ML20245E108
| ML20245E108 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 05/31/1988 |
| From: | Roller A PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20245E106 | List: |
| References | |
| NUDOCS 8812210343 | |
| Download: ML20245E108 (128) | |
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PIPE WHIP RESTRAINT CAP 3 PRESSUKlZER SURGE LINE MOVEMENT MAY 1988 i
Trojan Nuclear Plant Wuclear Plant Engineering Department Nuclear Division Portland General Electric Company 121 SW Salmon Street Portland Oregon 97204 Approved:
A. N. Roller Manager l
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.' l TABLE OF CONTENTS J Section Title gggg
1.0 INTRODUCTION
1 2.0
SUMMARY
2 2.1 Pipe Whip Restraint Caps 2 2.2 Pressurizer Surgw Line Movement 2 x
3.0 FIPE WHIP RESTRAINT CAPS 4 3.1 Background 4 3.2 Observations 4 3.3 Corrective Action 5 3.3.1 Piping System Evaluation 5 3.3.1.1 Cap Adequacy Determination 5 3.3.1.2 As-Found Interference Evaluation' S 3.3.1.3 Reshimming 6 3.3.1.4 Structural Integrity Analysis 6 3.3.2 Pipe Whip Restraint Evaluation 7 3.3.2.1 Existing Conditions 7 3.3.2.2 New caps 7 3.3.3 Inspection Program 8 3.4 Root Cause 8 4.0 PRESSURIZER SURGE LINE MOVEMENT 9 4.1 Background 9 4.1.1 Function 9 4 .1 -. 2 Design Basis 9 ,
4.1.3 Materials 10 !
4.1.4 Historical Background 10 i
4 Section Title Pate 4.2 Current Observations 11 4.3 Root Cause Evaluation 12 4.4 Corrective Actions 15 j.4.1 Inspections / Surveys 15 4.4.2 Piping Integrity Evaluation 16 4.4.3 Startup Monitoring Program 17 5.0 TABLES 5.1 Pipe Whip Restraints Inside Containment !
5.2 WDE Exams 5.3 Pressurizer Surge Line Movement' History 6.0 FIGURES 6.1 Pipe Whip Restraint Gap Evaluation Flowchart 6.2 Typical Pipe Whip Restraint 6.3 Isometric Drawing of Pressurizer Surge Line 6.4 Stratified Flow Profile 6.5 Pressurizer Surge Line Monitoring Instru-mentation Locations APPENDICES A. Quality Assurance Prosram B. Description of Pressurizer Surge Line Piping Integrity Analyses l
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1.0 INTRODUCTION
PUP. POSE
- i. This report documents the Portland General Electric (PCE) investigation and evaluation of (1) the inadequate gaps identified on the whip o
restraints (WR) for the non-reactor coolant loop (RCL) safety-related
. piping and (2)- the ' unexpected deflection of the pressurizer surge line at the Trojan Nuclear Plant. The report identifies the causes, corrective actions and future monitoring programs for these conditions.
BACKGROUND The Large-Bore Pipe Support Design Verification Program (LBPSDVP),
developed after the'1987 Refueling Outage, included an evaluation of the adequacy of the WR designs and their rock. bolt anchorages. As a result of this evaluation, thermal and seismic piping deflections and WR gaps were jalculated to'be different'from those used in the original design.
Field' measurements of WR gaps were obtained in order to evaluate the as-built conditions with the new calculations.
During the fleid nessurements of WR gaps, it was discovered that the pressurizer surge line WRs had gaps different from those previously measured and that WR 1.2 was in contact with the piping. The pressurizer surge line seps have been monitored and evaluated during refueling outages since 1983 when evidence of piping movement had been observed.
The Nuclear Regulatory Commission's (NRC) Safety Evaluation Report (SER)
J dated June 16, 1986 summarizes the efforts related to the pressurizer surge line movements to that time.
ThisPCEreportdealswithtwodistinck,butrelated, concerns:
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-a. Non-RCL pipe WR gaps and the dif ferences between current field ;f measured, latest. calculated, and original design values.
- b. The continued unexpected deflection of the pressurizer surge line.
1 2.0
SUMMARY
l 2.1 PIPE WHIP RESTRAINT CAPS (excluding the pressurizer surge line)
Using the data for pipe deflections from the latest calculations and (
the WR field measurements, it w&s determined that the piping umy '~'
contact the.WR on 55 of 134 WRs. There are a total of 134 WRs; the gaps on the remaining 79 WRs were adequate to preclude contact.
Evaluations will be completed prior to plant heat-up to demonstrate that the.affected piping system's design is in accordance with the Trojan Final Safety Analysis Report (FSAR). In cases where the gaps
'c Are .inadequat a, reshimming will be performed before plant heat-up.
An inspection program has been established to verify adequacy of the design approach for evaluating and setting WR gaps.
2.2 PRESSURIZER SURGE LINE MOVEMENT The dispideement of the pressurizer surge line and ultimate cratact with the WRs has been investigated. It is concluded that the most probable cause of pressurizer surge line displacement is thermal stratification of water within the surge line.
The gaps on the pressurizer surge line WRs will be set to the maximum allowable prior to heat-up to minimize their interferences with the surge line.
Conservative analyses have demonstrated the surge line integrity in accordance with the FSAR requirements for a minimum of eig,ht e.ddi'tional heatup and cooldown cycles. For reference, there have been 32 heatup and cooldown cycles since initial plant startup in 1975.
A monitoring program will be implemented to obtain temperature and displacement data during start-up from the current refueling outage.
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-The results~of the monitoring program will be.used to verify the assumptions'made in the piping integrity and thermal stratification analyses. and to provide data for further evaluations of the pressurizer surge line and surge line WRs for the full life of the-'
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3.0 PIPE WHIP RESTRAINT CAPS This section discusses the WR gaps on lines other than the RCL or the pressurizer surge line. The evaluation and necessary reshinsning of reactor coolant loop WRs were completed in 1986. The WRs on the pressurizer surge line.are discussed in Section 4.0.
.l This section describes the calculated and measured gaps between p* ping -
and pipe WRs and the corrective actions taken for those WRs where the gaps were inadequate. Also described are the evaluations performed to demonstrate piping system and pipe whip restraint design adequacy.
3.1 BACKGROUND
There are 134 WRs on non-RCLs. The original design (1974-75) of WRs was based on thermal movements from preliminary piping stress analyses performed by the Architect Engineer (A-E). After WR installation, gaps were measured in both the cold and hot conditions (preoperational hot functional testing) without shims in place.
Following these measurements, the shims were installed with the plant in a cold condition. Movements as a result of seismic and thermal displacements (for other than cold and hot standby conditions) were not considered in determining the shim size. The gap sizes were never reconciled with the final piping analysis. As part of the LBPSDVP, new piping stress analyses were performed. The new pipe thermal and seismic movements are being used to establish the required gaps in the WRs.
3.2 OBSERVATIONS From an initial comparison of new pipe deflections from the latest piping analysis with the existing WR gaps (as specified on drawings), 65 of 134 WRs were found to be in contact with the piping during thermal or seismic conditions. The remaining 69 were determined to have adequate gaps. As a result, field measurements of WR gaps were performed during the 1988 Refueling Outage. Piping
was found to be in contact with eight WRs in the cold condition.
Also, based on these field-measured gaps, it was determined analytically that piping could experience interference with existing Whip restraints under themal and/or seismic conditions at 47 WRs.
Of these 47, 25 were calculated to have contact for themal movement, whereas, 22 were calculated to have contact for seismic movement only. Table 5.1 is a complete listing of all non-RCL WRs (including WRs on the pressurizer surgo line) and identifies those WRs which were in cold contact or would have been in contact with the piping under thermal or seismic conditions.
