ML20029C140
| ML20029C140 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/31/1991 |
| From: | Palusamy S, Roarty D, Vora V WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17328A919 | List: |
| References | |
| WCAP-12851, NUDOCS 9103260253 | |
| Download: ML20029C140 (102) | |
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( WESTINGHOUSE CLASS 3 l
Structural Evaluation of Donald C. Cook Nuclear Plant Units 1 and 2 Pressurizer Surgo Lines, Considering the Effects of Thermal Stratification January 1991 T. H. Liu L. M. Valasek S. Tandon
, P. L. Strauch R. Brice-Nash M. A. Gray M. Yu Verified by: ( # M /s / '
Verified by:
- D. H. Roarty ' V. V. Vora A
Approved j; -
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- 5. 5. Palusamy. Manager R. 5. Patel, Manager Diagnostics and tionitoring System Structural Analysis Technology and Development VESTINGHOUSE ELECTRIC CORPORATION
. Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania IS230-272d e 1991 Westinghouse Electric Corp.
5100s/021H1 10
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TABLE OF CONTENTS Section Title Page Executive Summary iii 1.0 Background and Introduction 1-1 1.1 Background 1-1 1.2 Descriptien of Surge Line Stratification 1-3 1.3 Scope of Work 1-4 2 .<0 Surge Line Transient and Temperature Profile Development 2-1 2.1 General Approach 2-1 2.2 System Design Information 2-2 2.3 Development of Normal and Upset Transients 2-3 2.4 Monitoring Results arid Operator Interviews 2-4 2.5 Historical Operation 2-6
, 2.6 Development of Heatup and Cooldown Transients 2-7 2.7 Axial Stratification Profile Development 2-10
. 2.8 Striping Transients 2-l's 3.0 Stress Analysis 3-1 3.1 Surge Line Layouts 3-1 3.2 Piping System Global Structural Analysis 3-2 3.3 Local Stresses - Methodology and Results 3-4 3.4 Total Stress from Global and Local Analysis 3-6 3.5 Thermal Striping 3-7 4.0 Displacements at Support Locations 4-1 e
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, -5.0. Fatigu~e . Usage f 5-1
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5;1L 4 thodology 5-l--
_ 5.2:_ ; Fatigue Usage' Factors; 5 :
< 5.31 ' Fatigue Due to Thermal Striping .5-9
_ Fatigue: Usage Results
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l 6.0 :Sumary and Conclusions ~ 1
- 57.0.' . References L7 . .
' Appendix A: Computer Codes _- . A-1 .i
'A~ppendii!B; lUSNRCl Bulletin;88-11-- B-1: '
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'4 _ Appendix C; ~ Transient Development. Details- C ,
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EXECUTIVE
SUMMARY
Thermal stratification has been identified as a concern which uan af' ct the structural integrity of piping systems in nuclear plants 'nce 1979, when a leak was discovered in a PWR feedwater lins. In the pressurizer surge line, stratification can result from the difference in densities between the not leg water and generally hotter pressurizer water. Stratification with icrge temperature differences can produce very high stresses, and this can lead to integrity concerns. Study of the surge line behavior has concluded that the largest temperature differences occur during certain modes of plant heatup and Cooldown.
This report has been prepared to demonstrate compliance with the requirements of NRC Bulletin 88-11 for Donald C. Cook Nuclear Plant Units 1 and 2. Prior to the issuance of the bulletin, the Westinghouse Owners Group had a program in place to investigate the issue, and recommend actions by member utilities.
.;.at program provided the technical basis for the plant specific trantient development report-d here for the Cook Nucicar Plant, Units 1 and 2.
This transient development utilized a number of sources, including plant
. operating procedures, surge line monitoring data, and historical records for each unit. This transient information was used as input to a structural and stress analysis of the surge line for the two units. A review and comparison of the piping and support configurations for the units led to the conclusion that the surge lines are nearly identical, and thus one analysis could be done to apply to both units, for the stratification transient aevelopment.
The results of the structural analysis, and the fatigue analysis which followed, showed that the Cook Units 1 and 2 meet the stress limits and cumulative usage factor requirements of the ASME Code for the licensed operation (32 Effective Full Power Years) of the units. The support loads and displacements resulting from stratification have also been provided, for use in re evaluating the adequacy of the supports, and ensuring proper gaps in pipe whip restraints to allow free pipe movement at all thermal conditions.
This work has led to the conclusion that the Cook Units 1 and 2 are in full compliance with the requirements of NRC Bulletin 88-11.
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SUMMARY
Cf RESULTS, AND STATUS OF 88-11 QUALIFICATION Cook Unit 1 Cook Unit 2
,n Operating History up to'1990 Date ofLinitial RCS heatup 3/14/74 .8/13/77- -
Date of. commercial operation 8/23/75- 7/1/78 Years of; water-solid heatups 3 0 Years of steam-bubble heatups 14* 13 1/2 l System delta T limit 320F 320F Number ,of 320F.exceedances Two two 4
Maximum Stress and Usage Factor
- Results for-32 Effective Full Power Years E 35.1/53.0 35.1/53.0
-(quation 12=
ksi) -~ Hot stress / allowable-Leg-Nozzla.
. Fatigue usage / allowable 0.3/1.0' O.3/1.0 Hot-Leg Nozzle.
Pressurizer Surge Nozzle Results Maximum stress intensity range / 35.0/57,9** 35.0/80.1**
allowable (ksi) '
Fatigue usage / allowable 0,1/1.0 0.1/1.0 4
Suoport Modification Required- None None-Status of'B8-11 Requirements All analysis .
All analysis requirements met requirements met
- Unit'l years of- steam bubble heatups February 1977 until present
- Unit 1 material is SA-216, Grade.WCC (casting); Unit 2 material is- -
.SA-508, Class'2 (forging).
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'SECTION
1.0 BACKGROUND
AND INTRODUCTION 3
, Donald C. Cook Nuclear Plant Units 1 and 2 are four. loop pressurized water -)
c reactors, designed to be as nearly identical as practical, in both hardware and operation. This report has been developed to provide the technical basis and results of a plant-specific structural evaluation for the effects of thermal stratification of the pressurizer surge lines'for each of these units. ,
.The' operation of..a pressurized water reactor requires the primary coolant loops to be water solid, and this is accomplished through a pressurizer vessel, connected to one of the hot legs by the pressurizer surge line. A typical four loop arrangement is shown in Figure 1-1, with the surge line highlighted.
The pressurizer vessel contains steam and water at saturated conditions with
- the steam-water interface level typically between 25 and 60% of the volume depending on the plant 2 operating conditions. From the time the steam bubble ,
. is initially' drawn'during.the heatup operation to hot standby conditions, the level is maintained-at approximately 25% to 35%. During power ascension, the
- Unit 1 pressurizer-level is maintained at 40% and the Unit 2 level is increased from 25% to approximately 55% -The steam bubble provides a pressure cushion effect-in the event of sudden changes in Reactor Coolant System (RCS) mass inventory. Spray operation reduces system pressure by condensing come of the steam. Electric heaters, at the bottom of the pressurizer, are energized to raise the liquid temperature to generate additional steam and increase RCS pressure.
As-illustrated in figure 1-1, the bottom of the pressurizer vessel is
-connected to' the hot leg of one of the coolant loops by the surge line, a 14
. inch schedule 163 stainless steel- pipe, most of which is almost-horizontal, but slightly pitchec down toward the hot leg.
1.1 Backoround During the period from 1982 to 1988, a number of utilities reported unexpected movement of the pressurizer surge line, as evidenced by crushed insu htion, sioove m eiia 11
1 .
gap closures in the pipe whip restraints, and in some cases unusual snubber movement. Investigation of this problem revealed that the movement was caused by thermal stratification in the' surge line. ,,
Thermal stratification had not been considered in the original design of any .
pressurizer surge line, and was known to have been the cause of service-induced cracking in feedwater line piping, first discovered in 1979.
Further instances of service-induced cracking from thermal stratification surfaced in 1988, with a crack in a safety injection line, and a separate occurrence with a crack in a residual heat removal line, Each of the above incidents resulted in at least one through-wall crack, which was detected through leakage, and led to a plant shutdo '. Although no through wall cracks were found in surge lines, inservice inspectwns of one plant in the U.S. and another in Switzerland mistakenly claimed to have found sizeable cracks in the pressurizer surge line. Although both these findings were subsequently disproved, the previous history of stratified flow in other lines led the USNRC to issue Bulletin 88-11 in December of 1988. A copy of this bulletin is included as Appendix B.
