ML20086P842

Nonproprietary Structural Evaluation of Turkey Point,Units 3 & 4,Pressurizer Surge Lines,Considering Effects of Thermal Stratification

ML20086P842
Person / Time
Site:	Turkey Point
Issue date:	05/31/1991
From:	Palusamy S, Roarty D, Vora V WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML17348B302	List: ML20086P842
References
WCAP-12960, NUDOCS 9112270217
Download: ML20086P842 (109)

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• WCAP-12960 1

I Structural Evaluation of Turkey Point Units 3 and 4 Pressurizer Surge Lines. .

Considering tht Effects of Thermal Stratification May 1991 T. H. Liu L. H. Valasek M. Yu M. A. Gray P. L. Strauch S. Tandon

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Verified by: a Ml / Verified by: Y' YdV D. H. Roartr 9 V. V. Vora f' / )

Approve y2  !-:,_k Approved by: M. l _ b S. S. P~a'lusamy. Manager R. 5. Patel. Manager Diagnostics and Monitoring System Structural Analysis Technology and Development Work Performed under Shop Orders FGIP-964 and FGIP-145 HESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O.. Box 2728 Pittsburgh. Pennsylvania .15230-2728 e 1991 Westinghouse Electric Corp.

5341s/061091:10

. TABLE OF CONTENT $ t Section lille hge Executive Summary lii 1.0 Background and Introduction 1-1 1.1 Background 1-1 1.2 Description 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 Honitoring Results and Operational Practices 2-4

, 2.5 Historical Operation 8 2.6 Development of Heatup and Cooldown Transients 2-10

. 2.7 Axial Stratification Profile Development 2-13 2.8 Striping Transients 2-15 3.0 Stress Analysis 31 3 /1 Surge Line layouts 3-1 3.2 Piping System Global Structural Analysis 3-2 3.3 Local Stresses - Hethodology and Results 3-4 3.4 Total Stress from Global and Local Analysis 3-6 3.5 Thermal Striping 3-6 4.0 Olsplacements at Support Locations 4-1 s

5341s/061091:10 i

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TABLE OF CONTENTS (Continued) 4

^

SKiioll litle EA9k 5.0 ASME Section !!! 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 8 USNRC Bulletin 88-11 B-1 Appendix C Transient Development D ails C-1 5341s/061091:10. ii I

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EXECUTIVE

SUMMARY

Thermal Stratihcation has been identified as a concern which can affect the

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structural integrity of piping systems in nuclear plants since 1979, when a leak was discovered in a PHR feedwater line. In the pressurizer surge line, stratification can result from the difference in consities between the hot leg water and generally hotter pressurizer water. Stratification with large temperature differences can produce very high stresses, and this can lead to piping 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 Turkey Point Units 3 and 4. Prior to the issuance-of the bulletin, the Westinghous? Ownert Group had a program in place to investigate the issue and to recommend acticns by member utilities. That program provided the technical basis for the analysis reported herein for Turkey Point Units 3 and 4.

The transient development utilized a number of sources, including plant operating procedures, surge line monitoring data from other similar units, and historical records for each Turkey Point unit. This trensient information was used as input to a structural and stress analysis of the surge line for the two units. A review and ccmparison of the piping and support configurations for Turkey Point Units 3 and 4 led to the conclusion that the surge lines are nearly identical, and thus one bounding analysis could apply to both units.

[ The existing configurations for both Turkey Point' units have been analyzed as-l described in this HCAP. The results of P- ' lyses indicate no contact L between the pipe and pipe whip restraintf a unit for the worst case AT of 320'F, Actual pipe stresses were shvwn to be acceptable when compared to ASHE Code allowables for bo4 units under this existing configuration, if-travel allowance in the vertical spring hanger.can be adjusted in the future to accommodate the surge line displacement.

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With this spring hanger travel adjustment, the ASME Code stress limits and -

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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. *r

! This work has led to the conclusion that Turkey Point Units 3 and 4 are in full compliance with the requirements of NRC Bulletin 88-11 provided the i

spring hanger modifications discussed on the following page are implemented. 7 J

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SUMMARY

OF RESULTS, AND STATUS OF 88-11 QUALIF' CATION Unit 3 Unit 4

. Open11ng_tihton_.throvsh_t920 Date of commercial operation 12/14/72 9/7/73 Years of water-solid heatups 14 13 Years of steam bubble heatups 4 4 System delta T limit 320'F 320'F Number of 320'F AT exceedances None None tiaximum Stress and_UngtJJttiar Resul_t1 Equation 12 stress / allowable (ksi) 50.7/52.9 50.7/52.9 fatigue usage / allowable 0.90/1.0 0.90/1.0 Pressurizer Surge Nozzle Results Maximum stress intensity range / 32.9/57.9 32.9/57.9 allowable (ksi)

Fatigue usage / allowable 0.41/1.00 0.41/1.00 Ermainina Actions by Utilitici Spring hanger modification required Allow sufficient travel allowance on both units (Table 4-1)

StA115 of 88-11 Reauirementi All analysis requirements met with above modi fication 4

5341s/061091:10 v

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l SECTION 1.0 BACKGROUND AND INTRODUCTION l

Tur4ey Pcint Units 3 and 4 are three-loop pressurized water reactors, designed to be as nearly identical as practical, in both hardware and cperation. 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 both of these units.

l 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 three-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 interfaco level typically between 25 and 60% of the volume

l. 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 lines of both Unit 3 and Unit 4 are made of 12 inch schedule 140 stainless steel.

l 1

1.1 Background

During the period from 1982 to 1988, a number of utilities reported unexpected movement of the pressurizer surge.line, as evidenced by crushed insulation, 5341s/061091:10 1-1

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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 stratificaiton 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 lines led the USNRC to issue Bulletin 88-11 in necember 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 line. The program required both tisus) 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 HOG 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 justificat'on for continued operation was developed for use by each of the WOG plants, the basis of which was documented in references (1) and [2].*

' Numbers in brackets refer to references listed in Section 7.

~

5341s/061091:10 1-2

l The Westinghouse Owners Group implemented a program for generic detailed i l .

l. analysis in June of 1989, and this program involved individual detailed analyses of groups of plants. This approach permitted a more realistic .

- approach than could be obtained from a single bounding analysis for all l l

plants, and the results were published in June of 1990 (3).

l 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 NRC Bulletin 88-11 for Turkey Point Units 3 and 4.

I 1.2 Descriotion of Surge Line Thermal Stratif1 cat 10D ,

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 I stratification is increased as the difference in temperature between the l . 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 havn 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, n which case the effects of stratification are significant, and must be accounted for.

l. Ti.i mal tcratification in the surge line causes two effects:

o Bending of the pipe different.from that predicted in'the original design.

o Potentially reduced fatigue life of the piping due to the higher stress resulting from stratification and striping.

5341s/061091:10 1-3 1

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1.3 it0RE_01_Hath The primary purpose of this work was to develop transients which are applicable to the Tu. Key Point surge lines and include the effects of -

stratification, and to evaluate the structural integrity of the surge lines.

This work will therefore complete the demonstration of compliance with the requirements of HRC 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 Secon 2.

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, support load; and nozzle loads were datermined.