3.3 CORRECTIVE ACTION 3.3.1 Piping System Evaluation 3.3.1.1 Cap Adequacy Determination Pipe movements for normal, upset, and faulted conditions were determined in the current stress analysis calculations. Field gip measurements were obtained for all WRs and piping interferences in all conditions were detemined. The complete evaluation process is illustrated in Figure 6.1. A typical WR is shown in Figure 6.2.
3.3.1.2 As-Found Interference Evaluation Piping (and pipe supports) in cold contact or calculated to be in contact due to thermal or seismic conditions is being evaluated for the effects of restricted movement at WRs under all conditions.
The magnitude of piping interference in the cold condition was determined by removing the shims and noting the pipe movement. Piping stresses and pipe support loads were determined and evaluated for the f
L effects of cold contact and thermal contact. piping ,
with stresses in excess of code allowables is being I further evaluated by Nondestructive Examination (NDE). The NDE performed on the piping for each of 55 WRs in this category is shown in Table 5.2. The WDE inspections will be completed prior to heat-up.
The integrity of the piping system will be confirmed to be acceptable prior to heatup.
3.3.1.3 Reshimming The WR " gaps will be' adjusted by modifying shim dimensions to remove all interferences of piping-with WRs in cold and hot conditions except Where analyses demonstrate that code requirements are met for the piping system. The gaps will be increased to the maximum permitted by the WR structural design. Also, seismic movement interference will be removed except Where the piping system is shown to be acceptable. These shim modifications will be completed prior to heat-up.
3.3.1.4 Structural Integrity Analysis A structural integrity review will be performed for the increased loads on the pipe supports. Visual inspection of the pipe supports will be performed where required to ensure that no damage had occurred. This effort is ongoing and will be completed prior to plant heat-up. The integrity of the piping system will be confirmed in accordance with the FSAR. Assessment of Whether the piping systems were operable during previous operating cycles will be completed after restart.
+g l 3.3.2 pipe Whip Restraint (WR) Evaluation 1'
3.3,2.1 Existing Conditions The effects of a pipe in contact with a WR under cold, hot, or seismic conditions were determined to have an insignificant effect on the integrity of the WR and are bounded by the design loads resulting from a pipe break.
3.3.2.2 New Caps Reshimming will be completed to accommodate piping movement for all normal, upset, and faulted conditions. The new gaps resulting from reshimmir.g will be_ compared with the design gaps used in the WR design esiculations.
The assessment of the WR's adequacy in cases Where the as-left gap is larger tha.4 the gapc used in the WR design will be completed prior to heat-up. Any modifications identified to be necessary as part of this assessment will be performed before heat-up.
WRs Which were neither found to be in cold contact nor determined to be in contact with piping under thermal or seismic conditions, were not reshimmed.
The structural integrity of these WRs for which the as-left gap is larger than the gap used in the WR
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design will be evaluated after restart and any necessary modifications performed in 1989.
3.3.3 Inspection program An inspection program will be implemented to ensure that adequate WR gaps exist. During the 1988 Refueling Outage, gaps in cold conditions have been verified for all WRs. For selected WRs (where practical, considering radiation exposure '
and personnel safety), the gaps will be visually inspected at hot standby (on some systems, the inspections will be perfomed prior to hot standby if calculations show that a more restrictive condition exists prior to achieving hot standby). This inspection will confirm the adequacy of the '
design approach used to determine the WR gaps.
During the next outage, selected WR gaps will be measured in the cold condition.
3.4 ROOT CAUSES The root causes of inadequate gaps for 55 of 134 pipe WRs were det, ermined to be due to a cor.bination of the following:
- a. Failure to reconcile the A-E determined WR gaps with the final piping analyses.
- b. Failure to provide allowance for seismic induced movements when determining gap size.
- c. Incorrect gaps set during construction.
- 6. Cap change due to themal shakedown of piping system.
The gap settings for WRs had not been inspected or monitored since 1975.
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4.0 PRESSURI7ER SURCE LINE MOVEMENT y
4.1 BACKGROUND
This section describes the design basis of the pressurizer surge line, the history of observed movements of the pressurizer surge line since 1982, the cause of the observed movements, and the analyses performed to support'the cause of the movement and structural integrity of the line. This section also describes the monitoring program to be implemented.
4.1.1 Function The pressurizer surge line is designed as a flow path between the Reactor Coolant Loop (RCL) (hot leg) and the pressurizer for pressure control of the Reactor Coolant System (RCS).
4.1.2 Design Basis Design loads include pressure (2,485 psig), temperature (680*F), seismic, and design basis accident (pipe break) conditions. The American National Standards Institute (ANSI) B31.7 (1969) piping code was the original design and analysis code of record which included requirements for a detailed fatigue analysis. The American Society of Mechanical Engineers (ASME) Code,Section III (1977 Edition through 1979 Winter Addenda) has been used for analysis purposes and has been rsconciled with ANSI B31.7 analysis and design requirements. In developing loads for this analysis.
it was assumed that the line and attached notries would experience a thermal transient from every plant event which resulted in an in-surge or out-surge of flow between the RCS and pressurizer. Thus, many cycles were assumed to occur 1
4 with differential temperature loadings ranging from about 300*F (during plant heat-up) to about 50*F during normal power operation. In the original design analysis, the in-surges or out-surges were assumed to " sweep" the fluid in the line, resulting in uniform thermal loadings at any particular location. At that time, stratification of flow in the pressurizer surge line was not considered in the design.
The pressurizer surge line is a 14-inch Schedule 160 stainless steel pipe. The piping layout, ethich is typical of most Westinghouse "ressuriter Water ReactLr designs, is basically a horizontal run with several pipe bends (for thermal expansion) from the hot les to jist below the pressurizer centerline, where the line r16ts vertically about 7 feet 7 inches to the pressurizer nozzle. The horizontal run is sloped upwcrd to tha pressurizer with an angle of about 0.6 degrees, which res.dts in an 8.3-inch vertical increase over 63 feet. This overall layout is depicted in Figure 6.3. The line is insulated with 4-inch-thick stainless steel encapsulated mirror-type insulation.
4.1.3 Materials The following materials were used in the pressurizer surge line design:
Surge line piping: SA 376 Tp316 Surge line nozzle (RCS): SA 182 F316 Pressurizer nozzle: SA 508 Class 2 Pressurizer nozzle safe end: Sk 182 CD F316L 4 1.4 Historical Background Since 1982 PCE began monitoring the pressurizer surge line in 1982 following removal of the pressurizer surge line thermal
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sleeve from the reactor coolant loop nozzle. Themal sleeve removal was required because the sleeve welds had failed.
The MSSS Vendor demonstrated by a revised fatigue analysis that the sleeve was not necessary to protect the reactor coolant loop nozzle from themal transient effects.
To remove the thermal sleeve, the surge line was cut at two locations: adjacent to the reactor coolant loop at the nozzle-to-surge line weld and at a second. point several feet away toward the pressurizer beyond a 45-degree band. ,once thic section of piping had been removed, the thermal sleeve was extracted and the line welded back in place. pCE was advised at that time that some amount of thermal shakedown (ie, permanent deformation of the pipe as a result of relaxation of internal stresses) could be expected, and therefore, it was necassary for the line to be observed and, if needed, adjustments made at the hangers and at the pipe WRs.
The surge line WRs were monitored over the next six outages.
Contact was noted between the surge line and some of its WRs each year through the 1986 Refuellog Outage. For each contact, a root cause was postulated and an evaluation of the piping system was performed. In all cases, the piping system and its restraints were shown to satisfy the design limits.
