ML20059G568: Difference between revisions
StriderTol (talk | contribs) (StriderTol Bot insert) |
StriderTol (talk | contribs) (StriderTol Bot change) |
||
| Line 2: | Line 2: | ||
| number = ML20059G568 | | number = ML20059G568 | ||
| issue date = 11/04/1993 | | issue date = 11/04/1993 | ||
| title = Evaluation of Pressurizer Surge Line for Stratified Flow Conditions | | title = Evaluation of Pressurizer Surge Line for Stratified Flow Conditions | ||
| author name = | | author name = | ||
| author affiliation = SOUTHERN CALIFORNIA EDISON CO. | | author affiliation = SOUTHERN CALIFORNIA EDISON CO. | ||
Latest revision as of 21:51, 6 January 2021
| ML20059G568 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/04/1993 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20059G361 | List: |
| References | |
| IEB-88-011, IEB-88-11, NUDOCS 9311080233 | |
| Download: ML20059G568 (111) | |
Text
{{#Wiki_filter:_ .. - - . - _ _ - _ - - - _ - _ _ _ _ _ _ _ 1 i ! i 1 I l ENCLOSURE l i l Evaluation of the Pressurizer Surge Line i for Stratified Flow Conditions ! (Response to NRC Bulletin 88-11) i i
- San Onofre Nuclear Generating Station Units 2 and 3 1
I l l 4 i 4 i 4 1 l i i 1 ) 9311080233 931104 PDR ADOCK 05000361 ? G PDR }{#3
l l i i EXECUTIVE
SUMMARY
l l l Thermal stratification in the pressurizer surge line in pressurized water i reactor power plants has been observed in the past. The stratification phenomenon can be explained by the difference in density between the i pressurizer water and the reactor coolant system hot leg water. Although the i potential for stratification is small during normal plant operation, it l becomes significant during insurge and outsurge transient events, which occur i during plant heatup and cooldown and during other modes of operation. Thermal stratification can be characterized by top-to-bottom thermal gradients in the pipe wall resulting in thermal stresses and pipe motion. The stresses and i deformations produced by thermal stratification were not considered in the ! original design of the plant, and they could impact the fatigue life of the piping, or cause damage to the attached pipe supports and other components. l l Recognizing the significance of thermal stratification in the surge line, and the other associated issues, the Nuclear Regulatory Commission (NRC) issued Bulletin 88-11 requesting that all utilities establish and implement a program to demonstrate the structural integrity of the surge line in view of thermal ' stratification. The Bulletin required utilities to perform inspections, conduct bounding analysis to provide justification for continued operation ; (JCO), collect plant data on thermal stratification and update the original design of the surge line, on a plant specific basis, to include the effect of l thermal stratification. Collective effort in responding to some of these requests was allowed by the Bulletin, provided that similarity in design and operation is demonstrated. I In accordance with the requirements of Bulletin 88-11, Southern California l Edison (SCE) has performed the requested inspection of the surge line, and the results were satisfactory. Also, SCE SONGS Units 2 and 3 surge lines were included in the collective effort by the Combustion Engineering Owners Group (CEOG). Two reports resulted from this collective effort documenting the data collection and reduction, structural and fatigue analysis of all surge lines (of the participating utilities) and shakedown analysis. These reports were used to meet the JC0 and the bounding analysis requirements for SONGS Units 2 and 3. Furthermore, in accordance with the requirements of the NRC, SCE has l performed a plant specific evaluation of the pressurizer surge lines including ' pipe supports and integral attachments, pressurizer surge nozzle, hot leg surge nozzles and the liquid sample line attached to the pressurizer surge , l line. _j-
l l l 1 This report documents in detail the response of SCE to all the requirements of NRC Bulletin 88-11. These requirements are listed individually in Sections IV and V with SCE's response to each of these requests and the actions taken. The report also includes SCE's response to NRC staff recommendations provided in a Safety Evaluation Report. A summary of the scope of CEOG's reports on surge line thermal stratification is provided in Section V of this report. Section VII (Attachment A) of the report is dedicated to the plant specific evaluation performed by SCE in response to Item 1.d of Bulletin 88-11. A detailed description of the analysis results, system description and design input and analysis methodology is provided in Section VII. Based on SCE's response to requirements of NRC Bulletin 88-11, it is concluded that all of these requirements are met for the licensed life of the plant. i l l l
,ii -
I TABLE OF CONTENTS l ; i-l Page No. l EXECUTIVE SUPG4ARY i ; TABLE OF CONTENTS iii 4 I. INTRODUCTION I-1 I.1 Background I-2 l II. Supt (ARY OF RESULTS II-1 ; III. NRC SAFETY EVALUATION OF CEOG REPORT CEN-387-P, REVISION 1, III-1 ;
" PRESSURIZER SURGE LINE THERMAL STRATIFICATION EVALUATION" [
t j III.1 SEC's Response to SER Recomendations III-l i
~
1 l IV. NRC BULLETIN 88-11 IV-1 , V. SCE'S RESPONSE TO BULLETIN 88-1I REQUIREMENTS V-1 l V.1 SCE's Response to Reporting Requirements V-8 , I l VI. REFERENCES AND FIGURES VI-1 1 l VI.1 List of References VI-1 l VI.2 List of Figures VI-2 VII. ATTACHMENTS VII-1 VII.1 httachment A - Analysis Sumary Report VII-1
- iii -
l
l e I. INTRODUCTION i l The purpose of this report is to. document the work performed by Southern j California Edison (SCE) in response to the Nuclear Regulatory Commission (NRC) Bulletin 88-11 (Reference 1) which addresses pressurizer surge. line. flow. _. stratification. The report documents the analyses performed and actions:taken : by SCE to ensure that Units 2 and 3 at San Onofre Nuclear Generating Station ] (SONGS) meet all the requirements of Bulletin 88-11. I Bulletin 88-11 was issued by the NRC on December 20, 1988 to address the issue- ! of thermal stratification in the pressurizer surge line. The. Bulletin ~ ! requested that utilities establish and implement a program to confirm the structural integrity of the pressurizer surge. line in pressurized. water-- reactors (PWRs) under the effects of thermal stratification. It also : requested that utilities inform the staff of the actions taken to resolve all ; the concerns associated with this issue. n An evaluation of the pressurizer surge line under stratified flow conditions ; a was performed by Combustion Engineering (CE) for the Combustion Engineering Owners Group (CEOG) in response to Bulletin 88-11 concerns. This evaluation j was documented in Report CEN-387-P, Revisions 0 and 1, which applies to all ! CE0G surge lines (References 5 and 6). The report responds generically to .the ; NRC concerns for SONGS Units 2 and 3, and it was demonstrated by CE in this ; report that the structural integrity of all pressurizer surge lines in CE o designed plants is maintained for the 40-year service life of the plant as , requested by NRC Bulletin 88-11. Revision 1 of the CEOG report was reviewed - , and accepted by the NRC in a Safety Evaluation Report (Reference 15). > In addition to generic work performed for the CE0G (documented in References 5' : and 6), SCE took actions to provide plant specific assurance of the structural ! integrity of SONGS Units 2 and 3 pressurizer surge-lines including the impact ! of the additional fatigue and stresses caused by surge line. thermal , stratification. This report summarized the actions taken and the plant- ' specific analytical evaluations performed by SCE. These evaluations include, in addition to the pressurizer surge line, all components impacted by surge. line thermal stratification. Attachment A of this report (included in Section VII) provides the details of..the analysis results, the analysis methodology, system description and design input for the surge line and all other impacted components. The scope of the analysis includes structural and fatigue evaluations, effect of thermal striping and a_ recent ASME Code update as required by Bulletin 88-11. 1-1
'I
I.1 BACKGROUND As shown in Figure I-1, the Reactor Coolant System (RCS) at SONGS Units 2 and 3 consists of two loops connecting the reactor vessel to the steam generators. Tha system also includes a pressurizer, a surge line connecting the pressurizer and the RCS hot leg, pressurizer safety valves and a relief tank (quench tank). The pressurizer contains water and steam at saturated conditions, with the water-steam interface varying according to plant conditions. The steam bubble in the pressurizer acts as a cushion against sudden changes in RCS pressure. A system of electric heaters and water spray nozzles provide pressure control in the pressurizer. ; i Thermal stratification in the pressurizer surge line results from the difference in density between the hotter pressurizer water and the cooler RCS hot leg water. The hotter and less dense pressurizer water tends to float on top of the cooler and more dense RCS hot leg water as shown schematically in Figure I-2a. The potential for stratification in the surge line at power is relatively small since the difference between the pressurizer temperature and
, is small (less than 50 F at SONGS). As the RCS hot leg temperature, explained in detail in Section AT,d.,1 of Attachment A of this report, the potential for stratification decreases as AT , sydecreases since Richardson number is proportional to AT,,s.
