ML20197E758

Rev 0 to Preliminary Thermal Expansion Evaluation of Reactor Coolant Loop for Trojan Nuclear Plant

ML20197E758
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	01/31/1986
From:	Bak W, Grubb R, Mcleod W ABB IMPELL CORP. (FORMERLY IMPELL CORP.)
To:
Shared Package
ML20197E724	List: ML20197E758
References
01-0300-1471, 01-0300-1471-R00, 1-300-1471, 1-300-1471-R, TAC-61405, NUDOCS 8605150310
Download: ML20197E758 (34)

Text

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e Mr. Steven A. Varga May 9, 1986' Attachment B Page 1 of 32 PRELIMINARY THERMAL EXPANSION EVALUATION OF THE REACTOR COOLANT LOOP FOR THE TROJAN NUCLEAR PLANT

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Prepared for:

! PORTLAND GENERAL ELECTRIC COWANY 4

Prepared by:

Impell Corporation 350 Lennon Lane Walnut Creek, California 94598 Impe11 Report No. 01-0300-1471 Revision 0 January, 1986 l' 8605150310 860509 PDR ADOCK 05000344 S PDR

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IWELL CORPORATION l

l REPORT APPROVAL COVER SHEET Client: Portland General Electric Project: Trojan Job Nunber: 0300-029-1356

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Report

Title:

Preliminary Tharmal Expansion Evaluation of the Reactor Coolant Loop for the irojan Nuclear Plant Report Number: 01-0300-1471 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 compliance with applicable qualilty assurance requirements.

Prepared By: 4/26e 2 8d 3, Date: //tJ/BG Walter R. Bak, Jr., Mh' nager, Applied Mechanics Section Reviewed By: Md M Date: I/13 /8(,

W iam S McLeod, Lea eni ' Engineer Approved By: cr Date: / 3

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Robert L. Grubb, Manager, Advanced Engineering Division REVISION RECORD J

1 Rev. Approval No. Prepared Reviewed Approved Date Revision i

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, TABLE OF CONTENTS Page Title Page i Report Approval Cover Sheet 11 Table of Contents iii 1.0 Background 1-1 2.0 Scope of Work 2-1 l

3.0 Methods of Analysis 3-1 4.0 Sumary of Results 4-1 5.0 Conclusions and Recomendations 5-1 6.0 References 6-1 Appendix A: Meeting Minutes for PGE/Impell A-1 Discussion on RCL Thermal Expansion Analysis and Surge Line Evaluation (PGE Letter RLS-1411-85, dated f November 22,1985)

Appendix B: Recomendations for Data to be B-1

l. Collected in Followup Inspections at Trojan Nuclear Plant 5

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1.0 BACKGROUND

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Impell Corporation has previously evaluated the surge line at the Trojan Nuclear Plant to attempt to define the cause of observed movement of the piping system and to determine the effects of the movement on the structural adequacy of the surge line. The results of this evaluation are documented in Reference 1.

The surge line was modified in June,1982 to allow for removal of the thermal sleeve at the surge line l connection to the reactor coolant loop (RCL). During this modification effort, it was noted that the clearance between the surge line and several rupture l restraints was not as originally designed. In some cases, there was contact between the pipe and the

, rupture restraints. Reshiming of the rupture restraints was performed to reestablish design clearances. Additionally, a program was started to

examine clearances at the surge line rupture restraints. Later inspections indicated'that the i surge line continued to " move" ( not returning to its ,

original cold position). Reshiming was done on two '

separate occasions to maintain design clearances. l Impell reviewed the displacement data of the surge line in May, 1985 and performed an evaluation to g identify the cause of the surge line movements.

g Based on a review of the surge line geometry and its system function, the potential for a stratified thermal layer in the piping system was identified. -

Subsequent temperature monitoring of the surge line I indicated the existence of the stratified thermal I layer. The evaluation of the piping system under the i

effects of gravity, thermal expansion, stratified thermal layer, and cold spring was performed.

However, this analysis did not reproduce the surge line displacement pattern.

Impell reviewed possible causes of the displacement pattern and showed that small bending rotations in the RCL could produce the observed deflection of the surge line. Based on the evaluation of the surge line, Reference 1 provided the following conclusions:

o RCL rotation provided the "best" fit to the  !

observed surge line movement. i 3

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1.0 BACKGROUND

f l 0 Stresses in the surge line in its observed deflected shape were well within ASME Code allowables.

However, the surge line evaluation did not conclu-sively identify the cause of the observed surge line displacenent.

