactor VesJe1 SUDDort

! The : reactor vessel. is supported by a cylindrical skirt

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integral to the reactor vessel bottom head. Below this integral support, additional supporting structures have been

i. designed by the customer's architect engineer per B&W requirements to provide horizontal and vertical support for the vessel. Spring rates of the reactor building under the skirt are considered in the mathematical model as external boundary conditions.

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.3 . 5 H2delina of Steam Generator Upper Supports The steam generator is supported horizontally at the upper tubesheet by a support structure located on the outside of the vessel which -is anchored to the interior walls of the building.' Vertical movement of the steam generator is allowed, as well as the radial growth 'ioward the reactor vessel) of the vessel due to thermal expansion. This support structure is completely reflected in the mathema-tical model. The stiffness of the steam generator shell at

= this elevation is represented by elements from the center-line model of the steam generator to the support structure attachments.

3.6 Modelina of Steam Generator Lower Support The' steam generator is. supported horizontally and vertically by a cylindrical skirt integral to the steam generator lower head and bolted to the building floor. In the mathematical model, the lower support is represented by a hollow cylinder of varying properties to account for the openings in the skirt.

3.7 Modelina of Reactor Coolant Pumo Supports The reactor coolant pumps and motors are supported horizon-tally and vertically at the elevation of the casing to motor stand connection. Movement of each pump will be induced by thermal expansion of the lower and upper cold legs. Rigid supports are placed to allow the resultant direction of pump movement. Slow movements resulting from thermal expansion are permitted by the snubbers. The struts and hydraulic l

snubbers restrain the pumps against the sudden dynamic loads that occur during an earthquake.

The support and restraint system as modeled represents the I actual system orientation. Special attention is given to l B- 10

- - - .--. -e-------wr + + = r-y,---y = +v u- - - - - - - - - - - - -w>-*----- e v

y a a the appropriate bar elements that are required to model the actual freedoms of movement. The bar suppcrts are hinged at the pump casing ring and wall anchors, allowing them to i rotate and carry only axial forces.

The weight of these non-integral support are included in the mathematical model by adding their effective weight to the weight of the pump casing and motor stand.

3.8 Modelino of the Reactor Buildino (RB) Interior Walls The steam generators and pumps are supported horizontally by the interior walls to which they are attached by steel bars and hydraulic snubbers. Because of the structural interac-tion between the components, their supports, and the walls, the interior concrete structures must be included in the mathematical model of the RCS. The basemat provides a common input point for termination of the model.

An illustration of the interior concrete structural model can be observed from Figure 4. Superstructures are used at the elevations of all major component supports. Each superstructure models the geometry of the wall plan at the elevation in which it is located. The superstructures consist of stiff beam elements that transfer the static and dynamic loads from the component supports to the beam element centerline model of the interior concrete walls.

The required properties, such as bending and torsional moments of inertia, cross-sectional area, and shear deflec-tion factors, are input for each beam section to represent the stiffness of the structure. These properties, in addition to lumped masses, are furnished to B&W.

The superstructures are given no mass because the mass of the building is distributed in the concentric beam model.

In the case of seismic loading, interactions between the B ll

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inner walls 'and components are transmitted directly by the superstructures. The loading on the subject component depends. on the stiffness of its support and the method of support, both of which are reflected in the mathematical model.

3.9 Modelina Combined RB Interior Concrete and the RCS-

- All of . the modeling described to this point is combined to develop the mathematical model'of the reactor building interior concrete together with the reactor coolant system.

3.10 Justification of Half-System Model In the plan view of.the RCS, the system loops A'and B consist of identical components except that loop A contains the pressurizer and the surge and spray lines. It has been verified by previous analyses that the effects of the pressurizer, surge and spray lines joining into loop A are not significant and the RCS and building considered together are symmetric about the X axis (see Figure 2).

To model the concrete and reactor vessel, which are located in the model on the axis of symmetry, half of the full cross sectional properties of these two structures are input to the'model. More precisely, each element in the building and reactor vessel model is given exactly half of its full cross-sectional properties, namely, section area, bending moments of inertia in two perpendicular directions, and torsional moment of inertia. This approach, along with the appropriate boundary conditions, represents-the influence of the unmodeled loop B. This method has been verified by comparing the results from a half-system model to a model containing both loops.

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3.11 DEADWEIGHT ANALYSIS METHOD The deadweight analysis is performed to determine component and support structure loadings as a result of the deadweight of the components and piping. The STALUM program accepts concentrated loads at joints only, and for this reason, some

. joints in the mathematical model are chosen as mass joints.

The mass joints are distributed primarily to provide adequate information on the system behavior during dynamic excitation; they also serve as positions at which to apply the deadweight leading. The system is loaded by imposing a downward vertical force- (in the Y direction) at each of the mass joints, the value of this force being equal to the proportion of the deadweight of the component or piping lumped at this particular mass joint. All weights bearing on the components are included in the deadweight analysis.

All masses reflect the weights of the component or piping, insulation and enclosed fluid. Approximate weights of the snubbers and link bars are included in the analysis.

The weight of'the RV includes only half of the total mass.

The weight of the concrete wall has no effect on the piping and component loadings and is omitted.

'The boundary conditions for joints lying on the axis of symmetry are the same as those for the thermal analysis.

These boundary conditions apply only to those joints in the building inner wall model to which superstructures are attached since the remaining building joints will tend to move only in the Y direction. Any movement of the basemat under deadweight loads is prevented by representing the building foundation as fixed in all three translational and all three rotational directions.

All snubbers and link bars are modeled as disconnected to permit free movements due to deadweight loadings. Two constant support hangers per pump provide 66,000 lbs support B- 13 2

per hanger, or 1.'2,000 lbs per pump. The hangers are attached to the pump motor stand just below the top flange.

This is simulated in the mathematical model by imposing an upward vertical force (in the Y direction) at the hanger elevation on the pump centerline.

3.12 SEISMIC ANALYSIS . METHODS

_The response spectra method is used in the generation of OBE seismic loads. These are generated at 100% power operating conditions. The response spectra for the Crystal River Unit 3 site was used in the seismic analysis.

Acceleration values for damping values (2.0% and 5.0%) are input ~to the analysis.

The component damping values used are per U.S. Regulatory Guide 1.61. These are 2.0% for the OTSG, RV, pumps, and motors, and 4.0% for the interior concrete wall. The piping damping value is per ASME Nuclear Code Case N-411. This allows the use of 5% critical damping for piping up to 10 hertz, 2% above 20 hertz, and linearly decreasing from 5% to 2% for frequencies between 10 and 20 hertz. This is accomplished by increasing the accelerations for the 5%

damping spectra from 10 hertz to 20 hertz by linear inter-polation. The 5% and 2% spectra are set equal from 20 hertz on up. The assigned percentage of critical damping in conjunction with the response spectra curves are used to determine the acceleration to which the compo.. ant will be subjected.

Each mass joint of the system may have up to six freedoms of motion - translations X, Y, and Z and rotations x, y, and

z. In order to represent each component eith a sufficient ;

number of masses, B&W has conducted sensitivity studies in which the component models contained many masses. These refined models were analyzed, and the results were used B-14

later for comparison purposes on less elaborate models,

. which were obtained by subsequent reductions of mass points. The reduction process was continued until an optimum number of. mass points was reached, which yielded results in agreement with the more refined models.

Consequently, unnecessary or insignificant degrees of freedom are not specified. This is normal practice in all dynamic calculations.

