ML20073P546
| ML20073P546 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/31/1991 |
| From: | Tilda Liu, Ching Ng, Tandon S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML18095A926 | List: |
| References | |
| WCAP-12913, NUDOCS 9105220073 | |
| Download: ML20073P546 (103) | |
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l Westinghouse Proprietary Class 3 HCAP-12913 Structural Evaluation of Salem Nuclear Plant Units 1 and 2 Pressurizer Surge Lines, Considering the Effects of Thermal Stratification March 1991 T. H. Liu C. K. Ng S. Tandon P. L. Strauch L. M. Valasek Verified by: ha v Verified by:
M./ ~A. Gray S V. V. Vora m s h Approved by: - A~~ bk[- Approved by: c b
. S. S.,Palusamy, Manager R. B. Patel, Manager Diagnostics and Monitoring System Structural Analysis Technology and Development Work Performed under Shop Orders PJBP-964 and PJBP-145 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 e 1991 Hestinghouse Electric Corp.
5231s/041691:10 9105220073 910513 PDR ADOCK 05000272 P PDR
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TABLE OF CONTENTS Section ILtle Elge Executive Summary 111 1.0 Background and Introduction 1-1 1.1 Background 1-1
'! . 2 Description of Surge Line Stratification 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 and Operator Interviews 2-4 2.5 Historical Operation 2-7 2.6 Development of Heatup and Cooldown Transients 2-8 2.7 Axial Stratification Profile Development 2-11 2.8 Striping Transients- 2-13 3.0 Stress Analysis 3-1 3.1 Surge Line Layouts 3-1 3.2 Piping Sy: tem 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- 3-6 4.0 Displacements at Support Locations 4 5231s/041891:10 i
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TABLE OF CONTENTS-(Continued)
Section Title Eage 5.0 ASME Section III Fatigue Usage Factor Evaluation 5-1
. 5.1 Methodology 5-1 5.2 Fatigue Usage Factors 5-7 5.3 Fatigue Due to Thermal Striping 5-9 5.4 Fatigue Usage Results 5-10 6.0 Summary and Conclusions 6-1 7.0 - References 7-1 Appendix A Computer Codes A-1 Appendix B USNRC Bulletin 88-11 B-1 Appendix C Transient Development Details C-1 l
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EXECUTIVE
SUMMARY
Thermal stratification has been identified as a concern which can affect the structural in:egrity of piping systems in nuclear plants since 1979, when a leak was discovered in a PWR feedwater line. In the pressurizer surge line, stratificati(n can result from the difference in densities between the hot leg water and ge1erally hotter pressurizer water. Stratification with large 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 the Salem 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. That program provided the technical basis for the plant specific transient development reported here for the Salem Nuclear Plant Units 1 and 2.
This transient development utilized a number of sources, including plant operating procedures, surge line monitoring data from other similar units, 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 development.
The existing configuration for both Salem units have been analyzed in this WCAP. The results of the analyses indicate small contacts at one pipe whip restraint and " bottoming-out" of variable spring hangers for the 320*F AT.
ASME Code allowables were shown to be exceeded-for both Salem units under this existing configuration. However, by taking credit for insulation crushability '
and modifying spring hangers to avoid " bottomed-out" condition, the ASME Code 5231s/041891:10 111
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stress limits and cumulative usage factor requirements have been shown to be acceptable for the remainder of the licensed operation of both units. No whip restraint gap modifications are necessary to show Code acceptance if the spring can travel allowances are satisfied.
The spring hanger and whip restraint displacements resulting from normal thermal and stratification have also been provided in Section 4 for both the existing and future support configurations, for future verification of proper gaps in the pipe whip restraints and sufficient travel allowance in the spring hangers, to allow for pipe movement at all thermal conditions. The structural analysis which resulted in this recommendation is discussed in Sections 3 and 4.
This work has led to the conclusion that Salem Units 1 and 2 are in full compliance with the requirements of NRC Bulletin .88-11, provided tne spring hanger modifications discussed on the following page are implemented, and a limitation of 320*F on the temperature differential between the pressurizer liquid space and RCS hot leg is imposed.
Until the spring hanger modifications are implemented, the existing Justification for Continued Operation (JCO), Reference 14, is still valid for the Salem units, in view of this plant specific analysis.
The existing JC0 was based on the assessment without detailed reanalysis and indicated acceptability of continued operation for an additional ten heatup and cooldown cycles from the date of the Westinghouse Owner's Group (H0G) bounding evaluation, i.e. June 1989.
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SUMMARY
OF RESULTS, AND STATUS OF 88-11 QUALIFICATION Salem Unit 1 Salem Unit 2 ODerating History through 1990 Date of commercial operation 6/30/77 10/13/81 Years of water-solid heatups 0 0-Years of steam-bubble heatups 14 10 System delta T limit (assumed) 320'F 320'F Number of 320*F exceedances None One Maximum Stress and Usaae Factor Results Ecuation 12 stress / allowable * (ksi) 49.6/52.9 50.0/52.9 Fatigue usage / allowable 0.6/1.0 0.6/1.0 Pressurizer Surge Nozzle Results Maximum stress intensity range / 47.3/57.9 47.3/57.9 allowable (ksi)
Fatigue usage / allowable 0.50/1.00 0.50/1.00 Remainina Actions by Utilitv Spring hanger modification required Allow sufficient travel allowance on both units (Table 4-1A)
Status of 88-11 Reauirementt All analysis requirements met with above modi fications
'Results for future configuration. See Table 3-2 for results for existing configuration l
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SECTION
1.0 BACKGROUND
AND INTRODUCTION The-Salem Nuclear Plant Units 1 and 2 are four loop pressurized water 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 stratif_ication of the pressurizer surge lines for both of these units.
The operation of a pressurized water reactor requires the primary coola.7t i
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% cf the volume depending on the plant 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 pressurizer level varies between 22% and 50% depending on reactor thermal power. 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 some 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. The surge line of Unit 1 is 14 inch schedule 140, and the surge line of Unit 2 is 14 inch schedule 160. Both are stainless steel.
1.1 Backaround During the period from 1982 to 1988, a number of utilities reported unexpected movement of the pressurizer surge line, as evidenced by crushed insulation, 5231s/041691:10 1-1
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l gap closures in the pipe whip restraints, and in some cases unusual snubber movement. Investigation of thic 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 shutdown. Although no through wall cracks were found in surge lines, inservice inspections 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 liner 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 estaolish and implement a program to confirm the integrity of the pressurizer surge line. 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 H0G plants to' those for which a detailed
-plant specific analysis had been performed. Since this evaluation was unable to demonstrate the full design life for all plants, a generic justification for continued operation was developed for use by each of the HOG 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 more realistic 5231s/041691:10 1-2
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approach than could be obtained from a single bounding analysis for all plants, and the results were published in June 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 Salem Units 1 and 2.
1.2 Dngig.t)sn of Surge Line Thermal Stattf_i.ntion 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 pressurizer and the hot leg increases and as the insurge or outsurge flow rates decrease.
At power, when the difference in temperature between the pressurizer and hot 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 be 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 Staat of Work The primary purpose of this work was to develop transients applicable to the Salem units which include the effects of stratification and to evaluate these effects on the structural integrity of the surge lines. This work will therefore complete the dernonstration 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, operator interviews, and historical data on plant operation. This process is discussed in Section 2.
The resulting tran;ients were used to perform an analysis of the surge line, j wherein the existing support configuration was carefully modeled, and surge l line displacements, stresses, support loads 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 SURGE LINE TRANSIENT AND TEMPERATURE PROFILE DEVELOPMENT 2.1 General ADDroach i
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 from review of the plant operating procedures, operator interviews, monitoring data 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 the Salem Units 1 and 2 surge lines are provided in Table 2-1. For each of the transients the surge line fluid temperature was modified from the original design assumption of . .iform temperature to a stratified distribution, according to the predicted temperature differentials between the pressurizer and hot leg, as listed in the table. The transients have been characterized as either insurge/outsurges (I/O in the table) or fluctuations (F).
