ML20079F918

Structural Evaluation of Indian Point,Units 2 & 3 Pressurizer Surge Lines,Considering Effects of Thermal Stratification

ML20079F918
Person / Time
Site:	Indian Point
Issue date:	05/31/1991
From:	Bond C WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML100331749	List: ML20079F918 ML20079F926
References
WCAP-12938, NUDOCS 9110080323
Download: ML20079F918 (103)

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HESTINGHOUSE CLASS 3 HCAP-12938 l

Structural ivaluation of Indian Point Units 2 and 3 Pressurizer Surge Lines, Considering the Effects of Thermal Stratification May 1991 H. A. Gray P. L. Strauch S. Tandon T. H. Liu L. H. Valasek H. Yu h"

Verified by:

00 4' W Verified by:

C. B. Bond V. V. Vora (n

Approved by:

X 4

Approved by:

S. S Palus~amy, Manager R. B. PateI,# Manager Diagnostics and Monitoring System Structural Analysis Technology and Development Work Performed under Shop Orders IHXP-964 IBBP-964, IHXP-145 and IBBP-145 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 e 1991 Westinghouse Electric Corp.

5409s/081491:10

TABLE Of CONTENTS e

Scit10D Iltle EAge Executive Summary iii 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 Monitoring Results and Operational Practices 2-4 2.5 Historical Operation 2-8 l

2.6 Development of Heatup and Cooldown Transients 2-9 2.7 Axial Stratification Profile Development 2-13 2.8 Striping Transients 2-15 1

3.0 Stress Analysis 3-1 3.1 Surge Line Layouts 3-1 3.2 Piping System Global Structural Analysis 3-2 3.3 Local Stresses - Methodology and Results 3-4 3.4 Total Stress from Global and Local Analyses 3-6 3.5 Thermal Striping 3-6 4.0 Displacements at Support Locations 4-1 5409s/081491:10 i

TABLE Of CONTENTS (Coatinued)

Sstilan lilla Eage 5.0 ASME Section III Fatigue Usage factor Evaluation 5-1 5.1 Hothodology 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 ~

Analysis Summary 6-1 7.0-

-References 7-1 Appendix A List of Computer Programs A-1 Appendix B USNRC Bulletin 88-11 B-1 Appendix C Transient Development Details C-1 l'

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-5409s/081491:10 li l

EXECUTIVE

SUMMARY

Thermal stratification has been identified as a concern which can affect the structural integrity of piping systems in nuclear plants since 1979, when a leak was discovered in a PWR feedwater line.

In the pressurizer surge line, stratification can result from the difference in densities 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 integrity concerns.

Study of the surge line behavior has concluded that the largest t.mperature differences occur during certain modes of plant heatup and cooldown.

This report has been prepared to support compliance with the requirements of NRC Dulletin 08-11 for Indian Point Units 2 and 3.

Prior to the issuance of the bulletin, the Westinghouse Owners Group had a program in place to investigate the issue, and recommand actions by member utilities.

That program provided the technical basis for the analysis reported here for Indian Point Units 2 and 3.

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 unit.

This transient information was used as input to a struttural and stress analysis of the surge line for the two units.

A review and comparison of the piping and support configurations for the units led to the conclusion that the surge lines are nearly identical, and thus one analysis could be done to apply to both units, for the stratification transient development and structural evaluation.

The existing configurations for both Indian Point units have been analyzed in this HCAP.

The analysis results are provided in Section 3 for ASME code stress, Section 4 for piping displacements at support and restraint locations and Section 5 for ASME Code fatigue cumulative usage factors.

5409s/081491:10 iii

SECTION

1.0 BACKGROUND

AND INTRODUCTION Indian Point Units 2 and 3 are four-loop pressurized water reactors, designed to be as nearly identical as practical, in both hardware and operation.

This report br* 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.

The operation of a pressurized water ru-tor requires the primary coolant loops to be water solid, and this is accomplished through a pressurizer

-vessel, connected to one of the hot legs by the pressurizer surge line.

A typical four-loop arrangement is shown in Figuie 1-1, with the surge line highlighted.

The pressurizer vessel contains steam and water at saturated conditions with

'the steam-water interface level typically between 25 and 60% of the volume depending on the plant 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 presture cushion effect in the event of sudden changes in Reactor Coolant Systen (RCS) mass inventory.

Spray operation reduces systera pressure by condensing some of the steam.

Electric heaters, at the _ bottom of t'ne pressurizer, are energized to raise the liquid temperature to generate additional steam and increase RCS pressure.

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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 Units 2 and 3 are both 14 inch schedule 140 stainless steel.

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, 9

5409s/081491:10 1-1

l gap closures in the pipe whip restraints, and in some cases unusual snubber movement.

Investigation of this problem revealed that the movement was caused by thermal stratification in the surge line.

Thermal stratification had nr,t 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 December of 1988. A copy of this bulletin is included as Appendix B.

The bulletin requested utilities to establish and implement a program to confirm the integrity of the pressurizer surge line.

The program required both visual inspection of the surge line and demonstration that the design requircments of the surge line are satisfied, including the consideration of stratification effects. Visual inspections were conducted in accordance with task la of the Bulletin at both Indian Point Units 2 and 3 [16), [17).

Prior to the issuance of NRC Bulletin 88-11, the Westinghouse Owners Group had implemented a program to address the issue of surge line stratification.

A bounding evaluation was performed and presented to the NRC in April of 1989.

This evaluation compared all the WOG plants to those for which a detailed plant specific analysis had been performed.

Since this evaluation w;s unable to demonstrate the full design life for all plants, a generic justification for continued operation was developed for use by each of the WOG plants, the basis of which was documented in references [1] and [2].*

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

5409s/081491:10 1-2

The Westinght Owners Group implemented a program for generic detailed analysis in June 1989, and this program involved Individual detailed analyses of groups of plants.. This approach permitted a more realistir.

approach.than could be obtained from a single bounding analysis for all plants, and the results were published in June of 1990 (3).

The followup to the Westinghouse Owners Group Program is a performance of evaluations which could not be performed on a generic basis, The goal of this report is to accomplish these followup actions, and to therefore support L

completion of the requirements of NRC Bulletin 80-11 for Indian Point Units 2 and 3.

1.? EelCI.1Rt101.Af_Sgtge Line Thermal Stratification It will be useful to describe the phenomenon of stratification, before dealing with its effects.

Thermal stratification in the pressurizer surge line is the i

direct result of the difference in densities between the pressurizer water and the generally cooler RCS hot-leg water.

The warmer, lighter pressurizer water tends to float on the cooler, heavier hot leg water.

The potential for stratification is increased as the difference in temperature between the pressurizer and the hot leg increases and as the insurge or outsurge flow rates decrease.

At power, when tne difference in temperature between the pressurizer and hot leg is relatively small,_ the extent and effects of stratification have been

' observed to be small.

However, during certain modes of plant heatup and cooldown, this difference in system temperature could be as large as 320'F, in which case the effects of stratification are significant, and must be accounted for.

Thermal stratification in the surge line causes two effects:

o Bending of the pipe different 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.

5409s/081491:10

.1 -3

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1.3 keptoilior.k The primary purpose of this work was to develop transients applicable to the Indian Point surge lines which include the effects of stratification, and to evaluate the structural integrity of the surge lines.

This work will therefore support the demonstration of compliance with the requirements of NRC Bulletin 88-11.

The transients were developed following the same general approach originally established for the Westinghouse Owners Group.

Conservatisms inherent in the original approach were refined through the use of monitoring results, plant operating procedures, operator interviews, and historical data on plant operation.

This process is discussed in Section 2.

I 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 loads and nozzle loads were determined.

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

The stresses were used to perform a fatigue analysis for the surge line, and the methodology and results of this work are discussed in Section 5.

The summary and conclusions of this work are summarized in Section 6.

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SECTION 2.0 SURGE LINE TRANSIENT AND TEMPERATURE PROFILE DEVELOPMENT 2.1 GetitallpnoKh The transients for the pressurizer surge line were developed from a number of sources, including the most recent Hestinghouse Systems Standard Design transients.

The heatup and cooldown transients, which involve the majority of the severe stratification occurrences, were developed from review of the design transients, plant operating procedures, operator interviews, monitoring data and historical records for each unit.

The total number of heatup and cooldown events specified remains unchanged at 200 each, but a number of transient events within each heatup and cooldown cycle have been defined to reflect stratification effects, as described in more detail later.