For the seismic analysis, a single flexibility matrix is calculated for both X and Y earthquake directions since the same boundary' conditions apply to each case.- During the calculation of the inertia forces resulting from the. earth-quake, accelerations are applied separately to obtain results for X and Y direction earthquakes. For the Z earth-quake, an independent flexibility matrix is generated.

Inertial forces are applied to obtain results for a Z direction earthquake.

The flexibility matrices are generated and together with the mass matrices, the system frequencies and mode shapes are computed. For each natural frequency of the system, all components will be participating; however, the degree of

! participation is different for various components. For each frequency, the normalized mode shapes (deflection of each l mass point with respect to the maximum deflection) and I

participation factors are calculated. A composite damping value for the whole system is obtained for each system frequency using the specified component critical damping values. The acceleration value from the response spectra curves for a particular frequency is based on a composite

[

damping value obtained for the complete system and the considered mode. The frequencies of the system, together

- with the corresponding acceleration from the response spectra curves and the modal weighted damping, are used to B-15 i

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determine the inertia forces on the system masses. The

-inertia forces for each mass joint are applied as equivalent static loads to the mathematical model on a mode-by-mode

-basis. The recombination of modal results for each direction of excitation incorporates the effects of closely

' spaced modes in accordance with U.S. Regulatory Guide 1.92.

The results of the three simultaneous earthquakes, two horizontal and one vertical, are combined by' the absolute sum metrod-maximum of x+y or y+z earthquakes (FSAR 5.4.5.2);

pump _ loadings were combined by SRSS of the x, y, z direction earthquake loads (Regulatory Guide 1.92) . The OBE results are multiplied by two to obtain'the SSE results. The X earthquake corresponds to east-west, the Y earthquake is vertical, ind the Z earthquake corresponds to north-south (see Figures 1 and 2).

3.13 THERMAL ANALYSIS METHODS The thermal analysis of the RCS and the interior concrete is performed to determine the displacements, forces, and moments existing in the plant at 0% power, 15% power and 100% power steady-state operating conditions. The thermal analysis calculates the thermal displacements and associated forces and moments throughout the system.

All the information that is required to give the thermal loading on the structure is included in the material properties input for each element. Each element in the structure is given a modulus of elasticity, poisson's ratio, coefficient of thermal expansion corresponding to the represented material, and the temperature increase from the cold 70 deg F. condition to the steady-state hot condition.

The program calculates the length of each element from the coordinates of the joints; then the free thermal expansion of each element is calculated. The resultant displacements, B- 16

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resultant forces and moments for each element are calcu-la ted .

The elements representing the snubbers are modeled as structurally disconnected to permit free thermal expansion.

All link bars are active. To give the half-system model the same characteristics as the complete system, boundary conditions for the joints on the axis of symmetry are specified. The foundation or basemat is fixed in all translational and rotational directions.

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-4. 0 REFERENCES

1. Generic Letter 84-04, D.G. Eisenhut to PWR Licensees, Construction Permit Holders and Applicants for Con-struction Permits, dated February 1,1984.

2. . Letter 3F0285-02, G.R. Westafer to H.R. Denton,

" Request for Exemption From a Portion of 10CFR50, Appendix A, General Design Criteria 4" dated February 1, 1985.

3.. Babcock G.Wilcox Owners Group Report," Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping of B&W Designed NSS," B&W Topical

' Report BEW-1847, dated September 1984.

4. ' ACRS Letter, J.J. Ray to W.J. Dircks, " Fracture Mechanics Approach to Pipe Failure," date June 14, 1983.

5. Ravindra, M.K. et. al. , Probability of PiDe Failure in the Reactor Coolant LooD of Babcock and Wilcox PWR Plants. Volume 2: Guillotine Break Indirectly Induced by Earthauakes. Lawrence Livermore National Laboratory, UCRL-53644, NUREG/CR-4290, Vol. 2, (1985).

6. NRC letter, D.M. Crutchfield to L.C. Oakes of the B&W Owners Group, subject " Safety Evaluation of B&W Owners Group Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," dated December 12, 1985.

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'5.0' RCS COMPONENTS STRESS RESULTS

SUMMARY

Sections 5.1 and 5.2 present the applicable ASME Code Stress

' Report stresses for the reactor vessel and steam generator

. nozzles and component supports. The ratio of new loads due to. the reconfigured RC pump supports to the loads in the existing stress report are given.

The analysis / evaluation for each ccaponent nozzle or support is. contained in a separate section. The sequence of analysis / evaluation steps is shown on the following pages and is applicable for each section.

This basic method employed is to determine the resultant loads (i.e., axial force, transverse force, overturning moment and torsional moment) for both the existing RC pump support configuration and the revised RC pump support configuration and calculate the ratio of revised-to-existing loads. Then the maximum of these ratios is multiplied times the existing mechanical load stress in order to represent the mechanical load stress associated with the revised RC pump support configuration.

This method is considered to be very conservative in that the maximum ratio of increasel (if any) for the resultant loads is multiplied times the total mechanical loading stress from the existing Stress Report. This revised mechanical loading stress is added to the existing thermal transient and pressure stresses in order to represent the revised ASME Code Stress Report values.

Notes:

1) No ratio of less than 1.0 was used (i.e. , stresses are apaused to not be reduced for lower loads).

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ANALYSIS / EVALUATION STEPS

1. Review existing ASME Code Stress Report

2. Determine what portion of total stress is due to mechanical loads (Check all appropriate locations on nozzle - not just maximum total stress)

3. Compare existing stress report loads versus loads associated with reconfig-ured RCP support scheme (compare "SRSS" loads in local nozzle co-ordinate systems)

4. Find the maximum factor for any increase of "SRSS" load to existing stress report loads

5. Multiply the mechanical load stress portion of the existing Stress Report times the load factor from 4 above .

and recombine with thermal transient and pressure stresses of existing stress report.

6. Summarize new ASMC Code Stress Report

. values r

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LOAD COMPARISON

1. Tabulate existing Stress Report loads and loads associated with optimized RCP supports configuration (Transform loads to local nozzle coordinate system)

2. Perform load combinations as shown in existing Stress Report

3. Determine the " Square Root of the Sum of the Squares" (SRSS) of the loads

4. Determine the. ratio of the SRSS loading associated with the reconfigured supports to the existing Stress Report loads.

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m Loads corresponding to existing RCL Pump support scheme and associated stresses are taken from the respective stress report.

The Design Code used is per the original Stress Report Certifica-tion

1) Vessels - ASME Code,Section III, 1965 Edition with Addenda through Summer 1967.

2) Piping - USAS B31.7, Nuclear Power Piping Code, 1969 Edition.

Global Co-ordinate System Sign Conventign h +Y (vertical - up)

= /

Z (horizontal - parallel to RV out)

X (horizontal - parallel to RV core flood nozzle centerline)

LDeal Co-ordinate System Sign Convention

+Y (parallel to meridian) b i o h ','

1, , I

I T i Z (tangent to circ.)

/

X (co-linear with nozzle centerline)

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5.1 REACTOR VESSEL STRESS RESULTS

SUMMARY

5.1.1 RV Outlet Nozzle The following is a tabulation of the ratio of the SRSS loads for the new support configuration to the existing stress report loads. A value less than 1.0 indicates that the new loads are lower than the stress report values and similarly, a value greater than 1.0 indicates an increase over the stress report load values.