The bulletin requested utilities to establish and implement a program to confirm the integrity of the pressurizer surge 1ine. The program required
- both visual inspection of the surge line and demonstration that the design requirements of the surge line are satisfied, including the consideration of' stratification effects.
Prior to the issuance of NRC Bulletin 88-11, the Westinghouse Owners Group had implemented a program to address the issue of surge line stratification. A bounding evaluation was performed and presented to the NRC in April of 1989.
This evaluation compared all the WOG plants to those for which a detailed plant-specific analysis had been performed. Since this evaluation was unable
.to demonstrate the full des'gn life for all plants, a generic justification for continued operation was developed for use by each of the WOG plants, the basis of which was documented in references 1 and 2.
The Westinghouse Owners Group implemented a program for generic detailed analysis in June of 1989, and this program involved individual detailed analyses of groups of plants. This approach permitted a n4cre realistic l
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.- .- - - -- - - - . _ . . = . . - - - - - -. ..
approach _than could be obtained from a single bounding analysis for all plants, and the:results were published.in lune of 1990 (3)
- The followup to the Westinghouse Owners Group Program is a performance of
- evaluations which could.not be performed on a generic basis. The goal of this report is to accomplish these followup actions, and'to therefore complete the requirements of the'NRC Bulletin 88-11 for the Cook Units 1 and 2.
1.2 Descriotion of Surge Line Thermal Stratification
-It will be useful to describe the phenomenon of stratification, before dealing with'its effects. Thermal stratification in.the pressurizer surge line is the direct result of the difference in densities between the pressurizer water and the generally cooler RCS hot leg water. The warmer lighter pressurizer water tends to. float on the cooler heavier hot leg water. The potential for stratification is-increased as the difference in temperature between the pressuri:er and the hot leg increases and as the insurge or outsurge flow r'ates decrease.
-At power,1when the difference in temperature between the pressurizer and hot t leg is relatively small, the extent and effects of stratification have been observed to be small. However, during certain modes of plant heatup and cooldown,.this' difference in system temperature could be as large as 320*F, in which case-the effects of stratification are significant, and must lie accounted for.
Thermal: stratification in the surge line causes two effects:
-o. Bending of the pipe different than that predicted in the original design.
o Potentially reduced-fatigue life of the piping due to the higher
, stress resulting from stratification and striping.
- Numbers in brackets refer to references listed in Section 7.
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1.3 Scope of Work The primary purpose of this work was to develop transients applicable to the -
Cook units which include the effects of stratification and to evaluate these effects on the structural integrity of the surga lines. This work will '
therefore complete the demonstration of compliance with the requirements of NRC Bulletin 88-11.
The transients were developed following the same general approach originally established for the Westinghouse Owners Group. Conservatisms inherent in the original approach were refined through the use of monitoring results, plant operating procedures and operator interviews, and historical date on plant operation.
The resulting transients'were used to perform an analysis of the surge line, wherein the existing support configuration was carefully modeled, and surge line displacements, stresses, suoport loadings and nozzle loads were determined. This analysis and its results are discussed in Section 3 and 4.
The stresses were used to perform a fatigue analysis for the surge line, and ,
the methodology and results of this work are discussed in Section 5. The summary and conclusions of this work are summarized in Section 6.
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-SECTION 2.0
- SilRGE LINE TRANSIENT AND TEMPERATURE PROFILE DEVELOPMENT 2.1 Overall Approach The-transients for the pressurizer surge line were developed from a number of sources ' including the most recent systems standard design transients. 'The heatup and cooldown transients, which involve the majority of the= severe
" stratification occurrences, were developed frem review of the plant operating _ ,
. procedures, operator interviews, monitoring d:da and historical records for
-each unit. The total number of heatup and cooldown events specified remains unchanged at 200 each, but a number of sub-events within each heatup and cooldown cycle have been defined to reflect stratification effects, as described in more detail later.
The normal and upset transients, except for heatup and cooldown, for Cook
= Units 1 and 2 surge lines are provided in Table 2-1. For each of the ~
-transient's the surge'line fluid temperature was modified from the original
. design a'ssumption of uniform temperature to a stratified distribution, according to the predicted temperature differentials between the pressurizer-a and hot 11eg, as listed in the table. The transients have been characterized as either-insurge/outsurges (1/0 in the-table) or fluctuations (F).
- Insurge/outsurge transients-are. generally more severe, because they result in
-the' greatest. temperature change in the top or bottom of the pipe. Typical
~
temperature prof _iles for-insurges.and outsurges are shown-in Figure 2-1.
Transients identified as fluctuations (F) typically involve low surge flow rates and smaller temperature differences between the pressurizer _and hot leg, so the resulting stratification stresses are much lower. This type of cycle is important~ to include in the analysis, but is generally not the-major contributor to fatigue-usage.
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- In addition to the plant specific operating history discussed above, the j development of transients which are applicable to Cook Units 1 and 2 was based )
on the'wurk already accomplished under programs completed for the Westinghouse ,
Owners Group (1,2,3). In this work all the h stinghouse plants were grouped based on the similarity of-their response to stratification. The three most -
important factors influencing the effects of stratification were found to be the structural layout, support configuration, and plant operation. -
The transient development for the Cook units took advantage of the similarity in the' surge line layout for the two units, as well as the similarity of the operating procedures. A detailed comparison of the piping and support configurations for the units appears in Section 3.1.
The transients developed here, and used in the structural analysis, have taken advantago of the monitoring data collected during the WOG program, as well as operator interviews _and historical operation data for the Cook Units. Each of
-these will-be discussed in the sections which follow.
2.2 System Design Information *-
-The thermal design transients for a typical Reactor Coolant System, including the pressurizer surge line, are defined in Westinghouse Systems Standard Design Criteria.
The' design transients for the surge line consist of two major categories:
~
(a) _Heatup and Cooldown transients
-(b) Normal and' Upset operation transients (by definition, the emergency
- and faulted transients are not considered -in the ASr4E Section III
. fatigue life assessment of components).
I
(.
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.. . ..- - - - - -- . - - -.- - _ . ~ . -
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t i lntf.eevaluationofsurge_linestratification,thetransientevents i
Teonsidered encompass the-normal and upset design events defined in FSAR- s chapter 4.1.- '
i 1 The total number,of currant.heatup-cooldown cycles-(200) retains enchanged, l-However,L sub-events and:the associated number of occurrences (" Label",'" Type"
"=
and " Cycle" columns of tables-2-1 and 2-2-have been defined to reflect-Estratification'effcets, as-described-later.
2.3 Stratification Effects Criteria and Development of Normal and Voset Transients
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_................................ __.ja,c.e 2.4 Monitoring Results and Operator Interviews 2.4.1 Monitoring I
Monitoring was performed at Cook Unit 1. The monitoring program used temporary sensors on-the surge line piping, as shown in figure 2-2.
Monitoring information ecliected as part of this program was utilized in this analysis.
The pressurizer surge line monitoring program utilized externally mounted temperature sensors (resistance temperature detectors). The temperature
. sensors were attached to the outside surface of the pipe at various circumferential and axial locations,s In all cases these temperature sensors were securely clamped to the piping outer wall using clamos taking care to ,
properly insulate the area against heat loss due to thermal convection or radiation. .
stoo mnoiao 2-4
The typical temperature sensor configuration at a given pipe location consists of two to five sensors mounted as shown in figure 2-2. Temperature sensor configurations were mounted at various axial locations. The multiple axial locations give a good picture of how the top to bottom temperature
, distribution may vary along the longitudinal axis of the pipe, in addition, displacement sensors were mounted at two axial locations to detect horizontal and vertical movements, as shown in figure 2-2. Typically, data were collected at (----------Ja,c.e intervals or less, during periods of high system delta T.
Existing plant instrumentation was used to record various system parameters.
These system parameters were useful in correlating plant actions with stratification in the surge line. A list of typical plant parameters monitored is given below.
[.. ...................................
.. .................................................)a,c.e Data from the temporary sensors was stored on magnetic floppy disks and converted to hard copy time history plots with the use of common spreadsheet software. Data from existing plant instrumentation was obtained from the Cook
. plant computer.