The normal and upset transients, except for heattp and cooldown, used for the Indian Point Units 2 and 3 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 between the pressurizer and hot leg, as listed in the table.

The transients have been characterized as either insurge/outsurges (I/0 in the table) or fluctuations (F).

Insurge/outsurge transients are generally more severe, because they result in the greatest temperature change in the top or bottom of the pipe.

Typical temperature profiles f or insurges and outsurges are shown in Figure 2-1.

Transients identified as fluctuations (f) typically involve low surge flow rates and smaller temperature differences between the pressurizer and hot leg, so the resulting stratification stresses are much lower.

This type of cycle is important to include in the analysis, but is generally not the major contributor to fatigue usage.

5409s/081491:10 2-1

i In addition to the p) ant specific operating history discussed above, the development of transients which are applicable to Indian Point Units 2 and 3 was based on the work already accomplished under programs completed for the Westinghouse Owners Group [1,2,3).

In this work all the Westinghouse plants were grouped based on the similarity of their response to stratification.

The three most important fac1 ors influencing the effects of stratification were found to be the structural layout, support configuration, and plent operation.

The transient development for the Indian Point units took advantage of the similarity in the surge line layout for the two units, as well as general similarities in the operating procedures. A detailed comparison of the piping and support configurations for the units appears in Sectior 3.1.

The transients developed here, and used in the structural analysis, have taken advantage of the monitoring data cellected during the WOG program, as well as historical operation data for the Indian Point units.

Each of these will be discussed in the sections which follow.

2.2 Eyitem Desian Informg11gn The thermal design transients for a typical Reactor Coolant System, including the pressuriztr surge line, are defined in Westinghouse Systems Standard Design Criteria.

The design transients for the surge line consist of two major categories:

(a) Heatup and Cooldown transients (b) Normal and Upset operation transients (by definition, the emergency and faulted transients are not considered in the ASME Section III fatigue life assessment of components).

5409s/081491:10 2-2

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

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2.4 Bonitorina Relul.ts and Optrational_fra.ttites 2.4.1 Monitoring Monitoring information collected as part of the Westinghouse Owners Group generic detailed analysis [3] w>e utilized in this analysis.

The pressurizer i

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 propcrly insulate the area against heat loss due to thermal convection or radiation.

l 5409s/081491:10 2-4

The typical temperature sensor configuration at a given pipe location consists of two to five sensors mounted as shown in Figure 2-7.

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 (

aJ.c.e intervals or less, during periods of high system delta 1.

Existing plant instrumentation was used to record various system parameters.

These system parameters were useful in correlating plant actions with stratification in the surge line. A list of typical plant parameters menitored is given below.

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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 f rom the utility plant computer.

2.4.2 Operational Practices Based on a review of the Indian Point Units 2 and 3 heatup and cooldown operating procedures and operational interviews conducted at a number of H0G 5409s/081491:10 2-5

i utilities (inzluding Indian Point Unit 2 in September,1989), it was determined that both units neat up and cool down in manners somewhat similar

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'to other plants that heat up with a steam bubble in the pressurizer.

Heatups

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and cooldowns are used here to characterize plant operation because they

,l represent the periods during which the temperature dif ference between the pressurizer and the hot leg is potentially the greatest.

Brief descriptions of the Indian Point units' heatup and cooldown procedures follow.

U.alL2 The heatup and cooldown procedures at Indian Point Unit 2 are generally similar to those used at other steam bubble plants, but utilize a ettrogen j

bubble to maintain RCS pressure during periods early in the heatu9 or late in the cooldown. The heatup procedure begins_with the RCS full and the pressurizer level at 75-85L At this point, a nitrogen bubble is established i

in the pressurizer at a pressure of 400-450 psig.

The pressurizer level and RCS pressure are maintained in these ranges during nitrogen bubble operation using charging flow letdowr flow and nitrogen supply. After the nitrogen i

bubble is established, the RCS is vented, and at least ene reactor coolant pump (RCP) is started before the RCS temperature exceeds IBO'F.

Oxygen levels are also checked and hydrazine added, if necessary, with the RCS held between 170'F and 190'F.

After this, a steam bubble is esta's11shed in the pressurizer by energizing the pressurizer henters,-reaching taturation temperature at RCS pressure of 400-450 psig. After the steam bubble is formed, pressurizer spray is begun.

The nitrogen is returned into solution and transferred from the pressurizer to the volume control tank.

Nortral pressurizer level is established, and RHR is

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isolated, once the RCS temperature is approximately 350*F.

The remaining RCP's are started to complete the heatup withic. :,wcified limits.

Plant procedures allow a maximum RCS heatup rate of 50'F/hr in the range of 350*F, and-a-maximum allowable pressurizer heatup rate of 100*F/hr.

During the entire process, a limit of 320*F is imposed on the difference between pressurizer and spray fluid temperatures.

This inherently imposes the same limit on the difference between pressurizer and hot leg temperature during this period of operation.

5409s/081491:10 2-6

The cooldewn proce;s is essentially the reverse of the heatup. Once RCS pressure is below 450 psig and temperature is below 350'F, RHR is put in service, and the steam bubble is exchanged for a nitrogen bubble in the pressurizer.

Pressurizer spray and letdown are used to control the process.

The cooldown is completed using RHR, with at least one RCP operating until the RCS temperature is below 180'F.

VALL.3 The heatup and cooldown procedures for Indian Point Unit 3 tre typical of most plants that use the steam bubble method.

At the beginning of the heatup, the RCS is filled and vented, and pressurized to 400 450 psig, using charging ano letdow, ficw to maintain the pressure.

(In the past, nitrogen was also an option for controlling RCS pressure at this stage, but according to plant operations, has not been used extensively and is not used presently.) At least one RCP is started, and various checks (including chemistry and oxygen) are performed during the ensuing process prior to reaching itmits of 200'F and 250'F.

After the RCP is started, with the RCS solid and pressure greater than 400 psige a steam bubble is established in the pressurizer.

This is accomplished by closing the spray valves and energizing the pressurizer heaters until the steam bubble is drawn.

RCS pressure is maintained between 400-450 psig using letdown flow, charging and/or pressurizer heaters until the bubble is formed, at which point the pressurizer level is decreased to 327. using charging and letdown.

The RCS pressure is then controlled using pressurizer spray and/or heaters. Once the RCS pretsure is stabilized (between 400 and 450 psig) and pressurizer normal water level is established, the RHR is shut down and isolated.

From this point, the plant heatup is commenced by starting the remaining RCP's and using pressurizer heatert, if necessary, to supplement the heatup rete.

The cooldown procedure is basically the reversa of the huatup.

During these processes, administrative limits are imposed as follows:

RCS allowable heatup rate of 50*F/hr; pressurizer allowable heatup rate of 100*F/hr; maximum allowable delta T bet',,een the pressurizer and reactor I

5409s/081491:10 2-7 1

1

coolant loops of 320'F; spray should not be used if delta T between the pressurizer and the spray fluid exceeds 320*F.

Laux From the operating procedures, the possible ranges of the syster feita T could vary, but are bounded by the adtrinistrative limit of 320*F.

The actual impact of these pla'nt opirating procedures on the analysis was determined in conjunction with review of the plants' past onerating histories, and is discussed in the following section.

2.5 lii. sip.tinl_DAtration Historical records from the plants (oparator logs, surveillance test reports, etc.).:ere reviewed in December,1990, supplemented with confirmatory investigations of the data by the utilities [14, 15].

The purpose of the review was to obtain a distributtoa of maximum system delta T, and to identify heatup or cooldown evu where the maximum system delta T exceeded the 320*F limit.

To date (as of April,1991), Unit 2 has experienced 85 heatups and 85 cooldowns, and Unit 3 bas experienced 37 heatups and 36 cooldowns.

The data available represents only a portion of these events.

Therefore, the delta T distribution is expressed in terms of the events in a predetermined range as a percentage of the total number of events for which data was available.

A surnmary of the results for available data is presented below, Un1t 2 Ull.t_3 Number of Number of System AT Heatups or

% of Heatups or i ef hun _ff.1 Cnolduni 10111 CoWnat Iatal I

ja.c.e Total events 7

28 5409s/081491: 10 2-8

for Unit 2, pressurizer temperature was not specified in the data initially available for review, which represented 58 events, either heatups or cooldowns.

Since a nitrogen babble is used to maintain RCS pressure during periods where system delta T is typically maximum in steam-bubble plants, the corresponding saturation temperature at the indicated RCS pressure could not be used.