A description of the surge line and WR history from 1982 to the present is provided in Table 5.3.
4.2 CURRENT OBSERVATIONS (MAY 1988) e When the plant was shut down in 1988, a general inspection of pipe WRs was performed as part of the Large-Bore pipe Support Design
Y' Verification program. The pressurizer surge line WRs 'were included 1 I
within this program. Contact was observed on pressurizer surge ~line l
.WR 1.2. Measurement revealed an uplift force of 4,771 pounds, and following removal of the contacting shim, the pipe moved upward an additional 3/8 inch.
4.3 ROOT CAUSE EVALUATION There were seversi proposed explanations provided as to the cause of the surge line movement and contact with the WRs. These possible explanations include: ,
Thermal shakedown following the removal of the thermal sleeve in 1982.
Design and construction errors of the surge line whip restraints.
Abnormal movement of the reactor coolant loop or the pressurizer.
Thermal stratification of water in the surge line.
Each of these proposed causes was reviewed and conclusions on their credibility as reasons for surge line movement are discussed below.
Thermal Shakedown
. Removal of the thermal sleeve on the RCL nozzle required removal of a saction of piping on the surge line. When the surge line was cut, cold spring in the line occurred and the piping moved outward. When the piping was rewelded, a cold pull of approximately 7,000 lbs was required to bring the surge line into alignment with the nozzle.
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i some amount of shakedown was' expected for one or two cycles after welding. The thermal sleeve was removed in 1982 and six heatup/cooldown cycles have occurred since that time. Since the cold spring shakedown would have occurred in one or two cycles, it was not considered to be a factor during,this outage.
Desian and Construction Errors
'The pressurizer surge line WRs have been monitored regularly over the past six years, and the gaps have been reviewed. The gaps were reset consistent with the original design criteria each time contact was found or when they were deterzined to be inadequate. The design settings of the gaps are correct and are not a cause of surge line movement. There have been no indication that the shims for the gaps were inappropriately installed.
Abnormal Movement of the Reactor Coolant Loop or the Pressurizer i
The RCL motion which was observed in 1986 could have resulted in unexpected movements at the RCL surge line nozzle. This movement would have caused unexpecteo deflection of the surge line. The correct gap settings on the steam generator and hot leg whip restraint were established and the results of the monitoring program demonstrated that the RCL moved as expected and is therefore not a cause of the unexplained surge line motien. The pressurizer is designed to expand in the vertical direction only. Inspections of the pressurizer during the 1988 Refueling Outage indicate it is moving as designed and would not be a cause of the unexpected surge line movement. ;
Thermal Stratification of Water in the sures Line Thermal stratification has previously F n identified in the Trojan surge line. However, during evaltetion of this effect, the surge line motion was not explained. A ICJ'F differential temperature (which was assumed at that time) across the line was not adequate to
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explain the movement. Also, the RCL restrained thermal growth was identified during this time (see above), and further evaluation of stratified flow was discontinued.
Industry experience since 1987 has indicated that sign'ficant .
thermal stratification in the pressurizer surge line as poss*ble.
preliminary thermal-hydraulic calculations confirm this for typical flow rates in the surge line. piping stress analysis modeling stratified flow has been performed which shows significantly more deflection than obtained from analysis assuming uniform temperature. This deflection increases with increasing differential temperature between the top and bottom of the pipe. The evaluation indicates that the line under stratified conditions would deflect downward, contact WRs 1.2 and 1.4, and undergo plastic deformat' ion ~~
c b ould result in the.cpid set of the pipe above its original 1,r;c a tion. This agrees with the observed vertical set of the'line at these locations in the cold condition.
Operating conditions which. produce stratification occur during
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heat-up, cooldown, and steady-state operation of the plant. The 1985 assessment of thermal stratification had focused on hot standby and power operation during which there was a lower temperature difference because these were considered to be fla limiting conditions at that time. Current efforts have focused on plant conditions during heat-up when the temperature differences between the RCL and pressurizer are larger.
During a typical plant heat-up, water in the pressurizer is heated
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to a teinperature of approximately 440*F, thermal expansion of the water occurs, and a bubb1_e_ is_ formed,in the pressurizer. Loop temperature is gradually increasing. As the water flows f rom the
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pressurizer to the loop (out-surge), the hotter water rides on a layer of cooler water, causing the upper part of the pipe to be heated to a higher temperature than the lower part of the pipe.
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T'his condition is shown in Figure 6.4. The differential temperature could be as high as 300*F, based upon plant operating limitations.
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. i Under these conditions, analysis has shown that differential thermal expansion of the pipe metal causes the pipe to bow, resulting in either upward or downward deflection at any point on the line, depending upon the stratification distribution and piping configuration. A description of the analysis to verify the root cause is provided in Appendix B, Section B.2.
4.4 CORRECTIVE ACTION 4.4.1 Inspection and Surveys During the 1988 Refueling Outage, the following inspections and nondestructive examinations (NDEs) were performed on the pressurizer surge line:
+ Ultrasonic (UT) and dye penetrant (PT) examinations of all
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pressurizer surge line circumferential welds from the RCS nozzle-to-pipe we16 to the pressurizer nozzle-to-pipe weld. Table 5.3 describes the UT and PT examinations.
+ All pressurizer surge line WR gaps were measured. Contact was observed at WR 1.2.
+ A visual inspection revealed no discernable distrass in the piping line.
- The pressurizer surge line WRs and hangers were visually 7 inspected and no structural damage was found.
+ The Reactor Coolant Loop Thermal Expansion Program results were reviewed to confirm proper movement of the RCL. In oddition, visual inspections were performed on the steam generator seismic restraints and upper support rings. No a!.tormal conditions were found.
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. The pressurizer anchor bolts and'oeismic supports ~were
' checked. No abnormal conditions were found.
4 PCs contacted other utilities with.pressurl.:ed water reactors :
(PWRs). Several utilities confirmed the presence of thermal stratification in the pr vssurizer surge line and/or contact with restraints or unusual' surge line motion.
4.4.2 piping Integrity.gvaluation The root'cause analysis identifies thermal stratification as the cause of the surge line movement. ~The analysis predicts downward motion of the surge' lice due to stratified flow' conditions resulting-in interference with the pipe WRs. The restrained thermal growth of the pipe causes a yielding in
, some sections of piping Which produces a permenent upward set in the piping when cooldown occurs, which is consistent with the observed pipe movements.
The.following analyses were performed to demonstrate.the integrity of the surge line under the stratified flow conditions and existing WR gaps:
. Elkstic piping analysis of the surge-line to evaluate 1
l thermal expansion stresses in the piping system.
. Fatigue analycis of the piping and nozzles.
. Preliminary leak-before-break assessnent of the surge line piping to demonstrate leak detection and flaw stability under normal and faulted loads.
A detailed discussion of the piping integrity analyses is provided in Appendix B.
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The resultsaof these evaluations concluded that the piping and nozzles satisfy ASME Code Section III limits under. [
i t -thermal stratification and all other desian loadings for g'40plant heat-up/cooldown cycles.
The plant has currently experienced 32 heat-up/cooldown-cycles. '~
During plant startup, a detailed monitoring program (see Section 4.4.3) on the pressurizer surge line will verify the loading assumptions'(of the near-term analysis) and provide input for-final stress analyses and fatigue-life calculations.
The American Society of Mechanical Engineers (AEME) Code qualification of this piping for the past and future design life operation will be revised.