- - RATIO OF SRSS LOADS ***

FX FR MX MR COMBINATION NO. 1 DW + OBE 0.15 1.20 0.59 0.88 COMBINATION NO. 2 DW + THERMAL 0.51 1.11 0.96 0.90

+ OBE COMBINATION NO. 3 DW + SSE 0.13 0.84 0.60 0.59 COMBINATION NO. 4 DW + TH ERM AL 0.39 1.03 0.85 0.71

+ SSE B- 23

o RV OUTLET NOZZLE Below is a table containing the recalculation of the stresses applicable to the resupported RCL pump configuration. To be conservative, new stresses are not calculated if the new loads are less than those used in.the stress report.

TYPE OF STRESS: PRIMARY STRESS INTENSITY LOCATION: NOZZLE END STRESS NEW CODE REPORT TOTAL ALLOWABLE TOTAL STRESS STRESS LOAD CASE (ksi) (ksi) (ksi) 1 13.8 13.6 19.5

2. 14.3 15.0 29.25 3 14.0 14.0 23.4 4 17.0 17.3 33.1 LOCATION: NOZZLE-TO-SHELL STRESS NEW CODE REPORT TOTAL ALLOWABLE TOTAL STRESS STRESS LOAD CASE (ksi) (ksi) (ksi) 1 15.6 15.9 40 2 17.2 17.5 40 3 16.7 16.7 48 4 18.1 18.2 48 B_24

RV OUTLET NOZZLE (Cont.) {'

TYPE OF STRESS: PRIMARY PLUS SECONDARY STRESS INTENSITY RANGE STRESS NEW REPORT TOTAL ALLOWABLE LOCATION TOTAL STRESS STRESS Nozzle end 35.2 36.6 58-Nozzle-to 29.0 29.1 80 Shell juncture ,

c FATIGUE CONSIDERATIONS:

The increase in the level of the primary plus secondary stress on the cumulative usage factor must be considered. The maximum increase is 1.4 kai.

Stress Report Usage Factor = 0.29 < 1.0 = Code Allowable Revised Usage Factor = 0.38 < 1.0 = Code Allowable B- 25

5.1.2 RV INLET NOZZLE The following is a tabulation of the ratio of the SRSS loads for the new support' configuration to the existing stress report loads. A value less than 1.0 indicates the new loads are lower than the stress report values and similarly, a value greater than 1.0' indicates an increase over the stress report values. The tabulation is the maximum of any RV inlet nozzle. *

- - RATIO OF SRSS LOADS ***

F

_______________________X_________F_R_________M_X_________M_R_

, COMBINATION NO. 1 i

DW + OBE 0.77 0.72 0.51 0.69

\

COMBINATION NO. 2

DW + THERMAL 1.14 0.87 0.59 0.89

+ OBE COMBINATION NO. 3 DW + SSE 0.84 0.73 0.60 0.70 i

\

COMBINATION NO. 4 DW + THERMAL 1.09 0.85 0.62 0.84

+ SSE s

__s___________________ ________________________________

'k ,

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% d RV INLET NOZZLES (Cont.' )

Below is a table containing the recalculation of the stresses ,

applicable to the resupported RC pump configuration. To be conservative, new stresses are not calculated if the new11oads are less than those used in the existing stress report.

TYPE OR STRESS: PRIMARY STRESS INTENSITY LOCATION: NOZZLE END STRESS NEW CODE REPORT- TOTAL ALLOWABLE TOTAL STRESS STRESS LOAD CASE (ksi) (ksi) (ksi) 1 13.0 13.0 19.5 2 17.6 19.0 29.25 3 14.2 14.2 23.4 4 18.8 19.6 35.1 LOCATION: NOZZLE-TO-SHELL JUNCTURE STRESS NEW CODE

REPORT TOTAL ALLOWABLE l TOTAL STRESS STRESS LOAD CASE (ksi) (ksi) (ksi) 1 16.7 16.7 , ,

40

[ 2- 17.9 18.3 40 L 3 17.0 17.0 48 4 18.1 18.3 48 l

y _________________________________________________

l i

B_27 L

RV INLET NOZ ZLE (Cont.)

TYPE OF STRESS: PRIMARY PLUS SECONDARY STRESS INTENSITY RANGE STRESS NEW REPORT TOTAL ALLOWABLE LOCATION ' TOTAL STRESS STRESS Nozzle end 35.2 37.0 58 Nozzle-to 42.6 44.4 80 Shell juncture FATIGUE CONSIDERATIONS:

The increase in the level of the primary plus secondary stress on the cumulative usage factor must be considered. The maximum increase is 1.8 ksi.

Stress Report Usage Factor = 0.36 < 1.0 = Code Allowable Revised Usage Factor = 0.37 < 1.0 = Code Allowable i

B- 28 l-t a

4 %

5.1.3 REACTOR VESSEL SUPPORT SKIRT The following is a tabulation of the ratio of the SRSS loads for the new support configuration to the existing stress report loads. A value less than 1.0 indicates the new loads are lower than the stress report values and similarly, a value greater than 1.0 indicates an increase over the stress report values.

RATIO OF LOAD COMBINATIONS:

___________________ L__________ E____________ E_______ E.

COMBINATION NO. 1 DW + THERMAL 0.89 0.37 15.47(1) 0.35

+ OBE COMBINATION NO. 2 DW + THERMAL 0.94 0.33 21.99(1) 0.34

+ SSE e

Note: 1)

This load is a torsional moment. The effect of this load increase is to generate additional shear stresses in the skirt flange shear pins. The. existing stress report value for this stress is 0.09 ksi. This stress report value was modified by the above factors and added to the shear stresses due to other shear forces.

B- 29

REACTOR VESSEL SUPPORT SKIRT (Cont.)

TABLE OF STRESSES:

LOAD CASE 1: DW + THERMAL + OBE A) SKIRT MAXIMUM TENSILE STRESS:

Stress Report Value = 1.8 ksi Revised Value = 1.8 ksi Allowable Stress = 34.8 ksi B) SKIRT MAXIMUM COMPRESSIVE STRESS:

Stress Report Value = 5.62 ksi Revised Value = 5.62 ksi Allowable Stress = 34.8 ksi e

C) MAXIMUM SHEAR PIN SHEAR STRESS:

Stress Report Value = 7.65 ksi Revised Value = 9.0 ksi Allowable Stress = 52 ksi D) MAXIMUM BENDING IN FLANGE:

Stress Report Value = 2.4 ksi Revised Value = 2.4 ksi Allowable Stress = 34.8 ksi E) MAXIMUM SHEAR STRESS IN FLANGE:

Stress Report Value = 1.87 ksi Revised Value = 1.87 ksi Allowable Stress = 13.92 ksi l B- 3 0 l

REAQIOR VESSEL SUPPORT SKIRT (Cont.)

TABLE OF STRESSES:

CAS E 2 : DW + THERMAL + SSE A) SKIRT MAXIMUM TENSILE STRESS:

Stress Report Value = 5.28 ksi Revised Value = 5.28 ksi Allowable Stress = 41.6 ksi B) SKIRT MAXIMUM COMPRESSIVE STRESS:

Stress Report Value = 9.11 ksi Revised Value = 9.11 ksi Allowable Stress = 41.6 ksi C) MAXIMUM SHEAR PIN SHEAR STRESS:

Stress. Report Value = 11.50 kai Revised Value = 13.5 ksi Allowable Stress = 52 ksi D) MAXIMUM BENDING IN FLANGs

~ Stress Report Value = 14.1 ksi Revised Value = 14.1 ksi Allowable Stress 41.6 ksi E) MAXIMUM SHEAR STRESS IN FLANGE:

Stress Report Value = 3.04 ksi Revised Value = 3.04 ksi Allowable Value = 16.7 ksi B 31 i

5.2 STEAM GENERATOR STRESS RESULTS

SUMMARY

5.2.1 OTSG OUTLET NOZ ZLES

The'following is a tabulation of the ratio of the SRSS loads for the new support configuration to the existing stress report loads. A value less than 1.0 indicates the new loads are. lower than the stress report values and similarly,

- a value greater than 1.0 indicates an increase over the stress report values. The tabulation is the maximum of

.either outlet nozzle.