2.4.2 Operational Practices An' operations interview was conducted at Cook Nuclear Plant on October 4, 1989. Both units operate with similar if not identical procedures, so the
~
following is applicable to the two units. Since the maximum temperature difference between the pressurizer and the reactor coolant loop occurs during v oo,,can o io 2-5
3 the pl'antLheatup and cooldown, operations during these events were the main
- topic of-thesinterview. Figure 2-3 describe the heatup process, and figure 2-4.is the' corresponding plot for the cooloown process. ,
H As;a result of this interview, it was discovered that_the Cook units perform the fill and vent operation with a steam bubble in the pressurizer. This
- practice is unique to Cook in the WOG program. and has been considered in the thermal stratification thermal transient development.
I In both heat'upland cooldown, the units have operating limits-(based on technical ~ specification) of 320'F on the maximum temperature difference cbetween pressur'izer_ steam' space and pressurizer spray water. The reliance of
-this restriction to limit the temperature difference between the pressurizer
-liquid = space -and the RCS hot leg is valid because the pressurizer steam space
-temperature:is always the same or greater than the liquid space temperature (except-for a short period-just prior to bubble formation', when initial heatup of pressurizer inventory occurs). Additionally, the spray water temperature is always less than or equal to the RCS hot leg water temperature. This applies under both: normal and auxiliary spray. conditions. -
l 2.5E Historical Operation- *!
4 A review of. historical records from each unit (control room charts, operator logs,; surveillance test reports, etc.) was. performed. From this review, two
. pieces _of information were extracted; a characteristic maximum system delta T !
forieach heatup'and cooldown recorded:and-the number of maximum delta T
'exceedences. The number-of heatup and cooldowns for which data was considered-in the analysis and their associated system delta temperature are listed below l
-for-each' unit.
p 1
~
Cook Unit 1 ,
Number of l
. System AT Heatup & Cooldown u - Dange.('F) Experienced to Date- '
L Heatuos Cooldowns b- .... . . . -
[...... ,
......__ ___e ____
hee @eee ae ee eeeeeee ee ee
....... .. ._)a,c,e stoo,mneno 2-6
Cook Unit 2 Number of System AT Heatup & Cooldown Range (*F) Experienced to Date Heatups Cooldowns
[......
..]
A summary of the information given above is provided in Figure 2-5.
2.6 Development of Heatup and Cooldown Transients The heatup and cooldown transients used in the analysis were developed from a number of sources, as discussed in the overall approach. The transients were built upon the extensive work done for the Westinghouse Owners Group (1,2,3),
coupled with plant specific considerations for Cook Units 1 and 2.
The transients were developed based on monitoring data, historical operation
, and operator interviews conducted at a large number of plants, including Cook. For each monitoring location, the top-to-bottom differential
- temperature (pipe delta T) vs. time was recorded, along with the temperatures of the pressurizer and hot leg during the same time period. The difference between the pressurizer and hot leg temperature was termed the system delta T.
F, she pipe and system delta T information collected in the WOG(1,2,3) effort, individual plants' monitoring data were reduced to categorize stratification cycles (changes in relatively steady-state stratified conditions) using the rainflow cycle counting method. This method considers delta T range as opposed to absolute values.
4 1
1 si m eca nsiio 2-7
E
[................
j f
.......................................................)a,c.e The resulting distributions (for 1/0 transients) were cycles in each RSS range above 0.3, for each mode .(5,4,3 and 2). A separate distribution was determined for the reactor coolant loop nozzle and for a chosen critical pipe location.
Transients, which are represented by delta T pipe with a corresponding number of cycles were developed by combining the delta T system and cycle distributions. For mode 5, delta T system is represented by a historical system distribution developed from plant operating records.
v .
For modes 4, 3 and 2, the delta T system was defined by maximum values. The values were based on the maximum system delta T obtained from the monitored plants for each mode of operation. An analysis was conducted to determine the average number of cycles per cooldown relative to the average number of cycles "i
per heatup. (--------------------------------------------------------------
i
....................)a,c,e The' transients for all modes were then enveloped in ranges of aTp$p,, i.e., all cycles from transients within each aTp$p, range were added and ccsigned to the pre-defined-ranges. These cycles were then applied in the fatigue analysis with the maximum aT P i pe
.for each~ range. The values used are as follows:
O i
l l
l l
l l
l uno,m oo 2-8
^
For Cycles Within Pipe Delta-T Range- Pipe Delta T
_[.......... ...
. ... .. ... ... 4
...)a.c.e
.This grouping was done to simplify the fatigue analysis. The actual number of
- cycles used for'the analysis of the heatup and cooldown is'shown in Table 2-2.
- As previously discussed, the Cook units perform the fill and vent operation with a steam bubble-in the pressurizer. The effect of this practice is high cyclic-activity relative to other.WOG units. Consequently, the Cook k
trans'ients are more severe than those of other WOG plants. Historical records indicate-that for ten percent of the heatups, the Cook units heat-up in a-a manner'similar~to other WOG units.- Therefore the surge line transiente used in this analysis include 90 percent of the Cook transient cycles combined with h -10 percent of the WOG transient cycles at each pipe delta T. _
m .
The-final result of this complex process is a table of transients j -corresponding lto the subevents of the heatup and cooldown-process. A mathematical-description of the methodology used is given-in Appendix C.
' The total: transients for heatup and cooldown are identified:as HC1. thru HC10 for the pipe, and hcl thru HC8 for the RCS hot leg nonle as shown in tables 2,(a)-and:2-2(b)respectively. Transients HC9.thru HC10 for the pipe and HC8 for the nozzle represent transients which occur during later stages of the heatup, d [....................................................................
ej ........... ..................................................................
..................ja,c,e stoovostoomo 2-9 W; ,
, , ~ . - - - . - - . .. .. - - . ~ - .- . -
-Because of mainEcoolant pipe flow effects', the stratification transients xioadings at the RCS hot leg nozzle are different. These transients have been applied ~to the main body of the-nozzle as well as the pipe to nozzle girth butt weld.
Plant' monitoring included sensors located near the RCS hot leg nozzle to surge line pipe weld. Based on the monitoring, a set of transients was developed for.the nozzle region to reflect conditions when stratification could occur in the nozzle. The primary factor affecting these transients was the flow in the main coolant pipe. Significant stratification was noted only when the reactor coolant pump-in the loop with the surge line was not operating. Transients-were then developed using a conservative number of " pump trips.'
The-fatigue analysis of the RCS hot leg nozzle was then performed using the
" nozzle transients" and the pipe transients. The analysis included both the-stratification loadings from the nozzle transients,.and the pressure and bending loads-from-the piping transients. ,
As indicated in Section 2.5, based on a review of Cook Units 1 and 2 operating .
records, there were events in which the system delta-T exceeded the transient basis. upper limit of (-------------------------------------------------------
...........ja,c,e 2.7 Axial-Stratification Profile-Develooment In addition to transients, a profile of the-(--------------------------------
i ............................................................................
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. ,i Two types of profile were generated to envelope the stratified temperature distributions observed and predicted to occur in the line. These two profiles are l----~~~~~~~-------~~-------------------~~-------------------------------
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iThese:three: configurations are illustrated in Figure 2-9. The Cook. units fall T
?under:the category of (--------------------------------------------------
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................................................w....................... '
m m
...................................................................]a,c.e LRev'iewiand. study of the monitoring data revealed a consistent pattern of-
~
W development of delta:T:as ~a function of distance from.the hot leg
- intersection.- This pattern was consistent throughout the heat-up/cooldown processa for a given: plant geometry. This pattern was used along with' Cook plant operating' procedures:to provide a realictic yet somewhat~ conservative y
--. portrayal of- the pipe; delta T.along the surge line.
. ;The combination of the hot / cold interface and pipe delta T as functions of distance along the; surge line forms a profile for each individual unit-
, ; analyzed. Since Cook Units-1 and 2 have.similar surge line configurations, ,
the-profile applies to both ut.its.
sioo.,o:inuo 2-12
2.8 Stripine Transients ,
~,
The transients developed for the evaluation of thermal striping are shown in table 2-3.
[...........................................................................
......................................................)a,c,e Striping transients use the labels HST and CST denoting striping transients (ST). Table 2-3 contains a summary of the HSTl to HST8 and CST 1 to CST 7 thermal striping transients which are similar in their definition of events to the heatup and cooldown transient definition.
. . These striping transients were developed during plant specific surge line evaluations and are considered to be a conservative representation of striping in the surge line(3). Section 5 contains more information on specifically how the striping loading was considered in _the fatigue evaluation.
-4 5100s/021191:10 2-13
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TABLE 2-1 SURGE LINE TRANSIENTS WITH STRATIFICATION NORMAL AND UPSET TRANSIENT LIST - COOK UNIT 1 OR UNIT 2 .