For a sutset of seven of these events, available plant computer data that did indicate pressurizer temperature was investigated, and it was found that actual system delta T's were less than 320'F for the events investigated (14).

Since plant procedures inherently specify a limit on system delta 1 of 320'F (see Sectica 2.4 2), it was assumed that all past events for Unit 2 had system delta T values below 320'F.

This assumption is conservative with respect to plant operations until May, 1977, when water-solid heatup and cooldown operations were used.

Typical niaximum system delta T for water-solid plant operation is 210'F (3).

For IJnit 3, there was some recorded data that could leMi to the conclusion that the 320'F system delta T limit had been exceeded.

An investigation was made by the utility (15), which concluded that the 320*f administrative limit had not been exceeded for past operation.

Thus, the events in question were assumed to occur with a maximum system delta T of 320*F.

Data for Unit 3 was available for 28 events, either heatups or cooldowns, that occurred in 17 of the past heatup/cooldown cycles.

The 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 oistribution is discussed further in Section 2.6.)

[

]a,c.e from this comparison, it is evident that if future operation for Unit 2 continues as assumed for the past, the analysis is conservative.

2.6 DnelopmeALaf_ Rain _qndlp.oldorn Transigah The heatup and cooldown transients used in the analysis were developed from a number of sources, as discussed in the overall approach.

The 5409s/081491:10

'-9

transients were built upon the extensive work done for the Westinghouse Owners Group (1,2,3), coupled with plant specific considerations for Indian Point Units 2 and 3.

The transients were developed based on monitoring data, hi'torical operation and operator interviews ;onducted at a large number of piants.

For each monitoring location, the top-to-bottom differential temperature (pipe valta 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 WOG(1,2,33 effort, individual plants' monitoring data was reduced to categorize stratification cycles (changes in relatively steady-state stratified conditions) using the rainflow cycle counting method.

This method considers delta T range as opposed to absolute values.

(

)a,c.e The resulting distributions (for I/O transients) were cycles in each RSS range above 0.3, for each mode.

(In the surge line analyses, RCS temperature ranges of 1200'F, 200-350*F, >350*F at hot standby, and >350'F during startup were labelled as modes 5,4,3 and 2, respectively).

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,. rare is margin of 100%

on the average number of cycles per heatup in each. node of operation.

5409s/081491:10.

2-10

Transients, which are represented by delta T pipe with a corresponding number of cycles, were developed by combining the delta T system and cycle dist'ibutions.

For mode 5 delta T system is represented by a historical distribution de'; eloped from plant operating records f rom a number of plants, and is represerted in Figure 2-3 as "used in analysis".

As discussed in Section 2.5, this historical delta T system distribution was assumed to encompass the prior operating history of the Indian Point

-units, and to account for future operation.

For modes 4, 3 and 2, the delta T system was defined by maximum va'ues.

The values were based on the maximum system delta T obtained from the monitored plants for each mode of operation.

An analysis was conducted to determine the average number of stratification cycles per cooldown relative to the averaga number of stratification cycles per heatup.

I J c.e The transients a

for all modes were then enveloped in ranges of ATpipe' i' "II range were added and cycles from transients within each ATpipe assigned to the pre-defined ranges.

These cycles were then applied in the fatigue analysis with the maximum AT for each range.

The values olpe used are as follows:

E01_C1cles Within PiggJgitit T Range Elp.qJgLtt'.

[

Ja c.e This grouping was done to simplify the fatigue analysis.

The final result of this complex process is a table of transients corresponding to the subevents of the heatup and cooldown process.

The actual number of transient cycles used in the analysis to represent 200 5409s/081491:10 2-11

heatup/cooldown event cycles is shown in Table 2-2.

A mathematical description of the methodology used is given in Appendix C.

[

1 l.c.e The critical location is the location with the highest j

a combination of pipe delta T and number of stratification cycles.

Because of main coolant pipe flow effects, 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 stratification was noted only when the reactor coolant pump in the loop with the surge line was not operating.

Transients were then developed using a conservative number of ' pump trips."

[.

aJ.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 frora the nozzle transients, and the pressure and bending loads from the piping transients.

The total transients for heatup and cooldown are identified as hcl thru HC9 for the pipe, and hcl thru HC9 for the RCS hot leg nozzle as shown in Tables 2-2(a) and 2-2(b), respectively.

Transients HC8 and HC9 for the 5409s/081491:10 2-12 l

~ - -

1 pipe and HC9 for the nozzle represent transients which occur during later stages of the'~heatup.

2.7 AxiaLS.traltficath03r.o.t.UI Hesslormtat

-In addition-to transients, a profile of the-(

D-3.c.e 4

Two -types of profile envelope the stratifitd terverature distributions observed 'and-predicted to occur in the line Dese two profiles are a

[

a3,c.e I

l, C

s j,c.e a

[

j-p.

Ja c.e l

l-5409s/081491.10 2-13 y

i

(

)

ja.c,0 C

\\

3 c.e.

a

' Review and study of the monitoring data for all the plants revealed a consistent pattern of development-of delta T as a function of distance from the hot. leg intersection.

This pattern was consistent throughout the heatup/cooldown process, for a given plant geometry. This pattern was used along with plant operating practices to provide a realistic yet somewhat

. conservative portrayal of the pipe delta T along the surge line.

The combination of the hot / cold interface and pipe delta T as functions of distance along the surge line forms a profile for each individual plant analyzed.

Since Unit 2 and Unit 3 have similar surge line configurations, the profile applies to both units.

[

3,c,e a

5409s/081491:10 2-14 m

i-2.8 $1rfging._Itan11co.t1 The transients developed for the evaluation of thermal striping are shown in Tacle 2-3.

L 3,c.e a

Striping transients use the labels HST and CST denoting striping transients (ST). Table 2-3 contains a summary of the H5TI 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 striping in the surge linef3).

Section 5 contains m..e informatun on specifically how the striping loading was considered in the fatigue evaluation.

5409s/081491: 10

?-15

TABLE 2-1 SURGE LINE TRANSIENTS HITH STRATIFICATION 1

NORMAL AND UPSET TRANSIENT LIST - INDIAN: POINT UNIT-2 OR UNIT 3 TEMPERATURES (*F)

MAX NOMINAL i

LABEL TYPE CYCLES ATStrat:

-PRZ T RCS T l

(

l i

i

=

4 l

I 1-

1. -

i.

l I

i l

t f

)a,c.e See' notes on next page t-5409s/081491:10 2-16

o TABLE 2-1 (Cont'd.)

SURGE LINE TRANSIENTS HITH STRATIFICATION NORMAL AND UPSET TRANSIENT LIST - INDIAN POINT UNIT 2 OR UNIT 3 TEMPERATURES (*F)

MAX NOMINAL lab;L TYPE CYCLES ATStrat PRZ T RCS T

[

3,c.e a

5409s/081491:10 2-17

TABLE 2-2a LSURGE LINE PIPE: TRANSIENTS HITH STRATIFICATION - INDIAN POINT UNIT 2 OR.3

- HEATUP/COOLDOWN (HC) - 200 CYCLES TOTAL

.. l TEMPERATURES (*f)

MAX NOMINAL

- LABEL

' TYPE

-CYCLES ATStrat PRZ T RCS T

[-

1 1-

.ja,c,e 3.

l 1-4 i

5409s/081491:10 2-18 t-Y a

y u,ett 4

g

TABLE 2-2b SURGE LINE N0ZZLE TRANSIENTS HITH STRATIFICATION - INDIAN POINT UNIT 2 OR 3 HEATUP/COOLDOWN (HC) - 200 CYCLES TOTAL TEMPERATURES (*f)

HAX NOMINAL LABEL TYPE CYCLES ATStrat PRZ T RCS T l

Ja,c.e 5409s/081491:10 2-19

l TABLE 2-3 SURGE LINE TRANSIENTS - STRIPING FOR HEATUP'(H) and C00LDOWN (C) - INDIAN POINT UNIT 2 OR 3

[

3,c,e a

5409s/081491:10 2-20

~ a,c.e

~

i l

l t

.J Figure 2-1.

Typical Insurge-Outsurge (I/0) Teraperature Profiles 5409s/081491:10 2-21

i

~

a,c.e n

Figure 2-2.

Typical Monitoring Locations t

5409s/081491:10 2-22

l C

e,c,e 2;

8%

T G

P E!

l 1

I mr Figure 2-3.