During discussions with Portland General Electric Company (PGE) in November,1985, PGE indicated that testing done on two steam generator snubbers on Loop D of the RCL during the 1985 refueling outage, showed the two snubbers to be in a degraded condition. In earlier efforts it had been noted that the hot leg rupture restraint on the B Loop had apparently been contacted at one time. Based on these discussions, Impell informed PGE that degraded snubbers could have produced the surge line movements and caused the hot leg rupture restraint contact. Impell also indicated that the snubber cor uition could impact the thermal expansion analysis of the RCL. PGE then requested that Impell perform a preliminary evaluation of the RCL and this report sumarizes the findings of the l RCL thermal expansion evaluation.

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2.0 SCOPE OF WORK This preliminary evaluation of the RCL, under the -

, effects of thermal expansion with degraded steam * '

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9enerator snubbers, is performed to demonstrate the structural adequacy of the RCL and its associated supports and attachments. The evaluation is performed using conservative procedures.in order to obtain a " quick" assessment of the effects of the degraded snubbers. The scope of work developed to r perform the preliminary evaluation consists of the i

j following tasks:

o Perform a thermal expansion analysis of the RCL hot leg assuming that the steam generator ,l

' snubbers are fully locked or frozen. <

II o Evaluate the stresses at the reactor pressure l1 vessel (RPV) hot leg nozzle and at the steam generator (SG) hot leg nozzle.

o Evaluate the structural adequacy of the RCL restraints.

o Review effects of assumed snubber lock-up on other piping systems attached to the hot leg and steam generator.

These steps were performed in the preliminary evaluation and the analysis procedures, assumptions,

' criteria, and results are reported in the following sections of this document. The basis of defining this scope is presented in Appendix A.

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3.0 METHODS OF ANALYSIS i

The preliminary RCL themal expansion evaluation is  ;

performed to demonstrate the structural adequacy of l the RCL piping, supports and equipment under the assumed condition of steam generator snubber lock-up. The analytical methods, assumptions and qualification criteria for the preliminary evaluation are described in this section.

The Trojan RCL piping was originally designed to the criteria of ANSI B31.7 (Reference 2). These criteria are comparable with the rules of NB-3600 of the ASME Code Section III (Reference 3). For the analysis of l the RCL thermal expansion, the ASME Code rules will be utilized.

3.1 RCL Thermal Expansion The RCL is designed to allow free thermal expansion l of the piping and equipment when the reactor coolant

' Evaluation system heats up for normal operation. The snubbers I on the steam generator are installed to provide <

l support for the reactor coolant system (RCS) during postulated seismic and LOCA events and to allow stress-free movement under thermal expansion conditions. If the snubbers lock-up during normal operation, there will be a significant impact on the thermal expansion stresses in the piping and on the support and nozzle loads.

Based on a review of the RCL design, the significant impact of steam generator snubber lock-up will be lg experienced by the hot leg (piping between the RPV and SG). Due to the support scheme at the reactor coolant pump (RCP), which allows the pump to " float",

the crossover leg and the cold leg are not expected to be significantly affected by the snubber inoperablility. Therefore, the preliminary evaluation will include only the hot leg.

The model used for the thermal expansion analysis is shown in Figure 3-1. The model includes the following components:

o Hot leg piping o Steam generator and its vertical supports I o RPV and its nozzle supports 1

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3.0 METHODS OF ANALYSIS l

1 o Steam generator snubbers o Hot leg rupture restraint Major modelling considerations are as follows:

o Steam generator vertical support flexibility is included.

o No movement of the RPV centerline is considered (each loop acts independently).

o No gap is assumed at the hot leg rupture restraint.

o Actual elbow dimensions are considered.

The thermal expansion analysis is performed for the model assuming an operating terrperature of 617'F, a design temperature of 650*F and an ambient temperature of 70*T. The evaluation is performed for the following cases:

o As-designed configuration

{ o Snubbers assumed frozen in the cold position and I hot leg restraint inactive o Snubbers assumed frozen in the cold position and hot leg restraint active o Snubbers assumed frozen in the hot position and hot leg restraint inactive The as-designed evaluatiorr was performed to assure that the Mt ' leg model produces results similar to I the original RCL analysis in order to demonstrate the aderdr;: of ihe model. The evnluations with inactive hot la y et;., a re restraints are performed to determine '

whetuer the not leg piping will impact the restraints. The evaluation with active hot leg rupture restraints completes the piping evaluation scenarios.