This analysis and its results are discussed in Sections 3 and 4.

l 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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i SFCTION 2.0 SURGE LINE TRANSIENT AND TEMPERATURE PROFILE DEVELOPMENT 2.1 GenenLAppIORh The transients for the pressurizer surge line were developed from a number of sources, including the most recent Westinghouse Systems Standard Design transients. The heatup and cooldown transients, which involve the majority of the severe stratification occurrences, were developed from review of the I

design transients, plant operating procedures, operator interviews, monitoring data and historical records fcr each unit. The total number of heatup and cooldown events specified reules unchanged at 200 each, but a number of transient events within each heatup and cooldown cycle have been defined to reflect stratification effects, as described in more detail later. 1 l

The normal and upset transients, except for heatup and couldown, used for the Turkey Point Units 3 and 4 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 uniform temperature to a stratified distribution, according to the predicted temperature differentials betwean the pressurizer and hot leg, as listed in the table. The transients have t'een characterized >

as either insurge/outsurges (1/0 in the table) or fluctuations (F).

Insurge/outsurge transients arc generally more severe, because they result in the greatest temperature change in the top or bottom of the pipe. Typical temperature profiles for insurges and outsurges are shown in Figure 2-1.

l Transients identified as fluctuations (F) typically involve low serge flow rates and smaller temperature differences between _the pressurizer and hot leg, so the resulting stratification stresses are much 13wer. This type of cycle is important to include in the analysis,-but is generally not the~ major contributor to fatigue usage.

9 5341s/061091:10 2-1

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9 A

In addition to the plant specific operating history discussed above, the .

developent of transients which are applicable to Turkey Point Units 3 and 4 I was based on the work aircady accomplished under programs completed for the Hestinghouse Owners Group (1.2.?]. 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.

I' 3

1ht: transient development for the Turkey Point units took advantage of the 5 milarity in the surge line layout for the two units, as well as general 8

riimMarities in the operating procedures. A detailed comparison of the piping and support configurations for the units appears in Section 3.1.

1he transients developed here, and used in the structural analysis, have taken advantage of the monitorirg data collected during the HOG program, as well as histori a l operation data for the Turkey Point units. Each of these will be discussud in the sections which follow.

2,1 System Desian Inform &Il0D The thermal design transients for a typical Reactor Coolant System, including the pressurizer surge line, are defined in Westinghouse Systems Standard Design Critnria.  !

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).

In the evaluation of surge line stratification, the traNient events considered encompass the normal and upset design events defined in the FSAR.

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2.4 Bon 11pring_R31ults and Qgerational PrAc111t1 2.4.1 Honitoring i Monitoring information collected as part of the Westinghouse Owners Group generic detailed analysis (3) was utilized in this analysis. The pressurizer surge line monitoring programs utilized externally mounted temperature sensors (resistance temperature detectors or thermocouples). The temperature sensors were attached to the outside surface of the pipe at various circumferential and axial locations. In all cases these temperature sensors were securely >

clamped to the piping outer wall. taking care to properly insulate the area against heat loss due to thermal convection or radiation..

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-[

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 i stratification in the surge line. A list _of typical plant parameters monitored is given below, i

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j 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 Turkey Point Units 3 and 4 heatup and cooldown opertting procedures and operational interviews conducted at a number of HOG uti] ties,itwasdeterminedthatbothTurkeyPointunitspresentlyheatup ana cool down in manners similar to other plants that heat up with a steam bubble in the pressurizer. (Turkey Point previously used the water-solid method of heatup and cooldown until 1986.) Heatups and-cooldowns are used here to characterize plant operation berAuse they represent the periods during which the temperature difference between the pressurizer and the hot leg is potentially the greatest. A brief description of the Turkey Point units' present heatup and cooldown procedures follows.

The heatup and cooldown procedures for Turkey Point Units 3 and 4 are typical of most plants that use the steam bubble method. With the RCS solid and l

pressure being maintained at approximately 350 psig by the charging pumps, heatup is initiated by energizing the pressurizer heaters, with subsequent l'

formation of a steam bubble in the pressurizer in Mode 5 (RCS temperature 5341s/061091:10 2-5

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1ess than 200*F). After the bubble is established, reactor coolant pumps (RCP's) are placed in service, as required. The Residual Heat Removal System (RHR) may be maintained in operation until the RCS pressure and temperature reach 450 psig and 350'F, respectively. When normal pressurizer level is established, the normal letdown flow path is aligned.

Prior to increasing RCS temperature above 200*F (Mode 4), a series of prerequisite checks is performed. With the RCS between 200'F and 350'F, RHR is isolated. The heatup is then completed using pressurizer hesters and reactor coolant pumps.

The cooldown process is basically the reverse of the heatup. The steam bubble

. is maintained and the RCP's are operated continually at least until the RCS temperature is less than 200'F (Mode 5). If possible, continuous pressurizer spray flow is maintained throughout the cooldown.

Pertinent temperature limitations on the Turkey Point heatup and cooldown procedures are as follows: -

1) Maximum RCS cooldown rate of 100'F/hr; maximum RCS heatup rate of 50*F/hr

• for Unit 3 (Technical Specification limit is 100'F/hr), and 100'F/hr for Unit 4.

2) Maximum allowable delta T between pressurizer and spray of 320'F.

3) Maximum allowable delta T between pressurizer and RCS (system delta T) of 320'F during cooldown (this is also inherently included for heatup in the spray differential limit above.)

4) Maximum pressurizer heatup rate of 90*F/hr, and cooldown rate of 190'F/hr.

5) Only six spray actuations with a 60 second maximum duration allowed during 5eatup or cooldown, when the system delta T is between 200'F and 320'F with a pressurizer steam bubble.

5341s/061091:10 2-6 I-- .- - --- _ _ - . -. _ . . _ . - _ . . - ._ - , - . . . . _ - _ .

. 6) Minimum allowable system delta T of 100'F before collapsing the pressurizer steam bubble.  !

From the operating procedures, the possible ranges of the system delta T could vary, but are bounded by the administrative limit of 320'f. The actual impact of these plant operating procedures on the analysis was determined in conjunction with review of the plants' past operating histories, and is i

discussed in the following section.

2.5 Historical Ooeration Historical records from Turkey point Units 3 and 4 (operator logs, "

surveillance test reports, etc.) were reviewed in September, 1990 and February, 1991. The purpose of the review was to obtain a distribution of maximum system delta T, and to identify heatup or cooldown events where the '

maximum system delta T exceeded the 320'F limit. The Turkey Point units performed heatup and cooldown operations using the water-solid method until

. late 1986, with an administrative limit of 200'F on system delta T. After ,

j this time, heatup and cooldown procedures were revised to use the steam-bubble method, with a corresponding 320*F limit on system delta T. It is known by experience that the lower delta T limit in the water-solid periods has much l 1ess impact on the stress and fatigue evaluations. Therefore, the February, l 1991 review concentrated on determining the distribution and exceedances for the steam-bubble period. Data was obtained for approximately 80% of the heatup and cooldown events that occurred in this period. The delta T distribution is expressed in terms of the number of events in a predetermined range as a percentage of the total number of events for which data was available. A summary of the results for available data is presented below, Unit 3 UnLt_4 Number of Number of System AT Heatups or  % of Heatups or  % of Ranao ('F) CooldowDi latAl Cooldowns Total

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This information was used to ensure that the translents analyzed for the Turkey Point units encompassed the prior operating history of the plants. .

Comparison of these past system delta T distributions to that used in the analysis is illustrated in figure 2-3. (Development of the analytical system -

delta T distribution is discussed further in Section 2.6.) (

3'.,c.e 2.6 Development of Heatyp.and Cooldown framigaty 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 Turkey Point Units 3 and 4, 2.6.1 General Method The transients were developed based on monitoring data, historical operation .

and operator interviews conducted at a large number of plants. For each monitoring locatien, 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 (1.2,3) 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 considcrs delta T range as opposed to absolute values.

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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 each RSS range by two.

Therefore, there is margin of 1001, on the average number of cycles per heatup in each mode of operation.

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 distribution developed from plant operating records from a number of plants.