Insurge/outsurge transients are generally more severe, because they resolt in the greatest temperature change in the top or bottom of the pipe. Typical l
j temperature profiles for insurges and outsurges are shown in Figure 2-1.
l Transients identified as fluctuations (F) typically invo!ve low surge flow
! rates and smaller temperature differences betweer 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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f In addition to the plant specific operating history discussed above, the development of transients which are applicable to Salem Units 1 and 2 was based on the work already accomplished under programs completed for'the Westinghouse _0wners Group [1,2,3). In this work all the Westinghouse 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 Salem 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 advantage of the monitoring data collected during the WOG program, as well as historical operation data for the Salem Units. Each of these will be discussed-in the sections which follow.
2.2 System Desian InformatiDD 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 ASME Section III fatigue life assessment of-components).
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In the evaluation of surge line stratification, the transient events considered encompass the normal and upset design events defined in FSAR thapter s.2.
The total number of heatup-cooldown cycles (200) remains unchanged. However, 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 stratification Sffects, as described later.
2.3 Development of Normal and Voset_Tran_sjnts C
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3a ,c.e 2.4 BQDi.torina Results and Ooerator Interviews 2.4.1 Monitoring Monitoring information collected as part of the Westinghouse Owners Group l generic detailed analysis [3] was utilized in this analysis. The pressurizer surge line monitoring programs utilized externally mounted temperature sensors
- l. (resistance temperature detectors or thermocouples). The temperature sensors were attached to the outside surface of the pipe at various circumferential and axial locatians. In all cases these temperature sensors were securely clamped to the piping outer wall using clamps, taking care to properly insulate the area against heat loss due to thermal convection or radiation.
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l 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, many pressurizer surge line monitoring programs utilized displacement sensors mounted at various axial locations to detect horizontal and vertical movements, as shown in Figure 2-2. Typically, data was collected at (
]a.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.
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3 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 utility plant computer.
2.4.2 Operational Practices Based on a review of the Salem Units 1 and 2 heatup and cooldown operating procedures and operational interviews conducted at a number of WOG utilities, l
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it was determined that both units heat up and cool down in a manner similar to other plants that heat up with a steam bubble in the pressurizer. Heatups and cooldowns are used here to characterize plant operation because they represent the periods during which the temperature difference between the pressurizer and the hot leg is the greatest. . A brief description of the Galem heatup and cooldown procedures follows.
At the beginning of the heatup, the reactor coolant system (RCS) is filled and vented and a steam bubble is drawn in the pretsurizer with the RCC pressure less than 325 psig. As the RCS temperature is being raised, the pressure is limited to 375 psig for temperatures less than 312 deg F. The RCS pressure is maintained between 325 and 350 psig using the pressurizer heaters and pressurizer spray. The pressurizer level is maintained at 22% by centrolling charging and letdown flow. The remainder of the heatup is performed within the 100*F/ hour limit.
In terms of system delta temperatures, the cooldown is basically the reverse of heatup. During RCS cooldown, all reactor coolant pumps (RCP's) are running until the RCS temperature reaches 400*F at which time one RCP is shut down.
At RCS temperature of 350*F, one additional RCP is turned off. Hnen the RCS temperature reaches 350*F and the RCS pressure is less than 375 psig, the Residual Heat Removal (RHR) system is activated. Upon reaching 250*F, another RCP is turned off. When the RCS is less than 200 deg F, and the chemistry check is completed, the last RCP is turned off. At this point, the pressurizer pressure and, therefore, the cooldown are controlled by the auxiliary spray operation.
This analysis is based on the assumption of a maximum temperature difference between the pressurizer liquid space and the RCS hot leg (" system delta T") of 320*F for future operation. Public Service Electric and Gas Company must confirm this assumption for both Salem Units.
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Historical records from the plant (operator logs, st veillance test reports, etc.) were reviewed. From this review a maximum system ds ' ca T distr.bution and the number.of delta 1 exceedances of 320'F was obtained. Use of this information in the analysis is described in Section 2.' The results of the review are listed below, as percentages of the total number of past heatup and cooldown events for which data was available.
UnLt_1 Unit 2 Number of Number of System AT Heatups & % of Heatups & % of Hanne_1'D Cnoldowns lotal Coollow_n1 101A1
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This table does not include data prior to 1984 since none could be obtained. It was assumed that the operational practices from 1984 to present adequately represent the prior operation of the units. The highest observed system delta T was 325'F, which was the only exceedance of 320'F.
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j The above information was used to ensure that the transients' analyzed for the Salem units encompassed tue prior operating history of the plant.
l Comparison of this information-to the distribution used in the evaluation confirmed applicability to the Salem Units. This is illustrated in figure 2-3. [
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2.6 Develocment of HPJLtup and Con)down Tran j igntji 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 Salem Units 1 and 2.
The transients were developed based on monitoring data, historical operation and operator interviews conducted at a large number of plants. 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.
From the pipe and system delta T information collected in the HOG [1,2,33 effort, individual plants' monitoring data was 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.
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3a ,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. Next, a representative RSS distribution was determined by multiplying the average number of occurrences in cach RSS range by two.
Therefore, there is margin of 1007. on the average number of cycles per heatup in each mode of operation.
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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. As discussed in Section 2.5, the historical delta T system distribution was shown to encompass the prior operating history of th Salem units. For medes 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 stratification cycles per cooldown relative to the average number of stratification cycles per heatup. [
3a ,c.e The transients for all modes were then enveloped I in ranges of AT pj pg, i.e., all cycles from transients within each AT pipe range were added and assigned to the pre-defined ranges. These cycles were then applied in the fatigue analysis with the maximum ATpipe for each range. The values used are as follows:
i- For Cycles Within Pioe Delta T Range Pine Delta T
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3a,c.e This grouping was done to simplify the fatigue analysis. The actual number of cycles used in the analysis for the heatup and cooldown events is shown in Table 2-2.
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b l The final result of this complex process is a table of transients corresponding to the subevents of the heatup and cooldown process. A mathematical description of the methodology used is given in Appendix C.
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Ja .c.e The critical location is the location with the highest combination of pipe delta T and number of stratification cycles.
Because of main coolant pipe flow effects, the stratification transients loadings 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 "
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l a.c.e Therefore, the fatigue analysis of the RCS hot leg nozzle was 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.
The total transients for heatup and cooldown are identified as hcl thru HC9 for the pipe, and hcl thru HC9 for the RCS hot leg nozzle as shown in Tables 2-2(a) and 2-2(b) respectively. Transients HC8 and HC9 for the pipe and HC9 for the nozzle represent transients which occur during later stages of-the heatup.
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As indicated in Section 2.5, based on a review of Salem Units 1 and 2
- operating records, there was an event in which the system delta T exceeded the
. transient basis upper limit of [
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2.7 Axial Stratification Profile Develooment 1 l
In addition to transients, a profile of the [
3a ,c.e Two types of profile envelope the stratified temperature distributions >
observed and predicted to occur in the line. These two profiles are a [
ya.C,e The second case [
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[
3a ,c.e These three configurations are illustrated in Figure 2-5. [
3a ,c.e Review and study of the monitoring data for all the plants revealed a consistent pattern of development of delta T as a function of distance from the hot leg intersection. This pattern was consistent throughout the heatup/cooldown process, for a given plant geometry. This pattern was used along with plant operating practices to provide a realistic yet somewhat conservative 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 plant analyzed. Since Unit I and Unit 2 have similar surge line configurations, the l
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- profile applies to both units. The resulting profile is illustrated in Figure 2-6 in which the term " location" is defined in Figure 3-5.