TEMPERATURES ('F)'
MAX NOMINAL LABEL- TYPE CYCLES AT PRZ T RCS-T Strat
[.......................
4
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TABLE 2-1 (Cont'd.)
SURGE LINE TRANSIENTS WITH STRATIFICATION NORMAL AND UPSET TRANSIENT LIST - COOK UNIT 1 OR UNIT 2 TEMPERATURES (*F)
MAX NOMINAL LABEL TYPE CYCLES 6T PRZ T RCS T Strat
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TABLE 2-2a
. SURGE LINE PIPE TRANSIENTS WITH STRATlf! CATION - COOK UNIT 1 OR UNIT 2 HEATUP/COOLDOWN (HC) - 200 CYCLES TOTAL 1
TEMPERATURES (*F)
MAX NOMINAL l.ABEL .- TYPE CYCLES AT PRZ T RCS T Strat mee me me. SOS .ee
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IABLE 2-2b SURGE LINE N0ZZLE-TRANSIENTS WITH STRATIFICATION - COOK UNIT 1 OR UNIT 2 HEATUP/COOLDOWN (HC) - 200 CYC'.ES TOTAL
'4 TEMPERATURES (*P)
MAX NOMINAL LABEL TYPE CYCLES AT PRZ T RCS T Strat
(... ... . ... ... ...
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TABLE 2-3 l SURGE LINE TRANS!ENTS - STRIPING- l FOR HEATUP (H) and COOLDOWN (C) - UNIT 1 OR UNIT 2
[..........
.... ... ... 4
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Figure 2-1. Typical Insurge-Outsurge (1/0) Temperature Profiles steo.mnoi in 2-19
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Figure 2-2. Monitoring Locations for Cook Unit 1 L
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__ _I Figure 2-5. . Summary of Historical Data from Cook Units 1 and 2, Compared to Design Heatup and Cooldown for 40 Years (32 Effective Full Fswer Years) voo.,co i io 2-23
. ~ . . . . . - - - , _ _ . _ . - -. , _ - . . _ . . . _ _ _ . . - . . . _ . . _. . . . _ . . - . . - - _.
a.c.e lt i Figure 2-6. Temperature Profile for High Flow Conditions for Cook Units 2 and 2 sia.m m eiio. 2-24
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Figure 2-9, Geometry Considerations siowensei in 2-27
. . . - , . . . , , . , , , . , , , ,. ,, - -,...,,,,,! n,, ,,
SECTION 3.0 STRESS ANALYSES The flow diagram (figure 3-1) describes the procedure to determine the effects of thermal stratification on the pressuri:er surge line based on transients developed in section 2.0. (----------------------------------------------
............]a,c.e 3.1 , Surae Line layouts The Cook Units 1 & 2 surge line layouts are documented in references 6 and 7 and the Unit I layout is shown schematically in Figure 3-2. The two Conk units are mirror images of each other along plant East-West. The support configurations of Cook surge lines are the same. Below is a table summarizing the existing Cook surge line support configuration.
Conk Units 1 and 2 Suoport Unit 1 Unit 2 Node Type 1-SU-L-900 2-SU-L-900 1370 Pipe Whip Restraint 1-SU-L-901 2-SU-L-901 1350 Pipe Whip Restraint 1-SU-L-902 2-SU-L-902 1340 Pipe Whip Restraint 1-SU-L-903 2-SU-L-903 1330 Pipe Whip Restraint 1-SU-L-904 2-SU-L-904 1300 Pipe Whip Restraint 1-SU-L-905 2-SU-L-905 1290 Pipe Whip Restraint 1-SU-t-906 2-SU-L-906 1190 Pipe Whip Restraint 1-SU-L-907* 2-SU L-907* 1180 Vertical Rigid Restraint 1-SU-L-908 2-SU-L-908 1170 Pipe Whip Restraint 1-SU-L-909 2-SU-L-909 1160 Pipe Whip Restraint 1-SU-L-910 2-SU-L-910 1140 Pipe Whip Restraint 1-SU-L-911 2-SU-L-911 1090 Pipe Whip Restraint 1-SU-L-912 2-SU-L-912 1060 Pipe Whip Restraint
. Drawing number 1/2-GRC-R-619 siowensei io 31
_ _ _ _ _ - . _ - . _ _ _ _ _ _ _ . . ~ . _ _ _ . _ _ _ _ _ _ - . _ . _ ___ -
1 l
It'can be seen from the table above that both of the Cook surge lines contain one equivalent vsrtical rigid support which usually causes higher thermal I l
loads from the effect of thermal. stratification. The piping size is 14 inch ,
i schedule 160 and the pipe material is stainless steel, SA 376-Type 316, for both units. The hot leg nozzle material is SA-182, F316 for both units. .-
Experience with the analysis of thermal stratification has indicated that i surge iine layout (-~~-----------------------------------------------------
.........,.ja t
c.e 3.2- Piping System Global Structural Analysis i The Cook Unit 1 piping system was modeled by pipe, elbow, and non-linear spring elcmonts using the ANSYS computer code described in Appendix A. The results from this model are applicable to Unit 2. The geometric and material parameters are included. (-------------------------------------------------
r 1
.....................................)a,c.e The hot-cold temperature interface along the length of a surge-line [------ ,
4
.....................................................................)a,c.e .
im.u.
1 a,c,e t
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c Figure 3-8. Surge Line Local Axial Stress on Inside-Surface at l (----Ja,c e Axial Locations-i.oo.mnei io 3-22
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Figur.e 3-9. Sarge Line Local Axial Stress on Outside Surface'at
]a,c,e Axial Locations 1
sioo.m n ei .io 3-23
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Figure 3-10. Surge Line RCL Nozzle 3-D WECAN Model: 14 Inch Schedule 160 -
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+ . e d e e
+ .
- a,c.e Figure 3-11. Thermal Striping Fluctuation l
-sioe mn i ia 3-25
i a.c.e t
9-n
+
-Figure 3-12. Thermal Striping Temperature Distribution t
h i
sioo.,emei ie 3 26
SECTION 4.0 DISPLACEMENTS AT SUPPORT LOCATIONS The Cook plant specific pipe displacements along the surge lines were
~
calculated under the thermal stratification loads. Table 4-1 shows the maximum values of the pipe displacements at all pipe whip restraint locations in the Unit 1 surge line. These data are also applicable for Unit 2 with proper refiection of the coordinates.
All pipe displacements listed in Table 4-1 have been verified by as built gaps provided by walk down, to ensure that enough gaps are maintained between the pipe outside surface and the whip restraint surface so that the pipe will be free to move during all normal and stratified thermal conditions.
S T
4 S
4 m o. m uniio 41
p u <
-s TABLE 4-1..
, ,. Cook Unit 1 Maximum Pipe Displacement at Restraint Locations
~Under Thermal Stratification ,/
Displacementi-(in)*
Support- : Node DX DY D2 a,c.e y
?
i-4 f
g t*E These displacements arelin the Coordinate' System in Figure'.3-2.. For Unit-2 Displacements-:(in.the same plant global coordinate system as shown in Figure 312) the sign of_DX must be reversed.'DY and^DZ are the sams as.-
~ shown : ab~ove. -
g i t
em mim io .
- 4-2 l
- -
-4 g SECTION 5.0 ASME SECTION 111 FATIGUE USAGE FACTOR EVALVATION 5.1 Methodology Surge line fatigue evaluations have typically beon performed using the methods of ASME Section 111, NB-3600 for all piping components (--------------------
................... ,.__.................................... 3a,c.e Because of the nature of the stratification loading, as well as the magnitudes of the stresses produced, the more detailed and accurate methods of NB-3200 were employed using finite element analysis for all loading conditions.
Applichtion of these methods, as well as specific interpretation of Code stress values to evaluate fatigue results, is described in this section.