Surreary of Historical Data Distribution from Indian Point Units 2 and 3, Compared to the Distribution Used in the Analysis

.~.... _

i l

I a,c.e i

e W

1 m

Figure 2-4. ~ Example: Axial Stratification Profile for Low Flow Conditions 5409s/081491:10 2-24

a.C.O l

i-i-

i l

i I

I Figure 2-5.

Geometry Considerations I

5409s/081491:10 2-25 I

e.

.a c e

t -

Il 1

,=

~

Figure-2-6.

Temperature: Profile Analyzed for Indian Point Units 2 and 3 VE

5409s/081491:10 2-26

=

=

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

SECTION 3.0 STRESS ANALYSIS 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.

(

i J

a3 c.e

-3.1-Surge _Line Lay.ouis The Indian Point Units 2 & 3 surge line layouts are documented in references 6 and 7, and the layout is shown schematically in figure 3-2.

The layout dimensions for th9 two Indian _ Point units are identical. The support configurations of-the two Indian Point surge lines are similar.

Below is a table summarizing the existing Indian Point surge line support configuration.

I Indian Point-Valts 2 and_3 Support Unit 2

.Unil 3

-Enda Type

-PWR-122 PHR-122 1500 Pipe Whip Restraint PHR-123 PHR-123 2100 Pipe Whip Restraint PWR-124 2700 Pipe Whip Restraint PWR-120 PWR-120 3400 Pipe Whip Restraint PWR-121 PHR-121 3500 Pipe Whip Restraiht PWR-125 PWR-125 4300 Pipe Whip Restraint RCH-76 PH-H-63-1 3100 Variable Spring _ Hanger RCH-78 3450 Swing ~ Brace S409s/081491:10 3-1

____.m It can be seen from the table above that both of the Indian Point surge lines contain one variable-spring hanger with Unit 2 having five pipe whip

~

restraints and Unit 3 having six pipe whip restraints.

In addition, there is a horizontal _ sway brace in the Unit 2 surge line.

In some cases these supports can cause higher thermal-loads if displacement from thermal j

stratification exceed available gap limits.

The piping sizes are 14 inch schedule 140, and the pipe material is stainless steel, SA 376-Type 316, for i

both units.

3.2 Elpilt1_Sntom GlqhA]_SitutEg1 Analysis The Indian Point Units 2 and 3 piping systems _were modeled by pipe, elbow, and non-linear spring elements using the ANSYS computer code described in Appendix A.

The geometric and material parameters are included.

[

a 3,c.e a

Each thermal profile loading defined in section 2 was broken-into [

]a.c.e Table 3-1-shows the loading' cases. considered in the analysis.

To encompass all plant operations, [

g,c.e a

5409s/081491:10 3-2

J c.e Consequently, all a

[

the thermal transient loadings defined in section 2 could be evaluated.

[

j

.C,0 a

In order to meet the ASME Section III Code stress limits, global structural models of the surge lines were developed using the information provided by references 6 and 7 and the ANSYS general purpose finite element computer code.

Each model was constructed using [

3,c.e to reflect the layout of straight pipe, bends and field welds a

as shown in. Figure 3-2.

For the stratified condition, [

3,c.e These temperature distributions were established from a

~

the transients, as discussed in section 2.0.

The maximum system delta T was taken as 320*F for the future condition.

This corresponds to [

3.c.e a

The global piping stress analysis was based on two structural models for the Indian Point Units.

The first model represents the existing support configuration of Unit 2 and the second model represents the existing support configuraticn of Unit 3.

The existing configuration has the actual gaps at al, whip restraints.

In the analysis, no spring can bottom-out condition was assumed.

This assumption will be assured and verified for Indian Point Units 2 and 3 by Con Ed and NYPA respectively.

In addition, the beneficial effect of insulation crushability was taken into account for the existing configuration.

The results of the ANSYS global structural analyses provide the thermal expansion moments.

The ASME Section III equation (12) stress 5409s/081491:10 3-3

intensity range was evaluated for both units.

For the Indian 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 [4] of 3Sm-for the existing configuration, without the spring hanger bottomed out.

,j 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 based on no spring bottomed out configuration, were also evaluated according to the requirements of the ASME Code [18).

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 2 and 3 pressurizer nozzles, the maximum intensity range is 44.9 ksi compared to the code allowable value of 57.9 ksi for a material of SA 216 Grade WCC. The maximum fatigue usage factor will be reported in Section 5.

It was found that the Indian Point p.essurizer surge nozzles met the code stress requirements.

3.3 Local Stresses-Re_thodology and Results

i 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.

[

3.c e a

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 medel with thermal boundary conditions is shown in Figure 3-5.

Due to 5409s/081491:10 3-4

J symmetry of the geometry and thermal loading, only half of the cross _section was-required for n, deling and an& lysis.

[

j i

t i

3.c.e a

E 3.3.3 Pipe Local Stress.Results 1

. Figure 3-6 shows the temperature distributions through the pipe wall-[

(..

I'.

f aj,c.e 3.3.4 RCL Hot Leg Nozzle Analysis Detailed; surge line nozzle finite element models were developed to evaluate the effects of thermal stratification.

The 14 inch schedule 140 model is shown in Figure 3-10.

Loading cases included [

J,c.e A summary of stresses in the RCL nozzle a

(location 1) due to thermal stratification is given in Tables 3-3A and 3-38.

A summary of representative stresses for unit loading is shown in Table 3-4.

5409s/081a91:10 3-5

l l

3,4 Total StrR11 f_ tom Global _and_LacAl_AnAlyici I

(

j,c.e a

[

g,c e a

3.5 Ittermal Stricina 3.5.1

Background

At the time when the feedwater line cracking probl9ms in PWR's were first discovered, it was postulated that thermal oscillations (striping) may significantly contribute to the fatigue cracking problemi. These oscillations were thought to be due to either mixing of hot and cold fluid, or turbuleni.e 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 mcdel t

5409s/081591:10 3-6

tests Results of the flow model tests were used to establish boundary conditions for the stratification analysis and to provide striping oscillation data for evaluating high cycle fatigue.

Thermal striping was also examined during water model flow tests performed for the Liquid Metal f ast Breeder Reactor (LMFBR) primary pipe loop.

The stratified flow was observed to have a dynamic interface region which oscillated in a wave pattern.

These dynamic oscillations were shown to produce significant fatigue damage (primary track initiation).

The same interface oscillations were observed in experimental studies of thermal striping which were performed in Japan by Hitsubishi 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 oscillation of the hot and cold stratified boundary.

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

[

3,c.e a

The peak strets range and stress intensity was calculated from a 3-0 finite element analysis.

[

]"'C'*

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 K to account for the worst stress concentration 3

for all piping elements in the surge line.

The worst piping element was the butt weld.

5409s/081491:10 3-7

i I

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

[

j

.C e a

5409s/081491:10 3-8

(

4

~-

,a,c.e 5409s/081491:10 3-9

TABLE 3-1 TEMPERATURE DATA USED IN THE ANALYSIS e

Max Type of System Analysis Pressurizer RCL T

T Pipe Top Bot Operation AT(*F) Cases Temp ('F)

Temp (*F)

(*F)

('F) AT (*F) l S,c.e a

P 5409s/081491:10 3-10

TABLE 3-2 Sumary of Indian Point Units 2 & 3 Surge Lines Thermal Stratification Stress Results ASHLfode Eauation Stress Code Allonble Unit 2 untt_3 (ksi) 12 41.4*

52.6**

52,9 13 46.6 46.6 50.1

• at 50 bend underneath the pressurizer nozzle

• • at pipe side of pressurizer nozzle safe end

[.-

L l

l:

l l-.

5409s/081491:10 3-11

TABLE 3-3A INDIAN POINT UNIT 2 SURGE LINE MAXIMUM LOCAL AXIAL STRESS AT ANALYZED LOCATIONS Profile loc 3l Axial Streli (nsi)

Location

• Eurface Maximum Lepsile Mqximua Comgressive

~

~~

a,c.e See Figure 3-5 RCL nozzle transition RCL nozzle safe end and weld 3,c e a

[

5409s/081491:10 3-12

TABLE 3-3B INDIAN POINT UNIT 3 SURGE LINE MAXIMUM LOCAL AXIAL STRESS AT ANALYZED LOCATIONS

~

local Axial Stress (psi)

Profile Location

• Surfare M ni mm Tensile Maximum Comoressive

~

a,c.e See Figure 3-5 RCL nozzle safe end

• • • RCL nozzle safe end weld

[

ja,c.e 5409s/081491:10 3-13

1 3

TABLE 3 !