The stresses produced in the piping are compared to the allowable stress of 3.0 Sm for thermal expansion stresses as calculated by Equation 12 of the ASME Boiler and Pressure Vessel Code,Section III, e

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.[ 3.0 METHODS OF ANALYSIS Subsection NB-3600. The Equation 12 calculation is:

S E = C2 M / Z 2E 3.0 Sm where: M = thermal expansion moment range C2 = secondary stress index s

Z = section modulus of pipe Sm = material allowable stress intensity

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The Code evaluation for fatigue damage is also

.j required. However, due to the limited number of I heatupandcooldowncycles(basedon30 cycles),the fatigue usage factor will be significantly below the allowable usage of 1.0.

4 The results of the evaluation are described in Section 4.1. However, the worst-case stresses occur in the hot leg elbow. This is the only location where the Code allowable stresses are exceeded. The evaluation of the elbow is described in the following paragraphs.

3.2 Hot Leg Elbow The thermal expansion stress limit in the ASME Code Evaluation is specified to prevent formation of a plastic hinge in the piping system and to prevent collapse of the piping. The Code rules are intentionally conservative to allow for appropriate design j margins. For the evaluation of the thermal expansion stresses in the hot leg piping due to assumed steam generator snubber lock-up, it is appropriate to develop a more realistic criteria.

The ASME Code Section III NB-3200 provides rules for l a detailed stress analysis of components. The rules l in NB-3228 provide limits for plastic analysis. The l plastic analysis rules require that:

o Strains be kept below acceptable limits.

o Fatigue usage limits be satisfied.

The strain limits.are not explicitly defined in the Code. However, Code Case N47, Appendix T, provides strain limits of 1% for membrane strains, 2% for 3-3 lj Impell Report 01-0300-1471 4

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3.0 ETHODS OF ANALYSIS 1

bending strains, and 5% for peak strains. Code Case N47 applies to components which operate at elevated temperatures; however, the limits are also applicable to normal temperature components. Therefore, a i strain-based criteria are appropriate for the

! evaluation of the elbow..  !

!l The determination of strains in the elbow will be  !

based on analytical calculations and compared to test i data. The analytical calculations will be based on the simplified elastic-plastic rules in NS-3200 and in Code Case N47. The test data comparison will be

J performed by examining the Greenstreet elbow test i data (Reference 5) for components with similar radius-to-thickness ratios. The results of this evaluation are provided in Section 4.2.

3.3 Nozzle Evaluation The effects of the themal expansion loadings on the RPV and SG are considered. The loadings which affect the nozzles are gravity and thermal expansion. The nozzle evaluation is performed by using the thermal expansion loads generated in the piping analysis and calculating the stresses in the nozzle using the ,

procedures of Welding Research Council Bulletin 107 (Reference 4). WRC-107 procedures provide a

.g simplified method of calculating stresses at the

nozzle-vessel intersection. '

The nozzle evaluations are performed for two locations:

o RPV hot leg nozzle

.f o SG hot leg nozzle The criteria of the ASME Code NB-3200 are used to evaluate the effects of the thermal expansion loads.

For the preliminary evaluation there are three criteria to be considered:

o Local membrane stress less than 1.5 S, o Local membrane plus bending stress less than 3.0 Sm o Usage factor less than 1.0 3-4 Impell Report 01-0300-1471 Revisien 0 i

l' 3.0 METHODS OF ANALYSIS l

Since there is a limited number of heatup and cooldown cycles (based on 30 cycles), the usage factor will be well below 1.0 (calculated usage less than0.1). Therefore, only the membrane and bending effects will be analyzed in detail. The results of this evaluation are included in Section 4.3.

3.4 Support Evaluation The critical support elements are evaluated for the

! effects of the thermal expansion loads. The support elements evaluated include:

o Steam generator snubbers i o Steam generator vertical supports o Hot leg rupture restraint o RPV nozzle supports The supports are evaluated using standard procedures for structural analysis. The elastic criteria for the structural elements and bolts specified in

[ Appendix XVII of the ASME Code Section III are I utilized in the support review. The results of the support evaluation are provided in Section 4.4.

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.g 3.5 Non-RCL Piping The effects of the assumed snubber lock-up on the Evaluation piping attached to the hot leg and steam generator are evaluated. The piping systems reviewed are:

o Main steam line 1

o Feedwater line o Safety injection line o RHR line o Pressurizer surge line o RTD line The evaluation of these systenis is performed in '

accordance with the following steps:

o Determine distance to first support on piping ,

g from the RCL or SG

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, 3.0 METHODS OF ANALYSIS

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o Determine the maximum movements at branch connection points o Calculate stresses in piping lines based on fixed-simple ended beams The stress criteria applicable to the non-RCL piping are provided in Equation 12 of NB-3600. The results of the non-RCL piping analysis are provided in Section 4.5.