Inputs to the fatigue evaluation included the transients developed in section 2.0, r.nd the global leadinga and resulting stresses obtained using the me. nods described in section 3.0. In general, the stresses due to stratification were categorized according to the ASME Code aathods and used to evaluato Code stresses and fatigue cumulative usage factors, it should be noted that, [---
....._.............. ........................_........)a,c,e 5.1.1 Basis The ASME Code, Sectier. III,1986 (Reference 4) Edition was used to evaluate fatigue on surge lines with stratification loading. This was based on the requirement of NRC Bulletin 88-11 (Appendix B of this report) to use the
. " latest ASME Section III requirements incorporating high cycle fatigue".
voovemn;ic '
5-1
Specific requirements for' class 1 fatigue evaluation of piping components are
.given in'NB-3653.- These requirements must be. met for Level A and level B type !
loadings according to NB-3653 and NB-3654.
lac ording_to NB-3611 and NB-3630, the methods of NB-3200 may be used in lieu of the NB-3600 methods. This approach was used to evaluate the surge line
-cog onents under stratification loading. Since the NB-3650 requirements and equations correlate to those in NB-3200, the results of the fatigue evaluation are reported'in terms of the NB-3650 piping stress equations. These equations and requirements are summarized in' Tables 5-1 and 5-2..
LThe rgethods used to evaluate these requirements for the surge line components are-described in the following-sections.
-5.1'.2 Fatigue Stress Equations
^
Stress Classification-The stresses-in a' component are classified in the ASME Code based on the
-nature of the stress, the loading that causes the stress,-and the geometric '
characteristics that influence the' stress. This classification determines the
-acceptable-limits on the stress values and, in terms of NB-3653, the respective equation where the stress should be included. Table-NB-3217-2 lprovidesguidanceforstressclassification-inpipingcomponents,whichis
- reflected in. terms of the NB-3653 equations.
.The-terms in Equations 10, 11, 12-and 13 include stress indices which adjust nominal' stresses to account for secondary and peak _ effects _for a given component. Equations-10,;12 and 13 calculate secondary stresses, which are obtained.from' nominal values-using stress indices C1, C2, C3 and C3' for pressure, moment-and thermal transient stresses. Equation 11 includes the K1, K2 and K3 indices in-the pressure,. moment and thermal transient stress terms g in. order to represent peak stresses caused by local concentration, such as -
notches and weld effects. The NB-3653 equations use simplified formulas.to um.mn i ie 5-2
4 '
determine nominal stress based on straight pipe dimensions. (--------
..............................................)a.c.e For the RCL nozzles, three dimensional (3-D) finite element analysis was used as described in Section 3.0. (-------------------------------------------
..............................................................ja,c,e Classification of local stress due to thermal stratification was addressed with respect to the thermal transient stress terms in the NB-3653 equations.
Equation 10 includes a Ta-Tb term, classified as "Q" stress in NB-3200, which represents stress due to differential thermal expansion at gross structural discontinuities. -(-----------------------------------------------------------
.............................._____ ....___.ja,c,e The impact of this on l-
-the selection of components for evaluation is discussed in Section 5.1.3.
l l
i e mwo2mu 5-3
+
- x. .
o
. Stress Combinations-The. stresses in a given component _due-to pressure, moment and local thermal '-
. stratification loadings were calculated using.the finite element models described in Section 3.0. [--------------------------------------------
1 e
..............;)a.c.e ;This was done for specific components as follows: ,
m.
[.. :.................................................................. ->
S k
-+
+ < ...................................................................
t
~.ti a .
na- .. ................................................................
'. E:
3 . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
........................................t +f.................... g g.
e 5 , .
L 1. ! .. ......... .....a.......................... ................. L i
-1
'....................................................a......
a.
-A'
(- / , ..............s.........................................
.................e..................................
...............................................................a*CeG- .
- k. b e
f c'
I .
7
~ h '
.. g bm+v ~~mM+ >==w> e w F) 6 .r"
~
[.. ..............................................................
+
.....................................................,a,c,e
- J From the stress profiles created, the stresses for Equations 10 and 11 could
- be determined for any point in the section. Experience with the geometries and loading showed that certain points in the finite element models consistently produced the worst case fatigue stresses and resulting usage factors, in each stratified axial location. (------------------------------
.......m......................__..e. ___...............................
..................................e.......................m............e..
t I
___...................................................) ace , .
i I
O mwune, ia 5-5
Ecuation 12 Stress Code Equation 12 stress represents the maximum range of stress due to thermal expansion moments as described in Section 3.2. This used an enveloping ,
approach, identifying the highest stressed location in the model. By evaluating the worst locations in this manner, the remaining locations were inherently addressed.
Equation 13 Stress Equation 13 stress, presented in Section 3.2, is due to pressure, design mechanical loads and differential thermal expansion at structural discontinuities. Bassd on the transient set defined for stratification, the design pressures were not significantly different from previous design transients. Design mechanical loads are defined as deadweight plus seismic OBE loads.
The "Ta-Tb" term of Equation 13 is only applicable at structural ,
discortinuities. (-------------------------------------------- ------
..........................................................................)a,c.e Thermal Stress-Ratchet The requirements of NB-3222.5 are a function of the thermal transient stress and pressure stress in a component, and are independent of the global moment loading. As such, these requirements were' evaluated for controlling components using applicable stresses due to pressure and stratification transients.
4 4
stoovostistio 5-6 l
l
Allowable Stresses Allowable stress, Sm, was determined based on note 3 of Figure NB-3222-1. For secondary stress due to a temperature transient or thermal expansion loads
(" restraint of free end deflection"), the value of Sm was taken as the average of the Sm values at the highest and lowest temperatures of the metal during the transient. The metal temperatures were determined from the transient definition. When part of the secondary stress was due to mechanical load, the value of Sm was taken at the highest metal temperature during the transient.
5.1.3 Selection of Components for Evaluation Based on the results of the global analyses and the considerations for controlling stresses in Section 5.1.2, (--------------------------------------
......................................)a,c.e The method to evaluate usage factors using stresses determined according to Section 3,0 is described below.
5.2 Fatigue Usage Factors Cumulative usage factors were calculated for the controlling components using the methods described in NB-3222.4(e), based on N3-3653.5. Application of these methods is summarized below.
Transient Lt.adcases and Combinations From the transients described in Section 2.0, specific loadcases were
-developed for the usage evaluation. [------------------------- ------------
.......................____.......................................__.)a,c e
. Each loadcase was assigned the number of cycles of the associated transient as defined in Section 2.0. These were input to the usage factor evaluation,
. along with the stress data as described above, voovowes ,o 57
L-Usage factors were calculated at controlling locations in-the component as follows:
-1) Equation 10, Ke, Equation 11 and resulting Equation 14 (alternating stress - Salt) are calculated as described above for every possible-combination of the loadsets.
2)- For each value of Salt, the design fatigue curve was used to determine the maximum number of cycles which would be allowed if this type of cycle were the only one acting. These values, N3, N ...N ,.were determined from Code Figures I-9.2.1 and I-9.2.2, 2 n curve C, for austenitic stainless steels.
-3) Using the actual cycles of each transient loadset, ny, n 2 *"n' calculate the usage factors U12 ' U ...U n II
- Ui * "i/N g. This is done for all possible combinations. Cycles are used up for each
-combination inlthe order of decreasing Salt. When N g is greater than'10 11 cycles, the value of V4 is taken as zero.
[................................................................. ,
c................)a,c,e-
-4) The cumulative usage factor, Ucum, was calculated'as Ucum = U1+
U2 + ... + U .n.To'this was added the usage factor due to thermal-striping, as described below, to obtain total Ucum. The-
-Code allowable value is 1.0.
$100s/02 titM0 5-8
5.3 Jatigue Due to-Thermal Striping
. The usage factors calculated using the methods of Section 5.2 do not include the effects of thermal striping. (-------------------------------------
...........................ja,c.e Thermal striping stresses are a result of differences between the pipe inside surface wall and the average through wall temperatures which occur with time, due to the oscillation of the hot and cold stratified boundary. This type of stress is defined as a thermal discontinuity peak stress for ASME fatigue analysis. The peak stress is then used in the calculation of the ASME fatigue usage factor.
[.......................................................
s
Ja,c.e lhe methods used to determine alternating stress intensity are defined in the ASME code. Several locations were evaluated in order to determine the location where stress intensity was a maximum.
k-sico,m noi in 5-9
Thermal striping transients are shown at a AT level and number of cycles. The striping AT for each cycle of every transient is assumed to attenuate and follow the slope of the curve shown on figure 5-2. Figure 5-2 is conservatively represented -
by a series of 5 degree temperature steps. Each step lasts (--]a c.e seconds.