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4.4.3 Monitoring program In order to establish the actual temperature distribution and
'line movements in the pressurizer surge if e, a monitoring program has been developed. This monitoring program is s
neces'ary to characterize the thermal loadings to accurately evaluate the fatigue life and pipe stresses in the pressurizer surge line. . prior to this program, all piping stress ana?,yses have included many conservative assumptions in order to ensure that the uncertainties in the loading are bounded. This approach is overly conservative for the design basis eva?,uation r~ the pressurizer surge line.
The monitoring program will acquire system operational (flow, temperature, and pressure) and displacement data. With this j data, the stratification temperature profiles, flow rates, and line movements will be identified and will be used as input to stress analysis calculations to evaluate pipe stresses and fatigue life, ine locations of resistance temperatures detectors (RTDs) and the linear potentiometers are shown in Figure 6.5.
< N, The actual cold gap' dimensions at all pipe whip restraints will be .obtained prior to plant heat-up.
'The data will be used to ensure that the assumptions
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- AP*ENDIX A OUALITY ASSURANCE PROGRAM To be supplied later.
A-1
1
,- 1 APPENDIX B .
l
. DESCRIPTION OF PRESSURIZER SURGE
! LINE PIPING INTECRITY ANALYSES
/
B.O INTRODUCTION This appendix provides a detailed description of the tasks performed to evaluate the piping integrity of the pressurizer surge line. The sections included in this appendix are:
- Overview of approach
+ Piping analysis to evaluate root cause
+ Piping analysis to evaluate thermal expansion stresses
- Fatigue analysis of RCS and pressurizer nozzles
+ Fatigue analysis of piping components
+ Leak-before-break assessment B.1 OVERVIEW OF APPROACH The purpose of the piping integrity analysis of the surge line was to verify the credibility of the postulated root cause and to verify integrity of the surge line piping.
The steps required to perform the verification included piping analysis and fatigue eva*.uation. A nonlinear piping analysis was ;
performed to justify the rrot cause postulatica of stratified flow effects. An elastic piping analysis was performed to avait ste thermal expansion stresses in the piping. Fatigue analysi s was performed for the surge line piping and the nozzles. Finally, a B-1
4 1
preliminary leak-before-break study was performed to demonstrate the 1
inherent safety margin in the surge line.
B.2 PIPING ANALYSIS TO EVALUATE ROOT CAUSE A nonlinear piping analysis has been used to justify that thermal stratification was a probable cause of the observed surge line conditions. Based on observed data, it has been postulated that the .
I pipe movement produced by thermal stratification is restrained by the whip restraints and resulted in a permanent upward set of the surge line. The nonlinear evaluation correlates these causes and effects.
The piping analysis was performed using a beam model of the piping system, which extends from the RCS hot les nozzle ' o the pressurizer nozzle. The boundary conditions are assumed fixe ( at the terminal ends since the hot les and pressurizer are much s.iffer than the surge line. Thermal anchor motions are applied at these fixed points to model the effects of RCS and pressurizer movement. The nonlinear analysis was performed using the ANSYS (Reference 4) computer program with the following elements:
+ STIF 20 - Elastic-plastic straight pipe elements
+ STIF 60 - Elastic-plastic curved pipe elements
- STIF 39 - Nonlinear gap elements i
The material stress-strain proper,tles are input as bilinear curves 3
in the nonlinear analysis. Curves were developed at 70* and 617'F.
The initial nonlinear analysis co~nsisted of evaluating thermal stratification in the pipe using a temperature differential across the pipe section of 300*F. The temperature differential was applied as a linear gradient across the pipe section (420*F at top to 120*F on bottom) for the surge line horizontal length. The piping system
, B-2 l
was initially assumed to be unrestrained. The results of the j analysis showed that the surge line displacements are much larger i than those predicted by the original design analysis which j incorporates uniform temperature thermal expansion. Therefore, large interferences would exist at several pipe whip' restraints.
Pipe whip restraint gaps were included in the nonlinear analysis, and the surge line was evaluated for several complete heat-up and a cooldown cycles. A complete cycle consists of the following loadings:
i i
Lower Upper Load Step No. Pipe Temp Pipe Temp Stratification AT 1 (heat-up) 120*F 120*F 0 2 120*F 320*F 200 3 120*F 345*F 225 4 120*F 370*F 250 5 120*F 395'F 275 6 120*F 420*F 300 7 350*F 450*F 100 8 550*F 600*F 50 9 (100% power) 615*F 615'F 0 10 (cooldown) 100*F 100*F 0 The evaluation also included gravity loading and thermal movements at the RCL and pressurizer nozzles.
The results of the evaluation showed that the surge line movements closed the gaps on several whip restraints. Due to plasticity during the heat-up cycles, the surge line moves above its initial position during cooldown. The general behavior of the surge line in this analysis correlates reasonably well with field observations of the line. Additional cycles were considered which indicate that the surge line retains an upward set in the cold condition.
B-3
,?
9 i e
B.3 PIPING ANALYSIS TO EVALUATE CODE COMPLIANCE Since the nonlinear analysis provides reasonable justification for l the observed motion of the surge line, the worst-case loading conditions were selected and an elastic analysis was perfomed for use in the qualification of the surge line. From the analysis, maximum bending stresses were obtained to verify compliance with the ASME Code (Reference 1) Equation 12 which limits thermal expansien stresses. The stratification effects are included in the load cases evaluated. A review of the surge line motion indicMed that the maximum stresses resulted from restraint of the thermal expansion produced by the 300*F stratification effects. The maximum stress was induced by providing a gap interference equivalent to the original gap interference plus the thickness of the added shims.
The results of the elastic piping analysis indicate that the thermal expansion stresses are less than the expansion stress limit of Code Equation 12. The remaining ASME qualification is described in Sections B.4 and B.S.
B.4 FATIGUE ANALYSIS OF RCS AND PRESSURIZER N0ZZLES This section summarizes the ASME Code (Reference 1) fatigue evaluation of the pressurizer surge line nozzle-to-hot les and nozzle-to-pressurizer. This analysis included consideration of the potential effects of thermal stratification and whip restraint gap closure as well as design transients. The pressurizer nozzle, which is not significantly affected by stratification, was evaluated by load conformance and qualified. The results of the fatigue calculations of the RCS nozzle indicate that ASME Code limits are satisfied for at least 40 heat-up and cooldown cycles, as well as other design transients. The plant has currently completed 32 heat-up and cooldown cycles. Data from the plant heat-up monitoring program will be used to confirm this analysis and to B-4 f
m
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'4 ll l determine accurate stresses and cycles for longer term fatigue calculations.
7 B.4.1~ Thermal Stratification Stresses l
Thermal stratification in the pressurizer surge line causes
- additional stresses at the reactor coolant loop branch nozzle. These stresses result primarily from the interaction between the thermal movements of-the line, and the restraint of that movement by the closure of gaps ac the pipe whip l restraints. Several piping analyses with various stratification profiles and gap closure conditions were used to obtain maximum loadings at the noszle location.
The branch nozzle stresses were obtained from these loadings, using the data provided in the WRC Bulletin 297 (Reference 2) .
to obtain maximum stress components in the nozzle. These stresses were combined with stresses from the thermal ,
transient analysis. The stresses from the reactor coolant pipe loadings (run side) were calculated using ASME Section III methodology and stress indices.
The resultw from these calculations were factored into the usage f act or determination as described in Section B. A.3.
B.4.2 Thermal Transient Stresses The stresses for thermal transient loadings were obtained from finite element analyses of the reactor loop nozzle.
Transient loadings were applied as defined in Westinghouse j System Standard 1.31 (Reference 3). This document provides )
conservative design estimates of maximum possible transient loadings and number of occurrences of these transients. Thir document, which had not been developed at the time of the ;
original Trojan design analysis, is considered to contain a B-5
l
- e. .
more realistic transient description for the RCL surge nortle and, therefore, was used for this analysis.