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3.0 METHODS OF ANALYSIS bY '

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l FIGURE 3-1: HOT LEG MODEL l 3-7 Impell Report 01-0300-1471 f* Revision 0

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4.0 St# MARY OF RESULTS The preliminary RCL evaluation results for thermal expansion loads, assuming locked-up steam generator snubbers, are presented in the section. The results are based on a conservative evaluation procedure for a single hot leg. The results assume complete lock-up of the snubbers in the cold and hot positions. The results associated with each task described in Section 3.0 are presented in the following paragraphs. The results are described in detail in Reference 6.

4.1 RCL Evaluation The piping thermal expansion stresses satisfy the ASME Code limits for all hot leg piping components except for the elbou at the steam generator. The stresses at the RPV nozzle hot leg piping weld, the steam generator hot leg nozzle region, and the hot leg elbow are presented in Table 4-1. The overstress at the elbow is considered acceptable to demonstrate piping structural adequacy based on the discussion in Section 4.2.

4.2 Hot Leg Elbow The strain calculated in the elbow due to thermal Evaluation expansion loads is less than the allowables developed in Section 3.2. The maximum strain determined from analysis is 0.8% and the maximum strain determined from the test data comparison is 0.2%. These strain values are less than 1% so that the elbow is considered structurally adequate for this worst-case loading.

Impell has also previously performed a fracture mechanics analysis of the RCL. The evaluation is

'li documented in Reference 7. The fracture mechanics analysis demonstrated the structural. integrity of the RCL in its current configuration (no snubber lock-up),

assuming a crack which results in detectable I leakage. This evaluation remains valid for the RCL and provides additional assurance of piping integrity since no leakage or cracks have been detected.

4.3 Nozzle Evaluation The stresses at the RPV hot leg nozzle and the SG hot leg nozzle are tabulated in Table 4-2. The results indicate that the stresses are much less than the Code allowables so that the thermal expansion load case has no adverse effect on the RPV and SG.

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4.0 SUMMRY OF RESULTS j 4.4 Support Evaluation The supports on the hot leg and steam generator were i evaluated and the critical results are presented in Table 4-3. Based on this preliminary evaluation, the following results are noted:

- There is a slight overstress versus ASME Code

.l allowables at the RPV nozzle support leveling screws. This stress result is very conservative since only one leveling screw at each nozzle is assumed to resist the snubber lock-up loads. A more realistic evaluation would distribute the load to the additional leveling screws and the

(i stresses would be less than Code allowables.

- The critical elements of the steam generator support are the steam generator support column pins, capscrews, and the baseplate bolts. These items are significantly above Code allowables; however, the stresses are below material ultimate strengths. While this evaluation shows these elements to be highly-stressed, a more rigorous evaluation of the RCL, considering all flexibilities in the system, will produce lower stresses in the steam ger.erator supports.

i! The support evaluations are based on the conservative

i evaluation techniques used for this preliminary assessment.

4.5 Non-RCL Piping The maximum incremental stresses produced by the Evaluation frozen steam generator snubbers are presented in Table 4 The results show that the incremental stresses are very small. Since the conservative analysis techniques used neglected rotational pipe flexibility, elbow flexibility, support gaps, and 4

support flexibilities, it is expected that the actual stresses will be less than those calculated and will provide additional margin. These low stresses indicate that previous qualifications of these lines would not be significantly impacted.

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4.0 SUP9tARY OF RESULTS TABLE 4-1 HOT LEG STRESS EVALUATION Expansion Unintensified Allowables Yield Location Load Case Stress (ksi) Stress (ksi) Stress (ksi) Stress (ksi)

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RPV Nozzle Weld 1 41.2 34.3 53.4 19.8

, 2 13.3 11.1 53.4 19.8 3 16.7 13.9 53.4 19.8 4 2.9 2.4 53.4 19.8 l SG Nozzle 1 20.5 17.1 66.0 33.1 i1 2 34.5 28.7 66.0 33.1 3 6.2 5.1 66.0 33.1 1 4 31.4 26.1 66.0 33.1

! t Elbow 1 62.0 24.7 56.1 20.8 2 79.6 31.7 56.1 20.8 3 .13.9 5.5 56.1 20.8 4 74.4 29.7 56.1 20.8 I