. and is represented in Figure 2-3 as "Used in Analysis". As discussed in Section 2.5, this historical delta T system distribution was shown to

, encompass the prior operating history of the Turkey Point units, and was assumed to account for future operation. 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.

l An analysis was conducted to determine the average number of stratification I

cycles per cooldown relative to the average number of stratification cycles per heatup. [

]"'C -The transients for all modes were then. enveloped.

in ranges of ATpgp,, i.e., all cycles from transients within each ATpgp, range were added and assigned to the pre-defined ranges. These ,

cycles were then applied in the fatigue analysis with the maximum AT pgp ,

l for'each range. The values used are as follows:

5341s/061091:10 9

For Cycles _Hithin PiptDaltLJ_ Range ElptDg]itl ,

i

• ja.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.

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.

(

Ja .c.e The critical location is the location with the highest combination of pipe delta T and number of stratification cycles.

2.6.2 Hot Leg Nozzle Because of main coolant pipe -flow ef fects, the stratification transient 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 stratificatica 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."

5341s/061091:10 10

. (

Ja.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. i 2.6.3 Final Plant-Specific Transients The general table of transients, each defined by maximum pipe delta T and corresponding cycles, was further refined to reflect the operating history of the Turkey Point units. This involved two major items: the change from water-solid to steam-bubble operating methods, t.nd a recommended hardware change that would affect system loads and associated moment loadings with each transient, figure 2-4 tilustrates the operating histories of Units 3 and 4 with respect to these items. The respective periods of operation are labelled

. HS (water-solid), SB1 (steam-bubble 1) and SB2 (steam-bubble 2). The SD2 period represents steam-bubble operations af ter the spring hanger modification

. discussed in Section 3.

For the HS period, the distribution of cycles at each maximum delta T value was maintained, but the total number of cycles was determined by ratio, based on tha number of heatups and cooldowns in the HS period. Also, the correspo'iding delta T values were adjusted to reflect the generally lower i values that occur for water-solid operations. The cycles were applied in the fatigue analysis with the maximum AT pipe in the following pre-defined ranges for the HS period;

! for CycleL}illbin Pipe QLLta T Range Pine Delta.]

(*F) ('f) l .

l 1-l Ja,c.e 5341s/061091:10 2-11

. , ..%',, ._,,,....,i,,,..,

For the SBl and SB2 periods, pipe delta T values for steam-bubble operations -

were used, as dis:ussed previously. However, to account for a recommended -

future adjustment in spring travel, and a corresponding change in global moment response, cycles for each delta T condition werc determined by ratio, '

based on the number of heatups and cooldowns predicted to occur in each period.

The number of heatup/cooldown event cycles shown in Figure 2-4 for each period was determined based on the historical data review. An additional 5 heatup/cooldown cycles were added to the present (end of 1990) number determined for the S81 period for each unit, to account for operattuns until the modifications are made. The total cycles at each ATpipe value were distributed among the operating periods as a function of the number of heatup/cooldowns in each period. Tht distribution from Unit 3 was used for analysis, since it represented the worse case witi .;spect to global moment ranges.

Tne total transients for heatup and cooldown are identified as hcl thru HC9 for the pipe, ard hcl thru HC9 for the RCS hot leg nozzle, distributed among the respective n. . ' and SB2 periods, as shown in Tables 2-2(a) and 2-2(b),

respectively. Trai. ' . HC8 and HC9 for the pipe and HC9 for the nozzle -

represent transients which occur during later stages of the heatup.

2.7 \xial Stratification _ Profile Develonment 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

~

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3a ,c.e 5341s/061091:10 2-12

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j ac.e

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]a.C.e

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_Ja,c.e

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) , j a .C,9-5341s/061091i10' 2-13

Review and study of the monitoring data for all the;plarts revealed a .

consistent pattern of development of delta T as a functivn 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 3 and Unit 4 have similar surge line configurations, the profile applies to both units. [

3a ,c.e 2.8 Strioino Transient 1 The transients developed for the evaluation of thermal striping are shown in Table 2-3.

[

j a,c.e Striping transients use the_ labels HST and CST denoting striping transients (ST). Table 2-3 contains a summary of the HSTI to HST8 and CST) 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 striving _

in the s'Jrga line[3]. Section 5 contains more information on specifically how- '

the striping loading was considered in the fatigue evaluat.on.

l 5341s/061091:10 2 -_

TABLE 2 -

SURGE LINE TRANSIENTS HITH STRATIFICATION NORMAL AND UPSET TRANSIENT LIST ' TURKEY POINT UNIT 3 OR UNIT 4 TEMPERATURES (*F)

HAX NOMINAL LABEL TYPE CYCLES AT5trat PRZ T RCS T

[

.)a,c,e 3ee notes _on_next page 5341s/061091:10' 2-15

TABLE 2-1 (Cont'd.) -

SURGE LINE TRANSIENTS WITH STRATIFICATION NORMAL AND UPSET TRANSIENT LIST - TURKEY POINT UNIT 3 OR UNIT 4 TEMPERATURES (*F)

WsX NOP NAL LABEL TYPE CYCLES ATStrat PRZ T RCS T

[

E l

j a,C,e ,

5341s/061091:10 2 (

TABLE 2-2a SURGE LINE PIPE TRANSIENTS WITH STRATIFICATION - TURKEY POINT UNIT 3 OR 4 HEATUP/COOLDOWN (HC) - 200 HC CYCLES TOTAL TEMPERATURES (*F)

STRESS MAX NOMINAL LABEL TYPE CYCLES ATStrat PRZ T RCS T C

)a,c.e 5341s/061091:10 2-17

TABLE 2-2b .

SURGE LINE N0ZZLE TRANSIENTS WITH STRATIFICATION - TURKEY POINT UNIT 3 OR 4 .

HEATUP/COOLDOHN (HC) - 200 HC CYCLES TOTAL TEMPERATURES (*F)

STRESS MAX NOMINAL LABEL TYPE CYCLES ATStrat PRZ T RCS T

[

l .

l 1

3a .c.e l'

.5341s/061091:10 2-18 e -w ,- m -

,= - -% , -,~,

l.

l TABLE 2-3 SURGE LINE TRANSIENTS - STRIPING FOR HEATUP (H) and COOLDOWN (C) - INDIAN POINT UNIT 2 OR 3

[

Labal j_,c.e a -

P f

I V

5341s/061091:10 2-19 :-

9 1

a,c.e 9

Figure 2-1. Typical Insurge-Outsurge (I/0) Temperature Profiles - .

5288s/051391:10 2-20

4 F

a,c,e

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i

- 1 Typica) Honitoring Locations j

- F1gure 2-2.

1 l l j

' 5288s/051391: 10 2-21 t.

a,c.e i

Y M

L I

4 Figure 2-3. Sumcary of Historical Data Distribution from Turkey Point Units 3 and 4 Steam-Bubble Operations Compared to the Distribution Used in the Analysis

S288s/OS1391

10 k - - ,

.~

~

a,c.e Y

U Figure 2-4. Turkey Point Units 3 and 4 Operating History Periods Considered for Transient Development 5288s/051391:10

]

l a,c.e 3

5 Figure 2-5. Example Axial Stratification Profile for Low Flow Conditions .i i

1

,i 1

1 5288s/051391:10 2 a,c.e 4

. Figure 2-6. Geometry Considerations 5288s/051391:10 2-25

i

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N m

J

, Figure 2-7. Temperature Profile Analyzed for. Turkey Point Units 3 and 4 528Bs/051391:10 0 9

SECTION 3.0 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 developed in section 2.0. [

3a ,c.e 3.1 Surge Line layoyls The Turkey Point Units 3 & 4 surge line layouts are documented in references 6 ard 7 and the layout is shown schematically in Figure 3-2. The layout dimensions for the two Turkey Point units are identical. The support configurations of the two Turkey Point surge lines are similar. Below is a

. table summariaing the existing Turkey Point surge line support configuration. ,

l

- Turkey Point Units 3 and 4 g

Syooort Unit 3 __ Unit 4

_ Hade Tvoe RC-F12 HR-48 142 Pipe Whip Restraint-RC-F13 HR-49 118 Pipe Whip Restraint RC-F14 HR-50 87 Pipe Whip Restraint VS-1G-12. SR-400 110 Variable Spring Hanger 1

It can be seen from the table above that both of the Turkey Point surge lines l contain one variable spring hanger and three pipe whip restraints. In some cases these supports can cause higher thermal loads if displacements from-thermal stratification exceed available gap limits, and/or exceed the travel allowance in the spring hanger.