Two types of nozzle configurations were considered, one with a thermal sleeve, and one without. The thermal sleeve was removed from the surge line nozzle in 1982 due to cracking of the attachment welds. All transients which occurred af ter this time were evaluated for the nozzle without a thermal sleeve. The number of transient cycles was determined based on the number of heat-ups and cooldowns and the actual years of operatic,n assuming design transients occurred at the rate provided in the Westinghouse system Standard 1.3K.
Based on realistic frequencies of occurrente, this approach generally results in a very conservative estimate of the number of transient cycles.
Peak and secondary stresses were combined with the stresses resulting from stratification in the surge line to obtain a total thermal stress. ;
B.A.3 Fatigue Usage Factor Fatigue usage factor was calculated for the RCS nozzle using ASME Code Section III techr.iques and the design fatigue curve for austenitic stainicss steel. To maximize plant operating flexibility, the number of heat-up/cooldown cycles and the stratification temperature difference was maximized until a usage factor just below 1.0 was achieved.
B.5 FATIGUE ANALYSIS OF PIPING COMPONENTS ]
A fatigue analysis of the critical piping components evaluated the potential effects of stratification and standard design transients.
B-6 l
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t j
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1 A bounding analysis was performed .n all girth butt welds by j developing an enveloping stratification stress.
Based on a three-dimensional finite element analysis, a temperature profile was selected which maximized the stratification stresses.
Stresses from other loading (normal thermal expansion, pressure, seismic, and thermal transients) were considered by comparison with generic fatigue calculations of this line.
The three-dimensional stratification model was constrained to prohibit any free rotation, which simulated the worst case constraints which could'have actually occurred at Trojan from contact with whip restraints. Therefore, enveloping stratification stresses could be obtained without using the stratification piping analysis (Section B.2). j The usage factor was calculated based on ASME Section III methods, including the design fatigue curve. This analysis demonstrated that usage factor limits could be met for approximately 40 heat-up and cooldown cycles as compared to the 32 currently completed.
B.6 LEAK-BEFORE-BREAK ASSESSMENT A preliminary leak-before-break assessment of the surge line pipios has been completed. This assessment demonstrates that a flaw in the surge line would not propagate to leakage in the near term. A l preliminary assessment was completed which demonstrated that a 0.25T (wall thickness flaw) would require approximately 100,000 cycles to
', propagate to a 0.75T-sise flaw which would still be stable under all loadings. The frequency of significant stratification cycles is low enough that many additional years of continued operation cou14 be justified.
B-7
{
2 p-4 L .- 1 In addition, it is judged that leak-before-break esn be demonstrated j with the recommended margins (margin of 16,with respect to leak l detection capability, margin of 2 on flaw size, and margin of 1.4 on loads). q REFERENCES
- 1. ASME Code,Section III, Division 1. Subsection NB.
- 2. WRC Bulletin 297, " Local Stresses In cylindrical Shells Due to External Loadings On Nor.tle - Supplement to WRC Bulletin 107",
Revision 1 September 198.
- 3. Westinghouse System Standard 1.3K, Revision 0, 1979 (Westinghouse Proprietary).
- 4. ANSYS Computer Program, Rev. 4.2.b. Swanson Analysis System. Inc.
. 1 B-8
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wa el TU ELECTRIC e 2' IPO-OO1 A/IPO-OO5A HEATUP/COOLDOWN PROCEDURES .
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MAJOR EVOLUTIONS PERTAINING TO HEATUP INITIAL CONDITIONS o RCS TEMP 130*F. 170*F o RCS PRESS 325 PSIG o PRZR TEMP 130*F. 170*F PROCEDURAL GUIDANCE o START NO I AND.4 REACTOR COOLANT PUMPS INITIATING PLANT HEATUP. (ONLY IF RCS, TEMP 1130*F) o INITIATE PRESSURIZER IIEATUP.
o DRAW PRESSURI:'ER BUBBLE.
o CHEMISTRY PARAMETERS IN SPEC.
o START NO 1 AND 4 REACTOR COOLANT PUMPS TO INITIATE RCS HEATUP.
IF NOT PREVIOUSLY DONE.
o RHR OUT OF SERVICE.
o START NO 2 AND 3 REACTOR COOLANT PUMPS TO ASSIST IN RCS HEATUP.
o INITIATE PRESSURIZER HEATUP AND PRESSURE INCREASE.
o START REMAINING REACTOP, COOLANT IF NOT PREVIOUSELY DONE.
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IPO-001A PLANT HEATUP PROCEDURE INITIAL BUBBLE RCP START PRZR H/U FINAL CONDITIONS 50*F/HR 50*F/HR 50*F/HR CONDITION RCS TEMP BETWEEN BETWEEN HEATUP HEATUP 130-170 130-170 130-300 300-557 TIME DURATION N/A N/A 4 HRS 6 HRS 557 PRESSURIZER BETWEEN HEATUP 130-170 130 to 440 440 440 to 657 TIME DURATION N/A 7 HRS N/A 5 HRS 657 PRESSURIZER PRESSURIZE PRESSURE 325 325 325 325 to 2235
?IME DURATION N/A 7 HRS N/A 5 HRS 2235 DELTA T 0 0 to 310 310 to 140 140 to 100 (PRZR.RCS)
TIME DURATION 7 HRS 4 HRS 6 HRS i
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INITIAL CONDITIONS o RCS TEMP 557'T l
! o RCS PRESS 2235 PSIC o 'PRZR TEMP 657'T PROCEDURAL GUIDANCE o STOP NO. 2 AND 3 REACTOR COOLANT PUMPS TO INITIATE RCS C00LDOW.
o INITIATE RCS DEPRESSURIZATION AND PRESSURIZER C00LD0kL o INITIATE RHR COOLING.
o STCP NO. I AND 4 REACTOR COOLANT PUMPS.
o FILL, C00LDOW AND DEPRESSURIZE PRESSURIZER, o CONTINUE PRESSURIZER C00LDOW.
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- DEPRESSURIZE RCS TEMP COOLDOWN C00LDOWN COOLDOWN COOLDOWN 557 557 to 530 530 to 500 500 te 300 300 to 140 TIME DURATION N/A 1 HR I HR S HRS 4 HRS 140 PRESSURIZER C00LDOWN C00LDOWN C00LDOWN TEMP 657 657 657 to 630 630 to 440 440 to 140 TIME DURATION N/A N/A 1 HR 3 HRS 4 HRS 140 PRESSURIZER 2235 DEPRESSURIZE DEPRESSURIZE DEPRESSURIZE 2235 2235 to 1960 1960 to 350 325 to 50 TIME DURATICt: N/A N/A 1 HR 3 HRS 4 HRS ATMOS.
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1 PRESSURIZER SURGE LINE MONITORING PARAMETERS LOCATIONS / DIRECTIONS I e LOCATIONS D1 THROUdH D4 AND T1 THROUGH T4 (+6")
{
t S LATERAL AND VERTICAL READINGS AT ALL LOOATIONS i
RECORDING INTERVALS .
