ML20198R985

Rev 0 to Evaluation of Reactor Coolant Loop for Restrained Thermal Expansion,Trojan Nuclear Plant, Summary Rept

ML20198R985
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	06/30/1986
From:	ABB IMPELL CORP. (FORMERLY IMPELL CORP.)
To:
Shared Package
ML20198R910	List: ML20198R908 ML20198R985
References
01-0300-1525, 01-0300-1525-R00, 1-300-1525, 1-300-1525-R, TAC-61405, NUDOCS 8606100274
Download: ML20198R985 (28)

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tbrcl TrojarNuckearPlant Mr. Steven A. Varga Docket 50-344 June 3, 1986 License NPF-1 Attachment E Page 1 of 25 SU M Y REPORT EVALUATION OF REACTOR C00LMT LOOP FOR RESTRAINED THERMAL EXPANSION TRQ3M NUCLEAR PLMT PREPARED FOR:

PORTLMD GENERAL ELECTRIC i

PREPARED BY:

INPELL CORPORATION 350 LENNON LME WALPAIT CREEK, CALIFORNIA 9459S REPORT NO. 01-0300-1525 REVISION 0 JUNE, 1986 l

1 8606100274 860604 1

i DR ADOCK 0500 4

IMPELL CORPORATION REPORT APPROVAL COVER SHEET Client:

Portland General Electric Project:

Trojan Nulcear Plant Job Number: 0300-031-1356 Report

Title:

Evaluation of Reactor Coolant Loop for Restrained Thermal Expansion Trojan Nuclear Plant Report Number:01-0300-1525 Rev.

O The work described in this Report was performed in accordance with the Impell Quality Assurance Program. The signatures below verify the accuracy of this Report and its c mpliance with applicable quality assurance requirements.

Prepared By:

k) YL/hd Date e/d/6 6, Reviewed By:.M d

Date:

[8[

Date: h b

Approved By: IA_

REVISION RECORD Rev.

Approval No.

Prepared Reviewed Approved Date Revision ii

TABLE OF CONTENTS Title Page i

Report Approval Cover Sheet 11 Table of Contents 111

1.0 Background

1 2.0 Scope of Work 5

3.0 Description of System 6

4.0 Acceptance Criteria 10 5.0 Method of Analysis 12 6.0 Results 21

l 7.0 Conclusions 24 8.0 References 25 i

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111

l.0 INTRODUCTION Inspections of the Reactor Coolant System (RCS) at the Trojan Nuclear Plant indicated that the design expansion of the system may have been impeded and that the system could have experienced a thermal expansion i

condition not considered in the original design of the reactor coolant system. Thore. fore, the reactor coolant loop at the Trojan Nuclear Plant was evaluated for the effects of restrained thermal expansion of the system. The pertinent background infomation and purpose of this summary report are described below.

1.1 Background

The Trojan Nuclear Plant is a 4-loop 1130 Mf Pressurized Water f

Reactor (PWR). The Nuclear Steam Supply System was designed by i

Westinghouse. The original RCS design (relative to themal expansion of the system) was based upon unrestrained growth of the system radially outward from the reactor pressure vessel (RPV) conter11ne.

The supporting scheme for each of the four reactor coolant loops (R(1.) was designed to allow for themal growth with minimal impact on piping and components.

A conceptual drawing of a typical loop is shown in Figure 1.

l The supporting structure for the RCL consists of:

o Vertical and lateral supports at 4 RPV nozzles which allow radial growth of the vessel.

o Vertical steam generator supports which are pinned-pinned columns to allow movement of the generator outward from the RPV.

o Vertical reactor coolant pump (RCP) supports which allow the pump to " float" to accommodate motion of the cold leg and crossover leg.

o Upper and lower steam generator support frames which act as lateral guides (with hot gaps between the steam generator and support) and pipe rupture restraints.

o Steam generator snubbers at the upper frame to provide seismic and pipe rupture support in a direction radially from the RPV centerline but pomit themal growth as the RCS is heated up.

i o Vertical pipe rupture restraint on the hot leg.

l o Crossover log failure restraints at the crossover leg elbows.

These supporting structures were designed with the intent to allow free thermal expansion of the R(1..

Appropriate design gaps were specified to prevent action of the rupture and seismic restraints during normal operation.