However, the potential for stratification increases considerably during plant heatup and cooldown where the temperature difference between the pressurizer and the hot leg (AT,.,) typically exceeds 300*F. Large values of AT,, result ; 3 in considerable difference in the density between the pressurizer wa,ter and the RCS water. Thus, the insurge (flow from the hot le and outsurge (flow from the pressurizer to the hot leg)g to the events pressurizer) occurring during plant heatup and cooldown could produce stratified flow conditions in the surge line. Quantitative assessment of the potential for thermal stratification is discussed in Section 5 of Attachment A in this report (Attachment A, Analysis Summary Report, is included in Section VII of this report). Thermal stratification in the pressurizer surge line results mainly in the following effects that were not considered in the original design of the plant:
. Global thermal bending stress affecting the surge line, pipe supports and surge nozzles, I -'2 i
l l . Potential reduction in fatigue life of the plant due to global bending stress and local (peak) stress resulting from stratification and ; thermal striping in the pipe (see Figure 1-2b). l A more detailed description of the above effects can be found in Attachment A l along with an evaluation of their impact. Concerns identified with thermal stratification in the pressurizer surge line were initiated by Safety Event Report (SER) number 25-87 issued by the Institute of Nuclear Power Operations (INPO) in September 1987. The report included examples of observed stratification effects in some PWR plants. These effects were in the form of excessive pipe top-to-bottom temperature differential and pipe motion. These concerns were addressed by NRC Bulletin 88-11 which requires utilities to resolve the issue of surge line thermal ; stratification on a plant-specific basis. In the following sections of this ! report, a detailed description of the requirements of Bulletin 88-11 is given i along with the actions taken by SCE in response. I-3 l 1 I t
- - - - + , ,, ,-
- - - . , - g - , ,
l ; , ' . _ A-u_ J _ Q Q3SA _ 3 d _ n - a b 2 S G . N _ O . S . t . a _ m . e t
- s y
S t _ n . - a R L l o o OE C TS _ dS r - AE o EV R t c a e R e . h _ t _ _ f . _ o-c i i r e z r
) t a
m e u h s c s S h [ P e r N h 1 1 e r u g _ J i F Q 3 A S r= I, '
q
)
l l l i
- T sat i '
e j LEVEL I PRESSURIZER l l
. /'"
l hot les Figure I-2a Schematic of Stratified Flow in the Surge Line High Inkrface Iaal Hot Water t1 s-._____ A o Cold Water Law Inbrfwm Imel Figure I-2b Schematic of Themal Striping in the Surge Line - I-5 _ _ _ _ .. , -_ . - _ . ~-
i : i { II.
SUMMARY
OF RESULTS In response to the requirements of NRC Bulletin 88-11, SCE has performed inspections and a plant specific evaluation of the pressurizer surge line at SONGS Units 2 and 3. The results of the evaluation and the inspection of the surge line demonstrate that all requirements of Bulletin 88-11 are : satisfactorily met. Results of the analysis show that'all the requirements of l the ASME Code are satisfactorily met for the _following components for the life ! i of the plant-i
. The surge line and its supports, !
. The RCS hot leg surge nozzle, ,
. The pressurizer surge nozzle,-and l
. The liquid sample line '
The results of the plant specific evaluation are.in agreement with the results : of CE0G Report CEN-387-P, Revisions 0 and I which concluded that all CE designed surge lines, including SONGS 2 and 3, meet all the applicable ASME ;
- Code requirements for the service life of. 40 years. Detailed results of the '
ASME Code evaluation of the surge line and all impacted components can be found in Attachment A (Section VII) of this report. Furthermore, all other actions and NRC requests and recommendations specified I l in Bulletin 88-11 have been completed as detailed in Sections IV and V of this j report. t i 2 l
.i J i
?
e 11-1 1 -t
l III. NRC SAFETY EVALUATION OF COMBUSTION ENGINEERING OWNERS GROUP (CEOG) REPORT CEN-387-P, REVISION 1 The NRC staff and its consultant, Brookhaven National Laboratory (BNL), have completed the review of the CEOG report CEN-387-P, " Pressurizer Surge Line Flow Stratification." The NRC concluded that the CE0G analysis adequately demonstrates that the bounding pressurizer surge line and surge nozzles meet all applicable American Society of Mechanical Engineers (ASME) Code stress and fatigue requirements for the 40-year design life of the facility considering the phenomena of thermal stratification and thermal striping. NRC Bulletin 88-11 required licensees to update their stress and fatigue analyses to ensure compliance with applicable ASME Code requirements. The CEOG report can be used to update plant specific Code stress reports. NRC's SER requires SCE to verify the applicability of the CE0G bounding analysis in CEN-387-P, Revision 1, to SONGS Units 2 and 3. In addition, the SER requires confirmation that ali actions required by Bulletin 88-11, including the updating of the plant specific stress and fatigue analyses, have been completed. It should be noted that the NRC's safety evaluation (Reference 15) states that due to the fact that the elastic-plastic analysis was necessary in performing the pressurizer surge line stress evaluation, the NRC concurs with BNL's recommendation for performing enhanced inservice inspections to provide additional confidence in the structural integrity of the surge lines. Therefore, the NRC recommends that SCE perform a volumetric examination of critical elbow components as part of the future ASME Section XI inservice examinations. Examinations of elbow bodies, as well as welds, should be performed to ensure that the most highly stressed areas have not sustained damage. These additional examinations are staff recommendations only. However, it is requested that SCE inform the NRC of its intentions regarding the implementation of these staff recommendations. III.1 SCE's Response to SER Recommendations Current ASME Code Section XI inspection requirements for Class I piping (Table IWB 2500) covers only welds. The ASME Section XI Task Group on ISI optimization is presently chartered to establish recommendations for surge III-I
i 1 line inspections, including welds and base metal. Progress has been delayed, with the concurrence of the NRC, for the NRC to complete review of the surge ' line analyses by the three vendor owner groups. Future inspection criteria for the surge lines can best be developed through the CEOG in cooperation with , the ASME Task Group. , Therefore, performance of base metal volumetric examination will be evaluated 1 to present Section XI requirements including future ISI programs on surge line , inspection when the ASME Section XI criteria are modified by ongoing Code efforts. Until then, current Section XI inspections will continue. It should be noted that CE0G member utilities are currently active in the Task Group on ISI optimization and in the interim member utilities will cooperate with the i ASME in developing future ISI requirements. i e n l l I d l III-2 l j l
. _ , . _ _ _ _ . . , .1
i l \ l I l IV. NRC BULLETIN 88-11 REQUIREMENTS l IV.1 Actions Requested l l Since thermal stratification was not addressed in the original design of the l pressurizer surge line, the effects of this phenomenon were not included in l the design basis analysis of any CE designed plant. Actions were already l underway by the Utility Owner's Groups to address this phenomenon when, in l December 1988, the NRC issued Bulletin 88-11 which requested that specific I actions be taken by the utilities to address the issues associated with thermal stratification. A summary of these requested actions follows:
- 1. For all licensees of operating PWRs:
A. Action 1.a Perform a visual inspection walkdown (ASME Section XI, VT-3) at the first available cold shutdown after the receipt of the bulletin which exceeds 7 days. fhis inspecticn should determine any gross discernable distress or structural damage in the entire pressurizer surge line, including piping, pipe supports, pipe whip restraints and anchor bolts. B. Action 1.b Perform a plant specific or generic bounding analysis to demonstrate that the surge line meets applicable design codes and other Final Safety Analysis Report (FSAR) and regulatory commitments for the licensed life of the plant. The analysis is requested within four months for the plants in operation over 10 years and within 1 year for plants in operation less than 10 years. If the analysis does not demonstrate compliance with these requirements, submit a justification for continues operation (JCO) and implement Actions 1.c and 1.d below. i C. Action 1.c Obtain data on thermal stratification, thermal striping, and surge line , deflections either by plant specific monitoring or through collective ! efforts among plants with a similar surge line design. If the collective effort option is selected, the licensee should demonstrate similarity in geometry and operation. 1 IV-1
i 5 l 1 i i D. Action l'.d l' i Perform detailed stress and fatigue analysis of.the surge line to ensure- ~ compliance with applicable ASME Code requirements, incorporating any ! observations from 1.a above. The analysis should be based on the , applicable plant specific or referenced data, and should be completed ; no later than 2 years after the receipt of the bulletin. If the detailed ; analysis is unable to show compliance, submit a' JC0 and description of .j i corrective actions for effecting long term resolution. ! l
- 2. For all applicants for PWR-Operating Licenses: l This action is not applicable to SONGS. -f
- 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- .j any subsequent corrective actions, and maintain these records in ,
accordance with 10CFR part 50, Appendix B and plant procedures, j i IV.2 Reporting Requirements I
- 1. Addressees shall report to the NRC any discernable distress and damage l observed in Action 1.a along with corrective actions taken or plans and schedules for repair before restart of the unit.