2.8 Stricina Transients l
The transients developed for the evaluation of thermal striping are shown in Table 2-3.
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j 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 CSTl 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.
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TABLE 2-1 SURGE LINE TRANSIENTS WITH STRATIFICATION NORMAL AND UPSET TRANSIENT LIST - SALEM UNIT 1 OR UNIT 2 TEMPERATURES (*F)
MAX NOMINAL LABEL TYPE CYCLES ATStrat PRZ T RCS T
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TABLE 2-1 (Cont'd.)
" URGE LINE TRANSIENTS WITH STRATIFICATION NORMAL AND UPSET TRANSIENT-LIST - SALEM UNIT 1 OR UNIT 2 TEMPERATURES (*F)
MAX NOMINAL LABEL TYPE CYCLES ATStrat PRZ T RCS T
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TABLE 2-2a SURGF LINE PIPE TRANSIENTS HITH STRATIFICATION - SALEH HEATUP/COOLDOWN (HC) - 200 CYCLES TOTAL TEMPERATURES ('F)
MAX- NOMINAL LABEL TYPE CYCLES ATStrat PRZ T RCS T
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TABLE 2-2b SURGE LINE N0ZZLE TRANSIENTS WITH STRATIFICATION - SALEH HEATUP/C00LDOWN (HC) - 200 CYCLES TOTAL TEMPERATURES (*F)
MAX NOMINAL LABEL TYPE CYCLES ATStrat PRZ T RCS f f
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t TABLE 2-3 SURGE LINE TRANSIENTS - STRIPING .
FOR HEATUP (H) and COOLDOWN (C) - UNIT 1 OR UNIT 2
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l 1
i 1 .
5231s/041691:10 2-18
a c.e 5
9 Figure _2-1. _ Typical Insurge-Outsurge (I/0) Temperature Profiles 5231s/041891:10 2-19
r I
- a,c,e i
1 5 L I-I l-Figure 2-2. -Typical Monitoring Locations-l l
-5231s/041891:10 2-20
- - - . . . , . . . . - - . , ~ . . _ . . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . . _ _ _ _ _ . _ _ _ . . _ _ . _ _ _ , _ . . . _ _ . _ . . - . . _
a,c.e j i
i l
i i
l ro
- i
'o r
i i
l l
1 i
! l l '
Figure 2-3. Summary of Historical Data Distribution from Salem Units 1 and 2,
'- Compared t3 the Distribution Used in the Analysis !
l i
t t
t 5231s/041891:10
_ . _ - - , . . _-...... ~
__. _ _ _ . _ _ _ . _ . . . _ . . . _ _ _ , . _ _ . . , _ _ _. . _ f
i i
I a,c.e
.t 9
1 i
1 i
i I i i
i j
Figure-2-4. Example Axial Stratification Profile for Low flow Conditions 1
5231s/041891:10 2-22 l
l' . , , , . . --..-,_--__m_-__....~..._., . . . - . . , , . . - . _ , . - , , , - _ _ _ _ , _ . - . - . . ~ . _ _ . - . . . - - _ . . . _ _ . . _ - _ . . _ _ . _ _ - _ . _ . _ . _ _ . - _ _ .
1 e
a,c.e
?
i' e
i r
1
}
4 i i h
n 4
P Figure 2-5 --Geometry Considerations 1
5231s/041691:10 2-23
i l
[ >
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J
- t a,0,e
. I 7
1 r
i 1
ll 5
i 1
Figure 2-6.- Temperature Profile Analyzed for Salem Units 1 and 2 ,
5231s/041691:10 2-24 l
i I
t SECTION 3.0 i
STRESS ANALYSES The flow diagram (Figure 3-1) describes the procedure to determine the effects of thermal stratification on the pressurizer surge line based on transients i developed in section 2.0. [ r j a.C,e 3.1 Surce Line layouts i
The Salem Units 1 & 2 surge line layouts are documented in references 6 and 7 and the Unit 1 layout is shown schematically in Figure 3-2. The two Salem ;
units are mirror images of each other along plant East-West. The support configurations of Salem surge lines are the same. 8elow is a table summarizing the existing Salem surge line support configuration.
Salem Units = Land _2 Suecort
__ Unit 1 _ Unit 2 M941 Type R-239776-1 R-240185-1 1040 Pipe Whip Restraint R-239776-2 R-240185-2 1070 Pipe Whip Restraint R-239776-3 R-240185-3 1090 Pipe Whip Restraint ;
R-239776-4 R-240185-4 1120 Pipe Whip Restraint '
R-239776-5 R-240185-5 1200 Pipe Whip Restraint R-239776-6 R-240185-6 1220 Pipe Whip Restraint R-239776-7 R-240185-7 1140 Pipe Whip Restraint <
C-PSH-1 2C-PSH-1 1170 Variable Spring Hanger C-PSH-2 2C-PSH-2 1110 Variable Spring Hanger C-PSH-3 2C-PSH-3 1060 Variable Spring Hanger 5231s/041691:10 3-1 t
I I
4 It can be seen from the table above that both of the Salem surge lines contain l three variable spring hangers and seven pipe whip restraints. In some cases these supports can cause higher thermal loads if displacement from thermal i stratification exceed available displacement limits. The piping sizes aro 14 j
.i inch schedule 140 for Unit I and schedule 160 for Unit 2, 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.
] 3.2 Elping_SyitesLG1obal Structural Analys!s l l The Salem Units 1 and 2 piping systems were modeled by pipe, elbow, and non-linear spring elements using the ANSY$ computer code described in Appendix A. The geometric and material parameters are included. [
i i
1 L
l )a,c.e Each thermal profile loading defined in section 2 was broken into ( ,
Ja,c.e Table 3-1 shows the loading cases consideret in the analysis. To encompass all plant operations, (
i ya .c.e 5231s/041691:10 3-2
( 3a ,c.e Consequently, all the thermal transient loadings defined in section 2 could be evaluated.
i j a.c.e In order to meet the ASME Section III Code stress limits, gloial structural
- models of the surge lines were developed using the informatirn provided by ]'
references 6 and 7 and the ANSYS general purpose finite elenent computer code. r model was constructed using (
3 U to reflect the layout of straight pipe, bends and field welds as shown in Figure 3-2.
For the stratified condition -(
i ya.c.e The global piping stress analysis was based on two models for the Salem Units. The first model represents the existing support configuration and the second model represents the future configuration. The existing configuration has the actual gaps at all whip restraints and actual spring travel allowances at all spring hanger locations (see Table 3-2). The future support configuration represents a modified configuration in which no spring hangers 4
will bottom-out, and whip restraint gaps are the same as those in the existing ,
configuration. In addition, the beneficial effect of insulation crushability was taken into account for the future configuration. The results of the ANSYS global structural analyses provide the thermal expansion moments. The ASME Section III equation (12) stress intensity range was evaluated for both configurations. A system delta T of ( Ja c.e was evaluated for the existing configuration in addition to 320*F for the future configuration. For l 5231s/041691:10 3-3
i 1
the Salem Units, the maximum ASME equation (12) stress intensity range in the surge line was found to be under the code allowable of 35m for the future configuration. Maximum equation (12) and equation (13) stress intensity j ranges are shown in Table 3-3.
1 The pressurizer nozzle loads from thermal stratification in the surge line J were also evaluated according to the requirements of the ASME code. The evaluation using transients detailed in Reference [13] plus the moment loading from this analysis calculated primary plus secondary stress intensities and the fatigue usage factors. For the Unit I and 2 pressurizer nozzles, the maximum stress intensity range is 47.3 ksi compared to the code allowable ,
value of 57.9 ksi. The maximum fatigue usage factor will be reported in Section 5. It was found that the Salem pressurizer surge nozzles met the code stress requirements.