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I NOTES:

(1) Load cases: 1 - Snubber locked cold; rupture restraint inactive 2 - Snubber locked cold; rupture restraint active 3 - As-designed j 4 - Snubber locked hot; rupture restraint inactive 1

(2) Expansion stress = C2 M/z (based on hot modulus of elasticity)

(3) Allowable stress = 35m (at temperature)

(4) Unintensified stress = M/Z l

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4.0 SUPMARY OF RESULTS TABLE 4-2 N0ZZLE STRESS EVALUATION I i PRIMARY + SECONDARY LOCAL LOCAL MEMBRANE (Pt ) MEMBRANE PLUS BENDING (Pt + Q)

LOAD STRESS ALLOWABLE STRESS ALLOWABLE LOCATION CASE U) INTENSITY (ksi) STRESS (ksi)(2) INTENSITY (ksi) STRESS (ksi)(3)

I RPV Nozzle 1 14.9 40.1 25.4 80.1 I

i 2 13.0 40.1 17.5 80.1 SG Nozzle 1 20.6 33.0 43.0 66.0 2 15.1 33.0 19.1 66.0 l

NOTES: (1) Load cases: 1-Snubber locked cold; rupture restraint active 2-As-designed

{ (2) Allowable Stress - 1.5 Sm (3) Allowable Stress - 3.0 Sm

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SUMMARY

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l TABLE 4-3

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SUPPORT EVALUATION RESULTS (2)

(1) CODE ULTIMATE LOAD CALCULATED ALLOWABLES STRESS COMP 0NENT ' CASE MATERIAL STRESS (KSI) (KSI) (KSI)

, RPV LEVELING SAS40 B24 l SCREWS 1 31.9 30.0 72.5(3) 4 2 0.2 SG SUPPORT A-36(ASSUMED)

COLUMN 1 20.1 20.6 58.0 2 5.2 SG COLUMN A193 87 PINS 1 41.0 20.7 50.0(3) 2 10.6 BASE PLATE BOLTS A354 BC 1 100.5 62.5 125.0 l 2 =<3cC62.5 1

SG CAPSCREWS A540 B23 CL1 1 137.7 82.5 165.0 2 <<8 2. 5 i

(1) Load cases: 1 - Snubber locked in cold position, rupture restraint active.

2 - As-designed (2) Ultimate Stress = Tensile stength except as noted.

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

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SUMMARY

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TABLE 4-4 NON RCL PIPING EVALUATION NOMINAL LINE STRESS (M/Z)

PIPING SYSTEM SIZE (in) (ksi)

RTO 1&2 5.09 RHR 14 2.60 SI 2 2.62 PSL 14 1.57 MS 28 4.12 FW 14 1.08 i

NOTES: (1) Minimum allowable Stress = 3Sm = 48.6 ksi (2) Nominal stresses given are incremental stresses due to changes

. in the hot leg or steam generator thermal expansion movements.

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5.0 CONCLUSION

S AND RECOPMENDATIONS The results of the preliminary evaluation indicate that the assumed snubber lock-up does not , result in conditions which compromise the structural integrity of the RCL. The evaluation conservatively assumes that the snubbers are completely frozen during the heatup and cooldown cycle. The snubbers are designed to move under a static load and since heatup and cooldown occur at a slow rate, it is probable that the snubbers were not completely frozen as stated in Reference 8. Also, since the snubbers have been refurbished, the existing condition is comparable to the as-designed configuration. Based on this preliminary evaluation, the RCL is considered structurally adequate under the assumed steam generator snubber lock-up loads although not meeting Code allowables. Its existing (snubbers operable) condition meets Code allowables.

In addition to the preliminary evaluation, Impell reconnends that a supplementary inspection be performed to determine if any distress occurred in the system. Impell's recommended inspections are listed in Appendix B. Based on the results of the preliminary evaluations, Impell reconnends that consideration be given to replacing a number of the steam generator capscrews for subsequent examinations.

Impell further celieves that more rigorous evaluation should be performed on the RCL. The additional 4 analysis should include a model of the entire RCL and i should include nonlinear piping effects to accurately determine the support loads.

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6.0 REFERENCES

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1. Impell Report No. 01-0300-1456, Revision 0, October 1985; " Evaluation of Pressurizer Surge Line Movements for the Trojan Nuclear Plant."