- 2. Addressees who cannot meet the schedule described in Items 1 or 2 of Actions Reauested are required to submit to the NRC within 60 days of !
receipt of the bulletin an alternative schedule with justification' for the requested schedule. L 3. Addressees shall submit a . letter within 30 days after the completion of .i these actions which notifies the NRC that_ the actions requested in Items 1.b, 1.d or 2 of Actions Reauested have been performed andLthat the results are available for inspection. The letter shall include the ; justification for continued operation, if appropriate, a description of-the analytical approaches used, and a summary of the results. l 1 IV-2 l l i I l
l l 1 l 1 i l V. SCE RESPONSE TO NRC BULLETIN 88-11 ACTIONS i i A. Action 1.a: Perform visual inspection , i The Bulletin requires that a visual inspection be performed in accordance with ASME Section XI, VT-3 at the first available cold shutdown that exceeds seven ; days. Inspection performance and results are described below. Inspection Performance ASME Section XI, VT-3 requires that:
. Visual examination shall be conducted to determine the gereral mechanical and structural conditions of the components and their supparts, such as the presence of loose parts, debris, or abnormal corrosion products, wear, ,
erosion, corrosion and the loss of integrity at bolted or velded I connections.
. Visual inspection may require as applicable to determine structural :
integrity, the mechanical measurement of clearances, detection of physical displacement, structural adequacy of supporting elements, connections between load carrying structural members and tightness of bolting.
. For component supports and component interiors, the visual inspection may be performed remotely with or without optical aids to verify the structural integrity of the component.
This inspection was accomplished in accordance with SCE approved CE visual examination procedure for preservice and inservice inspections, Reference 7. 1 In this procedure, VT-3 examination standards are identified per the ASME Code requirements. Any relevant conditions which need to be recorded will be on l the inspection form. ) Inspection Results The inspection was done by a level II examiner on January 22,1989 for Unit 2, and on April 18 and 19,1990 for Unit 3. The results of the visual inspection showed that there is no discernable stress or structural damage to the pipe and pipe supports. The results of the inspection were satisfactory. i V-1
l l
- l
.i l
1 4 B. Action 1.b: Bounding Evaluation i 4 Action 1.b of Bulletin 88-11 requests that a bounding' analysis be performed which demonstrates acceptability-for the licensed life of the plant. In SCE's. ; initial response to NRCB 88-11 (Reference 3), SCE stated that it.would provide , the NRC with the results of our plant specific analysis required by the '{ bulletin. This letter also informed the NRC that SCE was participating in a i CEOG program that would provide a generic bounding analysis upon which SCE's : plant specific analysis would be based. CEOG submitted this analysis, CEN-4 387-P, Revision 0, dated July, 1989 (Reference 5), and the Revision 1 version, dated December 1991 (Reference 6). .[ q l l CEOG Response to Action 1.b In July 1989, CE0G submitted a report (Revision 0 of the CEN-387-P), which j contained a bounding generic analysis performed using generic loading ! conditions and plant specific surge line data. This report documented the j results obtained for a pressurizer surge line flow stratification evaluation. ! l This evaluation addressed the impact of surge line thermal stratification and ! themal striping as reported by INPO SER 25-87 and NRCB 88-11. The results of I this evaluation demonstrate that the structural integrity of.all CE0G l pressurizer surge lines is maintained for their forty year design life as j required by Bulletin 88-11. l l The CEOG program consisted of collecting and reducing data (outside pipe wall- i temperature and displacements), developing thermal hydraulic models, defining l new generic themal load definitions, performing a stress and fatigue analysis and determining the fatigue life of the surge line. Based on measured outside wall temperatures, conservative thermal hydraulic models were developed for a the thermal striping, stress and fatigue evaluations in a manner that was I consistent with the test data. The data noted above, was temperature data collected from SONGS Unit 3 (August
- 1988) surge line. SCE collected temperature data with surface mounted
; thermocouples during heatup. In addition, SCE recorded temperature data during the drawing of the steam bubble.
1 V-2 I 4
4 I I . i. i i ; ! l 5 j As a result of the analysis performed on the surge lines of the participating 1 utilities, all of the surge lines exceeded the 35m elastic stress limit of i Equation 12 of the ASME Code, Section III, Subsection 3650 (Sm is the i allowable stress intensity). An elastic-plastic analysis was performed on the 3 most highly stressed surge line to demonstrate that all surge lines shakedown, i i.e., progressive distortion will not occur. Although the ASME Code stress limits were exceeded, shakedown was proven to occur and the fatigue usage factor is less than 1.0. All CEOG surge lines have a fatigue usage factor j less than 1.0, and hence a fatigue life greater than 40 years. l Revision 0 of the report was divided into seven sections of analytical ! evaluations. The first two were associated with the data collection and i j reduction. This task consisted of a study of the pressurizer and reactor - ! . coolant system and the selection of thermocouple locations. Data was then ! collected and reduced to a manageable format. Transients, temperature ranges and generic loadings were then determined from the data and specific design r conditions. The third area of study concentrated on thermal hydraulic models. The purpose of this evaluation was to obtain an understanding of the relationship between ! the surface mounted thermocouples and the fluid conditions inside the surge line piping. The purpose of this task was to develop thermal hydraulic models : that would conservatively calculate the pipe wall temperature distributions consistent with test data. In this task, the generic nature of the thermal loadings on the CEOG member surge lines was addressed as well as the , correlations between the collected data and plant operations. Generic thermal ) load definitions were also developed. < t ') The fourth evaluation consisted of structural analyses for the purpose of demonstrating shakedown, after some initial plastic deformation has occurred, and that progressive distortion does not occur throughout the operating life of the plant. An elastic-plastic analysis of the bounding surge line piping was performed to investigate its behavior with respect to thermal ratcheting, and to provide the strain range for the fatigue evaluation of the piping i elbows, i.e., the piping component most severely affected by stratified flow
- loadings.
The fifth section of the analysis addressed the issue of thermal striping. This analysis used a one-dimensional heat transfer and structural finite element model to investigate the thermal stresses due to the oscillations at the hot-cold interface during stratified flow. 1 V-3
+
{ i The sixth and seventh sections of the report address the cyclic fatigue life of the surge line for all transients including thermal stratification and thermal striping. This involved modeling of the surge lines to include supports and then applying all generic loadings. - The results and conclusions of the CE0G report are applicable to SONGS. Therefore, thermal loading characteristics are generic for all of the CEOG participants. In addition, the largest temperature difference between the pressurizer and the RCS for all CE-designed plants participating in the program (system AT) was around 320 to 340 F based on plant heatup conditions. This limiting system AT was then applied to all plants configuration. The NRC reviewed this revision 0 of the report and issued a letter in August ! 1990 with a number of questions and concerns. Based on the NRC concerns the ! CEOG approved the third task in October 1990 to resolve NRC concerns. During : April 1990, surge line cooldown data were collected at SONGS and other CEOG ; plant and analyzed. Subsequently, meetings and phone conversations were held : with the NRC, and consequently CEOG issued Revision 1 of the report to resolve the NRC questions. Revision 1 of the report was reviewed by the NRC staff and its consultant, Brookhaven National Laboratory, and staff concluded that: "the l CEOG analysis adequately demonstrates that the bounding surge line and nozzles ' meet ASME Code stress and fatigue requirements for the 40-year design life of - the facility considering the phenomenon of thermal stratification and thermal ! stri pi ng. " This NRC staff evaluation was documented in a letter from the NRC to H. B. Ray of SCE (Reference 15). i Therefore, all ASME Code requirements were met and shakedown was demonstrated. To confirm that the intent of the Code requirements have been met, an ASME Code Inquiry has been submitted, and the response was favorable. All CEOG i surge lines have a fatigue usage factor of less than 1.0 and hence a fatigue l life of greater than 40 years. This analysis included the phenomena of thermal stratification and thermal striping in the fatigue and stress evaluations. For SONGS, it is confirmed that all specific surge line support capabilities are within the range assumed in this analysis. V-4 i s i J
, w--, -- - , - w, , - - - - - - , ,-- ,, , - - - -
+ .-- .~. .- .. . -~ -- - - - - .