- - RATIO OF SRSS LOADS ***

EF x FR MX MR r . COMBINATION NO. 1 7

DW + OBE 0.64 0.56 0.19 0.53 COMBINATION NO. 2 DW + THERMAL 1.08 0.77 0.41 0.77

+ SSE i

i r-i l

I B- 32 I

t.

~

OTSG OUTLET N05ELES (Cont.)

Belowlisc a: tabulation of the' recalculated stresses applicable to the'.resupported RC pump configuration. To be conservative, the existing _ stress report values were not reduced if the new loads are lower.

TYPE OF STRESS - Primary Stress Intensity Stress New Report Total Allowable Stress Value Stress Stress Location (ksi)

(ksi) (ksi)

Nozzle-to-Head 18.5 18.5 40.0

' Juncture Nozzle-End Inside 24.5 24.5 26.0-Outside. 21.3 21.3 26.0 TYPE OF. STRESS - Primary plus Secondary Stress Intensity Range Nozzle to Head Juncture 3 Inside 18.0 18.4 -80.0 Outside 24.0 25.0 80.0

' Nozzle End.

Inside 29.0 30.6 52.0 Outside 29.0 30.8 52.0 4

FATIGUE CONSIDNRATIONS:

. The' maximum increase in Primary .plus Secondary stress intensity range is 1.8-ksi (outside - 0 nozzle end) .

!The cumulative usage factor is:

Stress Report Value = 0.03 << l.0 = Code Allowable

- Revised Value = 0.03 << l.0 = Code Allowable 1

B-33 4

.m.'

. . , ., ,_. ..~ ,e - ,,.,-_,.,,_.,,...,..,,,,,,v..s. - . - . . , . . ,

5.2.2 OTSG INLET NOZZLE

.The following is a tabulation of the ratio of the SRSS loads for the new support configuration to the 9xisting stress report .

loads. A value less than 1.0 indicates.the new loads are lower than the stress report values and similarly, a value greater than 1.0 indicates an increase over the stress report values.

- - - - RATIO *****

F X F R M X M

_-_______--___-_-________---________-_-______----___-_____R_

Combination No. 1 DW + OBE- 0.21 1.08 1.27 0.19 Combination No. 2 DW + THERMAL + SSb 0.55 0.88 1.53 0.51 E

e B- 34 i

OTSG INLET NOZZLE (Cont.)

Below is a tabulation of the recalculated stresses applicable to the resupported RC pump configuration. To be conservative, the previous stress report values are not reduced if the new loads are lower.

TYPE OF STRESS - Primary Stress Intensity Stress Report Total Allowable Stress Value Stress Stress Location (ksi) (ksi) (ksi)

Nozzle-to-Head 15.5 16.4 40.0 Juncture

. Nozzle End Inside 20.3 20.3 26.0 Nozzle End Outside 13.3 14.4 26.0 B- 35 l

=-

OTSG INLET NOZZLE (Cont.)

TYPE OF STRESS - Primary plus Secondary Stress Intensity Range Stress Report Total Allowable Stress Value Stress Stress Location (ksi) (ksi) (ksi)

Nozzle-to-Head Juncture

-Inside 19.5 21.6 80.0 Outside 29.6 24.3 80.0 Nozzle End Inside 14.3 18.5 52.0 Outside 17.9 23.1 52.0 FATIGUE CONSIDERATIONS:

The maximum increase in primary plus secondary stress intensity range was considered.

- The cumulative usage factor is:

Revised.Value = 0.024 << 1.0 = Code Allowable 4

i l B- 3 6 1

5.2.3- OTSG SUPPORT SKIRT The'fo11owing is a tabulation of the ratio of the SRSS loads .

for the new support configuration to the existing stress !

report loads. A value less than 1.0 indicates that the new l loads are lower than the stress report values and similarly, a.value greater than 1.0 indicates an increase over the stress report load values.

RATIO OF LOAD COMBINATIONS l

1 1

Fy F M

______________________________R_________My_________________R___

Combination No. 1 DW -E THERMAL' + OBE 1.15 0.58 15.37(1) 2.08 Combination No. 2 DW + THERMAL + SSE 1.14 0.62 23.06(1) 1.29 t

Note: 1)

This load is a torsional moment. The greatest impact of this moment will be on the flange anchor bolts in the form of a shear stress. The new bolt shear stress due to this moment is 0.7 ksi.

B-37

i, . .

OTSG SUPPORT SKIRT . (Cont.)

TABLE OF STRESSES CASE 1: DW:+ THERMAL + OBE X A) SKIRT MAX. STRESS-INTENSITY:

j ['

Stress Report Value = 6.6 ksi Revised Value = 13.7 ksi

$6 ~

Allowable Stress = 40.1 ksi B) SUPPORT FLANGE:

The revised bearing stress = 1.27 kai CASE 2: DW + THERMAL + SSE A) SKIRT MAX. STRESS INTENSITY:

Stress Report Value = 8.4 ksi Revised Value = 10.8 ksi Allowable Stress = 48.1 ksi B) SUPPORT FLANGE:

The revised bearing stress = 1.41 ksi B-38

5.3

SUMMARY

OF RCS PRIMARY PIPING STRESSES The results of the RCS primary piping stress analysis are summarized in Tables 5.3.1 and 5.3.2. The results tabulated do not reflect stress and usage factors for any attached piping connections. Stress results for the affected nozzles are tabulated separately. A discussion of the tabulated results is presented following each table. The stress equations evaluated and the allowable stress were teken from the original design piping code of construction (USAS B31.7, Nuclear Power Pipino).

TABLE 5.3.1 COLD LEG PIPING STRESS

SUMMARY

Number Piping of Joints Maximuml Code Stress Exceeding Stress Joint 3 Condition Ecuation Allowable Allowable Ratio Number Seismic Combination (X + Y)

Design ( 9) 1.5 Sm 0 0.99 19 Normal / Upset (10) 3.0 Sm 42 1.26 168 Usage Factor (11) U < 1.0 0 0.05 168 Emergency ( 9) 2.25 Sm 0 0.69 19 Seismic Combination (Y + Z)

Design ( 9) 1.5 Sm 0 0.99 20 Normal / Upset (10) 3.0 Sm 42 1.26 168 Usage Factor (11) U < 1.0 0 0.05 168 Emergency ( 9) 2.25 Sm 0 20 Notes: 1) Maximum Stress Ratio equals calculated stress / allowable stress.

2) Actual stresses were justified by performing a code Simplified Elastic-Plastic Discontinuity Analysis per I-705.4.

3) For joint numbering see math model figures 1 thru 3.

B- 39

. -o 5.3 32HMAhY OF RCS PRIMARY PIPING STRESSES (cont)

Cold teo Primarv Pipina The following is a discussion of the results of the stress analyses for the cold leg as_ summarized in Table 5.3.1. The maximum stress at any location for each loading condition is shown. For-joint numbering see figures 1 thru 3.

Desian Conditions As indicated in the summary, no locations exceed the Equation (9) allowable stress for the Design Conditions. The maximum stress

. ratio is 0.99 ani occurs at joints 19 and 20. Thus, the combina-tion of design pressure, deadweight and OBE seismic is acceptable for the cold leg piping.