1 9 EVERY 10 MINUTES DURING HEATUP-O EVERY 60 MINUTES AT EACH TEMPERATURE PLATEAU RECORDING SYSTEM CAPABILITIES i O 10 SECOND INTERVALS FOR A PERIOD OF 10 MINUTES 4 O PROVIDE LABELED PLOTS OF TEMPERATURES AND DISPLACEMENTS AS A FUNCTION OF TIME O AVERAGE ALL TEMPERATURE READINGS AT EACH MEASUREMENT TIME I
J
ENCLOSURE 4
$ % VoJ E 2 9&f-DUQUESNE LIGHT COhfPANY BEAVER VAILEY UNIT 2 PRESSURIZER SURGE LLVE STRATIFICATION S. L MUKHERJEE NUCLEAR ENGINEERING DEPARTMENT
BACKGROUND e ORIGINAL CONFIGURATION OF SURGE LINE (PRE EFT)
TANDEM 6 KIP MECE. SNUBBERS INSTALLED WEIP RESTRAINTS RENOVED PER WHIPJET PROGRAM 2 VARIABLE SPRING EANGERS INSTALLED 1 PERMANENT RTD INSTALLED OPERATING LIMITS - MAX TEMP. DIFF. 3000F e STRATIFICATION FIRST BECAUSE APPARENT DURING HTT
- UNUSUAL MOVEMENT OF SNUBBERS NOTED
- ADDITIONAL INSTRUMENTATION INSTALLED TO REMOTELY MONITOR DISPLACEMENT DURING FUTURE TESTING e DURING PREOP AND POWER ASCENSION TESTING FURTHER DISPLACEMENT OF A CYCLIC NATURE WERE OBSERVED
- MAX. RECORDED DISPLACEMENT HAS - 3.0 INCHES VERSUS .4 IN. CALCUI.ATED DISPLACEMENT
- MAX. DISPLACEMENT OCCURRED DURING REAT UP AND COOL DONN
~
~
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- FROM EVALUATION OF DATA CONCLUDED TEAT RIGE 1' ' DISPLACEMENTS WERE A RESULT OF FLUID THERMAL STRATIFICATION DATA COLLECTION e - PIPING DISPLACEMENTS APPEARED TO VARY AS A FUNCTION OF TEMPERATURE DIFFERENTIAL BETNEEN THE PRESSURIZER AND THE RCS BOT LEG e - ADDITIONAL INSTRUMENTATION INSTALLED TO REMOTELY MONITOR TEMPERATURE e - CONTINUED PLANT TESTING WITE CONTINUOUS MONITORING OF DISPLACEMENTS AND TEMPERATURE ACTIONS e - IMPOSED REAT UP/ COOL DONN LIMITATION OF AT dE 200 0F BETNEEN PRESSURIZER AND RCS BOT LEG e - REMOVED SNUBBER TRAPEZE ON SURGE LIif2 UTILIEING ASME CODE CASE N-411 e - ANALYSIS CONDUCTED TO QUANTIFY EFFECTS OF STRATIFICATION
SPECIFIC ANALYSIS e - SIMPLIFIED 'ANSYS' MODEL OF SURGE LINE GENERATED
- MODEL COMPRISED OF ELASTIC BEAM ELEMENTS
- FIRST BENCE MARKED 'ANSYS' MODEL TO ORIGINAL
'NUPIPE' MODEL
- ALL THERMAL CONDITIONS APPLICABLE TO SURGE LINE WEFE REVIENED TO DETERMINE THE PRESENCE OF THERMAL STRATIFICATION AND POTENTIAL EFFECTS ON THE EXISTING PIPING ANALYSIS
- THESE THERMAL CONDITIONS INCLUDED e NORMAL POWER OPERATION e REAT UP AND COOL DONN e MAXIMUM STRATIFICATION FOLLOWING PLANT
. EXCURSION /1?77; 4
o 'ANSYS' MODEL INCORPORATED ACTUAL RECORDED .
IN-PLANT TEMPERATURE DATA DURING THE ABOVE THERMAL CONDITIONS TO QUANTITY THERMAL STRATIFICATION EFFECTS 1
e - RESULTS OF 'ANSYS' MODEL (FORCES, MOMENTS) FOR EACE APPLICABLE THERMAL CONDITION WAS i RE-EVALUATED BY THE ORIGIMPL 'NUPIPE' MODEL
- e - NUPIPE MODEL THEREFORE COMBINED THE EFFECTS OF THERMAL STRATIFICATION WITH ALL OTHER LOADING CONDITIONS e
4
4 WHIPJET REVIEW e EFFECT OF REVISED ANALYSIS ON SURGE LINE EVALUATED LEAKAGE SIZE FLAW RECALCULATED STABILITY EVALUATION CONDUCTED FATIGUE CRACK GROWTH RATE MALYSIS REVIEWED e MAINTAINED MARGINS PER NUREG 1061 VOL. 3
BV-1 SURGE 7,INE REVIEN I
e GEOMETRY CLOSELY APPROXIMATES BV-2
- THIS INCLUDES PIPE ROUTING, SIZE, ENTRANCE INTO BOT LEG AND SLOPE e SUPPORTING ARRANGEMENT SIMILAR TO BV-2 NO RUPTURE RESTRAINTS ON SURGE LINE PIPING SUPPORTED BY TNO VARIABLE SPRING BANGERS o OPERATIONAL RESTRICTIONS IMPOSED DURING BEATUP/
.COOLDONN SIMILAR TO BV-2 e PIPING LOADS DUE TO STRATIFICATION FROM BV-2 ANALYSIS BEING EVALUATED
CONCLUSIONS L.
e STRATIFICATION' EFFECTS WERE QUALIFIED AND INCORPORATED INTO TEE ANALYSIS e ANALYSIS SHOWS TEAT CODE ACCEPTABILITY MAINTAINED UNDER ALL CASES e WHIPJET REANALYSIS OF SURGE LINE DEMONSTRATES PIPE LBB CAPABILITY UNDER REVISED LOADING e OPERATIONAL RESTRICTIONS OF AT $ 200D' IMPOSED DURING HEATUP AND COOLDOWN ON BOTE BV-1 AND BV-2 o SIMILARITY OF BOTE UNITS SHONS BV-2 STRATIFICATION EFFECTS / LOADS CAN BE APPLIED TO BV-1 4
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B&WOG PRESSURIZER SURGE LINE PRESENTATION TO NRC
. t SEPTEMBER 29, 1988 l
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4 AGENDA l
NRC OPENING COMMENTS i
MEETING OBJECTIVES SURGE LINE HISTORY B&WOG PROGRAM M-K EVALUATION B&WOG PHASE 1 B&WOG PHASE 2 OBJECTIVES B&WOG PHASE 3 ACTIVITIES
SUMMARY
COMMENTS
1 1
MEETING OBJECTIVES TO ACQUIRE INFORMATION ABOUT THE NRC's CONCERN REGARDING PRESSURIZER SURGELINE THERMAL STRATIFICATION TO INFORM THE NRC ABOUT THE B&WOG SURGELINE PROGRAM TO DETERMINE IF THE NRC CONCERN IS ENVELOPED BY THE