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Recent inspections of the reactor coolant loop indicated that thermal expansion of the RCL may have been impeded during plant operation. These indications are:

o Potential inoperability of the steam generator snubbers.

o Erratic govement of the surge line (attached to the B RCL hot leg).

o Cold gap usasurements at various Ra supports which were less than the design gaps predicted by thermal expansion analyses.

o Contact between the RCL hot leg piping and RCL hot leg piping whip restraint (it.WR) as indicated by crushed graphite shtms and by pullout of a horizontal stabilizing brace (on RCL B).

The evidence leads to a scenario which indicates that snubber lockup and/or lower steam generator support contact (due to inadequate gaps) caused a rotation of the steam generator and subsequent downward movement of the hot leg. The hot leg movement could have led to surge lir.e movements unexplained by design and normal condition operations and could have led to damage to the HLWR horizontal stabilizing brace.

A preliminary evaluation of these conditions was performed and documented in Reference 1.

The preliminary review showed that thermal expansion stress limits were exceeded at the hot leg elbow. The remainder of the Ra satisfied acceptance criteria for the piping and components. The review stated that a more detailed evaluation would resolve piping and component qualification issues.

Portland General Electric (PGE) developed a two-part program to add ass I

the restrained thermal growth issue.

The program consists of che following:

o Restoration of the RCL to original design conditions and development of a monitoring program to ensure that restrained thermal growth does not occur.

o Evaluation of the Ra for the most limiting case (bounding conditions) of restrained thermal growth event to demonstrate acceptability of the system.

This two part program ensures the structural adequacy of the RCL and demonstrates that design clearances and snubber operability are provided.

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1.2 Purpose The purpose of this summary report is to document the results of the RQ. evaluation for the restrained thermal growth event. This summary report addresses the second part of PGE's program to fully resolve this issue. This report presents the scope of work, method of analysis, acceptance criteria and results of the RCL evaluation.

This evaluation considers past conditions to demonstrate that plant operation was not comprcaised and that there are no adverse effects on future plant operation.

Since the plant is restored to original design conditions, the original design stresses and loads of Reference 2 are valid.

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2.0 Scone of Work The scope of work to resolve the restrained thermal growth event is based on the results of the preliminary evaluation (Reference 1).

The scope of work for the RQ. thermal expansion analysis consists of the following steps:

1.

Nonlinear analysis of a typical loop to demonstrate structural adequacy of the RQ. relative to fatigue considerations.

2.

Nonlinear evaluation of the hot leg elbow (shell analysis) to accurately determine component strains and fatigue damage.

3.

Component and support qualification to demonstrate compliance with acceptance criteria.

The detailed procedures and methods used to perfonn this scope of work are provided in Section 5.0 and the results are described in Section 6.0.

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3.0 Descrintion of System The RCL is the major cooling water system for the reactor. The water is circulated through the RPV and passes through the steam generator to provide the steam source for power generation.

A reactor coolant pump drives the flow through the system.

The geometry of the RCL is described in this section.

3.1 Geometry The reactor coolant system consists of 4 loops connected to the reactor pressure vessel.

A conceptual sketch of a typical loop is shown in Figure 1.1.

The geometric characteristics of the system components are provided in Figure 3.1.

3.2 Material s The piping and components are fabricated from the materials satisfying the ASTM specifications listed in Table 3.1.

3.3 Ooarating conditions The design operating conditions consist of pressures and temperatures for the various parts of the RCS. The pressures and temperatures associated with the operating conditions are listed in Table 3.2.

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FIGURE 3 1 SYSTEM GE0 METRY 1

ASTM SPECIFICATION COMP 0NENT FOR MATERIAL

-l RPV Shell SA533 Gr B C1. 1 RPV Nozzles SA508 C1.2 4

Elbows A351 Gr CF8M Pipe A351 Gr CF8A SG Nozzles SA216 WCC RCP Casing A351 Gr CF8 l

Supports A36 Pipe Branch Fittings A182 F316 SG Bottom Chamber SA216 WCC SG Shell and Head SA533 Gr A C1. 2 1

TABLE 3. 1 MATERIALS

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l Normal Hot Leg 2235 616.8 Operation Crossover Leg 2235 552.3 Cold Leg 2235 552.5 1

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4.0 Accentance criteria The Trojan reactor coolant system was originally designed to the requirements of ANS 831.7. The evaluation of the RCS for the restrained thermal expansion was performed to the rules of the ASE Boiler and Pressure Vessel Code Section PS (1983 edition). The use of the ASE Code was considered acceptable since the material properties and composition as defined in the ASTM standards for these materials are not changed significantly to affect the use of the ASE Code.