Normal and UDset Conditions (Level A & B)

For Equation (10), the maximum stress ratio for any location is 1.26 and occurs at joint 168. Although several locations exceed the allowable stresses, all locations were shown to be acceptable when a . simplified elastic-plastic analysis per I-705.4

-of reference 34 was performed for each fatigue stress condition.

The maximum usage factor is 0.05 and occurs at joint 168.

Therefore the combination of operating pressure, OBE seismic and operating thermal expansion is acceptable in the cold leg piping.

'Emercency Conditions (Level C)

No ' joint locations exceed the Equation (9) allowable stress for the Emergency Conditions. The maximum stress ratio of 0.69

-occurs at joints 19 and 20, Thus, the combination of operating pressure, deadweight and SSE seismic is acceptable for the cold leg piping.

4 B 40

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

g - , -

~

5.3

SUMMARY

OF RCS PRIMARY PIPING STRESSES (CONT)

TABLE 5. 3.2 HOT' LEG PIPING STRESS

SUMMARY

Number Piping of Joints Maximum 1 Code Stress Exceeding Stress Joint condition Ecuation Allowable Allowable Ratio Number Seismic Combination (X + Y)

Design ( 9) 1.5 Sm 0 0.64 179 Normal / Upset (10) 3.0 Sm 0 0.87 30 Usage Factor (11) U < 1.0 0 0.02 30 Emergency- ( 9) 2.25 Sm 0 0.44 32 Seismic Combination (Y + Z)

Design ( 9) 1.5 Sm 0 0.64 32 Normal / Upset (10) 3.0 Sm 0 0.87 30 Usage Factor (11) U < l.0 0 0.02 30 Emergency ( 9) 2.25 Sm 0 0.44 32

-Notes: 1) Maximum Stress Ratio equals calculated stress / allowable stress.

9 B- 41 e

m . , , . - e- ,,.m,- -+ -,,-r-,,, --,www- -

r,ew,, w-, --

,r,- - - - - . . - - - - - - - , - - - .

5.3

SUMMARY

OF RCS PRIMARY PIPING STRESSES (CONT)

Hot Leg Primary Piping The following is a discussion of the results of the stress analyses for the hot leg as summarized in Table 5.3.2. The maximum stress at any location for each loading condition is shown.

Desian Conditions As' indicated in the summary, no locations exceed the Equation (9) allowable stress for the Design Conditions. The maximum stress Jratio is 0.64 and occurs at jcints 32 and 179. Thus, the combination of design pressure, deadweight and OBE seismic is acceptable for the hot leg piping.

Normal and UDset Conditions (Level A & B)

For Equation (10), the maximum stress ratio for any location is 0.87 and occurs at joint 30. The maximum usage factor is 0.02 and occurs at joint 30. Therefore the combination of operating. pressure, OBE seismic and operating thermal expansion is acceptable in the hot leg piping.

Emeroency Conditions (Level C)

No joint' locations exceed the Equation (9) allowable stress for the Emergency Conditions. The maximum stress ratio of 0.44 occurs at joint 32. Thus, the combination of operating pressure, deadweight and SSE seismic is acceptable for the hot' leg piping.

B-42

v 5.3

SUMMARY

OF PRIMARY PIPING STRESSES (Cont.)

, c

/

HIGH PRESSURE INJECTION NOZZLE fBPI)

The following is a tabulation of the ratio of the SRSS moment loads for the new support configuration to the existing stress report loads. A value less than 1.0 indicates the new loads are lower than the stress report values and similarly, a value greater than 1.0 indicates an increase over the stress report values. The tabulation only lists the maximum of the four HPI nozzles.

LOAD COMBINATION Ratio DW + OBE 0.85 THERMAL + OBE 0.70 DW + SSE 1.05 The only load combination increase from the stress report is

IMf + SS E. Primary stress results for this condition are:

Stress Report Value = 20.0 ksi Revised Value = 20.1 ksi Code Allowable Value = 31.3 ksi B- 43

5.3

SUMMARY

OF -PRIMARY PIPING 9 TRESSES (Cont.)

RC PRESSURIZER SPRAY LINE NOZZLE The sprayline nozzle was reanalyzed for the new RC loop piping loadings. The following is a summary of the stress results.

TYPE-OF STRESS - Primary Stress Intensity Stress Value Allowable Stress Location (ksi) (ksi)

Nozzle end 6.6 24.75 Nozzle-to-Cold Leg 18.0 24.75

I TYPE.0F STRESS - Primary plus Secondary Stress Intensity Range Stress Value Allowable Stress

, Location (ksi) (ksi)

Nozzle end 34.5 49.5 (1)

Nozzle-to-Cold Leg .66.41 49.5 Note 1) Actual stress was justified by performing a code Simplified Elastic-Plastic Discontinuity Analysis per I-705.4 of B31.7. !

l FATIGUE CONSIDERATIONS:

LThe revised fatigue usage factor at the nozzle-to-cold leg

-pipe juncture is:

Usage factor = 0.53 < 1.0 = Code Allowable B- 44

t 1

5.4 RC PUMP EVALUATION A three dimensional finite element structural model of the reactor coolant pump was used in the determination of stresses in the casing as a result of the new mechanical

. loadings. Unit loads to simulate the mechanical and pressure loads on the pump casing, are applied separately to the three dimensional pump model. The' stresses produced by these loads are used to formulate loading conditions on the pump case.

The load cases created were:

1) Design Pressure (2,500 psi)

2) Bolt Preload

3) Deadweight

4) Largest Thermal Nozzle Loads (15% Power)

5) OBE loads for X, Y, Z Earthquakes A review of the existing Stress Report loads versus new loads revealed the load components for certain joint locations increased as a result of the new loadings and a revision of the stress analysis of the pump casing due to the revised loadings.was performed. The two loading condition categories that required analysis are:

1) Design Condition Category

2) Normal and Upset Condition Category By nature, the thermal expansion and deadweight are statical-ly balanced load sets. The seismic modal response loads were evaluated. The response for each direction of earthquake which includes the effect of closely spaced modes was obtain ed . The three directions of earthquake response are combined using the SRSS (Square root sum-squared) method.

Inertia forces at mass joints within the reactor coolant pump boundary are also used for the seismic modal loads to represent a statically balanced condition of loading on the j pump casing.

I B- 45

i-REACTOR COOLANT PUMP (Cont.)

- The general procedure was to locate the highly stressed areas of the structure using the finite element stress summaries.

The loadings causing these stresses were determined and the

- structural behavior discussed. It should be noted that

- the pressure stresses throughout the casing were conser-

- vatively evaluated as primary.

^

Desian Catecorv Primary stress limits are: 1

,. Pm < l.0 S,= 18.7 ksi PL + Pb < l.5 S m = 28.05 ksi

.I In. general,'the case exhibits very moderate centroidal and surface stresses due to pressure and mechanical loadings.

The only highly' stressed areas for the Design Category are the vanes:(especially the tips). The results of the current analysis compare very well with those of the previous analysis. This compatibility of results is expected, since the stress due to design pressure is the primary contributor

to the total stress. The maximum primary stress !

intensity results are:

Stress Report Value = 16.7 ksi Revised Value = 17.6 ksi Allowable Stress = 18.7 ksi The primary membrane plus bending stress intensities are:

J Stress Report Value = 18.8-ksi Revised Value = 19.8 ksi Allowable Stress = 28.05 ksi B 46

7 REACTOR COOLANT PUMP (Cont.)

EgI, gal /UDset Catecorv Stresses-due to Normal and Upset Conditions must satisfy the secondary, peak and the special stress limits of the code.