Fluctuations are then calculated at each temperature step. Since a constant frequency of [------Ja,c.e is used in all of the usage f actor calculations, the total fluctuations per step is constant and becomes:
[.............................................ja,c,e Each stripi.,g transient is a group of steps with (----Ja,c,e fluctuations per s t ep,. For each transient, the steps begin at the maximum AT and decreases by
(--Ja.c,e steps down to the endurance limit of AT equal to (-----Ja,c e The cycles for all transients which have a temperature step at the same level were added together. This became the total cycles at a step. The total cycles were multiplied by (----)a.c.e to obtain total fluctuations. This results in total fluctuations at each step. This calculation is performed for each step plateau frem (-------------------------------Ja c.e to obtain total fluctuations. Allowable fluctuations and ultimately a usage factor at each plateau is calculated from the stress which exists at the AT for each step. ,
The total striping usage factor is the sum of all usage factors from each plateau.
The usage factor due to striping, alone, was calculated to be a maximum of
[-----Ja,c,o This is reflected in the results to be discussed below.
5.4 Fatigue Usage Results NPC Builetin 88-11 (14) requires fatigue analysis be performed in accordance with the latest ASME III requireraents incorporating high cycle fatigue and thermal stratification' transients. ASME fatigue usage factors have been calcuiated considerir.g the phenomenon of thermal stratification and thermal striping at various locations in the surge line. Total stresses included the ,
me,u ns t")
5-10 l
[............................................................................
..........................ja.c.e The total stressos for all transients in
=
the bounding set were used to form combinations to calculate alternating stresses and resulting fatigue damage in the manner defined by the Code. Of this total stress, the stresses in the 14 inch schedule 160 pipe due to
[.........................................................................
.................................................................ja,c,e The maximum usage factor on Cook surge lines occurred at (-------------------
...)a.c.e it is also t luded that the Cook pressurizer surge nozzles will meet the code stress allowables under the thermal stratification loading from tae surge line and the transient detailed in reference (13), and meet the fatigue usage requirements of ASME Section 111, with a maximum cumulative usage factor equal to 0.1.
b e
~
I L*
l 5100s 12 191-10 5-11 l
a- . ..
1j TABLE 5'1 ,
CODE / CRITERIA 1,- o ASME B&PV Code, Sec. III 1986 Edition ~I
' - -NB3600
-: NB3200 o : Level A/B~ Service Limits w! -
l-Primary Plus Secondary Stress Intensity 1 3Sm-(Eq. 10)
~
Simplified Elastic-Plastic Analysis
~
-Expansion Stress, S, 5 J'm (Eq. 12) - Global Analysis
-~ ! Primary Plus Secondary Excluding Thermal Bending < 35m
.m .
(Eq. 13)- -
W --
Elastic-Plastic Penalty Factor 1.0 5- K, 3 3.333
- Peak- Stress '(Eq. - 11)/ Cumulative ' Usage Factor (Ucum)
-Salt
- K ep S /2 -_(Eq. 14).
- Design Fstigue Curve nucum-1~1.0,
=
I 4
l-1 i
[
r
$ g
. . - _ -_.m . - . _ _- . . .
Y o.3 s. .
1 TABLE 5-2
SUMMARY
0F ASME FATIGUE REQUIREMENTS ;
Parameter l- Descrfption- Allowable <
' (if applicable)
Equation 101 Primary.plus' secondary stress-intensity; < 3Sm lifexceeded,simplifiedelasticplastic '
analysisLmay.be performed .
K,- Elastic plastic penalty factor; required ,
for simplified elastic plastic-analysis
- = when'Eq. 10 is exceeded; applied to
- alternating stress intensity'
[Eq0ation 12 -ExpansionJstress;-required for simplified < -3Sm . -
Lelastic plastic analysis when Eq. 10 is exceeded ~
Eq0ation'13 Primary plus secondary.. stress intensity < 35m excludinguthermal bending. stress;. required for1 simplified elastic plastic 1 analysis: '
when Eq._10 is exceeded
- Therm'al- JLimit on radial-thermal gradient stress-to
- Stress. prevent 1 cyclic distortion; required'for use
'Ratcheti _ < of Eq.l13L
- Equation 11 Peak stress intensity - Input to Eq. 14-Equation-14L . Alternating stress
- intensity - Input' to Ucum Ucum : Cumulative usage factor (fatigue damage)' < 1.0-
,- m
/sanearneuo 13 .
f f -
rm - , + - - , ..m..,e . ~ -,-s ...n-8-- e -s- - - - e -
a,c.e t
4 f
4
-Figure 5-1. Striping Finite Element Model v oogainecio 5-14
(- .
-S'. 9
~
~~
a,c.e Figure-5-2. Attenuation of Thermal Striping Potential by Molecular Conduction (Interface Wave Height of One Inch) sioo.miisi io 5-15
SECTION 6.0
SUMMARY
AND CONCLUSIONS The subject of pressurizer surge line integrity has been under intense investigation since 1988. The NRC issued Bulletin 88-11 in December of 1988, but the Westinghouse Owners Group had put a crogram in place earlier that year, and this allowed all members to make 6 timely response to the bulletin.
The Owners Group programs were completed in June of 1990, and have been followed by a series of plant specific evaluations. This report has documented the.results of the plant specific evaluation for Cook Units 1 and 2.
Following the general approach used in developing the surge line stratification transients for the WDG, a set of transients and stratification profiles were developed specifically for Cook Units 1 and 2. A study was made of the historical operating experience at the Cook Units, and this information, as well as plant operating procedures and monitoring results, was used in development of the transients and profiles.
The results of this plant specific analysis demonstrated acceptance to the requirements of the ASME Code Section !!I, including both stress limits and fatigue usage, for the licensed operation (32 Effective Full Power Years) of
-the units. The displacements and support loadings for the surge line have been verified and found acceptable _(14). This report demonstrates that Cook Units 1 and 2 have now completely satisfied the requirements of NRC Bulletin 88-11.
xoovemene 6-1
SECTION
7.0 REFERENCES
l'. : Coslow, B. J., at al., " Westinghouse Owners Group Bounding Evaluation for
.. Pressurizer Surge Line Thermal Stratification" Westinghouse Electric Corp.
1 WCAP-12277,-(proprietary class 2) and WCAP-12278 (non proprietary), June 1989.
- 2. Coslow, B. J., et al., Westinghouse Owners Group Pressurizer Surge Line Thermal. Stratification Pror, ram MUHP-1090 Summary Report," Westinghouse i Electric Corp. WCAP-12506 (proprietary-class 2) and WCAP 12509 j (non proprietary), March 1990.
- 3. Coslow, B. J., et al., " Westinghouse Owners Group Pressurizer Surge Line Thermal Stratification Generic Detailed Analysis Program MUHP-1091 Summary Report," Westinghouse Electric Corp. WCAP-12639 (proprietary class 2) and WCAP-12640 (non proprietary), June 1990.
.. 4._ ASME B&PV Code Section III, Subsection NB, 1986 Edition.
-* 5. " Pressurizer Surge Line Thermal Stratif'. cation," USNRC Bulletin-88-11, December 20, 1988.
- 6. .AEP Letter, from George P. Roulett to John C. Hoebel, 6-8-90.
- 7. 'AEP Letter, from'T. R.- Satyan-Sharma to John-C. Hoebel, 8-21-90.
- 8. " Investigation of Feedwater Line Cracking-in Pressurized-Water Reactor Plants,":WACP-9693, Volume'1,-June 1990 (Westinghouse' Proprietary Class 2).
- 9. Woodward, W. S., " Fatigue of LMFBR Piping due to Flow Stratification,"
ASME Paner 83-PVP-59,-1983.
- 10. Fujimoto, T., et al., " Experimental Study' of Striping at the Interface of y Thermal Stratification" in Thermal Hydraulics in Nuclear Technology, K. H.-Sun, et al., (ed.) ASME, 1981, pp. 73..
5100s/021H1 10 '
7.{
- 11. Holman,- J. P., Heat Transfer, McGraw Hill Book.Co.,1963.
- 12. Yang, C. Y., " Transfer Function Method For Thermal Stress Analysis: ,
Technical Basis," Westinghouse Electric Corporation WCAP-12315 (ProprietaryClass2). -
- 13. Series 84 Pressurizer Stress Report, Section 3.1, Surge Nozzle Analysis, December 1974, Westinghouse Proprietary,
- 14. AEP l.etter, G. P. Roulett to P. Strauch, 2/14/91.
e b
e 0
voo.,ome r io 72
APPENDIX A LIST OF COMPUTER PROGRAMS This appendix lists and summarizes the computer codes used in the pressurizer surge line thermal stratification. The codes are:
- 1. WECAN
- 2. STRFAT2
- 3. ANSYS
- 4. FATRX/ CMS A.1 WECAN A.1.1 Description WECAN is a Westinghouse-developed, general purpose finite element program, it contains universally accepted two-dimensional and three-dims r, W el isoparametric elements that can be used in many different types rf finite element analyses. Quadrilateral and triangular structural e n,en are used i
for plane strain, plane stress, and axisymmetric analyses, Brick and wedge structural elements are used for three-dimensional analyses. Companion heat conduction elements are used for steady state heat conduction analyses and transient heat conduction analyses.