SUMMARY

OF FRESSURE-AND BENDING INDUCED STRESSES IN THE-SURGE LINE RCL N0ZZLE FOR UNIT LOAD CASES All Stress in osi Linearized' Stress Pe.tk Stress Intensity Ranat.

-Intenstty RangL Diametral Unit Loading

- Location Location Condition Inside Outside Inside Outside a,c.e j -.

4 4

1 4

~

i b

5409s/081491:10 3-14

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

TABLE 3-5 STRIPING FREQUENCY AT 2 HAXIMUM LOCATIONS FROM 15 TEST RUNS j

Totrl Frequency (HZ)

Duration j

_ # Cycles i

L;th, in Min (Qurationi Max. (den"6 L Ava (QurQjs)___$ggytt j

.- ~~

e a.C.e

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I f

i i

L r

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5409s/081491:10 3-15

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..,...gm,,,..

,.,,,,,,,..,,,,,,,..y.:

,,,,~r.,,,,.7,,,9.

y

a,r,e e

1 figure 3-1.

Schematic of Stress Analysis Procedure 5409s/081491:10 3-16

,i!l l1

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1 lI!

lll!ill l

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a c.e l=

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i Figure 3-3.

l.ocal Axial Stress-in Piping Due to Thermal Stratification i

' 5409s/081491:10 3-18

---.-_,-.,,--e

-.-..i 1

i

t O'

j i

a,c,e 6

i i

i i

f i

b h

r 4

4 L

Figure 3-4.-. Local Stress - Finite E4 1ent Models/ Loading l

1 l-i 5409s/081491:10

~3 i.

..___e.

a,c.e-I i

?

i I

9 1

f w

E

-.'1 t

figure 3-5. Piping Local Stress Model and Thermal Boundary Conditions r

l-5409s/081491:10 3-20

,.+-4.-,-

,<,,v.,4

..we-w w e*w ee r1r e ~== w.-w e - w e w e

• - w r y,ew m-e--%wv=-,-w ww-e.-wwwnwm.--,.e-m e w-w ee c -w w w -er, e +%&f-e=e+--e-

- r e 's - --

--w*-t--

-m*'m'~r**T--7

=

• P"-"

- r e v

BeC,e

~

l 4

I l

t 9

A I

Y h

T

-Figure 3-6.

Surge Line Temperature Distribution at (

)"'C Axial.

Locations I

l' 5409s/081491:10 3-21

- ~...

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

4,C,0 3

'C

0 Figure 3-7.

Surge Line Local Axial Stress Distribution at [

Axial Locations 5409s/081491:10 3-22

a,c.e w

Figure 3-8.

Surge Line Local Axial Stress on inside Surface at

[

]^'C'0 Axial Locations

- 5409s/081491:10 3-23

a,c.e i

Figure 3-9.

Surge Line L.ocal Axial Stress on Outside Surface at

(

J,c.e Axial Locations a

5409s/081491:10 3-24

r i

l i

i

-i I

L i

P P

b l

a,c.o i

J 3

9 s

h figure 3-10.

Surge Line RCL Nozzle 3-D HECAN Model:

14 Inch Schedule 140 1

l l

--5409s/081491:10 3-25

l a,c.e Figure 3-11.

Thermal Striping Fluctuatinn 5409s/081491:10 3-26

_ a c.e 0

Figure 3-12.

Thermal Striping Temperature Distribution 5409s/081491:10 3-27

SECTION 4.0 DISPLACEMENTS AT SUPPORT LOCATIONS The Indian Point Units 2 and 3 plant specific piping displacements at the pipe whip restraints and hanger locations along the surge line were calculated under the thermal stratification and normal thermal loads.

Table 4-1 shows the maximum surge line piping displacements (from all stratified conditions) at whip restraint and spring hanger locations for Units 2 and 3.

Table 4-2 shows the maximum surg' line piping displacements from the normal operation thermal condition at whip restraint and spring hanger locations for both units.

The support configuration used in analysis is based on the existing whip restraint gaps and no bottomed-out spring hanger.

The pipe whip restraint gaps used in the analysis are listed in Table 4-3.

In the analysis of Unit 2, PHR-120 contacts the pipe when maximum system AT of 320*F is considered.

The resulting load on PHR-120 is 8.3 kips, which is i

judged to be small compared to its design capacity.

In the analysis of Unit 3, PHR-124 contacts the pipe when maximum system AT of 320*F is considered.

The resulting load on PHR-124 is 20.7 kips which is judged to be smail compared to its design capacity.

5409s/081491:10 4-1

TABl.E 4 I Maximum Piping Displacement Under Stratified Conditions

• _UtLi t 3

_..Mait 2.

N_qdfL!h Ux (in)

MY_1.1.01 VLilD1 UX_(lu).

ULii.0.)

UL(101

(

)a,c.e

• All stratified cases are defined in Table 3-1

• • Variable spring hanger location (RCH-76 for Unit 2 and PW-H-63-1 for Unit 3) 5409s/081491:10 4-2

TABLE 4-2

- Haximum Piping Displacement Under Normal Thermal Conditions i

j Uniti_2 & 3 i

Node No.

Ox (in)

Uy_Lini VLiinl

[

1 t

8 3

C,0

- 1 i

• Variable' spring hanger location (RCH-76 for Unit 2 and PH-H-63-1 for Unit 3) 4 i

1 y

i a

h t

5409s/081491:10 4-3 te

-,,.,.,,+w,v-,-,

-v..-,,,-

1,

..~, - -,.

4- #s-+.

..-,,...7,.,-e.,,,m.,..--,-,,.c,--,-..,,,,,,.,,--,I+-.w-.,,,,,,'y,,,r.,y,,.3.~,w--r v-~~'-----**t-a*'e'

TABLE 4-3 PIPE WHIP RESTRAINT GAPS Gap..M athes1 __

Em EMe*

UniL2 UnlL3

[

4

)a,c,e

'Looking toward pressurizer from RCL branch nozzle

• • This restraint limits pipe movement in other (vertical) directions 5409s/081491:10 4-4

SECTION 5.0 ASME SECTION 111 FATIGUE USAGE FACTOR EVALUATION S.I htth0dal02y Surge line fatigue evaluations have typically been performed using the methods of ASME Section III, NB-3600 for all piping components [

J,c.e Because a

of the nature of the stratification loading, as well as the magnitud?s of the stresses produced, the more detailed and accurate methods of NB-3200 were employed using finite element analysis for all 1cading conditions.

Application of these methods, as well as specific interpretation of Code stress values to evaluate fatigue results, is described in this section.

Inputs to the fatigue evaluation included the transients develJped 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. [

d.c,0 j

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 recommendation of NRC Bulletin 88-11 (Appendix B of this report) to use the " latest ASME Section III requirements incorporating high cycle fatigue". Specific 5409s/081591:10 5-1

requirements for class 1 fatigue evaluation of piping components are given in 48-3653.

These requirements must be met for Level A and Level B type loadings according to NB-3653 and NB 3654.

4 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 requiremerits and equations correlate to those in NB-3200, the results of the fatigue evaluation are reported in terms of the NB-3650 piping stress equations.

These equations and requirements are summarized in Tables 5-1 and 5-2.

The methods used to evaluate Code limits for the surge line components are described in the following sections.

5.1.2 Fatigue Stress Equations i

itI.011_ClL$11f.isitio.D l

The stresses in a component are classified in the ASME Code based on the nature of the stress,-the loading that causes the stress, and the geometric characteristics that influence tha 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, j

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 i

pressure, nioment and thermal transient stresses.

Equation 11 includes the K1, K2 and K3 indices in the pressure, moment and thermal transient stress terms in order to reptesent peak stresses caused by local concentration, such as notches and weld effects.

The NB-3653 equations use simplified formulas to 5409s/081591:10 5-2 a.----.-.-

-.. -. =. - - -

l r

i determine nominal stress based on straight pipe dimensions.

[

i aj.c.o for the RCL nozzles, three dimensional (3-D) finite element analysis was used as described in Section 3.0.

[

l J.C,e

[

a e

Classification of local stress due to thermal stratification was addressed f

-with-respect to the thermal transient stress terms in the NS-3653 equations.

Equation 10 includes a Ta-Tb term, classified as "0" stress in NB-3200, which represents stress due to differential thermal expansion at gross structural discontinuities.