3.3 Local Stresses-Methodoloav and Results 3.3.1 Explanation of Local Stress figure 3-3 depicts the local axial stress components in a beam with a sharply l nonlinear metal temperature gradient. Local axial stresses develop due to the restraint of axial expansion or contraction. This restraint is provided by the material in the adjacent beam cross section. For a linear top-to-bottom temperature gradient, the local axial stress would not exist. [
3a c.e 3.3.2 Finite Element Model of Pipe for Local Stress A short description of the pipe finite element model is shown in Figure 3-4.
The model with thermal boundary conditions is shown in Figure 3-5. Due to l
l l
5231s/041691:10 3-4
symmetry of the geometry and thermal loading, only half of the cross section was required for modeling and analysis. [
3a .C,e 3.3.3 Pipe Local Stress Results Figure 3-6 shows the temperature distributions through the pipe wall [
3a ,c.e 3.3.4 RCL Hot Leg Nozzle Analysis Detailed surge line nozzle finite element models were developed to evaluate the effects of thermal stratification. The 14 inch schedule 160 mo'jel i' shown in Figure 3-10, The schedule 140 model is similar. Loading case; included [
l a,c.e A summary due to thermal stratification is given in Tables 3-4A and 3-48. A summary of representative stresses for unit loading is shown in Table 3-5.
Iofs+ressesintheR 5231s/041691:10 3-5
3,4 Total Stress from Global and Lo n) Analvigi
[
3a .c.e
[
3a .c.e 3.5 Thermal Stri Ding 3.5.1 Background At the time when the feedwater line cracking problems in PHR's were first discovered, it was postulated that thermal oscillations (striping) may significantly contribute to the fatigue cracking problems. These oscillations were thought to be due to either mixing of hot and cold fluid, or turbulence in the hot-to-cold stratification layer from strong buoyancy forces during low flow rate conditions, (See Figure 3-11 which shows the thermal striping fluctuation in a pipe). Thermal striping was verified to occur during subsequent flow model 4
5231s/041891:10 3-6
b tests. Results of the flow model tests were used to establish boundary !
conditions for the stratification analysis and to provide striping oscillation data for evaluating high cycle fatigue.
Thermal striping was also examined during water model flow tests performed for the L1 quid Metal Fast Breeder Reactor (LMFBR) primary pipe loop. The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern. These dynamic oscillations were shown to produce significant fatigue damage (primary crack initiation). The same interface oscilli.tions were observed in experimental studies of thermal striping which wrre e performed in Japan by Mitsubishi Heavy Industries. The thermal striping evaluation process r was discussed in detail in references 3, 8, 9, and 10.
3.5.2 Thermal Striping Stresses i 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. (See Figure 3-12, which shows a typical temperature distribution through the pipe wall).
[
3a ,c.e ,
The peak stress range and strtss intensity was calculated from a 3-D finite element analysis. (
B l a,c.e The methods used to determine alternating stress intensity are defined in the ASME Code (4). Several locations were evaluated in order to determine the location where stress intensity was a maximum.
Stresses were intensified by K3 to account for the worst stress concentration for all piping elements in the surge line. The worst piping element was the butt weld.
5231s/041691:10 3-7
i j a.c.e 3.5.3 factors Which Affect Striping Stress The factors which affect striping are discussed briefly below:
(
l i
l l'
3a ,c.e a
5231s/041691:10 3-8
. _ . _ _ _ _ . . .. _ ._ _ . ..___ _ . _ _ __ _ _ .-._ _ _ = --. . . . . -
i t
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(
l a
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e L T I
f
=
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3 4
5231s/041691:10 3-9
i .
i !
TABLE 3-1 TEMPERATURE DATA USED IN THE ANALYSIS !
Hax I Type of System Analysis Pressurizer RCL T T Pipe Top Bot Operation AT(*F) Cases Temp ('F) Temp ('F) ('F) ('F) AT ('F)
I L
r j a.c.e i
i t
r 5231s/041691:10 3-10 D'ew+ ru-4--un - ,T=+ - - - , ar- gm.w- 9ge99,w-,go-ger.wpvy>-e- ,=>w-ggog,e-91+w,g- c- g +-- 4--%cew
- rw +7W F W-WM ,+ M$T 7 3-"4C- __yerw-h(re w773 yqr O' wW'
i I
4 d
I t.
i TABLE 3-2 i
i
~'. [ N\
j SALEM UNIT-1&2 SURGE LINE AS-ANALYZED GAPS Unit 1 As-Analyzed Unit 2 As-Analyzed I
ReJLtraint No. _Gao-(In.) Restraint No. ,_Gno (In. L l 4 I
! [ l i
i l
4 3- .
a l
l
! -i
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s !
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l l
r ,
a 3 .c.e I
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i l'
r r
5231s/041691:10 3-11 T
Mm . . w mem + reveNev e<g- vv wwww%evenem y-mvww me-w+ww.emw _ - -e-+*hewrs-w w. ~'----wm---~'+----*---*-
)
i TABLE 3-3 l Summary of Salem Unitt 1 & 2 Surge Lines Thermal Stratification Stress Results '
Analysis aire _ Code Ecuation Configuration .
Stress Code Allowghlt ;
Unit.1 unit _2 (ksi) l 7
12 Existing
- 55.8 58.4 52.9 Future + 49.6 50.0 52.9 13 Existing & ,
Future 45.0 45.0 50.1
- Existing configuration represents the effects of actual spring hanger bottomed-out, existing whip restraint gaps, and insulation crushability under all thermal loadings with maximum system delta T - 325'T
+ Future configuration represents no bottoming-out of spring hangers, existing whip restraint gaps, and insulation crushability under all thermal loadings with maximum system delta T - 320*F i
5231s/041691:10 3-12
TABLE 3-4A i l
SALEM UNIT 1 SURGE LINE i MAX 1 HUM LOCAL AXIAL STRESS AT ANALYZED LOCATIONS Profile local Axia] Stress (pju Location
- Surface Mulmum Tgasi1e Mnigua_Compffs11y.e f a.C,e ;
e t
See Figure 3-5 RCL nozzle transition RCL nozzle safe end and weld
_[
3a ,c.e I
5231s/041691:10 3-13 K er--iMm,+=Wa+c.r*y ,,y-t y . c a g..g.m wg ver+r,,. ,wm-- y-g,p p m w we, y , g.-w-gyg4w--eg 3- & g7 gm -% yer+,-eggm-y w , g y -9pi ,7ywy,y9ry-.w.+- q y vgyw,y wrwy g w.,9 m ,y-y T-iW
TABLE 3-4B SALEM UNIT 2 SURGE LINE MAXIMUM LOCAL AXIAL STRESS AT ANALYZED LOCATIONS Profile Locai Axial Stress (osi) l.ocation* Surface Maximym Tensile Maximum Comoressive
~
~
a,c.e
- See Figure 3-5
- RCL nozzle safe end
- RCL nozzle safe end weld
[ Ja.c.e 5231s/041691:10 3-14
TABLE 3-5
SUMMARY
OF PRESSURE AND BENDING INDUCED STRESSES IN THE SURGE LINE RCL N0ZZLE FOR UNIT LOAD CASES All Stress in esi Linearized Stress Peak Stress
.latensity Range _ _lntensity Range _
Diametral Unit Loading Location Location Condition Inside Outside Inside Outside a,c.e i !
1[ 1
- )
e l
i
_ J 5231s/041691:10 3-15
i I
1 TAP c 3-6 STRIPING TREQUENCY AT 2 MA:
+
. LOCATIONS FROM 15 TEST RUNS Total Frequency (HZ) Duration .