The piping sizes are 12 inch schedule 140, and the pipe material is stainless

. steel, SA 376-Type 316, for both units.

5341s/061091:10 3-1

_ .. _ .._ _ _ ___ -.. _. _ _ . _ . _ _ _ . . _ . . . _ . _ _ . . _ _ . _ . . _ _ _ _ ~ - . . _ . _

h 3.2 P_toing system Global St.tMetural Analyjit .

The Turkey Point Units 3 and 4 piping systems were modeled by pipe, elbow,. and non-linear spring elements using the ANSYS coraputer code described in Appendix -

A. The geometric and material parameters are included. [

ja,c.e ,

The thermal profile loading defined in section 2 was broken into l a.c.e Table 3-1 shows the loading cases considered in the analysis. To encompass all' plant operations, [

l l a.c.e Consequently, all the thermal transient loadings defined in.section 2 could be evaluated.

[

3a ,c.e .

1 5341s/061091:10 3-2

- !a order to moet the ASME Section III Code stress limits, a global structural

+

model of the surge lines was developed using the information-provided by references 6 and 7 and the ANSYS general purpose finite element computer -

]"'c.e to reflect the layout of straight pipe, bends and field welds ,

as shown in Figure 3-2.

For the stratified condition, [

]"'C These temperature distributions were established from the transients, as discussed in section 2.0. The maximum system delta T was taken as 320*F for the past and future conditions. This corresponds to (

3a .c.e The global piping s' tress analysis was based on two structural nodels for the Turkey Point Units. The first model represents the existing support configuration and the second'model represents the future support configuration of both units. The existing configuration has the actual gaps at all whip restraints and actual spring travel allowance from design. In the analysis, the spring can bottomed-out condition was predicted from the system AT of 320*F. The future configuration also has the actual gaps but no spring bottomed-out condition. In other words, the spring can will be adjusted to accommodate the thermal displacement. In addition, the' beneficial effect of insulation crushability was taken;into account for b'oth the_ existing and-future configurations.

The results of the ANSYS global structur11 analyses provide the thermal -

expansion moments. The ASME Section III equation (12) stress intensity range.

was evaluated for both units. For the Turkey Point units, the maximum ASME equation (12) stress intensity range in the surge line for a system delta T of

'320*F was found to be under the code allowable of 3Sm for tha. future-configuration, where spring travel allowance.will be modified to accommodate the thermal displacement from stratification. The conservative assumption off the spring can bottomed-out condition in the existing configuration causes an

~

EquationL(12) stress over the allowable. However, this is believed to be 5341s/061091:10 3-3

unrealistic since no evidence of a bottomed-out or overloaded condition was ,

observed from the plant inspections. Maximum equation (12) and equation (13) .

stress intensity ranges are shown in Table 3-2.

The pressurizer nozzle loads from thermal stratification in the surge-line 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 3 and 4 pressurizer nozzles, the maximum stress intensity range is 32.9 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 Turkey Point pressurizer surge nozzles met the code stress requirements.

3.3 Local Stresses-Heihadgjoav and- Results 3.3.1 Explanation of Local Stress Figure 3-3 depicts the local axial stress components in a beam with a sharply 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 symmetry of the geometry and thermal loading, only half of the cross section '

was required for modeling and analysis. [ Ja ,c e- -

5341s/061091:10 3-4

k

[

j a,c.e 3.3.3 Pipe local Stress Results Figure 3-6 shows the temperature distributions through the pipe wall [

,~ ~

3a ,c.e

[

3a c.e ,

5341s/061091:10 3-5

.. .. - . - . _ . . ..~ . - -. . .. . . . -. -.. . _ . _ - . . .

3.3.4 RCL Hot leg Nozzle Analysis

• Detailed surge line nozzle finite niement models were developed to evaluate the effects of thermal stratification. The 12 inch schedule 140 model is shown in Figure 3-10. Loading cases included [

Ja .c.e A summary of stresses in the RCL nozzle (location 1) due to thermal stratification is given in Table 3-3. A summary of representative stresses for unit leading is shown in Table 3-4, 3.4 10111 Stress from Global and Local Analyses

[

j a,c.e

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t j a,c.e 5341s/061091:10 3-6

, s __

._ ,.m. .,.. ,vr

4 3.5 Thermal Stricing 3.5.1 Background At the time when the feedwater line cracking problems in Ph'R's were first l

discovered, it was postulated that thermal oscillations (striping) may l significantly contribute to the fatigue cracking problems. These oscillations i were thought to be due to eithei 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 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 ' s examined during water model flow tests performed for the Liquid Metal F c Breeder Reactor (LMFBR) primary pipe loop. The stratified

. flow was observed to have a dynamic interface region which oscillated in a wave l

pattern. These dynamic oscillations were shown to produce significant fatigue damage (primary crack initiation). The same interface osciliations were observed in experimental studies of thermal striping which were performed in Japan by Mitsubishi Heavy Industries. The thermal striping evaluation process was discussed in detail in references 3, 8, 9, and 10.

3.5.2 Thermal Striping Stresses 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 osciliation of the hot and cold stratified boundary. (See Figure

, 3-12, which shows a typical temperature distribution through the pipe wall).

[

j a,c.e i .

5341s/061091:10 3-7

I The peak stress range and stress intensity was calculated from a 3-D finite ,

element analysis. [ .

Ja .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.

[

3a ,c.e 3.5.3 Factors Which Affect Striping Stress The factors which affect striping are discussed briefly below:

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5341s/061091i10 3-9

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TABLE 3-1 -

TEMPERATURE DATA USED IN THE ANALYSIS '

Max Type of System Analysis Pressurizer RCL T T Pipe Top Bot Operation AT(*F) Cases Temp (*F) Temp (*F) (*F). (*F) AT (*F)

[

3a ,c.e 4

T 5341s/061091:10- 3-10'

_- . . _ _ . - - . . - . - - - - . - - - . - . . . . - . _ . - - . - . ~ . . . . - ...- -.

TABLE 3-2' Summary of Turkey Point Units 3 & 4 Surge Lines

~

Thermal Stratification Strass .Results ASME Code _Eauation Stress Code Allowable UnLt_3 ' UALL4 (ksi)

[

3a ,c.e 5341s/061091:10 3-11

TABLE 3-3 TURKEY POINT UNITS 3 AND 4 SURGE LINES HAXIMUM LOCAL AXIAL STRESS AT ANALYZED LOCATIONS Profile Local Axial Stress (psi)

Location

• Surface Maximum Tentile Maximym Comoressive

[

j a.c.e b

See Figure 3-5 -

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l l

5341s/061091':10 3-12

l l

l . TADLE 3-4 l

SUMMARY

OF PRESSURE AND B;NDING INDUCED STRESSES

• IN A SURGE LINE RCL N0ZZLE FOR UNIT LOAD CASES

, All Stress in osi l Linearized Stress Peak Stress l

Intensity Ranag_ Intensity Rangg_

i Diametral Unit Loading Location Location Condition Inside Outside Inside Outside

[

l l

l ja c.e l

5341s/061091:10 3-13 I

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4 TABLE 3-5  !