[

i l.c.e The imp;ct of this on the a

selection of components for evaluation is discussed in Section 5.1.3.-

l.

\\*

5409s/081491:10 5-3 l

1

,i SILT.5LCombinations i

.l 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.

[

4 i

J.c.e This was done for specific components as foilows:

a

[

l 1

l

?

ag.c.e 5409s/081491:10 5-4

, _,... a

.___...<....._____.2_.._,

[.

l 4

)a.c.e From the stress profiles created the stresses for Equations 10 and 11 could be determined for any point it ~.ae section.

Experience with the geometries and loading showed that certain points in the finite element models consistently produced the worst case fatigue stresses and resulting usage factors, in each stratified axial location.

[

3,c.e a

I

(

5409s/081491:10 5-5

Equation 12 Streti Code Equation 12 stress represents the maximum range of stress due to thermal expansion moments as described in Section 3.2.

This used an enveloping approach, identifying the highest stressed location in the model.

By evaluating the mJrst locations in this manner, the remaining locations were inhet intly addressed.

LQualiOLll_S' til 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 transiants.

Design mechanical loads are defined as deadweight plus seismic OBE loads.

The "Ta-Tb" term of Equation 13 is only applicable at structural discontinuities.

[

3,c.e a

lhermal Stress Ratchel The requirements of NB-3222.5 are a function of the thermal transient stress and pressure stress in a component, and are independent of the global moment loading. As such, these requirements were evaluated for controlling components using applicable stresses due to pressure and stratification transients.

e 5409s/081491:10 5-6

l

'AllowahleJtiesici Allowable stress, Sm, was determined based on note 3 of figure 40-3222-1.

For secondary stress due to a temperature transient or thermal expansion Mads

(" restraint of free end deflection"), the value of Sm was taken as the average of the Sm values at the highest and lowest temperatures of the metal during the transient.

The metal temperatures were determined from the transient definition. When part of the secondary stress was due to mechanical load, the value of Sm was taken at the highest metal temperature during the transient.

5.1.3 Selection of Components for Evaluation Based on the results-of the global analyses and the considerations for controlling stresses in Section 5.1.2, (

L 3,c.e The method to evaluate usage a

factors using stresses determined according to Section 3.0 is described below.

5.2 f3119utusagLfAtton Cumulative usage factors were calculat'ed for the controlling components using the methods described in NB-3222.4(e), based on NB-3653.5. Application of these methods is summarized below.

Iransient_Loadcas1LanLCombinations from the transients described in Section 2.0, specific loadcases were developed for the usage evaluation.

[

j

,C,e a

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-wtth the stress data as described above.

5409s/081491:10-5-7

l Usage factors were calculated at controlling locations in the component as follows:

1)

Equation 10, Ke, Equation 11 and resulting Equation 14 (alternating stress - Salt) ara calculated as described above for every possible combination of the loadsets.

2) for each value of Salt, the design fatigue curve was used to determine the maximum number of cycles which would be allowed if this type of cycle were the only one acting.

These values N),

N * * *N, were determined from Code Figures I-9.2.1 and I-9.2.2, 2

n curve C, for austenitic stainless steels.

3)

Us**19 the actual cycles of each transient loadset, n),

n **"n' 2

calculate the usage factors U), U...U fr m Ug = n /N.

This 2

n g

g is done for all possible combinations.

Cycles are used up for each combination in the order of decreasing Salt.

When N is greater g

ll is taken as zero.

than 10 cycles, the value of Ug

[

j,c.e a

4)

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

+

U2 +...

• U.

To this was added the usage factor due to n

thermal striping, as described below, to obtain total Ucum.

The Code allowable value is 1.0.

s

)

Ei 5409s/081491:10 5-8

5.3 f111gutD.ue to ThRrRLStriplag The usage factors calculated using the methods of Section 5.2 do not include the effects of thermal striping.

(

- a.c.e j

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 J

c.e The methods used to determine alternating stress intensity a

are defined in the ASME code.

Several locations were evaluated in order to determine the location where-stress intensity was a maximum.

l 1

l 1.

I S409s/081491:10 5-9 l -.

l

. - ~.

Thermal striping transients are shown as a AT level and number of cycles.

The striping AT for each cycle of every transient is assumed to attenuate and follow the slope of the curve shown on figure 5-2.

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

Each step lasts (

Ja.c.e secondo.

(

)"' C d is used in all of the usage factor calculations, the total fluctuations per step is constant and becomes 1

aj,C,0

(

i Each striping transient is a group of steps with [

]"'C fluctuations per step.

For each transient, the steps begin at the maximum AT and d e reases by

(

J.c.e steps down to the endurance limit of AT equal to (

)"'C

The a

cycles for all transients which have a temperature step at the same level were added together.

This became the total cycles at a step.

The total cycles a

were multiplied by [

l.c.e to obtain total fluctuations.

This-results in total fluctuations at each step.

This calculation is performed for each step plateau from (

3"'C to obtain total fluctuations. Allowable fluctuations and ultimately a usage factor at each

.l plateau is calculhted 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, i

The usage factor due to striping, alone, was calculated to be a maximum of J.c.e -This is reflected in the results to be discussed below.

a

(

5.4 Fatiaue UsAae Res.ulti NRC Bulletin 88-11 (5) requests that fatigue analysis should be performed in l

accordance with the latest ASME III requirements incorporating high cycle 1

fatigue and thermal stratification transients.

ASME fatigue usage factors have been calculated considering the phenomenon of thermal stratification and l

thermal striping at various locations in the surge line.

Total stresses l

5409s/081491:10 5-10

included the (

)a.c.e The total stresses for all i

transients in the bounding set were used to form combinations to calculate alterr.ating stresses and resulting fatigue damage in the manner defined by the Code.

Of this total stress, the stresses in the 14 inch pipe due to

(

j.c.e a

The maximum usage factor oi. Indian Point surge lines occurred at (

J c.e In this thermal fatigue evaluation, weided l

a attachments at PHR-120 and'PHR-121 were also included, and it was found that the usage factors were smaller than the maximum value listed above due to lower total loadings and stresses at the lug locations.

It is also concluded that the Indian Point pressurizer surge nozzles meet the Code stress allowables under the thermal stratificution loading from the surge line, with no spring hanger bottomed out configuration and the transients detailed in reference (13).

They also meet the fatigue usago requirements of ASME Section III, with a maximum cumulative usage factor equal to 0.26 (18).

?

J 5409s/081491:10 5-11

I TABLE 5-1 i

CODE / CRITERIA

\\

o ASME B&PV Code, Sec. III, 1986 Edition NB3600 NB3200 L

o Level A/B Service-Limits Primary Plus Secondary Stress Intensity 1 35m (Eq. 10)

Simolified Elastic-Plastic Analysis Expansion Stress, S 1 35m (Eq. 12) - Global Analysis e

Primary Plus Secondary Excluding Thermal Bending < 3Sm

-(Eq. 13)

Elastic-Plastic Penalty Factor 1.0 1 K, I 3.333 Peak Stress (Eq.11)/Cun.ulative Usage Factor (Ucum}

Salt " K b /2 (Eq. 14) ep Design Fatigue Curve U

4 1.0 cum

~5409s/081491:10 5-12

TABLE 5-2

SUMMARY

OF ASME FATIGUE REQUIREMENTS Parameter Description Allowable (if applicable)

Equation

?timary plus secondary stress intensity;

< 3Sm if exceedci, simplified elastic-plastic analysis may be performed K,

Elastic-plastic penalty factor; required for simplified elastic-plastic analysis when Eq. 10 is exceeded; applied to alternating stress intensity Equation 12 Expansion stress; required for simplified

< 3Sm elastic-plastic analysis when Eq. 10 is exceeded Equation 13 Primary plus secondary stress intensity

< 3Sm excluding thermal bending stress; required for simplified elastic-plastic analysis when Eq. 10 is exceeded Thermal Limit on radial thermal gradient stress to Stress prevent cyclic distortion; required for use Ratchet of Eq. 13 Equation 11 Peak stress intensity - Input to Ea. 14 Equation 14 Alternating stress intensity - Input to Ucum Ucum Cumulative usage factor (fatigue damage)

< 1.0 5409s/081491:10 5-13

l

_ a,c,e 4

t Figure 5-1.

Striping Finite Element Model 5409s/081491:10 5-14

..-. -.~.

... ~...