8 3
) I i [
1 > I i l ) { i l Action 1.c: Plant Monitoring ) C. Requested Action 1.c of NRCB 88-11 states that utilities may obtain surge line
- monitoring data either by plant specific monitoring or through a collective !
effort. If the latter_ option is selected, similarity in geometry and operation should be demonstrated. CEOG, in a combined effort with SCE, obtained plant specific data through monitoring of the surge line at SONGS 4 Unit 3. ! I ! The pressurizer surge lines at SONGS Units 2 and 3 are approximately seventy- l J nine feet long. Each line is made up of a 12 inch schedule 160 pipe that runs ! primarily horizontal, attaching vertically to the bottom of the pressurizer i and the top of the RCS hot leg. Detailed description of the surge line ! geometry, support configuration and material properties can be found in l Attachment A of this report. ; i i ? J Pipe wall temperature monitoring was performed on the SONGS Unit 3 surge line. ! Two surface mounted thermocouples (T/C) were positioned-at the top (0*) and ! l bottom (180 ) of the surge line piping at location 1 (see Figure 4.9 of ! l Attachment A). Locations 2, 3, 4, and 5 were fitted with bands of six
- thermocouples that were 60* apart starting from the top (0*). Location 6 had !
! two thermocouples, one on the east and one on the west side of the pipe. The i T/C's used were Omega Type K with Type K wiring. The wiring was run outside !
of containment to the plant computer, a GOULD 32/9780. All thermocouple data l were recorded at two minute intervals by the plant computer. A more detailed j description of the pipe wall temperature monitoring and a summary of the ! results can be found in Section 4 of Attachment A. j Plant parameters were also collected at two minute intervals. These included l the following: i I - Cold leg temperature, !
. Surge line temperature. '
- Pressurizer pressure and level, and
. Reactor coolant pump status.
l j During heatup, data recording commenced prior to drawing the bubble in the pressurizer, and continued until the plant was at normal operating temperature and pressure. During cooldown, data were collected at 100% power and j continued until the bubble in the pressurizer was collapsed. The plant j j V-5
- ~~ - - . , . . - , . - - _ . - - . . - , - . . , . . . - - , , . , , - . - . . . . ,---,..,-,,.,n.--_,,,
i I i computer produced a hard copy of the recorded data. This data was entered on to floppy computer disks and plotted. The data described above was used in CEOG analysis and is included in their acceptability report for Items 1.b and 1.d. l D. Action 1.d: Update Stress and Fatigue Reports ) CEOG issued Revision 1 of Report CEN-387-P in response to additional NRC questions about Revision 0 of the same report. Revision 1 also serves as a t part of SCE response to Action 1.d. )' l l This Revision 1 effort was divided into eight tasks. The first task was data ; collection during plant heatup and cooldown (noted above). This consisted of a study to determine what plant dcta relevant to the issue of surge line flow i l stratification could be collected. i l l The second task was to reduce the plant data and generate time versus 1 l temperature plots. This task also involved determining which plant events resulted in surge line thermal transients. In this task, the generic nature of the thermal loadings on the CEOG member surge lines was addressed as well l as the correlations between the collected data and plant operations. l The third task concentrated on thermal hydraulic modeling. The purpose of this evaluation was to obtain an understanding of the relationship between the temperature measured by the surface mounted thermocouples and the fluid conditions inside the surge line piping. This study determined how to conservatively calculate the pipe wall temperature distribution using thermal hydraulic models. The fourth task was to develop revised thermal transients for the pressurizer surge line in view of thermal stratification. Generic thermal load I definitions were developed for use in both the structural and striping l analyses. The fifth analysis addressed the issue of thermal striping. This analysis used a one dimensional heat transfer and structural finite element model to investigate the thermal stresses due to the oscillations at the hot-cold interface during stratified flow. V-6
The sixth evaluation consisted of a structural analysis of the surge line. The first step in the evaluation was to model all CEOG surge lines. An elastic piping analysis of each surge line using the revised surge line thermal transients was then performed. The second step was to perform an elastic-plastic analysis for the bounding surge line (bounding was based on the results of the elastic analysis). The seventh task was performed to evaluate the cyclic fatigue life of the surge line for all transients including thermal stratification, thermal striping and Operating Basis Earthquake (0BE) seismic loading. This involved applying the results from the fifth and sixth evaluations to each plant surge line. All load states were included and the fatigue analysis was performed in accordance with the ASME Code cycle combination methodology. The eighth task was performed to determine the effect on both the pressurizer surge nozzle and the RCS hot leg surge nozzle. This nozzle evaluation was performed per the requirements of the applicable ASME Code. In addition to the CEOG evaluation of the surge line, SCE performed a plant - specific evaluation of SONGS Units 2 and 3. A summary description of this plant specific evaluation including analysis results, analysis methodology, system description and design input is documented in Attachment A (Section VII) of this report. As part of the plant specific analysis, design specifications for the surge line, pressurizer surge nozzles, and RCS hot leg surge nozzle are being updated to include the new plant design load cycles from thermal stratification. This effort is being performed by the CEOG in behalf of SCE. Additional NRC Requests Response to Reouest No. 2: This item requests information from all applicants for PWR Operating Licenses and is therefore not applicable since SONGS has its operating license. Response to Recuest No. 3: This item requests addressees to generate records to document the development and implementation of the program requested by Items 1 and 2. V-7
I l SCE has compiled records to document the development and implementation of the program requested by Actions 1.a. 1.b, 1.c, and 1.d, as well as any subsequent corrective actions. These records are maintained in accordance with 10 CFR ! Part 50, Appendix B and the plant procedures. l 1 i V.1 SCE Response to Reporting Requirements
- 1. This item requires the addressees to report to the NRC any discerrable distress and damage observed during the walkdown requested in Action 1.a along with corrective actions.
I SCE has completed the inspection of the pressurizer surge lires and the
- results were satisfactory. These inspections were performed on January 22, 1989 for unit 2 and April 18, 1990 for Unit 3. There was no evidence 1
of bowing damage, structural distress damage or visible degradation. All supports appeared to be intact, and no wear or fretting of the j pressurizer surge line was noted at support locations. l
- 2. This item requires the addressees who cannot meet the schedule described 1 in Items 1 and 2 of Actions requested to submit to the NRC within 60 days of receipt of the bulletin and alternative schedule with justification for the requested schedule.
i This item was addressed initially in Revision 0 of the CEOG report (Reference 5). This report addresses all of the NRC concerns as discussed I at the September 1990, May 1991, and September 1991 meetings between the l CEOG and the NRC.
- 3. This item requires the addressees to submit a letter within 30 days after the completion of the action items 1.b,1.d, or 2 to notify the NRC that the results are available for inspection.
On March 8,1989, via a letter to the NRC (Reference 8), SCE informed the NRC i of its participation in the CE0G program ( Reduction and Analysis of Pressurizer Surge Line Data Collected from CEOG Plants") and stated that it would provide the NRC with the results of SONGS 2 and 3 plant specific analysis required by NRCB 88-11 by February 4,1991. By Reference 9, the NRC acknowledged that it would be acceptable for SCE to work through the CE0G to address the issues discussed in Bulletin 88-11. On March 1, 1991 SCE submitted another letter (Reference 10) to the NRC to confirm that the V-8
original due date of February 4,1991, was superseded by the September 5,1990 meeting (documented in Reference 11), and the new schedule for completing our plant specific analysis is December 30, 1991. On May 7 and 8, 1991, the NRC staff audited the CEOG progress (Reference 12). The audit identified additional items that AB8/ Combustion Engineering needs to resolve. A second meeting between the NRC and the CE0G was held on September 18 and 19, 1991. This meeting resolved many of the NRC staff's concerns (Reference 13). In this meeting, the NRC agreed to a link between the plant specific analysis due date and the issuance of a favorable NRC Safety Evaluation Report (SER) for the CEOG bounding analysis. In addition, the NRC acknowledged that a Justification for Continued Operation (JCO) is not required from the utilities if the CE0G analysis report is issued before December 31, 1991. Therefore, SCE submitted a letter to the NRC (Reference
- 14) stating that our report will be submitted to the NRC within either 90 days :
after the issuance of a favorable SER, or by September 1,1992, whichever is ; later. With the submittal of the CEOG analysis which includes a description of the analytical approaches and a summary of the results, and this report, the action in Item 3 above is completed. i V-9
i VI. LIST OF REFERENCES AED FIGURES VI.1 REFEREHCES t
- 1. NRC Bulletin No. 88-11, Pressurizer Surge Line Thermal Stratification, dated December 20, 1988.