- B&WOG PROGRAM m.-
1 PRESSURIZER SURGELINE HISTORY.
l L -
12/84-USNRC INFORMATION NOTICE 84-87 " PIPING THERMAL DEFLECTION INDUCED BY STRATIFIED FLOW" l )
1 1 l-10/85-B&W WAS ADVISED OF THERMAL DATA RETRIEVED AT M-K 1
3/86-B&WOG STEERING COMMITTEE IS ADVISE OF M-K 1 CONDITION ]
3/86-B&WOG MATERIALS COMMITTEE ADVISED OF M-K CONDITION 6/86-B&WOG MATERIALS COMMITTEE REQUESTS AN INVESTIGATION 12/86-PROGRAM WORKSCOPE DEFINED IN RESPONSE TO B&WOG MATERIALS COMMITTEE REQUEST
- 4/87-DRAFT INPO SER ISSUED TO B&W, CE, W, FRAMATOME AND VBG IN GERMANY 6
9/87-INPO SER 25-87 ISSUED 7/88-B&WOG PRESSURIZER SURGELINE REPORT ISSUED
. DETAILING REVIEW OF M-K DATA AND RECOMMENDING INSTRUMENTATION OF A B&W 177FA SURGELINE
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INSTRUMENTED SURGELINE PROGRAM AT MUELHEIM-KAERLICH 0 - THERMOCOUPLE INSTALLED ON PRESSURIZER SURGELINE PRIOR TO STARTUP TESTING 0 STRATIFICATION OCCURRED IN HORIZONTAL PIPE RUN BELOW PRESSURIZER SURGE N0ZZLE L AT LOW FLOW CONDITIONS PRIOR TO REACTOR L COOLANT HEATUP (PRESSURIZER TEMPERATURE = 450F, PRESSURE = 435 PSIG) 0 TEMPERATURE SWINGS OF UP TO 3250F WERE MEASURED IN UP'PER PORTION OF SURGELINE DURING MINOR PLANT EVOLUTIONS 0 STRATIFICATION NOT COVERED IN DESIGN SPECIFICATION 0 RESULTS OF MEASUREMENT PROGRAM
. NECESSITATED ADDI,TIONAL STRUCTURAL ANALYSIS OF SURGELINE ND PRESSURIZER SURGE N0ZZLE TO ACCOUNT FOR ADDITIONAL LOADS DUE TO STRATIFICATION (E.G.,
CIRCUMFERENTIAL TEMPERATURE GRADIENT IN PIPE, ADDITIONAL PIPING MOMENT DUE TO TEMPERATURE DIFFERENCE OVER HEIGHT OF PIPE) i
CONCLUSIONS RESULTING FROM MUELHEIM-KAERLICH MEASUREMENT PROGRAM 0
EVALUATION OF THE THERMOCOUPLE DATA AND FATIGUE ANALYSES OF THE SURGELINE AND PRESSURIZER SURGE N0ZZLE REVEALED THAT ONLY COLD STARTUPS OF THE PLANT LINKED .
WITH THE RELEVANT COOLANT FLUCTUATIONS VIA THE SURGELINE CONTRIBUTE SIGNIFICANTLY TO THE FATIGUE USAGE FACTORS 0 THE CUMULATIVE USAGE FACTORS FOR THE SPECIFIED NUMBER OF DESIGN TRANSIENTS ARE ACCEPTABLE (U'< 0.5 < 1.0 ALLOWABLE) l l
i i
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i REVIEW 0F BBR DATA l
0 EVALUATION OF THERMAL HYDRAULICS TEMPERATURE FLUCTUATION IN SURGE LINE IS DIRECTLY RELATED TO PRESSURIZER LEVEL CHANGES (I.E., SURGE LINE FLOW).
T6 ALWAYS LEADS IN THE TEMPERATURE CHANGES.
THUS, FLUID MOTION OCCURS PREFERENTIALLY IN THE UPPER HALF 0F THE SURGE LINE.
T6 > T7 > T8 IS ALWAYS TRUE. THEREFORE, THE THERMAL STRATIFICATION APPEARS STABLE.
ROUGH CALCULATIONS SHOW THAT TYPICAL SURGE LINE YEL0 CITIES ARE IN THE RANGE OF 1/3 -
2/3 FT/S. THIS VELOCITY IS T00 SMALL TO QUICKLY ELIMINATE STRATIFICATION. _
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SURGELINE DATA MUELHEIM-KAERLICH:
STRAIGHT PIPE: 00 = 19.3" T= 1.8" ELBOWS: OD = 19.8" T= 2.0" MATERIAL: X10 CR NI TI 189 (1.4541) .
(STAINLESS STEEL)
B&W 177FA PLANTS:
STRAIGHT PIPE: OD = 10.75" T= 1.0" ELBOWS: OD = 10.75" T= 1.0" 1
MATERIAL: SA-376 TYPE 316 S.S.
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l HOST RECENT 177FA PLANT CONFIGURATION STRESSES AND CUMULATIVE USAGE FACTORS i o RESULTS FOR RECENT (1982) SURGELINE ANYALSIS o ANALYSIS BASED ON ASME SECTION III, SUMMER 1979 ADDENDA o
o STRATIFICATION NOT CONSIDERED o CUMULATIVE USAGE FACTORS BASED ON 360 HU/CD CYCLES (40 YEARS) o MAXIMUM PRIMARY & SECONDARY STRESS RANGES:
NODE 68: S = 30 KSI (STRAIGHT)
N0DE 105: S = 57* KSI (ELB0W)
NODE 148: S = 37 KSI (El.B0W)
ALLOWABLE = 3SM = 50 KSI
- SIMPLIFIED ELASTIC-PLASTIC ANALYSIS PERFORMED o MAXIMUM CUMULATIVE USAGE FACTORS:
NODE 68: U = 0.53 NODE 105: U = 0.55 ALLOWABLE = 1.0 l NODE 148: U = 0.16 NODE 149: U = 0.57 SEE FOLLOWING FIGURE FOR LOCATION OF N0 DES l
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HORIZONTAL PIPE RUN 0 SINCE POINTS OF MAXIMUM FATIGUE USAGE (AT N0ZZLE CONNECTIONS) ARE LOCATED IN OR NEAR VERTICAL RUNS, NO ADDITIONAL STRESSES WILL BE INDUCED AT THESE LOCATIONS DUE TO CIRCUMFERENTIAL TEMPERATURE GRADIENTS 0 ADDITIONAL BENDING MGMENTS AT N0ZZLES DUE TO STRATIFICATION IN HORIZONTAL PIPE RUN WOULD HAVE TO BE CONSIDERED, BUT MARGIN EXISTS TO ACCOMMODATE THESE 0 MAXIMUM USAGE FACTORS CALCULATED BASED ON 360 HEATUP/C00LDOWN CYCLES 0 NO B&W 177FA PLANT HAS UNDERGONE MORE THAN ~
120 HU/CD CYCLES S0 THAT ACTUAL FATIGUE USAGE IS LOW O MORE DATA IS NEEDED TO QUANITFY THE IMPACT ON OPERATION
B&W PHASE 1 SUMtiARY 0 THERMAL STRATIFICATION FOR THE B&W PLANTS IS POSSIBLE.
O THERMAL CYCLES ARE MORE FREQUENT THAN IN ORIGINAL DESIGN.
O THERMAL CYCLES IN SURGE LINE ARE DIRECTLY RELATED TO PRESSURIZER LEVEL CHANGES.
O FLUID MOTION OCCURS PREFERENTIALLY IN THE bvPER HALF 0F THE SURGE LINE.
O SURGE LINE FLUID WILL RESTRATIFY AFTER A LARGE SURGE.
0 THERMAL CYCLING HINIMIZED BY:
AUTOMATIC STEAM PRESSURE CONTROL CLOSE PRESSURIZER LEVEL CONTROL SEQUENCE OF EVENTS DURING HU/CD
{
94
/ OBJECTIVES OF PHASE 2 INSTRUMENTATION DETERMINE THE MAGNITUDE AND EXTENT OF POSSIBLE STRATIFICATION BY INSTRUMENTING THE OCONEE-1 PRESSURIZER SURGELINE RELATE HEATUP/C00LDOWN EVENTS TO PRESSURIZER N0ZZLE USAGE FACTOR COMPARE RESULTS TO THE REACTOR COOLANT SYSTEM (RCS)
FUNCTIONAL SPECIFICATION DETERMINE THE NEED FOR ADDITIONAL STRESS AND FATIGUE ANALYSIS, IF NECESSARY DETERMINE THE NEED FOR MODIFICATION OF OPERATIONAL PROCEDURES 1
DETERMINE IF MODIFICATIONS TO THE TRANSIENT CYCLE LOGGING PROGRAM ARE NECESSARY 1
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SUMMARY
COMMENTS l 0 WHEN INFORMED ABOUT THE GERMAN PLANT DATA, B&WOG INITIATED A PROGRAM.
o SOME ROUGH CALCULATIONS HAVE BEEN PERFORMED BASED ON WHAT WAS SEEN AT MK.
o MAIN CONCLUSION FROM THE ANALYSES WAS THAT ACTUAL PLANT TEMPERATURE DATA WAS NEEDED.
o A DATA COLLECTION PROGRAM WAS OUTLINED.