2. ANSI B31.7-1969 edition through 1971 addenda.

3. ASME B&PV Code,Section III, 1980 edition.

4. Welding Reseach Council Bulletin 107, March 1979 edition.

5. ORNL/NUREG-24, " Experimental Study of Plastic

( Responses of Pipe Elbows", by W. L. Greenstreet.

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6. Impell Calculation No. RCL-01, Revision 0; " Steam l Generator Snubber Parametric Evaluation."

l 7. Impell Report No. 01-0300-1395, Revision 0, June, ji 1985; " Leak-Before-Break Evaluation of the Reactor Coolant Loop."

f 8. Paul Monroe Letter to PGE, dated December 12, 1985; re " Conclusions regarding the operability of the Trojan steam generator hydraulic snubbers".

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APPENDIX A: Meeting Minutes for PGE/Impell Discussion on RCL Thermal Expansion Analysis and Surge

Line Evaluation (PGELetterRLS-1411-85 g

dated November 22.1985)

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November 22, 1985 ELS-1411-85

.g Mr. Elemune L. Herman, Project Manager

'g Impell Corporation 350 Lennon Lane Walnut Creek CA 94598 Attention Walter Bak/ William McLeod

-L Gentismen:

Subject:

TEOJAE NUCLEAR PLAET Transmittal of Meeting Minutes Pressuriser Surge Line Evaluation Attached are the sPeeting minutes for the PGE/Impell discussion on the Reactor Coolant Hot Les Analysis and surge Line Evaluation. This meeting was held at PGE on November 19, 1985.

'I If you have any questions / consents about the contents of the meeting j

minutes, please contact Jeff Wheeler at (503) 220-3047.

sincerely,

. AA_

R. L. Steele, Manager Nuclear Plant Engineering Department Buclear Division ELS /amh  !

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Attachment c: W. S. Orser A. Olmstead I C. P. Yundt W. L. Peregoy R. E. Fowler A. S. Cohlmeyer R. J. Wehage T. M. Mitts J. W. Lentsch T. E. Bushnell 1

l I 121 SW Samon Street. Portat. Oregon 97204 l

MIETING MINUTES Date: November 19, 1985

Subject:

Meeting Minutes for PCE/IMPELL Discussion on RCL Thomal Expansion Analysis and Surge Line Evaluation Attendees: PCE: C. P. Yundt IMPELL: Bob Grubb

R. E. Fowler Klesse Herman l

R. J. Wehage Bill McLeod i J. J. Wheeler Walter Bak i J. L. Lentsch I

U R. L. Steele 1- A. Olmstead W. L. Peregoy l l A. S. Cohlmeyer i i

g Rodger Wehage opened the meeting by describing the IMPELL evaluation of the surge line and his request for the RCL thermal expansion evaluation.

Walter Bak described the background of IMPELL involvement in the surge line and RCL thermal evaluation. He indicated that, based on the surge

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line work and PCE indications of degraded steam generator snubbers, the potential exists for thermal stress problems on the RCL.

i Rodger Wehage indicated that Paul Monroe performed testing on two snubbers and recorded no snubber movement under a 100-K load. They further indicated that the snubbers would operate under thermal expansion load. They stated that there is an orificed relief passage in the snubber that would allow for slow motion of the snubbers. Although Paul-Monroe indicated that foreign material was present in the hydraulic fluid, it was their conclusion that the bypass orifice would not have become completely clogged to the point where the snubber would be

" frozen". l Walter Bak described the IMPELL study on the RCL thermal expansion loads. A susmary of the description is attached as Enclosure A. He indicated that the " worst-case" scenario produced an overstress on the thermal expansion stress check of approximately 33 percent. However, he stated that this overstress should not constitute a real problem due to  ;

the displacement-limited loading. A plastic analyses should provide sufficient justification to ensure structural adequacy. He presented IMPELL's recommended interim solution (See Page 6 of Enclosure A).

PCE asked IMPELL to identify the tasks required to provide the interim and final justification to close the issue. IMPELL was also requested to provide a cost and schedule estimate.

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'" After a brief recess IMPELL presented a discu u lon of the 35-3200 criteria that would be used to perfom the interim and final analyses.

Jack Lentsch asked about the applicability of 5B-3200 (AINE Section III) to the 531.7 and Trojan FgAE elastic analysis requirements. This item will be investigated after the meeting.

IMPELL provided the following scope of work and schedule for resolving ll this issue.

Task 1: Finalize Preliminary Analysis 11/27 Task 2: Perform Plastic Component Analysis 11/27 Task 3: Perfom EPV/SC Bozzle Fatigue Evaluation 11/27 Task 4: Perfom Preliminary BCL support Evaluation 11/27 ,

I i Task 5: Perfom Fracture Mechanics Analysis of Highest-Stressed Hot-Les/ Steam ,

Generator Elbow 11/27  !