- Cycles
% % % Lgth, in '
Min (Duration) Max (Duration) Ava (Duration) Seconds j
- a,c.e
]
i f
e I
5231s/041691:10 3-16
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a t h
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L Figure 3-1. Schematic of Stress Analysis Procedure f ,
i I ?
1 5231s/041691:10 3-17
- L r
mwa, ,.-a-m,,---,.,a.-,r.--.a.emmn .,-e-_..- , , _ _ _ _ . , , . _n _ _, , - , , _ , - - w~w
~
a,c.e Figure 3-2. Pressurizer Surge Line layout 5231s/041891:10 3-18
1 i
d 4
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1 i
1 1
I j f
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a a,c.e i i
I I
f I i i
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Figure 3-3. Local Axial Stress in Piping Due to Thermal Stratification 1
5231s/041691:10 3-19 ,
t-
- ,...~,,.,.,-..-~.-,-..-. .- __ - - -
l E
i
-. .ma !
1 a,c.e l t
I l
l l
l-l L
Figure 3-4. Local Stress - Finite Element Models/ Loading
- i l
5231s/041691:10 3-20
. ...._ . _ , _ . . . . _ _ _ - _ _ , _ . - , _ _ . . _ - . . . _ _ _ _ _ _ _ _ . _ . _ . _ _ . _ . _ _ _ . . _ _ _ _ _ . _ _ _ _ . . _ . _ _ _ _ . . _ _1
l
~ ~
a,c.e Figure 3-5. Piping Local Stress Model and Tternal Boundary Conditions 5231s/041691:10 3-21 l
a,c.e Figure 3-6. Surge Line Temperature Distribution at [ l a.c.e Axial Locations
'5231s/041691:10 3-22 ;
-l c
~
a,c,e l
l i
Figure 3-7, Surge Line Local Axial Stress Distribution at [ ]d'C
Axial Locations 5231s/041691:10 3-23
. . - . ~ . . . . - . - . . _ . . .... .- ..--- . - - - . - . . - . - . - ... -. - ..- .. . - . .. . . .
a,c.e 4
f l
i i
E l,
f
[
Figure 3-8. Surge Line Local Axial Stress ~ on Inside Surface at
[' ]a,c e Axial Locations i
l u 5231s/041691:10 3-24
. . . _ . . _ ---_=_.~__ _ . _ . _ _ _ _ _ . _ . . _ _ _ . . . . _ - . _ _ . _ . . _ - _ , _ . _ . . . . _ _ _ . _ _ _ - _ . . _
a,e,e Figure 3-9. Surge Line Local Axial Stress on Outside Surface at
-[ ]a,c e Axial Locations 5231s/041691:10 3-25
l a,c.e
- l l
i Figure 3-10. Surge Line RCL Nozzle 3-D HECAN Model: 14 Inch Schedule 160 5231s/041691:10 3-26
i a,c.e -
-Figure 3-11. Thermal Striping Fluctuation-5231s/041691:10 3-27
a,c.e i
1 i
l i
l --
f l'
l-I l
Figure 3-12. Thermal Striping Temperature Distribution I
5231s/041691:10 3-28
SECTION 4.0 DISPLACEMENTS AT SUPPORT LOCATIONS The Salem Unit I and 2 plant specific piping displacements at the whip restraints / hangers along the surge line were calculated under the thermal stratification and normal thermal loads for both existing and future support configurations.
Tables 4-1A and 4-1B show the mtximum surge line piping displacements at the whip restraints and spring hangers for the future support configuration.
Future support configuration is bi,ed on the existing whip restraint gaps but with no " bottomed-out" spring hangers and taking credit for insulation crushability. Piping displacements at the whip restraints and spring hangers for the existing configuration are shown in Tables 4-2A and 4-28. The existing configuration is based on the existing whip restraint gaps with the effects of " bottomed-out" spring hangers and also taking credit for insulation crushability.
Based on the stress analysis in Section 3, no whip restraint gap modification is necessary to satisfy the ASME Code requirements, provided that the future whip restraint gaps are not staaller than those used in the analysis (Table 3-2) and confirmed by Ref. 7. However, the travel allowance in the spring hangers must be modified to accommodate the piping displacements shown in Tables 4-1A and 4-1B for the future configuration.
5231s/041691:10 4-1 e
__ m _. ..._ . _ _ _ . _ . _ ._ _ . _ _ _ _ _ . _ . _ _ . _ . . _ _ . _ . _ . _ _ _ . . _ _ . . . _ -
TABLE 4-1(A)
Salem Surge Line Maximum Piping Displacement at Restraint Locations Under Thermal Stratification (Future Configuration)
Disolacement (in)
Succort Node DX DY DZ Qa11 1 Unit 2
[~ a,c.e i
x
/
Note: (1) The displacements shown are for Unit 1 p (2) The displacements are in the coordinate system given in Figure l:
3-2. For Unit 2 displacements, the signs of DZ must be reversed L since Unit 2 is a mirror image of Unit 1 about the East-West 1
axis. DX and DY are the same as tabulated above for Unit 2.
l (3) The displacements tabulated are for a system delta T of 320*F.
l-5231s/041691:10 4-2 l
TABLE 4-1(B)
Salem Surge Line Maximum Piping Displacement at Restraint Locations Under Normal Operating Thermal (Future Configuration)
Disolacement (in)
Support Node DX DY DZ Unit 1 Unit 2 a,c.e Note: (1) The displacements shown are for Unit 1 (2) The displacements are in the coordinate system given in Figure 3-2.
For Unit 2 displacements, the signs of DZ must be reversed since Unit 2 is a mirror image of Unit I about the East-West axis. DX and DY are the same as tabulated above for Unit 2.
5231s/041691:10 4-3
TABLE 4-2(A)
SALEM SURGE LINE MAXIMUM PIPING DISPLACEMENTS AT RESTRAINT LOCATIONS UNDER THERHAL STRATIFICATION *
(Existing Configuration)
UNIT 1 UNIT 2 N0DE SUPPORT DX DY DZ SUPPORT DX DY DZ_ ,
a,c.e l
?
r t
l t
Note: Displacements in inches System Delta T equals 320*F 5231s/041691:10
J TABLE 4-2(B) l SALEM SURGE LINE MAXIMUM PIPING DISPLACEMENTS AT RESTRAINT LOCATIONS UNDER NORMAL OPERATING THERMAL (Existing Configuration)
UNIT I UNIT 2 DX DY DZ DX DY DZ SUPPORT N_QDE SUPPORT -~
a,c.e l
a di Note: Displacements in inches 5231s/041691:10
l SECTION 5.0 ASME SECTION III FATIGUE USAGE FACTOR EVALUATION 5.1 M_eiholohgy Surge line fatigue evaluations have typically been performed using the methods of ASME Section III, NB-3600 for all piping components [
Ja .c.e Because of the nature of the stratification loading, as well as the magnitudes or the stresses produced, the more detailed and accurate methods of NB-3200 were employed using finite element analysis for all loading conditions.
Application 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, and the global loadings and resulting stresses obtained using the methods described in section 3.0. In general, the stresses due to stratification were categorized according to the ASME Code methods and used to evaluate Code stresses and fatigue cumulative usage factors. It should be noted that, [
3a ,c.e 5.1.1 Basis The ASME Code,Section III, 1986 Edition [4] was used to evaluate fatigue on surge lines with stratification loading. This was based on the requirement of NRC Bulletin 88-11 (Appendix 8 of this report) to use the " latest ASME Section III requirements incorporating high cycle fatigue" 5231s/041691:10 5-1
Specific requirements for class I 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.
Accordin] to NB-3611 and NB- H30, the methods of NB-3200 may be used in lieu of the NB-3600 methods. This approach was used to evaluate the surge line components 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. Thesa equations and requirements are summarized in Tables 5-1 and 5-2.