IMPLEMENTATION IS PLANNED FOR EARLY 1989.
o DATA EVALUATION WILL START IN 4/89.
o FINAL RESULTS SHOULD BE AVAILABLE ABOUT 9/89.
o A FOLLOW ON PHASE HAS BEEN TENTATIVELY DEFINED IF HEEDED.
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p 9EACTOR HYDRODYNAMICS TABLE 4 IMPORTANT DIMENSIONLESS GROUPS FOR SIMILITUDE IN HYDRODYNAMIC TESTING Parameter Symbol Definition Significance Weisbach frcton, I DAP/2pVrt Pressure forcerinetta force factor 8
3 2. Cavitaton number b (P, - P,)/pV Pressure difference /inertta force l 3. Reynolds number Re pVD/g , inertia forceivacous force
- 4. Strouhal number Sr vp/V Vortex thecchng treauencyr inema force 8
- 5. Weber number We pDVla Inema forcersurfacqHenson force
- 6. Froude number Fr V'/gD . Inertia force / gravity force
- 7. Rchardson number Ri aggD/pV Buoyancy force /inetta force (Modified Froude number)
- 8. Euler number Eu AP/pV8 Pressure force /inetta force
- 9. Prandtl number Pr C/k Momentum diffusuty/ thermal ;
diffuswify 1
.10J Peclet number Pe pVDC/k Convectue heat transfert (Re x Pr) conductwe heat transfer
- 11. Grashof number Gr l'p'gsaT/g* Buoyancy force /vocous force
- 12. Rayleigh number Ra L8p*CgsaT/uk (Gr x Pr)
NOMENCLATURE:
C = specific heat g = acceleration of gravity l
! p = density P = pressure o = surface tension P, = state flud pressure k = thermal conducturty P, = flud vapor pressure S = volumetre expanson coefficient L.D = charactenste dimensons j
. AT = flud temperature change V = flud velocity
( ., esonex sheccing frequency u = viscosity l
l
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A OD T S
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W D B WG S O NE E DOI A L ATE H NBF F AT T AL P TUI ED E D ONN N L E E E HEI I g H T I TF . A0D S F NTI TS M9A .
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,L I P L T P D oE P E (C A T N C SE Y I RR L D EE L E TF A R EE R P MR E A N E RD E N B AN G O PA I N A T A CS T A C IE A C LS D I N UP F O AC T I I R S T T DR E A A YO T R C HF T I G S F LR N I AA I T ML T .
M U
S - - -
s 3
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. CEOG SURGE LINE DESIGN DATA pA -
WALL UNII MATERIAL DIAMETER E*
WATERFORD 3 SA-351 GR. CF8M 12" SCH. 160 ST. LUCIE 1 316SS A-351 GR. CF8M 12" SCH. 160 Sr. LUCIE 2 -
SA-351 GR. CF8M 12" SCH. 160 ANO 2 SA-351 GR. CF8M 12" SCH. 160 CALVERT CLIFFS 1 A-351 GR. CF8M 12" SCH. 160 CALVERT CLIFFS 2 A-351 GR. CF8M 12" SCH. 160 MILLSTONE 2 3165S A-351-65 GR. CF8M 12" SCH. 160 i MAINr YANKEE 3165S A-451 GR. CPF8M 12" SCH. 160 SAN ONOFRE 2 # ## Y 12" SCH. 160 SAN ONOFRE 3 SA 376 I[6 f 3Id 12" SCH. 160 PALO VERDE 1,2,3 SA-376 TYPE 304 (STR. RUN) 12" SCH. 160 SA-403 TYPE 304 (ELBOWS)
PALISADES A-376 12" SCH. 140
- FT. CALHOUN A-376 TYPE 316 10" SCH. 160 **
- 12" SCHEDULE 160 NOMINAL 00 = 12.75", NOMINAL WALL THICKNESS = 1.32"; 12" SCH.140 NOMINAL THICKNESS = 1.125"
- 10" SCHEDULE 160 NOMINAL 00 = 10.75", NOMINAL WALL THICKNESS = 1.125"
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OPERATIONAL TEMPERATURE DATA o TYPICAL HOT LEG TEMP. 550 F - 620 F o TYPICAL PRESSURIZER TEMP. 650 F o MAXIMUM PRESSURIZER HEAT UP Is 200 F/HR PLANTS HAVE TECH. SPEC OR ADMINISTRATIVE LIMITS WHICH LIMIT THE HEAT UP TO LESS THAN -
THIS MAXIMUM VALUE O MAXIMUM PRESSURIZER COOLDOWN IS 200 F/HR PLANTS HAVE TECH. SPEC. OR ADMINISTRATIVE LIMITS WHICH LIMIT THE HEAT UP TO LESS THAN THIS MAXIMUM VALVE
4 TYPICAL.CEOG PLANT HEATUP CURVE reo - -
d' soo . . h ge.hS = p L- .
r R.,k .100 M ~ '
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HEATUP TIME - HOURS Passsunzzen To hot LEG "oELTA T" 1
A) 320 0F s) 140 0F
. c) 283 0F -
o) 88 0F 4
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Ms.I.9.2.2 APPENDIX I 1986 Edidos 28 h
,, w 24
\ Curve A I 22 L d
% 20 h ' 5 3 %, s k> ,e \ '
A '
curve e 16 % curve c
~
14 12 1011 10B 108 1010 106 107 Number of cycles,N NOTE:
' E - 28.3 X 106p.3 Crheria for the Use of the Curves in This Figure (Notes (IMS))
assuc AnaWs of Destic Anaws of Matertal WWs and Adjacent 00er Than Welds and Rame Metal Curve Adjacent Base Metal A (P, + P + 0)% 5 27.2 ksi .
(P, + P, + @% > 27.2 ksi and (P,+P,+0)% 5 27.2 ksi B
S,is corrected for applied mean stress (P, + P + @% > 27.2 ks!
C (P, + P, + @.,,,, > 27.2 ksi NOTES:
(1) Range appres to the indMdual quantfues P,, P., and O and poplies to the set of cycles under consideration.
(2) Thermal bending stresses resulting from axial and rad 4at gradients are excW from O.
, (3) Curve A is also to be used wtth inelastic analysis with 5, = % A a, E, where a e,is the total effeco Ar strain range.
(4) The maximum effect of retained mean stress is included in Curve C.
(5) The ad)meent base metalis defined as three waft thicknesses from the center 16 FIG.1-9.2.2 DESIGN FATIGUE CURVE FOR AUSTEN! TIC STEELS, NICKEL-CH A87 ALLOY, NICKEL-IRON-CHROMIUM ALLOY, AND NICKEL-COPPER ALLOY FOR S, s; 21L2 ksi, FOR TEMPERATURES NOT EXCEEDING 800'F (For S, > 28.2 ksi, use Fig.19.2.1.)
Table I-9.2.2 Contains Tabulated Values for Accurate Interpolation of This Curve R
195 k
m
- - - - _ - _ - - - - - - - - - , - _ - _ _ - - _ - - _