Task 6: Perform Final Support Evaluation Later i

Task 7

Perform Final Theriaal Analysis and C1.1 code compliance Later i Task S: Perform Elastic System Analysis Later IMPELL indicated that Tasks 1 through 4 were necessary to confim, as

preliminary work indicates, that an operability problem does not exist.

I Task 5 was deemed prudent based on levels of stress indicated by preliminary evaluation. IMPELL proposed to complete these tasks for an estimated $40,000.

l IMPELL recommended that Tasks 6 through 8 be performed following l

verification of as-built conditions, and suggested that nondestructive aw==ination be perfomed for the high-stressed regions during the next outage.

' i PCE requested IMPELL to proceed with Tasks 1 through 5, and the meeting was adjourned.

FOLLOW-OW DISCUSSION lg Later in the day, the question was raised regarding modifications of pipe whip restraints on the pressuriser surge line (primarily PWE 1.4).

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IMPILL recommended that no modifications be made to any Pipe whip

! restraints until Task 7 is complete. At that time, recommendations will be made regarding this issue.

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INCLOSURE "A" l Preliminary BCL Thorinal 1 Evaluation

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PGE Meeting November 19. 1985 l

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s November 19, 1985 Page 1 of 6 i

Background .

• 3eview of surge line observed displacements (May 1985).

• Surge line evaluation perforimod for following cases:

- stratified thermal layer.

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- cold spring. <

, - RCL rotation.

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• Surge line report issued October 23, 1985 with the following conclusions:

- RCL rotation plus cold spring provided "best" fit to observed surge line movement.

p l - surge line stresses are within code allowables.

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IMPELL discussions with PCE during November indicated that two steam-i generator snubbers were on Loop D.

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• IMPELL informed PCE that " degraded" snubbers could have produced surge l j g. line movement and could impact RCL therinal analysis. i PGE requested IMPELL to perform preliminary analysis.

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4 November 19, 1985 Page 2 of 6 Preliminary Hot-Lem Evaluation - Loadina Condit h

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• Thermal expansion load with RCL temperature at 650*F (ambient temperature assumed 70*F).

• Thermal expansion cases considered:

1. As-designed configuration.

2. Snubbers assumed " frozen"; hot-les rupture restraint inactive.

3. Snubbers " frozen"; hot-leg rupture restraint active.

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November 19. 1985 Page 3 of 6 r

Preliminary Hot-Lee Evaluation - Modelina 36 Soubkh b

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• Assumptions.

- SC support flexibility considered.

- 1/16-in. gap assumed at rupture restraint.

- No movement of RpV considered (loop acts independently).

- Actual elbow dimensions used for stress analysis.

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{ o t- i Bovember 19, 1985 Page 4 of 6 Preliminary Hot-Les Evaluation - Criteria

• Hot-leg piping is subject to rules of ASME Code 1.Section III, E5-3600.

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• Themal expansion stresses and loadings affect two criteria:

1. Thermal expansion stress limit.

2. Patigue usage factor.

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• Thermal expansion stress limit is given by Equation 12:

5 12 = C2E/Z < 3.0 Sm where C2 = Secondary stress indice M = Thermal expansion moment range Z = section modulus Sa = Stress intensity allowable g

• Usage factor is determined by considering all stresses (pressure, g thermal transient, seismic, thermal expansion).

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Preliminary Results - Pipina Stresses I

i i Expansion Stress Allowable Location Load Case C2 M/Z (ksi) Stress (ksi)

J EPV Nozzle Wald 1 9.0 53.4

'u 2 36.9 53.4 3 1.9 53.4

. At Hot-Leg Rupture 1 1.5 53.4 Bestraint 2 25.2 53.4 l 3 15.2* 53.4 l SG Elbow / Nozzle 3 6.1 53.4 i Weld 2 45.4 53.4 I i' 3 71.0 53.4 i

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• Does not include local stresses due to rupture restraint contact.

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November 19, 1985 Page 5 of 6 Preliminary Results - Vessel Stresses at Bozzles Local Membrane Allowable Location Load Case Stress (ksi)* Stress (ksi)

EPV Nozzle 2 14.0 m22.0 SG Nozzle 3 16.9 =22.0 l

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• Includes thermal expansion loads only.

Preliminary Results - Support Loads Location Load Case Load (kios)

'i SG Snubbers 1 0 l 2 914 3 1244 .