The thermal transients applicable to the pump case are the same as those analyzed in the existing-Stress Report. The stresses in the Design Category (design pressure, deadweight and piping thermal expansion)-compared very well to the existing Stress Report values. Therefore,'the only change in the loadings for

.the Normal / Upset Condition category from the existing Stress Report analysis is the seismic loads.

The results from the existing stress report indicate that the critical stress areas are the upper flange solid elements. The stress on the corresponding finite element shell elements due to

. the revised OBE loads is 63 psi. It is obvious that the OBE loads do not contribute to the total primary plus secondary stress intensity range as determined in the stress report.

Therefore, the secondary and peak stress as determined in the stress report remain valid.

The stress report cumulative usage factor of 0.32 remains valid.

B- 47 4

< - . .n.. , , , ,

5.4.1 RC Pump Supports The following table lists the optimized RC Pump Support loadings and the design load capacity of each support.

RC Pump Snubbers Snubber Location SSE Load Design Load (kips) (kips) 2A 105.2 400 4C 80.4 400 8C 56.4 400 3D 105.2 400 RC PumD Struts SSE THERMAL DESIGN LOAD ' LOADING LOAD Strut Location (kips) (kips) (kips) 7A 62.8 85.0 1200 SB 56.4 27.8 2000 7B 80.4 51.3 1200 SD 62.8 85.0 1600 I

B- 4 8 I

l i

n l,' a,-

6. 0. LEAK-BEFORE-BREAK (LBB) EVALUATION

.The LBB evaluation :for B&W nuclear plants was performed using-piping" loads =at various locations on the hot and cold

legs of the'RCS main loop piping. These ' loadings were obtained from existing stress reports for each of the B&W

' Owners Group plants. From these loads, a single set of generic; loads was-evaluated. Table 4-1 and 4-6 of Reference 3 list the~ controlling generic load sets. Other load sets were evaluated, but these are limiting. The following page is a listing of the _ Crystal River 3 piping loadings

.resulting from the configuration of the RC pump supports-versus'the controlling generic load set evaluated in the LBB

' report. The Crystal River 3 ' loadings are considered to be r

L enveloped if the new Crystal River 3 maximum loads are less and the new Crystal. River 3 minimum loads are greater than

'those-loadings evaluated in the LBB report.

The maximum moments resulting from the reconfiguration of the RC' pump supports shown on the following page are within the' envelope of moments that have been justified relative to the Leak-Before-Break Topical Report. However, as was previously done_for the topical report evaluations, CR-3

-load' pairs (pairs with lower maximum and minimum moments) were also analyzed. They were analyzed using formulas

-developed for the topical evaluation and a safety factor for each load pair was calculated. The safety factors associated with these new CR-3 loads are greater than the limiting factors reported in the LBB Topical Report.

Therefore, the LBB Topical Report envelopes CR-3 and as such CR-3 is not a limiting plant with respect to LLB.

B-49

7.

_,. -3:

~

EAK-BEFORE-BREAK EVALUATION (Cont.)

CRYSTAL RIVER 3 VERSUS GENERIC EVALUATION PIPE SIZE. GENERIC MAXIMUM MOMENT CR-3 MAXIMUM MOMENT (FT-KIPS) (PT-KIPS) 28" I.D.

STRAIGHT 3098.0 1772 ELBOW. (1) 2822.8 1330 36"-I.D.

STRAIGHT 2376.5 2010 ELBOW 2376.5 2010

PIPE SIZE GENERIC MINIMUM MOMENT CR-3 MINIMUM MOMENT (FT-KIPS) (PT-KIPS)

'28" I.D.

STRAIGHT 560 1595 ELBOW (1) 1246 1157 36" I . D. .

STRAIGHT. 1010 1144 ELBOW 1010 1144 NOTE:

1) Load pair shown was the limiting case evaluated in the LBB

. report.. In addition, an alternate load pair maximum and minimum moment of 2389 and 871 respectively were evaluated in the supporting calculations for the LBB report. This load pair envelops the new CR-3 loads.

B-50

7.0 EVALUATION OF ATTACH ED PIPING This section contains a summary of the Amplified Response Spectra (ARS) for attachments to the RCS cold legs and pumps.. ARS is generated for these locations to be compared to the CR-3 interior concrete wall spectra. The procedure for computing ARS consisted of the following steps

1. A basemat acceleration time history was calculated by B&W open shop computer program TMH. This program uses

.the basemat spectra for CR-3 and the natural frequencies of the RCS math model up to 30 cps. An iterative technique is used which tries to generate a time history _which will produce a spectra identical to the input basemat spectra.

2. A basemat spectra is calculated for the above time history by the B&W computer code RESPECT. The time history is acceptable if the generated spectra cor-responds to the input spectra for the model frequencies.

3. ARS are calculated by RESPECT for the specific degrees of freedom where attachments are located using the basemat acceleration time history. The ARS peak accelerations are compared with the design basis spectra peaks to assure the existing piping analysis is conservative.

Table 7.1 lists the locations where ARS was generated and the spectra peaks. The spectra peaks are tabulated for 2 percent critical damping since ASME Code Case N-411 permits a minimum of 2 percent damping. The corresponding wall elevations for the building spectra are also shown.

B- 51

r The'CR-3 floor response spectra peaks are tabulated in Table 7.2. These spectra were used as the design basis for the attached piping _ seismic analyses. These peaks'are 1/2 percent critical damping spectra. The peak accelerations from Table 7.1 are clearly smaller than the peak accelerations of the existing design basis from Table 7.2. _This' comparison of peak accelerations at 1/2 percent versus'2 percent damping is valid because any future analysis of attached piping would use 2 percent damping.

Therefore, the existing seismic analysis of the attached piping lines remain valid.

4 b E B- 52 i_

r EVALUATION OF ATTACHED PIPING (CONT)

TABLE 7.1: COLD LEG ATTACHMENT OBE SPECTRA PEAKS ACCELERATIONS (G's) CORRESPONDING JOINT DESCRIPTION X DIR Y DIR Z DIR WALL ELEV(FT)

RCP-1A .

s

~

JOINT 55 CASING 0.604 ------

s0.228 '

135.0 JOINT 131' RESTRAINT RING 0.482 ------

0.217 135,0 JOINT 71 TCP OF CASING -----

0.128 ----- 135.0 JOINT 124 MOTOR 0.987 0.127 0.629 160.0 y.

RCP-1B JOINT 154 CASING 0.237 -----

0.219 135.0 g JOINT 138 RESTRAINT RING 0.215 -----

O.214 135.0 JOINT 139 TOP OF CASING -----

O.129 -----

135.0 JOINT 86 MOTOR 0.650 0.128 0.397 160.0 s.

ATTACHED PIPING -

JOINT 168 SPRAYLINE IB 0.223 0.130 0.228 135.0 JOINTS 56,64 Al HPI 0.379 0.128 0.220 135.0 JOINTS 181,188 B1 HPI 0.237-0.129 0.240 135.0 JOINT 8 Al DRAIN 0.195 0.126 0.194 >

118.5 JOINT 44 B1 DRAIN 0.197 0.127 0.192 118.5

'T

- s TABLE 7.2 FLOOR RESPONSE SPECTRA PEAKS IpTERIOR CONCRETE ,

ELEVATION ACCsLERATIONS i $

(FT) ,

'(G'S) 'k 118.5 'A 0.590 s ,,e t s

123'.,0 0.690 ,

13$).0 0.9 0 160.0 1.640 i

s

\.