- 2. SCE Calculation No. N-0220-027, Revision 0,
Subject:
SONGS Unit 2&3 Maximum Pressurizer and RCS Hot Leg Temperature Differential During Heatup.
- 3. Letter from F.R. Nandy (SCE) to Document Control Desk (NRC) dated March 8, 1989,
Subject:
Pressurizer Surge Line Thermal Stratification.
- 4. SCE letter to CE dated November 22,1991.
- 5. Combustion Engineering Owners Group (CE0G) report CEN-387-P, Revision 0, Pressurizer Surge Line Thermal Stratification Evaluation, July 1989.
- 6. Combustion Engineering Owners Group (CEOG) report CEN-387-P, Revision 1 Pressurizer Surge Line Thermal Stratification Evaluation, December 1991.
- 7. CE Visual Examination Procedure for Preservice and Inservice Inspection, 5023-ESS-066.
- 8. Letter from F. R. Handy (SCE) to Document Control Desk (NRC) dated March 8,1989;
Subject:
Response to NRC Bulletin 88-11.
- 9. Letter from L. W. Kokajko (NRC) to Messrs. Harold B. Ray (SCE) and G. D.
Cotton (SDG&E) dated July 25,1990;" Evaluation of Combustion Engineering Owners Group Bounding Analysis."
- 10. Letter from F. R. Nandy (SCE) to Document Control Desk (NRC), dated March 1,1991;
Subject:
NRC Bulletin, Pressurizer Surge Line Thermal , Stratification, SONGS Units 2&3. I
- 11. Letter from Edward C. Sterling (CE0G) to J. T. Larkins (NRC) dated October 24,1990, " Evaluation of Combustion Engineering Owners Group Bounding Analysis Regarding NRC Bulletin 88-11."
- 12. Letter from L. E. Kokajko (NRC) to Messrs. H. B. Ray (SCE) and G. D.
Cotton (SDG&E), dated September 13,1991," Pressurizer Surge Line Thermal Stratification, Bulletin 88-11." i VI-1
l l VI.1 REFERENCES - cont. l i , 13. Letter from P. J. Hijeck (CEOG) to CE0G Analysis Subcommittee Members dated September 30,1991,"NRC/CE0G Pressurizer Sur " meeting of September 18 & 19, 1991 (CEOG Task 662)ge Line Stratification
- 14. Letter from R. M. Rosenblum (SCE) to Document Control Desk (NRC) dated December 17,1991; NRCB 88-11 Plant Specific Analysis Submittal Schedule.
- 15. Letter from Mel B. Fields (NRC) to Harold Ray (SCE) dated August 2,1993.
Subject:
Safety Evaluation for CEOG Report CEN-387-P, Revision 1, i
" Pressurizer Surge Line Thermal Stratification Evaluation," (Bulletin 88- j 11). ,
1 l 4 < VI.2 LIST OF FIGURES l
)
i 2
- 1. Figure I-1 Schematic of the Reactor Coolant System at SONGS 2 and 3.
- 2. Figure I-2a Schematic of Stratified Flow in the Surge Line.
- 3. Figure I-2b Schematic of Thermal Striping in the Surge Line.
i 4 VI-2
l VII. ATTACHMENT A i s l i i VII.1 ANALYSIS
SUMMARY
REPORT THERMAL STRATIFICATION IN THE PRESSURIZER SURGE LINE 5 SAN ONOFRE NUCLEAR GENERATING STATION i ! UNITS 2 AND 3 l i
)
1 l l l i ( l I l VII-1 l l
i l i i 1
-i ATTACHMENT A - TABLE OF CONTENTS !
+
t t Sheet Number !
+
l i TABLE OF CONTENTS . . . . . . . . . . . . . .'. . . . . . . . VII-2 i 1 PURPOSE . . . . . . . . . . . . . . . . . . . . . . . . . . . VII-5 i 2 RESU LTS/ CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . VII-6 o t 3 ASSUMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . VII-12 _ t i 4 DESIGN INPUT ........................ VII-14 i 5 METHODOLOGY ........................ VII-39 I t 6 REFERENCES ......-................... VII-83 i 7 NOMENCLATURE ........................ VII-86 , i l i 5 VII-2 l
,,._--...m.. .-. . , , - . .
l 1 PURPOSE Item 'd' of Bulletin 88-11 requires plant specific update of the pressurizer surge line stress and fatigue analyses to ensure compliance with the (ASME) Code requirements. Accordingly, the pressurizer surge lines at SCE's SONGS Units 2 and 3 were reanalyzed to include the effect of thermal stratification. This effort comprises the tasks described in Sections 1.1 through 1.5. , It should be noted that in addition to the plant specific evaluation summarized in this attachment, a generic bounding evaluation was issued by CEOG. This report was reviewed and approved by the NRC, as explained in Section V of this report. The purpose of plant specific analysis required by the bulletin is to build upon the CEOG work, and to demonstrate the SONGS Units 2 and 3 are bounded by the results included in the CE0G report. 1.1 Pressurizer Surge Line Structural and Fatique Analysis The purpose of this task is to perform an analysis per ASME Code, Section NB- , 3650 on the pressurizer surge line including the effect of thermal ' stratification. The purpose of the structural analysis step of this task is i to generate the response of the surge line to unit thermal and mechanical loads. This response was obtained in terms of forces and moments at different locations. The thermal loads represent both stratified and non-stratified conditions. The results of the structural analysis are used in a subsequent ASME Code evaluation to ensure compliance with the Code under the updated , I design bases including thermal stratification. In this evaluation, applicable ! Code equations stresses are to be calculated and compared with the corresponding allowables. Fatigue usage factors were calculated based on the j design transients, listed in Section 4, for the service life of the plant. In addition to generating the data necessary for the Code evaluation, the structural analysis of the surge line is also used to calculate the following: (a) Reaction loads at the pressurizer and hot leg surge nozzles. l (b) Rigid support loads and displacements at all support locations. These results are generated for use in the pipe support evaluation. I (c) Calculate the thermal movement at the liquid sample line branch ! connection, which is used to update the sample line structural analysis. Calculation of support loads and thermal movement is conservatively based on the most severe stratified flow conditions, from past operation. VII-3
l
-i 1.2 Surce Nozzles The surge nozzles at the RCS hot leg side and the pressurizer side of the surge line are impacted by thermal stratification, and require re-evaluation. !