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Hot-L*5 1 0 Rupture Restraint 2 0 3 -2597 SG Support 1 -38/85 2 1401/-1110 3 1806/-594 a

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Wovember it. 1985

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Fue of 6 Re h ination of Code Criteris

• Fotential failure modes due to thermal expansion loads. , r i .

- fatigue I

! - elastic follow-up

• Thermal expansion loads are self-limiting, and actual displacements g

and rotations are limited.

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• Piping system satisfies code limits in current configuration (snubbers l not " frozen"; no hot-les restraint contact).

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• Fatigue usage satisfies code limits under worst-case loading.

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• Displacement and rotation at elbows will not produce failure under worst-case loads based on examination of test data.

• Frat wre analysis should be performed for worst-case loads since stresses exceeded code allowable.

• Review main steam and feedwater piping.

• Review fatigue at vessel nozzles due to high secondary stresses.

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b Trojan Nuclear Plant Mr. Steven A. Varga Docket 50-344 May 9, 1986 License NPF-1 Attachment C

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l SUPPLEMENTAL INSPECTIONS AND CORRECTIVE ACTIONS The following inspections and corrective actions are being taken to verify the affected piping systems and components are in the as-designed

-configuration. For those actions already completed, the results are also included.

1. RCS Hot Les Pipe Whip Restraints Inspection - The pipe whip restraints on each of the four RCS hot less will be inspected to verify the integrity of the horizontal support members. The clearance between the pipe and the whip I restraint will be measured and adjusted as necessary during both cold

• and hot conditions.

I I Results - Under cold conditions, the pipe whip restraint on the B RCS

hot leg was found in contact with the pipe. Several of the graphite i shims were crushed or broken. This condition is under evaluation and action is being taken to restore the specified clearance between the pipe and the whip restraint. The pipe whip restraint on the C RCS hot leg was inspected and a single point of contact was observed between the pipe and the whip restraint. Action is being taken to restore the required gap in this single location. Inspections of the A and D loop pipe whip restraints have not yet been completed.

2. Steam Generator Supports and Restraints Inspection - The vertical (column) supports on all four steam generators will be inspected to ensure no plastic deformation has occurred in base plate anchor bolts or support pad cap screws.

The specified gap between the steam generator seismic support ring

, girder and the bumper pads will be measured and adjusted as necessary i

under cold conditions. The movement of the steam generators, as a result of heatup, will be determined in horizontal and vertical 1 directions.

Results - The inspection of the base plate anchor bolt and cap screw have been completed on one column support for the B and C steam generators, and no degradation was found.

J l 3. Pressurizer Surge Line Pipe Hangers and Whip Restraints 4

j Inspection - For both cold and hot conditions, the pipe whip restraints and pipe hangers on the pressurizer surge line will be inspected to verify proper gaps and loads.

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Trojan Nuclear Plant Mr. Steven'A. Varga Docket 50-344 May 9, 1986 License NPF-1 Attachment C Page 2 of 2 Results - The inspection of the pressurizer surge line pipe whip restraint and hangers have been co2pleted in a cold configuration and reveal there has been no abnormal movement of the surge line as observed during previous years.

4. RCS Reactor Coolant Pumps (RCP)

Inspection - A visual inspection of potential interferences and RCP supports will be conducted.

5. RCS Crossover Pipe Whip Restraints 7 Inspection - The gaps between the RCS crossover pipe and its associ-ated whip restraints (two per loop) will be measured and adjusted as

, necessary under both cold and hot conditions.

6. RTD Bypass Manifold Piping Inspection - Visual inspections of the RTD bypass manifold piping and 4

supports will be conducted in cold conditions.

7. Main Steam Line Pipe Whip Restraints t

i Inspection - The gaps between the main steam line pipe whip restraints snd the main steam line at the exit of the steam generators will be measured and adjusted as necessary in a cold condition.

8. RCS Hot Les to Steam Generator Elbow Weld

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Inspection - A liquid penetrant nondestructive examination of the subject weld will be performed.

4 l Results - The liquid penetrant examination has been performed on the '

. B hot leg and no indications were discovered.

As an overall step to verify the conclusions and corrective actions, Bechtel Power Corporation has been retained to do an independent review of the RCS Thermal Expansion Analysis perforund by Impell Corporation and

, provide an assessment of the overall program. Westinghouse Electric j Corporation has also been asked to provide an independent overview of the program.

j The above actions will be completed prior to startup from the 1986 refueling outage.

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