STRIP!NG FREQUENCY AT 0 MAXIHUM LOCATIONS FROH 15 TEST RUNS 4 t

i >

.i

# Cycles  !

1 Frequency (HZ) Total  !

Duration

%  %  % in i Hin .(Luration) Max -(Duration) Ava (Duration) SS10Ddi ,

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l' figure 3-1. Schematic of Stress Analysis Procedure 5288s/051591:10- 3-15

.. .- __ .. - . - , = . . . . . . .~. . --- . . , , -- . , . - -.

1 l

. I 8,C, I

l l

~

Figure 3-2. Pressurizer Surge Line Layout: Turkey Point Units 3 & 4 --

' 5288s/051591:10 3-16

, _ , . . , y - --, ,--w , . - , . - . . - , . . , , ,

a.c.e

Figure 3-3. Local Axial Stress in Piping Due to Thermal Stratification l

l 3-17 5288s/051591:10

. . _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ . _ . . _ _ _ _ . _ _ _ - _ _ _ - - _ _ _ ~ _ _ _ _ _ _ _ - _ . . . _ _ - . . _ . - . . _ . . _ . . _ _ . _ _ - . _ _ _ . _ . . _ _ - _

l 4

a.Cie 1

i l

i I

l i

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Figure 3-4. Local Stress - Finite Element Models/ Loading .

5288s/051591:10 3 ._ -. . . .- , . . - . - , - - - . - .

b a.c.e i t

' ' Figure 3-5, Piping Local Stress Model and Thermal Boundary Conditions 5288s/051591:10 3 19 ~

1

a,c.e i

figure 3-6. Surge Line Temperature Olstribution at [ 1"'C Axial Locations 5288s/051591:10 3 '

I s.c.e 0 'C

Figure 3-7. Surge Line Local Arial' Stress Distribution at ( 3 Axial Locations 5288s/051591:10 3-21

a,c,e 1

9 l

i l

1 l

. 1 l

I 4

l a

Figure 3-8. Surge Line Local-Axial Stress on Inside Surface at ,

( Ja.c.e Axial locations-5288s/051591:10 3-22

t a,c c 1 I

h I

i Figure 3-9. Surge Line Local Axial Stress on Outside Surface at C l ac.e Axial Locations l

l 5288s/051591:10 3-23 1

i- _ _ _ - ~ . . . . _ _ _ , , _ , _, ,, ' ' * ' " ' ' ' ' =*c>y -- --,,,.....g ,, , , ,, _ ,_.

8,C,0 Figure 3-10. Surge Line RCL Nozzle 3-0 WECAN Hodel: 12 Inch Schedule 140 -

5280s/051591:10 3-24

4 a.c.e If I

l l

i Figure 3-11. Thermal Striping Fluctuation 5288s/051591:10 3-25

(

a.C.e Figure 3-12. Thermal Striping Temperature _01stribution .

528Bs/051591:10 3-26 l

. SECTION 4.0 DISPLACEMENTS AT SUPPORT LOCATIONS The Turkey Point Units 3 and 4 plant specific piping displacements at the whip ,

restraint and hanger locations along the surge line were calculated for the I thermal stratification and normal thermal loads.

Table 4-1 shows the surge line piping displacements, for the highest delta T conditions in Table 3-1, at the whip restraint and spring hanger locations for both units. The support configuration is based on the existing-whip restraint gaps, and both the existing and future spring hanger configurations, with and without " bottomed-out" spring hanger displacements, respectively. Table 4-2 shows the maximum surge line piping displacements from the normal operating thermal condition at the whip restraint and spring hanger locations for both units. t 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 7 whip restraint gaps are not smaller than those displacements listed in Tables

. 4-1 and 4-2. Furthermore, the travel allowance in_the spring hangers will be modified to accommodate the piping displacements shown in Tables 4-1 and 4-2 for the future operation of Units 3 and 4.

I e

5341s/06 109 :1 10- 4-1

_ _ . - _ _ _ _ _ . _ . ~ - - -

TABLE 4-1 .

Piping Displacement Under Maximum Stratified Conditions I

Stratified Condition with Pipe ai = 304*F Existina ConfiauratiDD II) [ylure Confiaura11RD( >

Node No. Ux (in) Uy_Lini ULiini Ux (ini Uy_U.nl Ur (in)

[

I I

l.

i 1

ja.c.e ,

l 1

5341s/061091:10- 4-2

P i

TABLE 4-2

. i r

Maximum Piping Displacement Under Normal Thermal Conditions i 1

i distina and _ Qinre Confiaurationt* _

t Hodg_t{o_,. Ox-(in) Uv (in) Qz_(101 i .

)

(

38 .c.e i

e

' Existing and future are the same support configuration for-this condition-I i

I r

-f 9

5 5341s/061091:10 4-3

,,,--~w-e__ ,, - m r- r e ------.w,,-w r--,.- ,- w n, w - mw m -,*e-t+=d- -* Va # se-+wr---*e ,e- y -r w -y.e r'-r--,--wv v < ww. vwn~r v,--v4,-~-y--wr.*<va

-- -----...m.. .. -

m-______- - - - ... -- m___---- -

_. .~_m.. ~~ . - _ . _ . . _ _ .

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a wrm,'-- A--,,-m-,,, ,,w,, .m.w--.mc, ,e .w w.er- ,.p,-tr.- .u... 6-..w-w,.-.,w.,--.~e,w, - . www .,,seo, ,e,-y%%- , vv yr y.ry~,----w- p .w+ +3 y c y.

SECTION 5.0 ASME SECTION III FATIGUE USAGE TACTOR EVALUATION ,

l 5.1 Mttttcul0102Y Surge line fatigue evaluations have typically been performed using the methods of ASME Section III, NB-3600 for all piping components [

I

) Because l of the hature of the stratification loading, as well as the magnitudes of the I stresses produced, tne more detailed and accurate methods of NB-3200 were employed using fini+e 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.

, 1 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, (

3 a 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 B of this report) to use the " latest ASME Section III requirements incorporating high cycle fatigue". Spech'ic requirements for 5341s/061091:10 5-1

- - .. - - - . . - ~ , . . , . -

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.

According 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 components under stratification loading. Since the NB-3650 requirements and equations correlate to those in ND-3200, the results of the fatigue evaluation are reported in terms of thG NB-3650 piping stress equations. These equations and requirements are summarized in Tables 5-1 and 5-2.

The methods used to evaluate these requirements for the surge line componente are described in the following sections.

5.1.2 Fatigue Stress Ecuations Stress Classification ThestressesinacomponentareclassifiedintheASkECodebasedonthe 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 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 teres in order to represent peak stresses caused by local concentration, such as notches and weld effects. The NB-3653 equations use simplified formulas to

• 5341s/061091:10 '5-2

determine nominal stress based on straight pipe dimensions. ( .

l t

ja.c.e i

for the RCL nozzles, three dimensional (3-0) finite element analysis was used as described in Section 3.0. (

j a.c.e 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 Yh-Tb term, classified as "Q" stress in NB-3200, which .

represents stress due to differential thermal expansion at gross structural ,

l discontinuities. (  ;

l Ja .c.e The impact of this on the selection of components for evaluation is discussed in Section 5.1.3.

1 5341s/061091:10 5-3

% " 9 y y -

T'f 9'YT -N-W W'M*'"*ME -'"i V 9

S.treulamJttations .