The methods used to evaluate these requirements for the surge line components are describe'd in the following sections.
5.1.2 Fatigue Stress Equations Stress Classification The stresses in a component are classified in the ASME Code baseo'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 ano, in terms of NB-3653, the respective equation where the stress should be included. Table NB-3217-2 provides guidance for stress classification in piping components, which is 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 in order to represent peak stresses caused by local concentration, such as notches and weld effects. The NB-3653 equations use simplified formulas to 5231s/041691:10 5-2
l determine nominal stress based on straight pipe dimensions. [
ya .c.e For the RCL nozzles, three dimensional (3-0) finite element analysis was used as described in Section 3.0. [
3a .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. [
a J c.e The impact of this on the selection of components for evaluation is discussed in Section 5.1.3.
5231s/041691:10 5-3
Stress Combinnlons 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. [
l a,c.e This was done for specific components as follows:
[
l i
Ja,c.e l
l 5231s/041691:10 5-4 a C " - - - + - ' - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - - - . _ _ _ -
- - .- - . - - - . ~ .. . - . - ..- . - - . . . - . - - - = - - - - . - . . -
[
L l
t 3a ,c,e 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. [
l-l l
l 3a ,c e 5231s/041691:10 5-5
0
[quation 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.
Eauation 13 Streil Equation 13 stress, presented in Section 3.2, is due to pressure, design
, mechanical loads and differential thermal expansion at structural discontinuities. Based 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 discontinuities. [
3a ,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 l components using applicable stresses due to pressure and stratification transients.
l l
5231s/041691:10 5-6
_. _ . __ _ . _ . . . . . . _ _.. _ - _ . _ . _ . _ . . . _ . __.__._._....m._.__ _
Ailowable Stressu Allowable stress, St, was determined based on note 3 of Figure NB-3222-1. For secondary :stre,$ dut to a temperature transient or thermal expansion loads
(" rest' air.t oi' fr?c M def?ection"), the value of Sm was taken as the average of the Sm valuct 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 i value of Sm was taken at the highest matal temperature during the transient.
l 5.1.3 Selection of Components for Evaluation l
Based on the results of the global analyses and the considerations for controlling stresses in Section 5.1.2, [
l a.c.e The method to evaluate usage factors using stresses determined according to Section 3.0 is described below.
l 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 NB-3653.5, Application of these methods is summarized below.
Transient Loadcases and Combinations l
From the transients described in Section 2.0, specific loadcases were developed for the usage evaluation. [
L l
l 3a ,c.e Each leadcase 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.
5231s/041691:10 5-7
Usage factors were calculated at controlling locations in the component as follows:
- 1) Ecuation 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, N j, N ...N , were determined from Code Figures I-9.2.1 and I-9.2.2, 2 n curve C, for austenitic stainless steels. l
- 3) Using the actual cycles of each transient loadset, nj , n 2 *"n' calculate the usage factors U), U2 ...U n from Ug = ng /Ng . This is done for all possible combinations. Cycles are used up for each combination in the order of decreasing Salt. When N) is greater j Il than 10 cycles, the value of Ug is taken as zero.
[
)a,c.e
- 4) The cumulative usage factor, Ucum, was calculated as Ucum - U) +
To this was added the usage factor due-to U2 + ... + Un .
thermal striping, as described below, to obtain total Ucum. The Code allowable value is 1.0. j l
5231s/041691:10 5-8
5.3 Fatiaue Due to Thermal Stricing The usage factors calculated using the methods of Section 5.2 do not include the effects of thermal striping. [
3a ,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.
[
3a ,c.e The methods used to determine alternating stress intensity are defined in the ASHE code. Several locations were evaluated in order to determine the location where stress intensity was a maximum.
l 5231s/041691:10 5-9
Thermal striping transients are shown as 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 cons".rvatively represented a
by a series of 5 degree temperature steps. Each step lasts [ J .c.e seconds.
Fluctuations are then calculated at each temperature step. Since a cor.:+ ant a
J .c.e is used in all of the usage factor calculations, the frequency of [1 total fluctuations per step is constant and becomes:
[ 3a ,c.e Each striping transient is a group of steps with [ Ja .c.e fluctuations per step. For each transient, the steps begin at the maximum AT and decreases by
[ ]"'C 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 [ l a,c.e to obtain total fluctuations. This results in total fluctuations at each step. This calculation is performed for each step plateau from [ 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.e This is reflected in the results to be discussed below.
5.4 Faticue Usaae Results NRC_ Bulletin 88-11 [5] requires fatigue analysis be performed in accordance with the latest ASME III requirements incorporating high cycle fatigue and thermal stratification transients. ASME fatigue usage factors have been calculated considering the phenomenon of thermal stratification and -thermal striping at various locations in the surge line. Total stresses included the 5231s/041691:10 5-10
l E l Ja ,c.e The total stresses 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 pipe due to [
3a ,c e l
The maximum usage factor on Salem surge lines occurred at [
3a ,c.e i It is also concluded that the Salem pressurizer surge nozzles meet the code stress allowables under the thermal stratification loading from the surge line and the transient detailed in reference [13), and meet the fatigue usage requirements of ASME Section III, with a maximum cumulative usage factor equal to 0.50, 5231s/041691:10 5-11 1-mw m g , - -a w w
. _ ~ . - _ . . . _ _ . - _ . _ - . .- . . - - _ _ . . . . . . - - . - . . . . .
^
TABLE 5-1 CODE / CRITERIA o -ASME B&PV Code, Sec. III, 1986 Edition
- NB3600
- NB3200 o Level A/B Service Limits
- Primary Plus Secondary Stress Intensity 13Sm (Eq.10)
- Simplified Elastic-Plastic Analysis
- Expansion Stress, Se 1 3Sm (Eq. 12) - Global Analysis
- Primary Plus Secondary Excluding Thermal Bending < 3Sm (Eq. 13)
- Elastic-Plastic Penalty Factor 1.01 K 1 3.333 e
Peak Stress (Eq. II)/ Cumulative Usage Factor (Ucum}
Salt " Kep S /2 (Eq. 14)
- Design Fatigue Curve U
cum i 1.0 I
1 5231s/041691:10 5-12
TABLE 5-2
SUMMARY
OF ASME FATIGUE REQUIREMENTS Parameter Description Allowable (if applicable)
Equation 10 Primary plus secondary stress intensity; < 3Sm if exceeded, simplified elastic-plastic analysis maj be performed K Elastic-plastic penalty factor: required e
for simplified elastic-plastic analysis when Eq. 10 is exceeded; applied to alternating stress intensity Equation 12 Expansion stress; required for simplified < 3Sm elastic-plastic analysis when Eq. 10 is exceeded Equation 13 Primary plus secondary stress intensity < 3Sm excluding thermal bending stress; required for simplified elastic-plastic analysis _
when Eq. 10 is exceeded Thermal Limit on radial thermal gradient stress to Stress prevent cyclic distortion; required for use Ratchet of Eq. 13 Equation 11 Peak stress intensity - Input to Eq. 14 Equation 14 Alternating stress intensity - Input to Ucum Ucum Cumulative usage factor (fatigue damage) < l.0 5231s/041691:10 5-13
~~ ~
a,c.e j
1 1
l 1
1 i
Figure 5-1. Striping Finite Element Model l-5231s/041691:10 E-14
'"" a,c.e Figure 5-2. - Attenuation of Thermal Striping Potential by' Molecular Conduction (Interface Wave Height- of One ~ Inch) l l-5231s/041691:10 5-15
i SECTION 6.0 SUMMA 3CLUSIONS The subject of pressurizer ser't 1 ", * ;grity has been under intense investigation since 1988. Th( Y- ,: Bulletin 88-11 in December of 1988, but the Westinghouse Owners Group ao put a program in place earlier that year, and this allowed all members to make a 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 Salem Units 1 and 2.