4.1 Pressure-Retaining haanants The restrained thermal expansion of the RQ. affects the Code Compliance of the piping and components in two ways o Thorwal expansion stress limit.

Thermal expansion limits are imposed for two reasons:

(1) to prevent formation of a plastic hinge which could lead to collapse of a component, and (2) to justify the simplified elastic-plastic f atigue rules in the ASE Code.

For the restrained growth of the RCL, the displacements are limited so that collapse of the system will not occur.

In order to demonstrate the integrity of the system, strain limits and the fatigue usage limits are the only requirements to be satisfied, o Fatigue usage limit.

The acceptance criteria for the pressure-retaining parts is usage f actor less than 1.0 The usage f actor will be calculated by adding the usage due to 28 cycles of the restrained thermal growth event to the usage factor calculated by the original design report.

o Strain Limits Strain limits are specified to ensure that the structural response of the systec is capable of withstanding the loadings imposed by the thermal expansion event. The strain limits used are based on ASE Code Case N-47, Appendix T, which specifies maximum local membrane strains of 15. Therefore, the strains in the RQ. will be limited to 15.

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4.2 suenorts lhe acceptance criteria for supports and non-pressure retaining parts is based on Section NF of the ASME Code.

Exceptions to the Code l

criteria will be justified when used. The exceptions will be taken for bolt tad pin allowables which exceed Code limits but will be limited to less than yield stress. Also for cases where Code allowables are exceeded due to f actors of safety, a reduced f actor of safety will be applied for the analysis of the supports. These exceptions are considered acceptable since the evaluation of the RCS is based on worst-case assumptions and loadings. _.

5.0 Method of Analvsts The evaluation of the reactor coolant system is performed by nonlinear analysis techniques.

A description of the methods of analysis are described in this section.

5.1 Nonlinear Ra Evaluation The nonlinear analysis considered a 1-loop model. A 4-loop model analysis showed that lockup of snubbers on various loops produced negligible displacements at the RPY centerline. Therefore, the loops act independently and there is no carryover effect on a loop due to constraints on another loop. Therefore, the use of a 1-loop model for the nonlinear analysis is justified. The evaluation considered the effects of nonlinear component behavior, gaps at critical locations, and bounding loading conditions.

5.1.1 Modell ing The nonlinear analysis model was developed using the MSYS computer program. The model includes the Ra piping, major components, and supporting structure and gaps. The model is shown in Figure 3.1.

The MSYS elements used for the evaluation are o Plastic pipe elbow element (STIF 60).

o Plastic pipe element (STIF 20).

o Elastic pipe element (STIF 9).

o Gap element (STIF 39).

The piping components were modelled with the plastic pipe and elbow elements and provide for elastic-plastic response of the l

piping system. The supports were expected to remain elastic and were modelled with elastic elements. The results of the support evaluation indicated that the elastic support response was valid. The clearances between piping and components and their supports were modelled using gap elements and provided for closure of the support clearances.

t 5.1.2 Material Pronertf as The nonlinear evaluation required formulation of a lower bound material stress-strain curve. For the RCL analyses, material properties were based on L.D. Blackburn's work ("Isochronous Stress-Strain Curves for Austenttic Stainless Steel") to define material properties for Code Case N-47.

The Blackburn correlations are used for hot temperature material I

properties. The ambient temperature properties are based on Ramberg-Osgood correlations.

4 The material correlations provide a means of defining material stress-strain curves for given temperatures and yield stresses. The Blackburn correlation provides a curve which best " fits" stainless steel properties at 600 F.

The curves used in the analyses were based on proportional 11m'ts which equal ASE Code values at the appropriate temperatures. The Code materials are lower bound properties.

The material curves used in the evaluation are shown in Figures 5.2 and 5.3.