A.1.2 Feature Used The' temperatures obtained from a static heat conduction analysis, or at a specific time in a transient heat conduction analysis, can be automatically input to-a static structural analysis where the heat conduction elements are replaced by corresponding structural elements. Pressure and external loads nn also be include in the WECAN structural analysis. Such coupled r 1-stress analyses are a standard application used extensively on an
,try wide basis.
sico.m neno A-1
A.I.3 Program Verification Both the WECAN program and input for the WECAN verificatior, problems, currently numbering over four hundred, are maintained under configuration ,
control. Verification prcblems include coupled thermal-stress analyses for the quadrilateral, triangular, brick, and wedge isoparametric elements. These problems are an integral part of the WECAN quality assurance procedures. When a change is made to WECAN, as part of the reverification process, the configured inputs for the coupled thermal-stress verification problems are used to reverify WECAN for coupled thermal-stress analyses.
A.2 ,.STRFAi2 A.2.1 Descriotion STRFAT2 is a program which computes the alternating peak strest on the insioe surface of a flat plate and the usage factor due to striping on the surface.
The program is applicable to be used for striping on the inside surface of a ,
pipe if the program assumptions are considered to apply for the particular pipe being evaluated. .
For striping the fluid temperature is a sinusoidal variation with numerous cycles.
The frequency, convection film coefficient, and pipe material properties are input.
The pregram computes maximum alternating stress based en 'he maximum difference between inside surface sk~in temperature and the average through wa'll temperature.
sno..un
A-2
A.2.2 Feature Used-The program is used to calculate striping usage factor based on a ratio of actual cycles of stress for a specified length of time divided by allowable cycles'of stress at maximum the alternating stress level. Design fatigue curves for several materials are contained into the program. However, the user has the option to input any other fatigue design curve, by designating -
that the fatigue curve-is to be user defined.
A.2.3 Program Verification
,STRFAT2 is verified to Westinghouse procedures by independent review of the stress equations and calculations.
A.3 ANSYS:
A.3.1 Description t ANSYS-is a public domain,= general purpoce finite element code.
. A~.3.2 Feature Used The ANSYS elements used for tha analysis of stratification effects in-the
~
surge:line are STIF 20 (straight pipe), STIF 60 (elbow and bends) and STIF14 .
(spring-damper-for. supports).
A 3.3 Program Verification As- described in sectior. 3.2, the application of ANSYS for-stratification has -
~
been independently verified by' comparison to-WEST 0YN (Westinghouse pipir.g]-
-analysis cocel and WECAN (finite element code). The results from ANSYS are also Lverified againsi closed form solutions for simple beam configurations.
y :.
ocovemeiao A-3 l-
,. . , , , . ,J, , - -
)
A.4 FATRK/ CMS A.4.1 Descr gi; e -
FATRK/ CMS is a Westinghouse d3veloped computer code for fatigue '*acking .
(FATRK) as used in the Cycle Monitoring System (CMS) for structural components l of nuclear power plants. The transfer function method is used for transient I thermal stress calculations. Thebendingstresses(duetoglobal -l stratification effects. ordinary the mal expansion and seismic) and the pressure stresses are also. included. The fatigue usage factors are evaluated in accordance with the. guidelines given in the ASME Boiler and Pressure Vessel Code,'Section !!!, Subsections NB-3200 and NB-3600. ,
The code can be used both as a regular analysis program or an on-line monitoring device. .
A 4.2 Feature Used FATRK/ CMS is'used as an analysis program for the present application. The *i input data which include the weight functions for thermal t, tresses, the unit ,l i bending stress, the unit pressure stress, the bending moment vs.
stratification-temperatures, etc are prepared for all locations and geometric conditions. These data, as stored in the independent files, can be
. appropriately retrieved for required analyses. The transient data files contain the time history of-temperature.' pressure, number of occurrence,-and additional condition necessary for data flowing. The program prints out the-total usage f actors, and the transients pairing information which determine the stress range magnitudes-and number of cycles. The detailed stress data may'also be printed.
~
A 4.3 Program Verification FATRK/ CMS is' verified according to Westinghouse procedures with several. levels ,
of independent calculations as described below * (1) transfer function method e<
me,e mo A-4
of thermal strsases as compared with direct WECAN finit6 element analyses.
(2) combined stresses as compared with hand calculations and WECEVAL,
. WCAP-9376 analyses. (3) The fatigue usage factor results as compared with WECEVAL analyses. '
O e O e
s k
D sioo.mnes io A-5
APPENDIX B
+
USNRC BULLETIN 88-11 In December of 1988 the NRC issued this bulletin, and it has led to an extensive investigation of surge line integrity, culminating in this and other plant specific reports. The bulletin is reproduced in its entirety in the pages which follow.
e 9
h
$100s /02116: 10 g.}
i
t
( OM8 No. 3150-0011 NRC8 88 11 UNITED STATES NUCLEAR REGULATORY COMMIS$10N OFFICE OF NUCLEAR REACTOR REGULATION ,
WASHINGTON 0.C. 20555 December 20, 1988 e!
NRC BULLETIN NO. 88-11: PRES $URIZER SURGE LINE THERMAL STRATIFICATION :
i Addressees:
All holders of operating licenses or construction pemits for pressurized water teactors (PWRs). .
F. '?.E2211 ,
LThe purpose of this bulletin is to (1) request that addressees establish and e implement a program to confim pressurizer surge line integrity in view of the i occurrence of themel stratification and (2) require addressees to infom the staff of the actions,taken to resolve this issue.
Descriotion of Circumstances:
The licensee for the Trojan plant has observed unexpected movement of the .
pressurizer surge line during inspections perfomed at each refueling outage ,,
since 1982, when monitoring of the line movements began. During the last ,
refueling outage, the licensee found .that.in addition to unexpected gap clo-sures in the pipe whip restraints, the piping actually coattacted two re- 'i straints. Although.the licensee had repeatedly adjusted shims and gap sizes ,
based _ on analysis of various postulated conditions, the problem-had not been '
. resolved. The most recent investigation.by the licensee confimed that the movement of piping was caused by thermal stratification in the line. This phenomenon was not considered in the original piping design. -On October 7.- *
~1988, the staff issued Infomation Notice 88 80, " Unexpected Piping Movement Attributed to Themal . Stratification," regarding the Trojan experience and indicated that further generic contaication may be forthcoming. The licensee 1 for leaver Valley 2 has also noticed unusual snubber movement and significantly larger-than-expected surge line dispiacement' during power ascension.
The concerns ratsed b NRC Su11etins 79-13Revision.2, (y the above-observations.are dated' October 16, 1979), similar to .thoseindescribed in
" Cracking i
Feedwater System Piping" and 88 08-(dated June 22, 1988), " Thermal Stresses-in !
Fiping Connected to Reactor Coolant Systems."
l 8812150118- i B-2
-.-..,..,._m.._.. .-_...-..i..,_m..~.-m,,,....,,,,. . . _ _ . , . , , , , - . - . . , .,,.-.-,c,. ,.t
. . . . - < . ...m, . . , , . - . . . .m., .m.- . ~ _ _
NRCB 88 11 December 20. 1988 Page 2 of 6 '
l
, Discussion: i
- Unexpected piping movements are highly undesirable because of potential high piping stress that may exceed design limits for fatigue and stresses. The j problem can be more acute when the piping expansion is restricted, such 45 through contact with pipe whip restraints. Plastic deformation can result, which can lead to high local stresses, low cycle fatigue and functional im-pairment of the line. Analysis performed by the Trojan licensee indicated that thermal stratification c.ccurs in the pressurizer surge line during beatup, cocidown, and steady ~icate operations of the plant.
During a typical plant heatup, water in the pressurizer is heated to abt.
440*F; a steam bubble is then formed in the pressurizer. Although the euct phenomenon is not thoroughly understood, as the hot water flows (at a very low flowrate) from the pressurizer through the surge line to the hot-leg piping, the hot 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 (see Figure 1).