Following the general approach used in developing the surge line stratification transients for the WOG, a set of transients snd stratification profile were developed specifically for Salem Units 1 and 2. A study was made of the historical operating experience at the Salem Units 1 and 2, and this information, as well as plant operating procedures and monitoring results (from similar plants), was used in development of the transients and profiles.
Based on the etress analysis in Section 3 and fatigue evaluation in Section 5, it is not necessary to modify the existing whip restraint gaps for ASME Code acceptability provided sufficient travel allowances are maintained in all the variable spring hangers.
The results of this plant specifi: analysis along with support modification demonstrate acceptance to the requirements of the ASME Code Section III, including both stress limits and fatigue usage for the full licensed life of the plant. This report demonstrates that the Salem Units have completely satisfied the regt' ements of NRC Bulletin 88-11.
5231s/041691:10 6-1
SECTION 7.0 ;
I REFERENCES
- 1. Coslow, B. J., et al., " Westinghouse Owners Group Bounding Evaluation for Pressurizer Surge Line Thermal Stratification", Westinghouse Electric
~
. Corp. 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 Program MUHP-1090 Summary Report," Westinghouse Electric Corp. WCAP-12508 (Proprietary Class 2) and ICAP.12309 (non-proprietary), March 1990. ;
1
- 3. Coslow, B. J., et al., " Westinghouse Owners Group Pressurizer Surge Line Thermal Stratification Generic Detailed Analysis Program MUHP-1091 Summary Report," Hestinghouse Electric Corp. WCAP-12639 (Proprietary Class 2) and WCAP-12640 (non-proprietary), June 1990.
C. " Pressurizer Surge Line Thermal Stratification," USNRC Bulletin 88-11, ;
December- 20, 1988, i
- 6. PSE&G Letter MEC-90-06S8, October 3, 1990, "NRC Bulletin 88-11 Pressurizer Surge Line Thermal Stratification".
t
- 7. PSE&G Letter MEC-91-1074, April 9,1991 " Pressurizer Surge Line Thermal Stratification Salem No. I and 2 Units." ,
- 8. " Investigation of Feedwater Line Cracking in Pressurized Hater Reactor Plants," HACP-9693, Volume 1. June 1990 (Proprietary Class 2).
- 9. Woodward, H. S., " Fatigue of _LMFBR Piping due to flow Stratification,"
ASME Paper 83-PVP-59, 1983.--
- 10. Fujimoto, T., et al., " Experimental Study of Strioing at the Interface of Thermal- Stratification" in Thermal Hydraulics in Nuch5I_ Te.chnology, K. H.
Sun, et al., (ed.) ASME, 1981, pp. 73.
S231s/041691:10 7-1
- 11. Holman, J. P. , heat 'trJLD$fu, McG. au Hill Book Co. ,1963.
- 12. Yang, C. Y., "iransfer function Method for Thermal 5 tress Analysis:
Technical Basis," Westinghouse Electric Corporation WCAP-12315 (Proprietary Class 2).
- 13. Series 84 Pressurizer Stress Report, Section 3.1, Surge Nozzle Analysis, December 1974.
k
- 14. PSELG Letter to NRC, NLR-N89069, dated 5/31/89 "NRC Bulletin 88-11 Justification for Continued Operation, Salem Generating Station Unit Nos. 1 and 2".
5231s/041691:10 7-2
i APPENDIX A LIST OF COMPUTER PROGRAMS This appendix lists and summarizes the computer codes us6d in the pressurizer surge line thermal stratification. The codet are:
- 1. WECAN
- 2. STRFAT2
- 3. ANSYS
- 4. FATRK/ CMS A.1 HECA!j A.1.1 Denr_tglion WECAN is a Westinghouse-developed, general purpose finite element program, it contains univti:P lyi accepted two-dimensional and three-dimensional isoparametric elements that can be used in many different types of finite element analyses. Quadrilateral and triangular structural elements are used 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 [1itEelnd 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 can also be include in the WECAN structural analysis. Such coupled thermal-stress analyses are a standard application used extensively on an industry wide basis.
5231s/041691:10 A-1
i A.l.3 Program _VUi f kat303 Both the HECAN program and input for the WECAN verification problems, currtntly numbering over four hundred, are maintained under configuration control. Verification problems include coupled thermal-stress analyses for
. mril:*cral triangulct, 'rici and wedge isorararetrit elements. T h e". e problems are an integral part of the HECAN quality assurance procedures. When a change is made to HECAN, as part of the reverification process, the configured inputs for the coupled thermal-stress verification problems are used to reverify HECAN for coupled therm 41-stress analyses.
A.2 STRFAll A.2.I De.sg h tipf1 STRFAT2 is a program which computes the alternating peak stress on the inside 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 program computes maximum alternating stress based on the maximum difference between inside surface skin temperature and the average through wall temperature.
5231s/041691:10 A-2
A.2.2 [e3tytellstd 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 Et02EA"LYerif_lC3110D STRfAT2 is verified to Westinghouse procedures by independent review of the stress equations and calculations.
A.3 A3515 A.3.1 QegIlpli.on ANSYS is a public domain, general purpose finite element code.
A.3.2 E.cAturLUird The ANSYS elements used for the 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 Etogram Verification As described in section 3.2, the application of ANSYS for stratification has been independently verified by comparison to HESTDYN (Westinghouse piping analysis code) and WECAN (finite element code). The results from ANSYS are also verified against closed form solutions for simple beam configurations.
5231s/041691:10 A-3
l A.4 EAIRKlCMS A.4.1 QeltLiRiiCD FATRK/ CMS is a Westinghouse developed computer code for fatigue tracking (FATRK) as used in the Cycle Monitoring System (CMS) for structural components of nuclear power plants. The transfer function method is used for transient thermal stress calculations. The bending stresses (due to global stratification effects, ordinary thermal 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 III, Subsections NB-3200 and NB-3600.
The code can be used both as a regular analysis program or an on-lino nonitoring device.
A.4.2 Feature Used FATRK/ CMS is used as an analysis program for the present application. The input data which include the weight functions for thermal stresses, the unit bending stress, the unit pressure stress, the bending moment vs stratification temperatures, etc. are prepared for all locati.... and geometric conditions. These data, as stored in the independent files, can be appropriately retrieved for required analyses. The tr rsient data files contain the time history of temperature, pressure, number of occurrence, and additiot.a1 condition necessary for data flowing. The program prints out the total usage factors, 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 P_rogr r a !Lyjerificatinn FATRK/ CMS is verified according to Westinghouse procedures with several levels of independent calculations.
5231s/041691:10 A-4
[
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l APPENDIX B l l l 1
USNRC BULLETIN 88-11 l
1 In December of 1988 the NRC issued this bulletin, and it has led to an l
extensive investigation of surge line integrity, culminating in this and other j l plant specific reports. The bulletin is reproduced in its entirety in the i pages which follow. l l
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I 5231s/041691:10 B-1 !
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CPB No. 3150 0011 NRCB E8 11 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTCR REGULATICN WASHINGTON, D.C. 20!55 December 20, 1988 NRC BULLETIN NO. 88 11: PRESSUR!ZER SURGE LINE THERMAL STR.AT!FICATION Addressees:
All holders of operating licenses or construction permits for pressurized water reactors (PWRs).
Purpose:
The purpose of this bulletin is to (1) request that addressees establish and implement a program to confim pressurizer surge line integrity in view of the occurrence of thermal stratification and (2) require addressees to inform the staff of the actions taken to resolve this issue.