The proportional limits for these l

materials match ASE Code yield stresses. The actual material properties for the hot leg elbows vary from a yield stress of 34.5 ksi to 48.0 ksi. Therefore, conservative material properties have been considered in the evaluation.

5.1.3 Loading Conditions There are several possible loading conditions to be considered. Tne worst-case loadings are obtained by providing limited ette'acements and gaps. The assumptions used in establishing bounding conditions which were considered are:

o Locking snubbers in the cold and fully hot position.

o Varying gaps at the hot leg whip restraint from 0-0.25".

(Cases were run at 0", 0.125" and 0.25").

The variation of gaps at the hot leg whip restraint is considered because the loadings at the hot leg elbow and RPV hot leg nozzle are affected by this gap. The design gap is specified as 0.125 and field measurements indicated gaps partially closed at one restraint. An increased gap of 0.125" is used to account for crushing of the graphite shims.

Variation of the HLWR gap from 0-0.25" provides a bound of all possible gap conditions at the restraint. The physical conditions represented by the evalution are:

Zero gap between pipe and restraint in cold position.

Maximum 0.25" cold gap between the pipe and restraint in the cold position.

Uplift of the pipe due to restraint rotation of 0.125 ".

o Using minimum measured gaps between the steam generator and l

Its lower and upper support frames.

o Using minimum measured gaps on the crossover leg whip restraints.

i A listing of the loading conditions evaluated is provided in Table 5.1.

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5.2 Elbow Evaluation The nonlinear analysis indicated that the location of the highest stresses in the RCL was the elbow at the steam generator on the hot l eg.

Since this component experiences the largest strains and in order to verify the use of the piping model elbow element, this component was evaluated using 3-dimensional finite element analysis (shell analysis).

5.2.1 Geometry The elbow was modelled using 20-node isoparametric elements.

The ANSYS STIF 95 element is used in the evaluation. A one-half model of the elbow was used since the in-plane action dominates the RQ. response. The model is shown in Figure distance of (Rt)ppgel is extended in the hot leg side by a 5.4.

The elbow to preclude end effects on the elbow resul ts.

The nozzle end is extended to the nozzle taper and this end is assumed to remain round since the nozzle stiffness is large.

5.2.2 Materi al s The material properties used previously in 5.1.2 remain applicable for this evaluation.

5.2.3 Loading The loading on the elbow consisted of internal pressure and applied displacements from the nonlinear piping analysis. The maximum difference in rotation and displacements from one end of the elbow to the other is calculated, factored by 1.25 for conservatism, and applied at the steam generator nozzle end of the model. The other end of the model is fixed against rotation and displacement. A conservatist. factor of 25% was used to compensate for variations in the displacements and rotations assumed in the inputs.

5.3 Cannonents Analysis Fatigue evaluations were performed for the following components (refer to Figure 1.1 for identification):

o RPY hot leg nozzle.

o RPY hot leg bimetallic weld o Hot leg elbow.

o SG hot leg bimettalic weld.

o SG hot leg nozzle.

o SG crossover leg nozzle.

o Pump crossover nozzle. _.

The components considered were the locations which were likely to experience the maximum usage f actors. The nozzles are evaluated using Bijlaard analysis techniques by taking the maximum loadings from the nonlinear analysis and calculating the usage factor based on ASME Code NB-3200 simplified elastic-plastic methods.

The Bijlaard analysis techniques are an empirical method of calculating stresses in cylinder to cylinder intersections and cylinder to sphere intersections documented in WRC Bulletins 107 and 297.

The bimetallic welds were evaluated using loadings from the nonlinear analysis and combining these with temperature mismatch effects to determine the usage due to the restrained growth event. The ASE Code simplified elastic-plastic rules are utilized.

This approach is valid since the average strains at the bimetallic weld are only slightly above yield strains.

Fatigue evaluations have not been completed for branch connections to the RCS.

Fatigue stresses in the branch connections were not considered to be limiting because they are usually governed by transients. The thermal expansion stresses in the run pipe would normally not factor into the stress range calculation which produced the highest usage factor at the branch connection.

The hot leg elbow was evaluated by using the strains predicted by the elbow model and directly calculating the increased usage factor due to restrained thermal growth. Restrained thermal growth was assumed to be caused by failure of the steam generator hydraulic snubbers in a locked condition.