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. (

8 'C

1 This was done for specific components as follows:

[

[

,)a c.e 5341s/061091:10 5-4

l 4I

, [

l  !

. i l

l l

, j a .C,0 from the stress profiles created, the stresses for Equations 10 and 11 could b tec;emined fae any point in the section. Experience with the geometries and 103 ing showed that certait points in the finite element models v,nsisd ently oroduced the worst case fatigue stresses and resulting usage fe srs, .in each stratified axial location. (

j a .C,0 l

i L

5341s/061091:10 5-5 l

____._.____-a_m _.__ __m____._-__ _ _

________.____________.______.m___.-_-_.2--.-_________.___________________.____m . _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . - _ . _ . . _ , _ , _ _ _ . . _ . . _ _ _ _

. ., -,-~

[quation__12 Stre11 .

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 highast stressed location in the model. By evaluating the worst locations in this manner, the remaining locations were inherently addressed.

Egyation 13 S1I311 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 c OBE loads.

The "T6-Tb" term of Equation 13 is only applicable at structural ,

discontinuities. (

j a .C,0 1hermal_itttit_Ratchti '

The requirements of ND-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 compenents using applicable stresses due to pressure and stratification transients.

5341s/061091:10 5-6

- _ - . . - - _ - . . - - - . - _- _.-.. . . - - . - . _ . _ _ _ - - _ - =

t

. Allowable Stresiti ,

Allowable stress, Sm. was determined based on note 3 of figure NB-3222-1. For

~

secondary stress due to a temperature transient or therrSI expansion loads

(" restraint of free end deflection"), the value of Sm was taken as the average of the $m values at the highest and lowest temperatures of the metal during the transient. The metal temperatures were deter:nined from the transient definition. When part of the secondary stress has 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 analyfos and the considerations for controlling stresses in Section 5.1.2, [

l a,c.e The method to evaluate usage tactors using stresses determined according to Section 3.0 is described below.

5.2 Fatigue _Lliftge_. Lac tors Cumulative usage factors were calculated for the controlling components using the methods described in NB-3222.4(e), based on NS-3653.5. Application of -

these methods is summarlzed below.

Transient Loadufes and Combinatintts from the transients described in Section 2.0, specific loadcases were developed for the usage evaluation. (

l l

3a ,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.

5341s/061091:10 5-7

Usage factors were calculated at controlling locations in the component as .

folicss: -

1) Equation 10, Ke, Equation 11 and resulting Equation !4 (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 Ng ,

N ...N , were determined from Code figures I-9.2.1 and I-9.2.2, 2 n curve C, for austenttic stainless steels.

3) Using the actual cycles of each transient loadset, ng, n2 '"n' calculate the usage factors Ug , U 2 ...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 g is greater than 10 II cycles, the value of U gis taken as zero. .

[ .

8 ,C,0 3

4) The cumulative usage factor Ucum, was calculated as Ucum . U) +

U2 + ... + Un . To this was added the usage factor due to thermal striping, as described below, to oMain total Ucum. The Code allowable value is 1.0.

5341s/061091:10 5-8

. 5.3 Ea11 gut 0utto Thermal Stripjng 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.

[

\

l a,c.e The 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.

l l

l.

5341s/061091:10 5-9

- , .-,v-.v,.w-r e- - . ,-------#.*=v- r= 1 =r,-r'--*fv e** " #W+

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 Tigure 5-2. Figure 5-2 is conservatively represented by a series of 5 degree temperature steps. Each '

step lasts [ 3a.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 factor calculations, the total fluctuations per step is constant and becomes:

( j a.c.e Each striping transient is a group of steps with [ l a.c.e fluctuations per step. For each transient, the steps begin at the maximum AT and decreases by

[ ]"' 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 ths total cycles at a step. The total cycles were multiplied by [ Ja ,c.e to obtain total fluctuations. This results in total fluctuations at each step. This calculation is performed for each ,

step plateau from ( 1 8 'C

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

( la.c.e This is reflected in the results to be discussed below.

5.4 [Ltigue Usage Resultt 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 '

5341s/061091:10 5-10

_ ..-._ _ _ _ . _ . _ - . _ _ _ . _ - . . _ . _ . _ _ _ _ . _ _ - . _ _ . ~ . _ _ _ _ . _ . _ . _ ,

(

8 'C

1 The total stresses for all transients in the bounding set were used to form combinations to calculate alternating stresse; and resulting fatirle damage in the manner defined by the Coie. Of this total stress, ihe st v sses in the 12 inch pipe due to (

3a .c.e The maximum usage factor on Turkey Point surge lines occurred at ( ,

ja.c.e It is also concluded that the Turkey Point pressurizer surge nozzles meet the code stress allowables under the thermal stratification loading from the surge line and the transients detailed in reference (13), and meet the fatigue usage requir6ments of ASHE Section III, with a maximum cumulative usage factor equal to 0.41.

F

\

l 53415/061091:10 5-11

- = - - . _ . -- .. .. -- . . - . - . - - - - .. .

TABLE 4-i CODE /CkI'iERI A o ASME B&PV Code, Sec. III, 1986 Edition

- NB3600 NB3200 o Level A/B Service Limits

- Primary Plus Secondary Stress Intensity 135m (Eq.10)

- Simplified Elastic-Plastic Analysis Expansion Stress, S, 1 3Sm (Eq. 12) - Global Analysis

- Primary Plus Secondary Excluding Thermal Dending < 3Sm (Eq. 13)

Elastic-Plastic Pent'ty Factor 1.01 K,13,333 Peak Strass (Eq. 11)/Cumuistive Usage Factor (Ucum)

S ait"*

/,1 (Eq. 14) p Design Fatigue Curve

- - U cum 11.0 8

5341s/061091:10 5-13

lABLE 5-2

SUMMARY

Of ASME FATIGUE REQUIREMENTS Parameter Description Allowable (if applicable)

Equation 10 Primary plus st.sadary stress intensity; < 3Sm if exceeded, simplified elastic-plastic analysis may be performed Elastic-plastic penalty factor; required for simplified elastic-plastic analysis when Eq.10 is exceeded; applied to alternating stress intensity

. Equation 12 Er 'nsion stress; required for simplified < 35m elastic-plastic analysis when Eq. 10 is-exceeded Equat' i 13 Primary plus secondary stress intensity < 3Sm excluding thermal bending stress; required for siniplified 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 5341s/061091:10 5-13

l J

l l

a,c.e 4

Figure 5-1. Striping Finite Element Hodel .

1 I

(

1 l

! 5288s/050991:10 5-14 1

l l

~ ~

a,c,e 9

h Figt'e 5-2. Attenuation of Thermal Striping Potential by Holecular Conduction (Interface Have Height of One Inch) 528Bs/050991:10 5-15

4 L

SECTIch 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 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 Turkey Point Units 3 and 4. .

Following the general approach used in developing the surge line strati-fication transients for the HOG, a set of transients and a stratification profile were developed specifically for Turkey Point Units 3 and 4. A study was made of the historical operating experience at Turkey Point Units 3 and 4,

. 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 stress a ulysis in Section 3 and fatigue evaluation in Sectior, 5, it is not necessary to modf fy the existing whip restraint gaps for ASME Code acceptability, provided sufficient travel allowances are maintained in the variable spring hanger in each unit. The analysis assumed five additional heatup/cooldown cycles will occur before the spring hanger modification is completed.

The results of this plant specific analysis, along with the above support modification in the spring hanger, 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 Turkey Point units have completely satisfied the requirements of NRC Bulletin 88-11.