Description of Circumstances:
The licensee for the Trojan plant has observed unexpected travement of the pressurtzer surge line during inspections perfomed at each refueling outage since 1902, 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 contacted two re-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 confirmed 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 Information Notice 88-80, " Unexpected Piping Movement Attributed to Thermal Stratification," regarding the Trojan experience and indicated that further generic consnunication may be forthcoming. The licensee for Beaver Valley 2 has also noticed unusual snubber movement and significantly larger-than-expected surge line displacement during power ascension.
The concerns raised by the above observations are similar to those described in NRC Bulletins 79-13 (Revision 2, dated October 16, 1979), " Cracking in Feedwater System Piping" and 88-08 (dated June 22,1988), " Thermal Stresses in Piping Connected to Reactor Coolant Systems."
8812150118 B-2
l NRCB 28 1; Cecemter 20, Mis Page 2 of 6 Discussion:
Urexpected piping movements are highly uncesirable tecause of cotential ri;n piping stress that may exceed design limits for f atigue and stresses. Se problem can be more acute when the piping expansion is restrictec, such as through contact with pipe whip restraints. Plastic defomaticn can result, which can lead to high local stresses, low cycle fatigue and functicnal m.
pairment of the line. Analysis performed by the Trojan licensee indicatec *. eat themal stratification occurs in the pressurizer surge line curing heatup.
cooldown, and steady-state operations of the plant.
Curing a typical plant heatup, water in the pressuri:er is heated to atout 440'F; a steam bubble is then formed in the pressurizer. Although the enact phenomenon is not thoroughly understcod, as the hot water flows (at a very 'cw 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 t*e pipe to be heated to a higher temperature than the lower part (see Figure 1),
The differential temperature could be as high at J00'F, based on expected conditions during typical plant operations. Under this conditien, differential themal expansion of the pipe metal can cause the pipe to ceflect signifi-cantly.
For the specific configuration of the pressurizer surge line in the Trojan plant, the line deflected downwarc and when the sur whip restraints, it uncerwent plastic deformation, ge line contacted resulting two pi e in gemanent deformation of the pipe.
The Trojan event demonstrates that themal stratification in the pressuri:er surge line causes unexpected piping movement and potential plastic defomation.
The licensing basis according to 10 CFR 50.55a for all PhRs requires that the licensee meet the American Society of Mechanical Engineer: Boiler and Pressure Vessel Code Sections 111 and XI and to reconcile the pint stresses and fatigt.e evaluation when any significant differences are observed between measured data and the analytical results for the hypothesized conditions. Staff evaluation indicates that the thermal stratification phenomenon could occur in all PKR surge line.
surge lines and may invalidate the analyses supporting the integrity of the The staf f's concerns include unexpected bending and themal striping (rapid oscillation of the themal boundary interface along the pioing inside surface) as they affect the overall integrity of the surge line for its design life (e.g., the increase of fatigue).
Actions Requested:
Addressees are requested to take the following actions:
- 1. For all licensees of operating PWRs:
a.
Licensees are requested to cenduct 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
1 i
NRCB 88 11 December :0, 17E8 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 cperation over 10 years (i.e., low power license prior to January 1, 1979) are requested to demonstrate that the pressurizer surge line meets the applicible 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 the fatigue and stress evaluations. This may be accomplished by performing a plant specific or generic bounding analysis. If t'n latter option is selected, licensees should demonstrate applicability of the referenced generic bounding analysis, l.icensees of plants in operation less than ten years (i.e., low power license after January 1,1979), should complete the foregoing analysis within ore year of receipt of this bulletin. Since any piping distrtss 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 perfom 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 shculd 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 duratien of the license, th 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 conscitments stated therein for the duration of the operating license, the licensee is requested to obtain plant specific data on themal stratification, thermal striping, and line deflet-tions. The licensee may choose, for example, either to inctall 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 reouested to update their stress and fatigue analyses to ensure compliance with applicable Code requirements, incorporating any observations from 1.a above. The analysis bould be completed no later than two years after receipt of this bulletin. If a licensee
- Fatigue analysis should be performed in accordance with the latest ASME Section !!! requirements incorporating high cycle fatigue, i
l B-4
l NRCB 28-11 Decemoer 20, 19E8 Page 4 of 6 is unable to show compliance with the applicable design codes and other FSAR and regulatory comitments, the licensee is reouested te submit a justification 'or continued operation and a description of the proposed corrective actions for effecting long tem resolution,
- 2. For all applicants for PWR Operating Licenses:
- a. Before issuance of the low cover license, applicants are requestec 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 cerforming a plant-specific or generic bounding analysis. Tne analysis should include consideration of thermal stratification and thermal stripin" toensurethatfatigueandstressesareincompliancewithapplicable code limits. The analysis and hot functional testing should verify that piping themal deflections result in no adverse consecuences, 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 op m tional alternatives 0" 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 themal stratification, beginning with hot functional testing, or obtain data through collective efforts to assess the extent of themal stratification, themal striping and piping deflections.
- d. Applicants are requested to update stress and f atigue 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, Appendix B and plant procedures.
Reporting Requirements:
- 1. Addressees shall report to the NRC any discernable distress and damage observed in Action 1.a along with corrective actions taken or plans and schedules for repair before restart of the unit.
- 1f compliance with the applicable codes is not demonstrated for the full duration of an operating license, the staff may impose a license condition sutn that normal operation is restricted to the duration that compliance is actually demonstrated.
B-5
NRCB 88-11 December 20, 1988 Page 5 of 6
- 2. Addressees who cannot meet the schedule cescribed in Items 1 or 2 of actions Recuested are required to submit to the NRC within 60 days of receipt of this bulletin an alternative schedule with justification for the recuested schedule.
- 3. Aderessees shall submit a letter within 30 days after the completion of these actions which notifies the NRC that the actions reauested in !ters Ib, ld or 2 of Actions Recuested have been perfomed and that the results are available for inspection. The letter shall include the justificatien fur continued creration, if appropriate, a description of the analytical approaches used, and a surrnary of the results.
Although not requested by this bulletin, addressets are encouraged to work collectively to address the tecnnical concerns associated with this issue, as well as e share pressurizer surge line data and operational experience. In addition, addressees are encouraged to review piping in other systems which ray ,
experience themal stratifict, tion and thermal striping, especially in light of the previously mentioned Bulletins 79-13 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 Commission, ATTN: Document Control Desk Washington, D.C. 20555, under oath or affirmation urder the provisions of Section 182a, Atomic Energy Act of 1954, as amended. In addition, a copy shall be submitted to the appropriate Regional Administrator.
This recuest is covered by Office of Management and Budget Clearance Number 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 raoiation exposure is approximately 3.5 person-rems per licensee response.
Cocinents on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Office of Management and Budget, Room 3208 New becu-tive Offica Building, Washington. 0.C. 20503, and to the U.S. Nuclear Regula-tory Comission, Records and Reports Management Branch, Of fice of Administration and Resource Management, Washington, D.C. 20555.
B-6
i l
NRCB 88-11 Cecember 20, 1988 Page 6 of 6 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 regional office.
bl{
.a es E. Rossi, Dir' ctor Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contacts: S. N. Hou, NRR (301) 492-0904 S. S. Lee, NRR (301) 492-0943 N. P. Kadambi, NRR (301) 492-1153 Attachments:
- 1. Figure 1
- 2. List of Recently Issued NRC Bulletins B-7
l!i!"ii'li
- .c. o r :0. ns Page i of Surc e _ine S':ra':i"ication CS __
PZR
<1 1 1 1 1 1 ;
Hot Flow from Pressurizer Thot = 425*F
(
Tgt Stagnant Cold Fluid Tcold = 125*F Figure 1
, B-8
l APPENDIX C TRANSIENT DEVELOPMENT DETAILS
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