5.4 Sunnorts The supports were evaluated by using standard linear elastic techniques for static loadings.

e LOAD CASE NO.

CONDITION 1

2 3

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GRAVITY X

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PRESSURE X

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HOT TEMPERATURE X

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SNUBBERS LOCKED-COLD X

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X SNUBBERS LOCKED-HOT X

HOT LEG WHIP RESTRAINT CONTACTED X

X X

X RADIAL S.G. BUMPERS CONTACTED X

X X

CROSS 0VER LEG FAILURE RESTRAINTS CONTACTED X

X TABLE 5.1 LOADING CONDITIONS.. - -..

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i FIGURE 5.2 NONLINEAR STRESS-STRAIN CURVES - PIPING

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Loop A: 48 ksi Loop B: 35 ksi Loop C: 42 ksi A

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6.0 Resul ts The results of the evaluation consist of fatigue damage at the critical components and structural evaluation of the support members. The component evaluation is summartzed in Table 6.1 and the support evaluation is described in Table 6.2.

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I COMPONENT RESULTS COMPONENT Ug au UTOT RPV Outlet Nozzle 0.2 0.02 0.22 RPV Outlet Nozzle Safeend Weld 0.63 0.14 0.77 Hot Leg /Elbcw Weld 0.2 0.1 0.30 Hot Leg Elbow 0.1 0.03 0.13 Safe Weld SG Inlet Nozzle 0.88 0.004 0.884 SG Outlet Nozzle 0.88 0.001 0.881 RCP Outlet Nozzle 0.20 0

0.20 RCP Outlet Nozzle 0.1 0.004 0.104 NOTES:

(1) Uw = original design usage factor Usage factor due to restrained thermal expansion event as u

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TABLE 6.1 COMPONENT RESULTS _ _.

SUPPORT RESULTS Code

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(KSI) (1)

SG Baseplate Bolts A354 BC 99.6 62.5 1 25.0 SG Column Pins A193 B7 47.6 20.7 50.0 (2)

SG Support Col umn A36 (assumed) 23.3 18.2 32.9 SG Capscrew AS40 823 CL1 136.4 82.5 165.0 RCP Baseplate A354 BC 72.1 62.5 125.0 Bol ts RCP Column Pins A193 B7 22.1 20.7 50.0 (2)

RCP Support Col umn A36 ( Assumed)

Interaction based on Code allowable exceeds 1.0 in tension; however, interaction based on operability limits is less than 1.0.

RCP Bolt AS40 B24 CL1 99.0 82.5 165.0 RCP 2" Bent Plate at Adaptor Shoe A36 ( Assumed) 24.0 16.2 36.0 NOTES:

1) Table is limited to components above code allowables.

2) Ultimate Stress = tensile stress or operability limit as appropriate except as noted.

3) Loaded in shear, ultimate stress limit for material is based on one half tensile strength (S /2).

u TABLE 6.2 SUPPORT RESULTS -

7.0 Concl usi ons The reactor coolant loop at Trojan satisfies the accceptance criteria established for the restrained thermal growth event.

Since the system has been returned to original design conditions, the RCL remains structurally acceptable for its intended use and there are no adverse effects in the system from the postulated restrained thermal 9rowth event.

The intent of the evaluation is to bound the worst possible loadings at all critical locations. The elbow on the hot leg and the RPV outlet nozzle safeend are the locations with highest strain and fatigue usage, respectively.

The evaluation of these components is bounded by:

o Bounding gaps at the hot leg whip restraint have been evaluated in order to obtain worst case loads at the RPY outlet nozzle and hot legelbow.

Snubber lockup in the fully cold and fully hot position has been considered.

Closure of gaps on the crossover leg whip restraint and steam generator support frame has also been reviewed.

l o Material properties have been developed which provide lower bound J

results on the RCL piping and components.

Based on a bounding of the physical restraints on the system and the component material properties, the RCL is demonstrated for the effects of the restrained thermal expansion loading.

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8.0 References 1.

Preliminary Thermal Expansion Evaluation of the Reactor Coolant Loop for the Trojan Nuclear Plant, dated January 1986 (Attachment of PGE to NRC letter dated May 9, 1986).

2.

Westinghouse Stress Report SD-112.

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