=

j

'i 4

a,c,e

~

1 W

2 Figure-5-2.

. Attenuation of Thermal Striping Potential by Molecular Conduction (Interface Have Height of One Inch) l I

l-l

.5409s/081491:10 5-15 u*

A r

e

,,4 y

m

SECTION 6.0 ANALYSIS

SUMMARY

The subject of pressurizer scrge 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 2 timely response to the bulletin.

The Owners Group programs were c. cipleted in June of 1990, and have been tollowed by a series of plant specific evaluations.

This report has documented the results of the plant specific evaluation for Indian Point Units 2 and 3.

Following the general approach used in developing the surge line stratification transients for the WOG, a set of transients and stratification profile were developed specifically for Indian Point Units 2 and 3.

A study was-made of the historical operating experience at the Indian Point Units 2 and 3, and this information, as well as-plant operating procedures and

{

monitoring results (from similar piants), was used in development of'the transients and profiles.

The analysis results are shown in Section 3.0 for -ASME code stress,- Section 4.0 for displacements at' supports and whip restraint locations and Section 5.0 for ASME code fatigue cumulative usage factors.

The results were conservatively ca'cula.+ed using the maximun design temperature differential and worst case assumptions for inducing thermal stratificetion to the system.

L 5409s/081491:10 6-1

l l

SECTION

7.0 REFERENCES

1.

Coslow, B.

J., et al., " Westinghouse Owners Group Bounding Evaluation for Pressurizer Surge Line Thermal Stratification", Westinghouse Electric Corp. HCAP-12277, (Proprietary Class 2) and NCAP-12278 (non-proprietary),

June 1989, 2.

Coslow, B.

J., et al., Westinghouse Owners Group Pressurizer Surge Line Thermal Stratification Program MUHP-1090 Summary Report," Westinghouse Electric Corp. HCAP '2508 (Proprietary Class 2) and WCAP-12509 (non-proprietary), March 1993.

3.

Coslow B.

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

L 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.

Consolidated Edison Co. of New York letter, Joseph Madia to T. H. Liu,

~

February 27, 1991.

7.

New York Power Authority letters, (i) NED CS-91-MM-M19, " Indian Point No.

3 Pressurizer Surge Line Stratification Analysis." February 11, 1991, and (ii) NED-CS-91-HP-M33, March 5, 1991.

8.

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

9.

Woodward, H. S., " Fatigue of LMF8R Piping due to Flow Stratification,"

ASME Paper 83-PVP-59, 1983.

5409s/081491:10 7-1

10. Fujimoto, T., et al., " Experimental Study of Striping at the Interface of Thermal Stratification" in Thermal Hydraulics in Nytlear Technoloav, K. H.

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

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

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

Technical Basis," Westinghouse Electric Corporation HCAP-12315 (Proprietary Class 2).

13. Series 84 Pressurizer Stress Report, Section 3.1, Surge Nozzle Analysi;,

December 1974.

14. Hayes, C.

V., " Pressurizer Temperature Differentials Relating to Surge Line Stratification," Consolidated Edison Indian Point 2 Station, Technical Services Department Technical Engineering Report No. TER 91-03,

-February 22, 1991.

L

15. New York Power Authority Memorandum 91-036, to R. Drake from F. Gumble, "Hestinghouse Surge Line Data Evaluation," 3/14/91.

16.-Con-Edison Memorandum IPQA#9-0354, " Pressurizer Surge Line Examination "

May 23, 1989.

17. New York Power Authority Memorandum IP-89-52, " Visual Inspection of IP3's Pressurizer's Surge Line," February 15, 1989,

18. Westinghouse Letter AEA-91-100, "IPP/ INT Pressurizer Surge Nozzle Analysis

~ for Pipe Loads due to Thermal Stratification," May 2,1991 (Hestinghouse central file IPP/ INT-160-2A3).

V d

5409s/081491:10 7-2

l 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 EeicrLation HECAN is a Westinghouse-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 for plane strain, plane stress, and axisymmetric analyses.

Brick and wedge structural elements are used for three-dimensional analyses.

Companion heat conduction elements are used for steady state heat conduction analyses and transient heat conduction analyses.

A.I.2 Feature Uttd The temperaturec obtained from a static heat conduction analysis, or at a specific time in a transient heat conduction analysis, can be automatically input to a static structural analysis where the heat conduction elements are replaced by corresponding structural elements.

Pressure and external loads can also be include in the WECAN structural analysis.

Such coupled thermal-stress analyses are a standard application used extensively on an industry wide basis.

1 5409s/081491:10 A-1

-A.l.3 P_togre verification Both the WECAN program and input for the WECAN verification problems, currently numbering over four hundred, are maintained under configuration control.

Verification problems include coupled thermal-stress analyses for the quadrilateral, triangular, brick, and wedge isoparametric elements.

These

. problems-are an-integral cart of the HECAN quality assurance procedures.

When a change is made to HECAN, as part of the reverification process, the configured inputs for the coupled thermal-stress verification problems are used to revertfy HECAN for coupled thermal-stress analyses.

A.2 STRFAT2 A.2.1 D1scriotion i

STRFAT2 is a program which computes the alternating peak stress on the inside surface of a flat plate and the usage factor due to striping on the surface.

The program is applicable to_ be used for striping on the inside surface of a pipe if the program assumptions are considered to apply for the particular pipe being evaluated.

6 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.

l l

5409s/081491:10 A-2

l A.2.2 EntEcJhed 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 mcterials 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 P_tqgnm Verifica11on STRFAT2 is verified to Westinghouse procedures by independent review of the stress equations and calculations.

A.3 6MSlS A.3.1 Reitrh11on ANSYS is a public domain, general purpose finite element code, A.3.2 Feature Usad The ANSYS elements used for the analysis of stratification effects in the surge line are STIF 20 (straight pipe), STIF 60 (elbow and beids) and STIF14 (spring-damper ior supports).

A.3.3 ProaramJeri fi cation As described in sectioi 3.2, the application of ANSYS for stratification has been independently verified by comparison to WESTDYN (Westinghouse piping analysis code) and HECAN (finite element code).

The results from ANSYS are also verified against closed form solutions for simple beam configurations.

4 5409s/081491:iG A-3

A.4 FATRK/ CMS A.4.1 Descriotion FATRK/ CMS is a Westinghouse developed computer code for fatigue tracking (FATRK) as used in the Cycle Monitoring System (CMS) for structural components of nuclear power plants.

The transfer function method is used for transient thermal stress calculations.

The bending stresses (due to global stratification effects, ordinary thermal exparsion ani seismic) and the pressure stresses are also included.

The fatigue usage factors are evaluated in accordance with the guidelines given in the ASME Boiler and Pressure Vessel Code,Section III, Subsections NB-3200 and NB-3600.

The code can be used both as a regular analysis program or an on-line

-monitoring device.

A.4.2 Feature Ustd FATRK/ CMS is used as an analysis program for the present application.

The input data which include the weight functions for thermal stresses, the unit bending stress, the unit pressure stress, the bending moment vs.

stratification. temperatures, etc. are prepared for all locations and geometric conditions. These data, as stored in the independent files, can be appropriately retrieved for required analyses.

The transient data files contain-the time history of temperature, pressure, number of occurrence,-and additional condition necessary for data flowing.

The program prints out the total usage factors, and the transients pairing information which determine-the stress range magnitudes and number of cycles.

The detailed stress data may also be printed.

A.4.3 Proaram Verification

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

l l

5409s/081491:10 A-4 l

l

.. ~.

APPENDIX B-USNRC BULLETIN 88-11 In December of 1988 the NRC issued this bulletin, and it has led to an extensive investigation of surge line integrity, culminating in this and other plant specific reports.

The bulletin is reproduced in its entirety in the pages which follow.

4 r.

1 5409s/081491:10 B-1 i

CPB No. 3150-0011 NRCB 80-11 UNITED STATES NUCLEAR REGULATOPY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.

20555 December 20, 1988 NPC 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 addressees establish and implement a program to confirm pressurizer surge line integrity in view of the occurrence of thermal stratification and (2) require addressees to inform the staff of the actions taken to resolve this issue.

Description of Circumstances.

The licensee for the Trojan plant has observed unexpected 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 contacted two re-straints.

Although the licensee had repeatedly adjusted shims and gap sizes based on analysis of various postulated conditions, the problem had not been resolved.

The most recent investigation by the licensee confirmed that the movement of piping was caused by thermal stratification in the line. _ This phenomenon was not considered in the original piping design.