5341s/061091:10 6-1

A

._ J.43.4. 4-'as#.-r*s41e 4 4 4 5.dSW9 4 e.- 4.FJM dd.4 -N. if .M 3ze*pM- M m.4ha4*+.$*r -+4 4n.4 3W+_-n.--dd 4AG==nMJ os h dE-,4. eD,a 4. 4 4 # D. 4. - - -'a 4L. We4 m4.pM---,dde-n f g 54w.,5- .5 34. Q a'e 4a@M.. 4-9 1

-Q .

i k

f k

.. 4 t

}

h 5

+

4 4 -

4 1

s f

e Et

-Q-

' il .

SECTION 7.0 REFERENCES

1. Coslow, B. J., et al., "Hestinghouse Owners Group Bounding Evaluation for Pressurizer Surge Line Thermal Stratification", Westinghouse Electric Corp. HCAP-12277, (Proprietary Clacs 2) and WCAP-12278 (non-proprietary),

June 1989.

2. Coslow, B. J., et al., Westinghouse Owners Group Pressurizer Surge Line Thermal Stratification Program HUHP-1090 Summary Report," Hestinghouse Electric Corp. HCAP-12508 (Proprietary Class 2) and HCAP-12509 (non-proprietary), March 1990.

3. Coslow, B. J., et al., " Westinghouse Owners Group Pressurizer Surge Line Thermal Stratification Generic Detailed Analysis Program HUHP-1091 Summary Report," Westinghouse Electric Corp. HCAP-12639 (Proprietary Class 2) and HCAP-12640 (non-proprietary), June 1990.

4

4. ASME B&PV Code Section III, Subsection NB, 1986 Edition.

5. " Pressurizer Surge Line Thermal Stratification," USNRC Bulletin 88-11, December 20, 1988.

6. FPL letter, JPN-PTN-91-5019 " Turkey Point Units 3 & 4. Pressurizer Surge Line Thermal Stratification Analysis, I/S Mod 1312" File #PTP 100-14, January 23, 1991.

7. FPL letter JPN-PTN-91-5190, " Turkey Point Units 3 & 4, Comments on Pressurizer Surge Line Thermal Stratification Analysis, NRC Bulletin 88-011 I/S Mod 1312," File # PTP 100-14, May 31, 1991.

8. " Investigation of Feedwater Line Cracking in Pressurized Hater Reactor Plants " HCAP-9693, Volume 1, June 1990 (Proprietary Class 2).

9. Hoodward, H. S., " Fatigue of LHFBR Piping due to Flow Stratification,"

ASME Paper 83-PVP-59, 1983.

5341s/061091:10 7-1

10. Fujimoto, T., et al., " Experimental Study of Striping at the Interface of '

Thermal Stratification" in Thermal Hvdraulics in Nuclear Technology, K. H. '

Sun, et al., (ed.) ASME, 1981, pp. 73. f

11. Holman, J. P., Heat Transfer, McGraw Hill Book Co.,1963.

12. Yang, C. Y., " Transfer function Method for Thermal Stress Analysis:

Technical Basis," Hestinghouse Electric Corporation NCAP-12315 (Proprietary Class 2).

13. Series 84 Pressurizer Stress Report, Section 3.1, Surge Nozzle Analysis, December 1974.

1 9

9 5341s/061091'10

7-2

- _ - . . - - - ~ ~

b

. 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. HECAN

2. STRFAT2

3. ANSYS

4. FATRK/ CMS A.1 BECAN A,1,1 Descriotion HECAN is a Westinghoute-developed, general purpose finite element program. It contains universally 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

. fm 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 conduc tion 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 automatica1_1y 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 HECAN structural analysis. Such coupled thermal-stress analyses are a standard application used extensively on an industry wide basis.

5341s/061091:10 A-1

.- , .. .~. _ -.. .

_ _._ .. . _ _ . . . _ . _ ____._ m._ ... _ _ _ _ . _ _ _ . _ - . . ._ .. . ..

A.l.3 Program Vertfjcation ,

Both the WECAN program and input for the HECAN verification problems, currently numbering over four hundred, are maintained under configuration .

control. Verification problems include coupled thermal 4 tress analyses for the quadrilateral, triangular, brick, and wedge isoparametric olaments. These problems are an integral part of the HECAN quality assurance procedurst. When a change is made to HECAN, as part of the reverifica'cica process, the configured inputs for the coupled thermal-stress vrrification problems are used to reverify WECAN for coupled thermal-stress analyset.

A.2 STRFAT2 ,

A.2.1 Egscription STRFAT2 is a program which computes the alternating peak stress on the inside surface of a flat plate and the usage factor due to sttiping o,n the surface.

The program is applicable to be used for striping on the inside surface of a pipe if the prog'am 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.

't

5341s/061091

10 A-2 i

. 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.

A3 ANSYS A.3.1 Qascriotion ANSYS is a public domain, general pt* pose finite element code.

A.3.2 Feature Used The ANSYS elements used for the analysis of' stratification-effects in the i surge line are STIF 20 (straight pipe), STIF 60 (elSow and bends) and STIF14 (spring-damper for supports).

A.3.3 Procram Verification 1-As described in section 3.2,. the application of ANSYS for stratification has been independently verified by comparison to HESTDYN-(Hestinghouse piping analysis code) and HECAN (finite el' ment code). The results from ANSYS are also verified against closed fo'in solutions for simple beam configurations.

5341s/061091:10 A-3

A.4 FATRK/ CMS A.4.1 DescriDLlon FATRK/ CMS is a Westinghouse developed computer code for fatigue tracking (FAIRK) 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 ana NB-3600.

The code can be used both as a regular analys's 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 input data which include the weight functions for thermal stresses, the unit ,

bending stress, the unit pressure stress, the bending moment vs.

stratification temperatt res, 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

• nts out the total usage factors, and the transients pairing information i :h determine ,

the stress range magnitudes and numbar of cycles. The detailed stress data may also be printed.

l A.4.3 Proaram Verification i l

FATRK/ CMS is verified according to Westinghouse procedures with several levels of independent calculations.

5341s/061091:10 A-4 i

l

. APPENDIX 8 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.

1 l

l l

l i

5341s/061091:10 B-1

CPB No. 3150-0011 NRCB 88 11 i

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555

( December 20, 1988 NRC BULLETIN NO. 88-11: PRESSURIZER SURGE LINE THERMAL STRATIFICATION 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 addressets establish and implement a program to confirm pressurizer surge line integrity in view of the occurrence of thermal stratification and (2) require addressets to inform the staff of the actions taken to resolve this issue.

Description of Circumstances:

- The licensee for the Trojan plant has observed unexpected movement of the pressurizer surge line during inspections performed 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 cortacted 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 i resolved. The most recent investigation by the licensee confirmed tha':-

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 communication 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

._ ~

NRCB 38-11 December 20, '328 Page 2 of 5 -

Discussion:

Unexpected piping movreents are highly undesirable because of Ootential rign piping stress that may exceed design limits for fatigue ano stresses. 'he problem can be more acute when the piping expansior, is restricted, such as through Contact with pipe whip restraints. Plastic deformation can resu t, which pairmentcanoflead the to high local stresses, low cycle fatigue and functional in-

'ine.

Analysis performed by the Trojan licensee indicatec t*at tnermal stratification occurs in the pressurizer surge line curing restuo.

cooldown, and steady-state operations of the plant, During a typical plant heatup, water in the pressurizer is heated to abcut 440'F; a steem bubble is then formed in the pressurizer. Although the exact phenomenon is not thoroughly understood, as the hot water flows (at a very 'ew flowrate) from the pressurizer through the tsrge line to the hot-leg piping.

the hot water rides on a layer of cooler water, causing the upper part of tre The differential temperature could be as hipipe to be heated to a higher temp conditions during typical plant operations.gh as 300'F, Under based this on expected condition, differential thermal expansion of the pipe metal can cause the pipo to deflect signifi.

cantly.