1he differential temperature could Le as hi conditions during typical plant operations.gh as Under 300*F, based this on expected condition, differential thermal expansion of the pipe metal can cause the pipe to deflect signifi-cantly.
For the specific configuration of the pressurizer suae line in the Trojan 4
plant, the line deflected <iewnward and when the sut line contacted two pipe whip restraints, it underwent plastic deformation, r:=esulting in pemanent defomation of the pipe.
The Trojan event demonstrates that thermal stratification in the pressurizer surge line causes unexpected piping movement and potential plastic defomation.
The licensing basis according to 10 CFR 50.55a for all PWRs requires that the licensee meet the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Sections III and XI and to reconcile the pipe stresses and fatigue evaluation when any significant differences are observed between measured data and the analytical results for the hypothesized conditions. Staff evaluation indicates that the thennel stratification phenomenon could occur in all PWR surge surge line.lines and may invalidate the analyses supporting the integrity of the The staf f's concerns include unexpected bending and thermal striping (rapid oscillation of the thennel boundary interface along the piping inside surface) as they affect the overall integrity of the surge lina for its design life (e.g., the increase of fatigue).
Actions Reouested:
Addressees are requested to_ t'.ke the following actions:
- 1. For all licensees of operating PWRs:
4
- a. 1.icensees are requested to conduct a visual inspection (ASME,Section XI, VT-3) of the pressurizer surge line at the first available cold shutdown after receipt of this bulletin which exceeds seven days.
B-3
NRCB 88 11 December to,1988 1 Page 3 of 6 This inspection should determine any gross discernable distress or structural damage in the entire pressurizer surge line, including piping, pipe supports, pipe whip restraints, and anchor bolts. '
- b. Within four months of receipt of this Bulletin, licensees of plants in operation over 10 years (i.e., low power license prior to l January 1, 1979) are requested to demonstrate that the pressurizer l surge line meets the applicable design codes' and other FSAR and !
regulatory comitments for the licensed life of the plant, consider- :
ing the phenomenon of thermal stratification and thermal striping in i the fatigue and stress evaluatioM. This may be accomplisheo by I performing a plant specific or generic bounding analysis. If the latter option is selected, licensees should demonstrate applicability of the referenced generic bounding analysis. Licensees of plants in operation less than ten years (i.e., low power license after January 1, 1979), should complete the foregoing analysis within one year of receipt of this bulletir. $1nce any piping distress observed by addressees in performing action 1.a may affect the analysis, the licensee should verify that the bounding analysis remains valid. If the opportunity to perform the visual inspection in 1.4 does not occur within the periods specified in this requested item, incorpora-tion of the results of the visual inspection into the analysis should be performed in a supplemental analysis as appropriate.
Where the analysis shows that the surge line does not meet the requirements and licensing comitments stated above for the duration of the license, the licensee should submit a justification for continued operation or bring the plant to cold shutdown, as appropri-ate, and implement items 1.c and 1.d below to develop a detailed analysis of the surge line,
- c. If the analysis in 1.b does not show compliance with the recuirements and licensing consvitments stated therein for the duration of the operating license, the licensee is requested to obtain plant specific data on thermal stratification, thermal striping, and line deflec-tions. The licensee may choose, for example, either to install instruments on the surge line to detect temperature distribution and thermal movements or to obtain data through collective efforts, such as from other plants with a similar surge line design. If the latter option is selected, the licensee should demonstrate similarity in geometry and operation,
- d. Based on the applicable plant specific or referenced data, licensees are recuetted to update their stress ano fatigue analyses to ensure compliance with applicable Code requirements, incorporating any observations from 1.a above. The analysis should be completed no later than two years after receipt of this bulletin. If a licensee i
i
- Fatigue analysis should be performed in accordance with the latest ASME Section 111 requirements incorporating high cycle fatigue.
B-4 _- . _ __ _
- O NRCB 88-11 December 20, 1988 Page 4 of 6 is urable to show compliance with the applicable design codes and other FSAR and regulatory corritmei,?s, the licensee is requested to O submit a justification for continued operation and a description of the proposed corrective actions for effecting long term resolution,
- 2. For all applicants for PWR Operating Licenses:
- a. Before issuance of the low power license, applicants are requested te demonstrate that the pressurizer surge line meets the applicable design codes and other FSAR and regulatory comitments for the licensed life of the plant. This may be accomplished by performing a plant-specific or generic bounding analysis. The analysis should include consideration of thermal stratification and thermal striping to ensure that fatigue and stresses are in compliance with applicable code limits. The analysis and hot functional testing should verify that piping themal deflections result in nc adverse consequences.
such as contacting the pipe whip restraints. If analysis or test results show Code noncompliance, conduct of all actions specified below is requested,
- b. Applicants are requested to evaluate operational alternatives or piping modifications needed to reduce fatigue and stresses to acceptable levels.
- c. Applicants are requested to either monitor the surge line for the effects of thennal stratification, beginning with hot functional testing, or obtain data through collective efforts to assess the extent of thermal stratification, thermal striping and piping deflections,
- d. Applicants are requested to update stress and fatigue analyses, as necessary, to ensure Code compliance.* The analyses should be completed no later than one year after issuance of the low power license.
- 3. Addressees are requested to generate records to document the development and implementation of the program requested by items 1 or 2, as well as any subsequent corrective actions, and maintain these records in accor-dance with 10 CFR Part 50, Append 1x B and plant procedures.
Reporting Requirements:
- 1. Addressees shall report to the hRC any discernable distress and d aage observed in Action 1.a along with corrective actions taken or plans and schedules for repair before restart of the unit.
'If compliance with the applicable codes is not demonstrated for the full duration of an operating license, the staff may impose a license condition such that nonnal operation is restricted to the duration that compliance is actually demonstrated.
B-5
NRCB 88-11 1 December 20, 1988 Page 5 of 6
- 2. Addressees who cannot meet the schedule oescribed in Items 1 or 2 of -
Actions Requested are required to submit to the NRC within 60 days of receipt of tnis bulletin an alternative schedule with justification for ,
the requested schedule.
- 3. Addressees shall submit a letter witnin 30 days after the completion of these actions which notifies the NRC that the actions recuested in Items i lb, id or 2 of Actions Regrested have been performed and that the results are available for inspection. The letter shall include the justification for continued operation, if appropriate, a description of the analytical approaches used, and a sumary of the results.
Although not requested by this bulletin, addressees are encouraged to work collectively to address the technical concerns associated with this issue, as well as to share pressurizer surge line data and operational ex.nerience. In addition, addressees are encouraged to review piping in other systems which may experience thermal stratification and thermal striping, especially in light of the previously mentioned Bulletins 79-1.1 and 88-08. The NRC staff intends to review operational experience giving appropriate recognition to this phenome-non, so as to detemine if further generic comunications are in order.
The letters required above shall be addressed to the U.S. Nuclear Regulatory Comission, ATTN: Document Control Desk, Washington 0.C. 20555, under oath or dffinnation under the provisions of Section 182a, Atomic Energy Act of 1954, as amended. In addition, a copy shall be submitted to the appropriate Regional
- Administratur.
This request is covered by Office of Management and Budget Clearance Number i
3150-0011 which expires December 31, 1989. The estimated average burden hours is approximately 3000 person-hours per licensee response, including assessment of the new requirements, searching data sources, gathering and analyzing the data, and preparing the required reports. These estimated average burden hours pertain only to these- identified response-related matters and do not include the time for actual implementation of physical changes, such as test equipment installation or component modification. . The estimated average raciation exposure is approximately 3.5 person-rems per licensee response.
Comments on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Of fice of Management and Budget, Room 3208, New Execu-tive Offica Building, Washington, D.C. 20503, and to the U.S. Nuclear Regula-tory Commission, Records and Reports Management Branch Of fice of Administration and Resource Mdnagement Washington, D.C. 20555.
b 8-6
i e e l
NRC8 88-11 December 20, 1988 Page 6 of 6 l
If you have any questions about this matter, please contact one of the techni-
- cal contacts listed below or the Regional Administrator of the appropriate 1 regional office.
) <
/f a es .Rossi,kr'cto Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contacts: S. N. Hou, NRR (301)492-0904
- 5. S. Lee, NRR '
(301)492-0943 N. P. Kadambi, NRR (301)492-1153 Attachments:
- 1. Figure 1
- 2. List of Recently Issued NRC Bulletins 4
-l 2
'gs B-7 1:
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December 20 '966 Page i of l
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