Structural and fatigue evaluations were performed on both nozzles to ensure : compliance with the Code requirements. ! 1.3 Integral Attachments ; Each of the pressurizer surge line at SONGS Units 2 and 3 has one rigid ! l support, four snubbers and two spring hangers as shown in Figure 4.2. The rigid support has two rectangular lugs welded to the surge line. Similarly,
- each snubber is attached to the surge line by means of a dummy pipe. Two l l dummy pipes are horizontal and the other two are vertical. The purpose of this analysis task is to perform a fatigue evaluation on the rectangular lug and the bounding dummy pipe. l 1.4 Liouid Sample Line Each of the pressurizer surge line at SONGS Units 2 and 3 has one sample l branch line as shown in Figure 4.2. One of the purposes of the structural analysis of the surge line (Section 1.1) is to produce updated thermal ;
movement at the liquid sample line branch connection under stratified flow ' conditions in the surge line. The purpose of this task is to re-analyze the liquid sample line using the calculated thermal movement to ensure compliance with the Code requirements when stratified flow conditions exist in the surge line. Conservatively, the evaluation is based on the most severe thermal - stratification conditions. , l 1.5 Pipe Support Evaluation l l The purpose of this task is to perform a pipe support re-evaluation in view of i thermal stratification in the pressurizer surge line (see the pipe support '
- configuration in Figure 4.2). This evaluation uses updated thermal loads and l displacements based on the results of the structural analysis of the surge l line (see Section 1.1 above) including the most severe thermal stratification
( condition. l l VII-4 ,
l l 2 RESULTS/ CONCLUSIONS Structural and fatigue analyses of the pressurizer surge line were performed ! per the ASME Code to evaluate the effect of thermal stratification on the l pressurizer surge line, the RCS hot leg surge nozzle, the pressurizer surge l nozzle, and the surge line integral attachments. The evaluation includes the l effects of thermal striping. Thermal stress ratchet analysis was performed per the ASME Code when required. Results also include an evaluation of the pressurizer surge line support loads and displacements. Furthermore, a re-analysis of the liquid sample line was performed since it is affected by thermal stratification in the surge line. A summary of these results is given in Sections 2.1 through 2.4 2.1 Pressurizer Surae Line Pipina and Intearal Attachments Structural and fatigue analysis of the pressurizer surge line were performed per the ASME Code to include the effect of thermal stratification. Results of the analysis are summarized in Tables 2.1 and 2.2. These tables summarize the results of the stress and fatigue evaluations at different piping locations (see Figure 4.2 for node point definition). Specifically, a total of 18 l piping locations, identified by the ANSYS node point numbers shown in Figure 4.2, were evaluated per NB-3600 rules. The evaluation was performed on the inside and outside surfaces at each ANSYS piping node point location. l l The stress limit of equation (10) of ASME Code, Subsection NB-3653 was , exceeded at ten of the 18 locations shown in Figure 2.1 on the inside diameter i l of the pipe. Simplified elastic-plastic analysis rules were invoked at these ! ten locations, and Code Equations 12 and 13 stresses were evaluated. Also, a K, penalty factor was used in the fatigue analysis as required by the Code. Furthermore, a thermal stress ratchet was performed in accordance with NB-3653.7 requirements (see Section 2.1.3). Similarly, Equation (10) stress limit was exceeded at 12 outside diameter locations. Simplified elastic-plastic analysis rules were invoked for 8 of these 12 locations, and Code Equations 12 and 13 stresses were evaluated. Also, a K, penalty factor was used in the fatigue analysis as required by the Code. Furthermore, a thermal stress ratchet was performed in accordance with i NB-3653.7 requirements (see Section 2.2.3). The remaining 4 locations that did not meet the simplified elastic-plastic requirements met the Code i requirements for shakedown. The Combustion Engineering Owners Group (CE0G) evaluation of the pressurizer surge line (Reference 3) has demonstrated shakedown for all surge line locations. Analysis results, however, show that ; the 4 locations that did not meet the simplified elastic-plastic requirements ! are bounded by the Reference 3 evaluation. Accordingly, it is concluded that l l the shakedown analysis of Reference 3 is bounding. l 1 VII-5
l 1 4 a d i 4 I ! i j Table 2.1 Stress and Fatigue Evaluationo).(r) _ Inside Diameter ! Location on the Eq (10)/ Eq (12)/ Eq (13)/ Fatigue , j surge line S, S, S, usage factor l Node point: 3 3.00 1.98 1.33 0.536 , 1 4 3.88 2.82 1.49 0.981 i 12 3.99 2.84 1.39 0.787 ; 14 3.85 2.85 1.41 0.815 22 3.78 2.80 1.40 0.705 23 2.00 1.10 0.96 0.371 28 1.91 0.98 1.10 0.453 i , , 29 3.37 2.23 1.35 0.438 ] 37 3.31 2.17 2.17 0.438 > e 38 1.61 0.75 1.08 0.440 44 1.61 0.76 1.02 0.340 1 1 46 2.80 1.87 1.39 0.422 4
- 54 3.18 2.07 1.39 0.431 2
57 3.19 2.09 1.39 0.433 65 3.06 1.98 1.37 0.429 67 3.07 1.89 1.90 0.993 ] j 68 2.15 1.13 1.49 0.697 I l 69 2.29 1.24 1.44 0.536 l l Notes: (1) See Figure 4.2 for node point definition l l (2) See Section 5.3.1 of this attachment for details of stress limits i l and applicable ASME Code criteria I i l 1 i s VII-6
l t Table 2.2 Stress and Fatigue Evaluationo).(r) . l Outside Diameter Location on the Eq (10)/ Eq (12)/ Eq (13)/ Fatigue surge line S, S, S, usage factor Node point: 3 3.50 2.49 1.39 0.096 4 4.59 3.55 1.58 0.401 . t 12 4.72 3.58 1.46 0.247 14 4.55 3.59 1.48 0.271 22 4.46 3.53 1.47 0.191 23 2.27 1.39 0.96 0.003 28 2.16 1.23 1.15 0.005 ; 29 3.94 2.80 1.40 0.018 37 3.87 2.73 1.46 0.015 38 1.79 0.94 1.12 0.004 44 1.79 0.96 0.96 0.002 46 3.22 2.35 1.46 0.004 54 3.71 2.61 1.46 0.007 57 3.72 2.64 1.46 0.009 65 3.56 2.50 1.42 0.008 67 3.38 2.16 1.96 0.089 68 2.48 1.42 1.56 0.024 69 2.69 1.57 1.53 0.015 Notes: (1) See Figure 4.2 for node point definition (2) See Section 5.3.1 of this attachment for details of stress limits and applicable ASME Code criteria. Section 5.3.1 also addresses the cases where Equation 12 limit is exceeded l 1 l l l VII-7
N l l l l r I 2.1.3 Thermal Stress Ratchet In accordance with the rules of ASME Code, Subsection NB-3653.7, thermal i stress ratchet check was performed at locations where Code Equation (10) stress limit is exceeded. Similarly, the rules of NB-3222.5 require a thermal stress ratchet check to be performed. The methodology of the this check is summarized in Section 5 of this summary report. Results of the thermal ratchet check indicate that all ASME Code requirements are satisfied. 2.1.4 Thermal Stripina Thermal striping analysis was performed, and the fatigue usage factor due to l thermal striping was shown to be less than 0.01. It occurs at the hot / cold ) fluid interface at the middle of the pipe, while the accumulated usage factor i due to all other loadings occurs at either the top or the bottom of the pipe. ! Therefore, the thermal striping fatigue usage factor was not included in the ! usage factor calculations. It should be noted, however, that the fatigue ! usage factor of 1.0 is met at all locations even when the thermal striping is ; included, i Description of the methodology of thermal striping fatigue analysis can be found in Section 5 of this attachment. i I I l VIl-8 l
2.2 Surae Nozzles The pressurizer surge nozzle and the RCS hot leg surge nozzle were evaluated per Subsection NB-3200 of the ASME Code (Design by Analysis). Results are summarized in Tables 2.3 and 2.4. Table 2.3 Summary of Pressurizer Surge Nozzle Evaluation Maximum Maximum (l) Fatigue Usage l (Pt+Pe+Q)/S, (P +P,+Q)/S, t Factor 1 Nozzle Inside Diameter 3.15 1.00 0.252 Nozzle Outside Diameter 3.01 1.00 0.009 1 l Notes: (1) Excluding thermal bending. Table 2.4 Summary of Hot Leg Nozzle Evaluation Maximum Maximum (2) Fatigue Usage (P t+P,+Q) /S, (Pt +Pe +Q)/S, Factor Nozzle Inside Diameter 3.28 1.01 0.767 i Nozzle Outside Diameter 3.43 1.01 0.062 Notes: (1) Excluding thermal bending. It should be noted that in Tables 2.3 and 2.4 the limit on (P +P o 3 +Q)/S, is 3.0 per NB-3222.2. However, as shown in Subsection NB-3228.5 of the ASME Code, this limit can be exceeded provided that: . l (a) (P t+P g +Q)/S , excluding thermal bending, is lass than 3.0 (b) A penalty factor, K,, is applied to the alternating stress, S,, in the fatigue analysis, and (c) The fatigue usage factor is less than 1.0. Details of the applicable criteria can be found Section 5.3.1 of this ; attachment. Accordingly, it is concluded that the requirements of NB-3200 are met. VII-9
2.3 Intearal Attachments - Section 2.3 Each surge line at SONGS Units 2 and 3 has five integral attachments, as shown in Figure 4.2. These attachments include rectangular lugs at the rigid support (node point 38), and dummy pipes at all snubber locations (node points 26, 28, 38 and 44). The effect of thermal stratification is greater on horizontal dummy pipes since they are subjected to greater thermal gradients during stratified flow conditions in the surge line. Therefore, a horizontal dummy pipe was conservatively selected for analysis. Integral attachments were evaluated per Subsection NB-3650 in combination with Code Cases N-122-1 and N-391-1. The methodology of this evaluation is described in Section 5.3.4 of this attachment. Tables 2.5 and 2.6 summarize the results obtained for the integral attachments on the inside and outside diameters of the pipe, respectively. Table 2.5 Integral Attachments EvaluationI2) - Inside Diameter Maximum Eq 10 Maximum Eq 12 Maximum Eq 13 Fatigue Stress Ratio Stress Ratio Stress Ratio Usage Rectangular 2.25 0.75 1.28 0.740 lug rei Dummy pipe () 4.15 0.76 1.37 0.562 Notes: (1) See Section 5.3.1 of this attachment for details of the application of Code Cases N-122-1 and N-391-1 (2) Code Case N-122-1 used (3) Code Case N-391-1 used Table 2.6 Integral Attachments Evaluation ll) - Outside Diameter Maximum Eq 10 Maximum Eq 12 Maximum Eq 13 Fatigue Stress Ratio Stress Ratio Stress Ratio Usage Recgngular 2.46 0.94 1.32 0.055 lug Dummy pipef3) 4.33 0.96 1.39 0.064 Notes: (1) See Section 5.3.1 of this attachment for details of the application i of Code Cases N-122-1 and N-391-1 (2) Code Case N-122-1 used (3) Code Case N-391-1 used VII-10
i 2.3 Liauid Sample Line Structural Analysis l The maximum displacement at the pressurizer liquid sample branch connection ; was calculated for a maximum recorded system temperature differential (AT,,, = 386*F) per reference 16. This evaluation is similar to the existing evaluation of the branch line, except that the thermal displacements at the l branch line connection include the effect of thermal stratification in the surge line. The structural analysis of the liquid sample line was performed for SONGS Units 2 and 3 under the thermal stratification conditions described above. Results of the analysis show that all ASME Code requirements are met. Pipe supports were evaluated for the new loads, and were found adequate under the new loads. l l 2.4 Pipe Support Loads As shown in Figure 4.2, each pressurizer surge line at SONGS has one rigid support, four snubbers and two spring hangers. The Rigid support load and ! displacements at all snubber and spring hanger locations were calculated under maximum thermal stratification conditions in the surge line. The rigid support was evaluated, and was found capable of sustaining the new faulted load. Similarly, the displacements at all snubber and spring hanger locations are within range, and no resetting is required. l l VII-11
l l' l l 3 ASSUMPTIONS
- 1. Thermal stratification, in the horizontal part of the pressurizer surge line, is constant in the axial direction of the pipe. This assumption is conservative, since an axially varying stratification profile results in lower bending of the pipe, and consequently lower stresses.