B- 53 l

l

. g; j

8.0 DISCUSSION OF BRANCH PIPING GUILLOTINE EFFECTS U The analysis of postulated breaks in the large RCS main loop 3,

?

-piping'is eliminated with the adoption of LBB. Guillotine p ~' breaks at the nozzle to pipe weld of branch lines attached to the primary side of the NSSS must be reconsidered.

Previously the large RC loop piping breaks were enveloping I load cases for these branch line breaks. This discussion is limited to the effects on RC pump support loadings.

Guillotine breaks at the connection point of the following

. lines are considered.herein:

1) 21/2" high pressure injection

2) 2-1/2" pressurizer spray a

, 3) 2-1/2" drain connection All vent and instrumentation lines are 1" or less and therefore do not require analysis. The evaluation of these postulated branch line guillotine breaks must address the potential associated effects of pipe whip, jet impingement, pressurization of compartments, environmental conditions,

, and flooding. Circumferential breaks which result in full separation of the two severed pipe ends with unlimited

_ branch line motion are considered.

The subject breaks cause negligible disturbance to the steady state operation flow conditions in the primary piping; therefore the. associated loading is unchanged from normal operating conditions. The magnitude of a conserv-ative fluid thrust force (an instantaneous step change of thrust from 0 to 15 kips based on [2 x PA] applied at the break point) is insignificant compared to the other design loads. An important fect is that the RCS piping remains intact and thereby the stiffness of the piping assemblies does not change nor will it create a free moment arm to g

transfer force to the pump and its supports. Jet impinge-B-54

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

m

[

ment effects due to these breaks can be evaluated by engineering judgement. Only the guillotine at the spray line attachment could potentially result in a jet force on the pump or its supports. Considering the jet plume's small conical shape and vertical direction with respect to the limited amount of nonvertical potential targets, the jet impingement loads on the pump assemoly would be minor.

Shielding and shape factors of targets would also lower jet impingement loads. Compartment pressurization, environ-mental effects, and flooding due to these small breaks have negligible influence on the pump support loadings.

NOTES 1.- The HELBA for the branch'line side of the break to assess potential damage to cafety related equipment is not changed by adopting LBB and/or modifying the RC pump support arrangement; therefore the HELBA remains valid.

2. Secondary side breaks are not considered.

3. Breaks in small piping attached to major NSSS components (RV, OTSG, RC Pump, Pressurizer) result in comparably less loads due to the mass (inertia) and nearby rigid supports of the components.

8.1 DECAY HEAT, SURGE, & CORE FLOOD LINE GUILLOTINE EFFECTS Results of previous LOCA analyses of the RC system indicate that the pump support loads are not significantly influenced by guillotine breaks of the decay heat line at the hot leg nozzle, the surge line at the hot leg nozzle, or the core flood line at the RV nozzle. Minimal load passes beyond the RV support to the pump and pump supperts.

8.2 PUMP SEAL FAILURE LOADINGS The estimated flow through an annulus around the pump shaft caused by failure of the entire seal cartridge is B-55

~approximately 250 gallons per minute which is within the-capacity of the makeup pump. The resulting change in the total, pressure' thrust loading on the pump and its supports is insignificant.

e w.

1 P'

l' 4

t-B-56

.___t____

9.0 DISCUSSION OF ADDITIONAL LOADINGS CONSIDERED In addition to deadweight and thermal loadings, the NSS will experience loadings fres another normal operating condition

steady-state hydraulic loadings. These loads are produced primarily by pressure anJ, to lesser extent, by momentum changes of the reactor coolant as it flows past obstruc-tions and through components and elbows.

Under such loadings, the pipes and components attemp to expand and, because of the high structural redundancy of the system and its supports,-loadings similar to those produced by thermal expansion will result. To determine the effects of this loading condition, a static analysis similar to a deadweight-analysis is performed.

Steady-state hydraulic loads are applied to the model as static concentrated loads at locations throughout the system where resistance to flow is met. Computer code STALUM

-determines the resultant internal forces, moments, and displacements for each element. The boundary conditions imposed on the half-loop model during steady-state. loading are identical to those applied for the thermal analysis.

Similarily, all hydraulic snubbers are inactive in the mathematical model. The steady-state hydraulic loadings were evaluated for effect on the piping, compor,ents and pump supports.

In addition, known valaes of pump startup torque was evaluated for its effect on the RC pump supports and piping.

These loadings are within the design capacity of the pump assembly and supports.

B- 57

10.0 RCS COMPONENT SUPPORT SEISMIC CAPACITY FACTORS The requirement to design the Crystal River-3 Nuclear Power Plant (CR-3) for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant loop (RCL) piping has led to excessive design costs, interference with normal plant operation and maintenance, i and unnecessary radiation exposure of the plant maintenance personnel. NRC/ Lawrence Livermore National Laboratory (LLNL) sponsored a research program aimed at exploring whether the probability of DEGB in RCL Piping of nuclear power plants is acceptably small and if the requirements to '

design for DEGB effects (e.g., provision of pipe whip restraints) may be removed. Reference 5 describes the study performed for ten Babcock & Wilcox (B&W) plants of which l CR-3 was included. Reference 5 estimates the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of B&W supplied pressurized water reactor nuclear power plants in the United States.

The Reference 5 study used existing component support SSE i loadings to determine the seismic capacity factors for the

{

existing component supports. A listing a 1) the new CR-3 capacity factors (calculated in a manner similar to LLNL methodology), 2) the existing f actors calculated in the LLNL study for CR-3, and 3) the minimum SSE seismic capacity factor calculated for any B&W operating plant is listed on the next page.

Only the strength factor F s was recalculated. The probability of indirect DEGB at CR-3 due to new capacity factors and methods of analysis was not recalculated. The strength factor Ps, represents the ratio of the ultimate strength to the stress calculated for the design basis earthquake.

B-58 d

_ _ _ =-: _ - --

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l l

l 10.0 RCS COMPONENT SUPPORT SEISMIC CAPACITY FACTORS The requirement to design the Crystal River-3 Nuclear Power Plant (CR-3) for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant loop (RCL) piping has led to excessive design costs, interference with normal plant operation and maintenance, and unnecessary radiation exposure of the plant maintenance personnel. NRC/ Lawrence Livermore National Laboratory (LLNL) sponsored a research program aimed at exploring whether the probability of DEGB in RCL Piping of nuclear power plants is acceptably small and if the requirements to design for DEGB effects (e.g., provision of pipe whip restraints) may be removed. Reference 5 describes the study performed for ten Babcock & Wilcox (B&W) plants of which CR-3 was included. Reference 5 estimates the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of B&W supplied pressurized water reactor nuclear power plants in the United States.

The Reference 5 study used existing component support SSE loadings to determine the seismic capacity f actors for the existing component supports. A listing a 1) the new CR-3 capacity factors (calculated in a manner similar to LLNL methodology), 2) the existing factors calculated in the LLNL study for CR-3, and 3) the minimum SSE seismic capacity factor calculated for any B&W operating plant is listed on the next page.

Only the strength factor Fs was recalculated. The probability of indirect DEGB at CR-3 due to new capacity factors'and methods of analysis was not recalculated. The strength factor Fs, represents the ratio of the ultimate strength to the stress calculated for the design basis earthquake.

B-58

-~ , _ _ __ _

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/hY RCS COMPONENT SUPPORT SEISMIC CAPACITY FACTORS (Cont.)