For the specific configuration of the pressurizer surge line in the Trojan plant, the line deflected downward and when the sur '

whip restraints, it underwent plastic deformation, ge line contacted resulting two pi:e in permanent deformation of the pipe.

The Trojan event demonstrates that thermal stratification in the pressurizer The licensing basis according to 10 CFR 50.554 for all PW licensee meet the American Society of Mechanical Engineers Boiler and Pressure evalua~ tion when any significant differences are observ and the analytical results ser the hypothesized conditions. Staff evaluation indicates that the thermal stratification phenomenon could occur in all PWR surge lines and may invalidate the analyses supporting the integrity of the surge line.

The staff's concerns include unexpected bending and thermal striping (rapid oscillation of the thermal boundary interface along the piping inside surface) as they affect the overall integrity of the surge line for its design life (e.g., the increase of fatigue).

Actions Requested:

l Addressees are requested to take the following actions:  !

1. For all licensees of ope ating PWRs: i a.

Licensees 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. '

i 5

B-3

I NRCS 88 11 December 20, 1988 Page 3 of 6 i

i l ihis inspection shoulo detemine 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 January 1,1979) are requested to demonstrate that the pressuri:er I surge line meets the applicable design codes' and other FSAR and regulatory commitments for the licensed life of the plant, consider-t ing the phenomenon of- thennal stratification and thermal stricing in i

the fatigue and stress evaluations. This may be accomplisheo by rerforming 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 af ter January 1,1979), should complete the foregoing analysis within one year of receipt of this bulletin. Since 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 perfom the visual inspection in 1.a 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 consnitments 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 net show compliance with the.recuirements and licensing consuitment:

stated the. rein for the duration of the operating license, the licensee is requested to obtain plant specific data tions.onThe thermal stratification, thermal striping,.and line deflec-i 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 geometry is and selected, the licensee should demonstrate similarity in operation.

d.

Based on the applicable plant specific or referenced data, licensees l

-are recuested to update their stress and fatigue analyses to ensure compliance with applicable Code requirements, incorporating any observations from 1.a above.

later than two years after receipt of this bulletin.The analysis should be If a licensee

• Fatigue analysis should be performed in accordance with the latest ASME Section !!! requirements incorporating high cycle fatigue.

B-4

NRCB 28-11 Oecember 20, 1988 Page 4 of 6 is ur.6ble to show compliance with the applicable design codes and other FSAR and regulatory connitments, the licensee is requested 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 power license, applicants are requestec te demonstrate that the pressurizer surge line meets the applicable design codes and other FSAR and regulatory connitments 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 themel striping to ensure that fatigue and stresses are in compliance with applicable code limits. The analysis and hot functicnal testing should verify that piping thermal deflections result in no adverse consequences, such as contacting tha pipe whip restrain ~cs. 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 moniter the surge line for the effects of thermal stratification, beginning with hot functional testing, or obtain data througn 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, Appendix B and plant procedures.

Reporting Requirements:

1. Addressets shall report to the hRC 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.

• If compliance with the applicable codes is not demon 3trated for the full =

duration of an operating license, the staff may impose a license condition sutn that normal operation is restricted to tne duration that compliance is actually demonstrated.

B-5

NRCB 80 11 Decemt*r 20, 1988 page 5 of 6 2.

Addressees who cannot meet the schedule oescribed in items 1 or 2 of Actions Recu_e_sted are required to submit to the NRC within 60 days of receipt o'f this bulletin an alternative schedule with justification for the recuested schedule,

3. Addressees shall submit a letter within 30 days after the comoletion of these actions which notifies the NRC that the actions reouested in !ters Ib, la or 2 of Actions Recuested have beest perfomed and that the results are available for inspection. The letter shall include the justification fur continued cperation, if appropriate, a description of the analytical approaches used, and a sumary of the results.

Although not reouested by this bulletin, addressets 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 experience. In aedition, addressees are encouraged to review piping in other systems which ray experience themal stratification 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 determine if further generic comunications are in order.

The lettersATTN:

Comission, recuired above shall be addressed to the U.S. Nuclear Regulatory Document Control Desk, Washington, D.C. 20555, under oath

- or offirmation under the provisions of Section 1824, Atomic Energy Act of 1954 as amended.

Administrator.In addition, a copy shall be submitted to the appropriate Regional This request 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 pnysical changes, such as test equipment installation or component modification. -The estimated average raoiation exposure is approximately 3.5 person-rems per licensee response.

Coments 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 Execu-tive Office eutiding, Washington, D.C.

20503, and to the U.S. Nuclear Regula-tory Comission, Records and Reports Management Branch, Office of Administration and Resource Management,_ Washington, D.C. 20555.

l l

1 B-6

3 NRCB 88-11 December 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. .

['

a es . Rossi, Dir' ctor Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contacts: S. N. Hou, NRR (301) 492-0904 S. 5. Lee, NRR (301) 492-0943 N. P. Kadambi. NRR (301)492-1153 Attachments:

1. Figure 1

2. List of Recently Issued NRC Bulletins e

t' a-7 I

_ _ -_____ a

nww -

December .0, 153 Page i of Surge Line Strati"ication

'PZR

- 1 1 1 1 1 1 Hot Flow from Pressurizer Thot = 425'F

k. - '

Tst l

Stagnant Cold Fluid Tcold = 125*F Figure 1 B-8

. APPENDIX C TRANSIENT DE'/ELOPt4ENT DETAILS

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NRCB 86-11 December 20, 1988 Page 5 of 6 2.

Addressees who carinot meet the schedule oescribed in Items 1 or 2 of Actions Recuested are reouired to submit to the NRC within 60 days of receipt of this culletin an alternative schedule with justification for the recuested schedule.

3. Addressees shall submit a letter within 30 days after the completion of these actions which notifies the NRC that the actions recuested in lters Ib, id or 2 of Actions Requested have been performed and that the results are available for inspection. The letter shall include the N,tification fur continued operation, if appropriate, a description of tte analytical approaches used, and a summary of the results.

Although not requested by this bulletin, addressees are encouraged to work collectively to address the tecnnical concerns associated with this issue, as well as to share pressurizer surge line data and operational experience. In aedition, addressees are encouraged to review piping in other systems which ray experience thermal stratification 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 determine if further generic communications are in order, The letters ATTN:

Commission, reoutred above shall be addressed to the U.S. Nuclear Regulatory Document Control Desk, Washington, D.C. 20555, under oath

- or affirmation under the provisions of Section 182a, Atoc.ic Energy Act of 1954, as amended.

Administrator,In addition, a copy shall be submitted to the appropriate Regional This request 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 ted equipment installation or component modification. - The estimated average radiation exposure is approximately 3.5 person-rems per. licensee response.

Coments on the accuracy of this estimate and suggestions to reduc'e the burden may be directed to the Office of Management and Bud 3et Room 3208 New Execu-tive Offict Building, Washington, D.C.

tory Commission, Records and Reports Manageant Branch. Office of20503, and t

~

Administration and Resource Management. Washington, D.C. 20555.

I e

4 B-6

NRC8 88-11 December 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.

[f

. Rossi, Dir cto %r a es s Division of Operational Events Assessment Office of Nuclear Reactor Regulatien Technical Contacts: $. 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 on .. .

4 9

B-7 I

w n December 20, iBi Page I of

!- Succe Line Stratification

'PZR m1 1 1 .

Hot Flow from Pressurizer Thot = 425*F i

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. APPENDIX C TRANSIENT DEVELOPMENT DETAILS

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