! 2. Thermal stratification does not occur in the vertical runs of the j pressurizer surge line, or the nozzle at both ends of the line. The i difference in density between hot water and cold water should not produce ; a hot side and a cold side in a vertical pipe. This assumption is ; supported by test data recorded during plant heat up, which show that ; thermal stratification does not occur at location number 6 of the surge ' line which lies on the vertical run of the surge line below the { pressurizer. j l
- 3. Thermal anchor movement of the RCS hot leg is proportional to the i temperature rise above 70*F. The reactor cooling loop is basically ;
allowed to expand without restriction; therefore, the displacement due to temperature rise would be proportional to the magnitude of this rise. l
- 4. The ratio of the pipe wall AT to the system AT (ATwgtt /ATsrs) where AT3,3 = difference between pressurizer temperature and RCS hot leg ;
temperature, AT wntt = pipe wall top-to-bottom temperature differential, was calculated corresponding to the maximum value of AT of 308.51 F recorded at Unit-3 on 8/9/88 (Reference 17). Thiscalcbtated ratio (=0.95), however, was assumed constant for all values AT ggt , i.e., it is assumed that AT wntt can be obtained from AT 3ys at any given time by multiplying by a factor of 0.95 for all values of AT . This assumption represents a good approximation for the ratio AT watt / sys based on the available test results and the record of AT ,3 3
- 5. The hot / cold fluid interface is assumed to be at middle of the pipe cross section. This interface level should generate the highest bending (secondary) stress in the pipe that results in a conservative fatigue '
usage factor. VII-12
i l i l 3 ASSUMPTIONS - cont. I i
- 6. The fatigue usage factor due to thermal striping is calculated independently, and is not added to the total usage factor. This is based on the fact that the maximum bending stress occurs at the top and the bottom of the pipe when the fluid hot / cold interface is in the middle of the pipe. However, the stress due to striping occurs in the pipe wall at the hot / cold interface not at the top or the bottom.
- 7. Only part of the sample line, from the branch connection to the three-way l
support at elevation 40'-10", is analyzed. This segment of the line includes at least three supports in each direction. Limiting the analysis
- to this part of the line is considered acceptable since the only difference between the current analysis and the previous analysis is the l change in thermal movement at the branch connection. The impact of this difference on the line beyond the analyzed segment is negligible.
I I i i l l i l 1 l l VII-13 i._. .
4 DESIGN INPUT l ! 4.1 Geometry and General Description of Analyzed Components l l 4.1.1 Surae Line Description i I l Figure 4.1 shows a schematic of the arrangement of the reactor vessel, steam l generator, reactor coolant system (RCS) piping, the pressurizer and the i pressurizer surge line at SONGS 2 and 3. The pressurizer surge lines at SONGS l Units 2 and 3 can be described briefly as follows: ' j Pipe Size & Schedule : 12" sch 160 (outside diameter = 12.75", wall l l thickness = 1.312") l Pipe Material : SA-376 Gr TP-316 l Design Tempeiature
- 700*F l Operating Temperature: 653 F Design Pressure : 2485 psig Operating Pressure : 2235 psig 4.1.2 Surae Line Integral Attachments Each of the pressurizer surge lines at SONGS Units 2 and 3 are supported by one rigid support, four snubbers and two spring hangers as shown in Figure 4.2. Figures 4.3 and 4.4 show the details of the rigid support rectangular lugs (at node point 38) and the dummy pipe used at snubber locations (node points 26, 28, 38 and 44). Thermal and mechanical properties were obtained from the ASME material properties tables (ASME Code, Section III, Division 1, Appendix I). ;
1 l I l 4.1.3 Surae Nozzles Each of the pressurizer surge lines at SONGS Units 2 and 3 is connected by two surge nozzles to the pressurizer at one end and the RCS hot leg at the other end. Figures 4.5 and 4.6 show the details of the RCS hot leg surge nozzle, and the pressurizer surge nozzle, respectively. Thermal and mechanical properties were obtained from the ASME material properties tables (ASME Code, Section III, Division I, Appendix I). 1 VII-14
O y PRESSURIZER j SURGE LINE
, HOT LEG
/
P 8 PRESSURIZER STEAM GENERATOR DETAll A No.2 Et ATOR NO.i O O
\ /
l
\ / , , }_
y PUMP NO.1A y PUMP ll 2 ( i*ELEV.
! I 37'6"
\1 l
(Y lf
\
M l l REACTOR COOLANT i PIPING 4
\ REACTOR VESSEL SEE DETAIL A Figure 4.1 Reactor Coolant System Arrangement at SONGS 2 and 3 VII-15
Snubber [ H Pressurizer / Rigid Support [ Surge r 69 Nozzle i Spring Hanger D i 2 f 68 Instrument Line 67 66 l y ,, l 12 gg 56 RCS Ilot 12, Ln 13 l 54
; ])
11 gy
= 14 '
& 46 aJ ,
2 ,5 22 g yy e 45 s% j 23 44
, 24 42 41 25 26 27 f 40 28 39 ,
.37 38-l 29 o t'
,i i
! Figure 4 2 1sometric Drawing of the' Pressurizer Surge Line Showing Pipe.
Support s and I inite Element Model Node Numbers 3 b 4
, .u
1 4 1 4 i i i 8 l f
-kJct.a.
_ _ a h h . . T ~ 55. j ,_ r- , 12"+ )
$Rf .( 0 i i
I Figure 4.3 Rigid Support Rectangular Lug (Node Point 38) V il-17
1 i l 1 1 l l 1 i ? l \ l l Y L% 2 4 _ _ _ ' // " _ na
, *sv~5c , .s m -
5 I // V F-
' ( :
1
.. i i
W i ! I2"/ l i I i hmmy NIpc hr.ge, L. in e. l l f h /v - 7 l i' 1 1 l i I l Figure 4.4 Dumy Pipe at Snubber Locations l (Node points 26, 28, 38 and 44) l l V il-18 l
l 1 l l i i ! i l l - n
. 24,g
--4 ,
l
. - ... -o* *
- d' l e ssg '
e 3..
. iig- - -[-
l i s MGib- Q a. I _
- M
~
[f O/ ri L,i r I M'-) J ** T}'*Q. -surteen r
% is b* ol.
- l. * * ' ,b' R . -
J -7"**% w' ~ ~, N )
' 1
, $ .#,3 , [,./ #j',ma
- g g i i'
wt 9
==
g N N fe-sagsp'j '. v g *3 *e, y *e
' a* g e' *
.: %. .s
.g
.' -e'
.
- W. <
, . .3 -
.* *.. ,4 *.= .e,
?