W y SSE Capacity Factors for CR-3 Component Support New Old Min. B&W

-Reactor Pressure Vessel 28.1 5.56 3.76 Steam Generator ~ 67.4 118.0 2.57 Reactor Coolant Pump Snubber . 7.6 (1) 9.91

~'

. Reactor Coolant Pump Strut 29.8 (2) (3) 5 However, from Figure 2-8 of Reference 5 it can be inferred that a 50

-percent reduction in capacity would result in a probability-reduction of approximately one order of magnitude. Therefore, the new' support configuration at CR-3 has more than sufficient-margin

'when1 compared to PWR nuclear plants in' total.

. Notes:

~

.1 ) -The existing seismic ' capacity factor for the CR-3 pump snubbers was not calculated in the LLNL study. An approximate value would have been 12.

2)- The existing CR-3 pump-support configuration does not utilize rigid struts.

.3). The LLNL study did not calculate a factor for reactor coolant pump struts utilized at any B&W plant.

1 t B-59

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l ATTACHMENT B DRAWINGS (6)

B-60

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, . ,1 ATTACHMENT C INFORMATION ON BENCHMARKING ANT- .iRC AUDITINF- OF B&W JOMPUTER CODES e

Babcock & Wilcox computer codes are certified to B&W Quality Assurance Procedures. Administrative Procedures NPG-0902-06 and NPG-0903-03 define the requirements and responsibilities for developing certified computer programs and user manuals for execution in engineering design applications. These quality assurance procedures meet all the current licensing criteria of our customers.

The following B&W proprietary computer codes were used in the evaluation of the Crystal River-3 plant:

1) STALUM - General static, thermal, and dynamic analysis code for linear clastic and gap structures.

2) RESPECT - Calculates amplified response spectra due to structural amplification between a known point and an attachment point.

3) T3 PIPE - Performs a simplified piping stress analysis and fatigue evaluation of nuclear power piping in accordance with the design and analysis philosophy of USAS B31.7, 1969 Edition.

STALUM is the B&W standard analysis code used to develop static and dynamic loadings for the NSS components and piping. STALUM has been benchmarked against several structural test problems, calculations and/or other industry computer codes.

STALUM has not been benchmarked against the NRC's seven piping problems presented in NUREG/CR-1677 to comply with NUREG 0800.

Only problems 1 and 2 of the seven listed have been evaluated.

The STALUM results for problems 1 and 2 compare very favorably with NUREG/CR-1677.

Babcock & Wilcox is audited on a regular basis by the NRC. The following NRC audit inspection reports (performed over the past 5 years) were reviewed:

99900400/80-01, 80-02, 80-03, 80-04, 81-01, 81-02, 81-03, 81-04, 82-01, 82-02, 83-01, 83-02, 84-01, 84-02, 84-03 and 85-01.

C-1

).-

[.. ( ..

F F

I I. -From.the list of sixteen NRC Audits, the computer code certifica-tion files were audited in nine of these inspections. The p following is a listing of audits where either STALUM, RESPECT or

=T3 PIPE'Were specifically reviewed.

NRC InsDection ReDort ComDuter Code 99900400/80-01 STALUM and RESPECT 99900400/82-01 STALUM 99900400/83-02 STALUM 99900400/84-01 STALUM 99900400/85-01 T3 PIPE C-2 i

Qc-

.d 4-)

P ATTACHMENT D RELIABILITY AND SUPPORT CONFIGURATION CRYSTAL RIVER 3 GENERATING PLANT Prepared By FLORIDA POWER CORPORATION

-__..w

my _ .

s .:.

RC PUMP SUPPORT CONFIGURATION Present Optimized Quantity 32 Snubbers (8/ pump) 4 Snubbers (ael/ pump) 4 Rigid Struts Size 18" to 25" Bore Max. 14" Bore C&pacity 1000 Kip to 20P9 Kip 400 Kip OPTIMIZED COMPONENT RELIABILITY Snubbers-

. Integral Reservoir (Pressurized) i

. More Radiation Resistant Fluid

. Longer Seal Service Life

. Improved Control Valve Performance / Reliability

. Test-in-Place Capability

. Link Bars / Struts

. One-Time Installation .

.- Only Requires Visual Inspection i Large Component Supports

! . Acceptable R.V. and 0TSG Margins SYSTEM RELIABILITY

. Overall _ Flexibility of Piping System Increased

. Lower stress during thermal expansion of system - heat-up/ cool down.

(vs. present configuration if snubbers were assumed to fail in locked-up/ rigid condition)

'. Higher seismic stress (low probability of occurrence)

. Suffici ent design margins retained for piping and large canponent supports.

D-1

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(

ATTACHMENT E BASIS AND COSTS ASSOCIATED WITH ,

VALUE IMPACT ANALYSIS ,

CRYSTAL RIVER 3 GENERATING PLANT Prepared For FLORIDA POWER CORPORATION By BABCOCK & WILCOX COMPANY Lynchburg, Virginia

m.

y RADIATION COSTS RETAIN PRESENT CONFIGURATION (32 SNUBBERS)

TOTAL YEAR SITE TOTAL MHs EXPOSURE HOURS MAN REM- -RADIATION COST 1987 105,560 63,336 950 $ 4,770.000' 1994 105,560 63,336 950 4,770,000 2001 105,560 63,336 950 4,770,000 TOTALS 316,680 190,008 2850 $ 14,310,000 WITH LEAK B/ BREAK (# SNUBBERS REDUCED TO 8 W/4 RIGID STRUTS)

TOTAL ~

YEAR SITE TOTAL MHs EXPOSURE HOURS MAN REM RADIATION COST 1987 46,760 28,056 420 $ 2,100,000 1994 5,000 3,000 45 225,000 2001 5,000 3,000 45 225,000 TOTALS 56,760 34,056 510 $ 2,550,000 Typical cost of $5,000 per Man Rem has been established by FPC. Also, an average value of 15 Mr/Hr for time inside containment was used. To allow for ef ficiencies when working in a. radiation environment, a factor of 0.6 was applied to the total manhours to obtain exposure time.

(Example: Total Man-Rem = [(Site Total Hrs.) x 0.6] x 15 Mr/Hr)/1000 Exposure Hours ,

Note: All values are rounded "I off (August 1984 $)

4 w.

MAINTENANCE COSTS

( TOTAL COST = $11,518,400 L

! RETAIN PRESENT RC PUMP / MOTOR SUPPORT CONFIGURATION l (32 SNUBBERS, 8 ON EACH PUMP)

OTHER COSTS REBUILDING, SHIPPING, YEAR MATERIAL (SEALS / FLUID) LABOR HOURS LABOR COST ENGINEERING, SUPERVISION 1987 $ 40,000 105,560 $ 3,166,800 $ 716,000 1994 $ 40,000 105,560 3,166,800 591,000 l 2001 $ 40,000 105,560 3,166,800 591,000 TOTALS $ 120,000 316,680 $ 9,500,400 $ 1,898,000 l

l l WITH LEAK B/ BREAK ANALYSIS & ELIMINATION OF LOCA LOADS (TOTAL # OF SNUBBERS REDUCED TO 8, WITH ADDITION OF 4 RIGID STRUTS) TOTAL COST = $2,850,200 YEAR MATERIAL (SEALS / FLUID) LABOR HOURS LABOR COSTS DISPOSAL COSTS OTHER COSTS 1987 $ 450,000 47,750 $ 1,432,500 $ 32,500 $'555,200 1994 40,000 5,000 130,500 19.500 2001 40.000 5.000 130,500 19.500 TOTALS $ 530,000 57,750 $ 1,693,500 $ 32,500 $ 594,200 Note: All values are rounded '

off (August 1984 $)

E-2 L -

_ a

ML20137E719

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1.0 INTRODUCTION

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