On October 7, 1988, the staff issued-Information Notice P8-80, " Unexpected Piping Movement Attributed to Thermal Stratification," regarding the Trojan experience and indicated that further generic communication may be forthcoming.

The licensee larger-than-expected surge line displacement during power asc The concerns raised by the above observations are similar to those described-in

. NRC Bulletins 79-13 (Revision 2, dated October Feedwater System Piping" and 88-08 (dated June 16, 1979), " Cracking in Piping Connected to Reactor Coolant Systems."

22, 1988), " Thermal Stresses in 4

4812150118 B-2

NRCB E8-il Cecember 20, 'g88 Dage 2 of 6 Discussion:

Unexpected piping movements are highly undesirable because of potential ri;n i

piping stress that riay exceed design limits for fatigue and stresses.

The problem can be more acute when the piping expansion is restricted, such through contact with pipe whip restraints.

Plastic determation can resuit, as which can lead to high local stresses, low cycle fatigue and functional tm-pairment of the line.

Analysis performed by the Trojhn licensee indicated trat thermal stratification occurs in the pressurizer surge line during reatuo, cooldown, and steady-state operations of the plant.

During a typical pit.nt heatup, water in the pressuri:er is heated to about 440*F; a steam bubble is then formed in the pressurizer.

Although the exact phenomenon is not thoroughly understood, as the hot water flows (at a very 'ow flowrate) from the pressurizer through the surge line to the hot-leg piping, the hot water rides on a layer of cooler water, causing the upper part of the The differential temperature could be as high as 300'F, based on conditions during typical plant ope ~ations.

Under this condition, differential thermal expansion of the pipe metal can cause the pipe to deflect signifi-cantly.

For the specific configuration of the pressurizer surge lict in the Trojan plant, the line deflected downward and when the surge '4ne contacted two pipe e

whip restraints, it underwent plastic deformation, m.iting in permanent 7

defomation of the pipe.

The Trojan event demonstrates that thermal stratification in the pressurizer surge line causes unexpected piping movament and potential plastic deformation.

4 The licensing basis according to 10 CFR 50.55a for all PWRs requires that the licensee meet the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Sections III and XI and to recontile the pipe stresses and fatigue evaluation when any significant differences are observed between measured data and the analytical results for the hypothesized conditions.

Staff evaluation indicates that the 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:

Addressees are requested to take the following actions:

1.

For all licensees of operating PWRs:

Licensees are requested to conduct a visual inspection (ASME, Section 4.

XI, VT-3) of the pressurizer surge line at the first available coic shutdown after receipt of this bulletin which exceeds seven days.

B-3 h..

l NRCB 88 11 December 20, 1988 i

Page 3 of 6 This inspection should determine any gross discernable distress or structural damage in the entire pressurizer surge line, including piping, pipe supports, pipe whip restraints, and anchor bolts.

,1 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 pressurizer

.j surge line meets the applicable design codes' and other FSAR and

-regulatory comitments for the licensed life of the plant, consider-j 4

ing the phenomenon of thennal stratification and thermal striping in

{

the fatigue and stress evaluations.

This may be accomplished by 1

performing a plant specific or generic bounding analysis.

If the latter option is selected, licensees should demonstrate applicability of the referenced generic bounding analysis. Licensees of plants in j

operation less than-ten years (i.e., low power 'icense after January 1,1979), should complete.the forego..ig 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 valic, If the opportunity to perfonn the visual inspection in 1.a does not occur within the periods specified in this requested item, incorocra-tion of the results of the visual inspection ir.

he analysis-strould

.be performed in a supplemental analysis as te.

Where the -

• sis shows that the surge line dot not meet the requir r*-

.nd licensing comitments stated above for the duration of the av, the licensee should submit a justification for contint.

.veration or bring the plant to cold shutdown, as appropri-ate, ano implement Items 1.c and 1.d below to develop a detailed analysis of the surge line, If the analysis in 1.b does not show compliance with the requirements c.

and licensing convoitments stated therein for the duration of the operating license, the licensee is requested to obtain plant specific data on thermal stratification, thermal striping, and line deflec-I tions.

The licensee may choose, for example, either to install instruments on the surge line to detect temperature' distribution and thermal movements or to obtain data through collective efforts, such as. from other plants with a similar surge line design.

If the latter option is selected, the licensee should-demonstrate similarity in geometry and operation.

d.

Based on the applicable plant specific or referenced data, licensees-are reauested to update their stress and fatigue analyses to ensure compliance with applicable Code requirements, incorporating any observations from 1.a above The analysis should be completed.no later than two years after receipt of this bulletin.

If a licensee

.i

• Fatigue analysis should be performed in accordance with the latest ASME Section 111 requirements incorporating high cycle fatigue.

B-4

NRCB 88 11 December 20 -1998 Page 4 of 6 is unet,le to show compliance with the applicable design codes anct other FSAR and regulatory comitments, the licensee is requested te submit a justification for continued operation and a description of the proposed corrective actions for effecting long term resolution, 2.

For all applicants for PWR Operating Licenses:

a.

Before issuance of the low power license, applicants are requestea te demonstrate that the pressurizer surge line meets the applicable design codes and other FSAR and regulatory comitments for the licensed life of the pMnt.

This may be accompli 9ed by performing a plant-specific or generic bounding analysis.

The analysis should include consideration of thermal stratification and therinal striping to ensure that fatigue and stresses are in compliance with applicable code limits.

The analysis and hot functional testing should verit that piping themal deflections result in no adverse consequences,y such as contacting the pipe whip restraints.

If analysis or test results show Code noncompliance, conduct of 611 actions specified below is requested, b.

' Applicants are requested to evaluate operational alternatives or piping modifications needed to reduce fatigue and stresses to acceptaole levels, Applicants are requested to either monitor the surge line for the c.

ef fects of thermal stratification, beginning with hot functional 6

testing, or obtain data through collective efforts to assess ths

-extent of thennal 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 corre<:tive actions, and maintain these records in accor-dance with 10 CFR Part 50, Appendix B and plant procedures.

Reporting R,equirements:

L 1.

Addressees 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 demonstrated for the full duration of an operating license, the staff may impose a license condition such that normal operation is restricted to the duration that compliance is actually demonstrated.

B-5 l

NRCS 88-11; December 20, 1988 Page 5 of 6 2.

Addressees who car.not meet the schedule cescribed in items 1 or 2 of Actions Requested are required to submit to the NRC within 60 days of receipt of this bulletin an alternative schedule with justification for the requested schedule.

i

=

3.

Addressees shall submit a letter within 30 days after the completion of i

these actions.which notifies the NRC that the actions recuested in Items i

lb, ld or 2-of Actions Reouested have been performed and that the results are available'for inspection.

The letter shall include the justification for continued operation, if appropriate, a description of the analytical i

approaches used, and a summar*y of the results.

~ Although net requested by this bulletin, addressees are encouraged to work collectively to address the technical concerns associated with this issue, as well as to share pressurizer surge line data and operational experience.

In addition, addressees are encouraged to review piping in other systems which may experience thermal stratification and thermal striping, especially in light of the previously mentioned Bulletins 79-13 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 required above shall be addressed to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C.

20555, under oath or affirmation under the provisions of Section 182a, Atomic Energy Act of 195a,

s amended.--In addition, a copy shall be submitted to the appropriate Regional Administrator, b

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 partain only to these identified response-related matters and do not include the time for actual implementation of paysical changes, such as test equipment installation or component modification.

The estimattd average raciation exposure. is approximatlly 3.5 person rems per licensee response.

Comments on the accuracy of this estimate and suggestions to reduce the burden

-may.he directed to the Of fice of Management and Budget, Room 3208, New Execu-tive Office Building, _ Washington 0.C.

tory Commission, Records and Reports Management Branch, Office of20503, and Administration and Resource Management, Washington 0.C.

20555.

9 B-6

-~-

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

i l

NRCB 88-11 December 20, 1988 Pa9e 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 M(

afles E. Rossi. Director 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 Attachnents :

1.

Figure 1 2.

List of Recently Issued NRC Bulletins d

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B-7

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Thot = 425*F I

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ENCLOSURE C TO IPH 91-il38 Westinghouso: (1) Letter and affidavit regarding

" Application for Withholding Proprietary Information flom Public Disclosuro.*; (2) Propriotary inforrnatior. Notico; (3) Copywrito Notico.

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NEW YORK POWER AUTHORITY

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INDIAN POINT 3 NUCLEAR POWER PLANT DOCKET NO. 50-286 DPRPA a

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