! E Y. I, ! i3'(sto ,4 h
, Yk.
s ,m. - -. ---- $. , 7, % I
%l
- k *
[ >1_ ~ p s W- , j u ~b '
~
hai . l n .]as*:S
\
4m>Os i r ) ' P '*IM ei l'1 1 i 4 i 4 i Figure 4.5 RCS Hot leg Surge Nozzle Details (Reference 10) (IJode 7tcrt 3) i d 4 V(1-19 5
4 i j i i 4 i s b e u RS* 1 , ) go
- # 22 5 **
04E 4 as g 2 o- ~
,,J' l i l -
o MKDtTAttO,1-SD (
- l no i
nr e V \@ l
- Ih . *
- l i
, M-s,
[g ci,,, h_ w ,a
/
/
4 i p7 88 st e f- f- ,' h . --- ~
- . 2$, o
~ ^ - '
. r ,4
^
r
- - I " i, s w
.--_ ,4
\r i
< .. n ,., , , .. 1 - s ,r g v. g 1 m f. 3 y*
.a
! I- e. '- , . 7. 4 i a mmus l.ane ; 84
- c. *I '-as N. .'
-*=
j . enunurenn 1 M *.e, *g j g g 1
- Q- , e e y'"e e 113 3 \ d 2._, p. .. _
l q,,-s,.,.h,ee usj av Ag } ,
*1rs'$ / \
l r_ .; ,4, %i'5- 4 1 i i i l e j Figure 4.6 Pressurizer Surge Nozzle Details (Reference 11) YO.h C 05F 0 t 1 4 4 1 Vil-20 1 4 _ ~ -
4.2 Material Properties The material properties used in the analysis are: Young's modulus (E), Poisson's ratio (v), thermal conductivity (k), coefficient of thermal expansion (a) and heat capacity . These Appendix I of the ASME Code (Reference 6) properties were obtained from 4.3 Desian Transients and Past Operation System AT 4.3.1 Desian Transients Design basis transients for SONGS Units 2 and 3 are tabulated in Tables 4.1 through 4.5. These transients can be divided into the following: (a) Normal Transients Normal transients are listed in Table 4.1. In this table, the first column identifies each transient by an ID number ranging from la to
- 18. The second column gives a brief description of the transient and the fourth column gives the number of occurrences over the life of the plant. The third column gives the absolute maximum pressure during ;
the transient. The last column gives the temperature profile in terms of the reactor coolant system (RCS) temperature (Tacs),the pressurizer temperature (Tm), and the system temperature i differential ATsys, which is defined as the temperature difference between the pressurizer and the~ RCS. !
- It should be noted that some transients in Table 4.1 are classified as "No Stratification" transients. Other transients are considered to have a potential for thermal stratification in the surge line.
i 1 > (b) Upset Transients ! l Upset transients are listed in Table 4.2, which has the same arrangement as Table 4.1. These transients were included in the thermal stratification analysis of the surge line. (c) Test Conditions Table 4.3 lists the test transients for the life of the plant. These transients are divided into 10 hydro tests at 3125 psia pressure and 200 leak tests at the normal operating pressure of 2250 psia. Test conditions are included in the fatigue analysis. VII-21
(d) Operatina Basis Earthauake (OBE) Transients OBE seismic transients are listed in Table 4.4. These transients are ! included in the fatigue analysis of the surge line. t (e) Full Flow Water Sluo , Additional transients, defined as full water slug, are listed in Table 4.5. These transients are included in the fatigue analysis. ! i
)
Heat transfer coefficients during slug flow transient are given in Table 4.6 f (events 3, 4, 5, 6 and 7 in Table 4.1). . l l l l l l l VII-22
Table 4.1 Design Basis Transients - Normal , i i
! Press. No.of Rate Temperatures (*F)
Event Operation Occur- j (psia) rences. (* F/hr) T ucs Tm ATsys l la Steady state uniform 2250 2000000 0. 653 653 0(2) l lb Steady state strat. 2250 2000000 0. 621 653 32 i 2a Unload 2250 15000 -14.4 621- 653 32-593 60 2b Load 2250 .15000 14.4 593- 653 60-621 32 3a Turbing'IStep 2250 4120 0. 564 653 89 'l (inc.) , 3b Turbine Step 2250 4120 0. 564 653 89 ! (dec.)(') f 4a Turbing) Ramp 2250 17040 0. 573 653 80 i (inc.) { 4b Turbing) Ramp 2250 17040 0. 573 653 .80 .l (dec.) ! 5 Planned (2) 2250 9400 0. 573 653 80 i 6 Unpl anned(') 2250 200 D. 564 653 89 l 7 Below power (?) 2250 4580 0. 564 653 89 8a Heatup low (3) 410 75 100.0 100 440 340 l 8b Heatup high(') 2250 75 100.0 313 653 340 8c Heatup low (3) 410 375 100.0 190 440 250 8d Heatup low (3) 410 400 100.0 240 440 200 8e Heatup low (3) 410 500 100.0 290 440 '150 8f Heatup higb(') 2250 375 100.0 403 653 250 t 8g Heatup high(') 2250 400 -100.0 453 653 200 8h Heatup high(') 2250 500 100.0 503 653 150 9a Cooldown high(') 2250 75 -100.0 313 653 340 9b Cooldown low (') 410 75 -100.0 100 440 340 (1) No stratification. (2) See Table 4.6 for the heat transfer coefficients during these events. (3) AT,y, in the table represents maximum system AT. VII-23 l l
,m - ,e- . - = _ - - - , _ . - .-,....--,......-.,,......-e..,.m-. . , , - - , ~ . . . . - _ _ , , - - - - ,
I l Table 4.1 Design Basis Transients - Normal -- cont. Press. No.of Rate Temperatures (*F) t Event Operation Occur- ! (psia) rences. (*F/hr) Tacs T nz AT3,3 9c Cooldown higb(3) 2250 375 -100.0 403 653 250 9d Cooldown high(3) 2250 400 -100.0 453 653 200 9e Cooldown high(3) 2250 500 -100.0 503 653 150 9f Cooldown low (3) 410 375 -100.0 190 440 250 , 99 Cooldown low (3) 410 400 -100.0 240 440 200 9h Cooldown low (3) 410 500 -100.0 290 440 150 10a Heatup hot standby (3) 2250 87710 100.0 563 653 90 10b Cooldown hot 2250 87710 -100.0 563 653 90 standbyl3) 11 Start heatup(') 15 500 100.0 100 100 0(2) 12 Pressure low (3) 410 500 100.0 100 440 0(2) 13 Heatup no strat.(3) 410 500 100.0 313 653 0(2) 14 Pressure high(3) 2250 500 100.0 313 653 0(2) 15 Cooldown no strat.I3) 2250 500 -100.0 313 653 0(2) 16 Cooldown delta P(3) 410 500 -100.0 230 440 0(2) 17 Cooldown uniformI3) 15 500 -100.0 100 100 0(2) 18 Shutdown 15 500 0 70 70 O)l2 (1) No stratification. (2) See Table 4.6 for the heat transfer coefficients during these events. (3) AT,3 in the table represents maximum system AT. I VII-24
l Table 4.2 Design Basis Transients - Upset l i l Press. No.of Rate Temperatures (*F) l Event Operation Occur-l (psia) rences. (* F/hr) T acs T raz AT3,3 l 1 Heat removal 2250 70 0. 621 653 32 2 Decay heat 2250 95 0. 621 653 32 3 Decay RCS 2250- 30 -50/-44 621- 653- 32-0 ; 410 403 403 4 Reactor AnI2) 2000- 40 0, 564 653 89 2450 5 Inc RCS InI2) 2250 30 0. 621 653 32 6 Inc RCS OutI3) 2250 5 0. 621 653 32 II) Reactor An stands for Reactivity and Power Distribution Anomalies which is broken into 10 occurrences of uncontrolled CEA withdrawal from subcritical or low power,10 occurrences of uncontrolled CEA withdrawal at power and 20 occurrences of control rod misoperation, system malfunction, RCPS inadvertent operation or operator error. l I2) Inc RCS In stands for an Increase in Reactor Coolant System Inventory such as when there is a loss of component cooling water to the letdown heat exchanger or when there is a CVCS malfunction that increases RCS inventory. (3) Inc RCS Out stands for a Decrease in Reactor Coolant System Inventory such as during a sample line break or failure of other small lines that carry coolant outside containment. l Table 4.3 Design Basis Transient - Test Press. No.of Rate Temperatures (*F) Event Operation Occur-(psia) rences. ( F/hr) T ac3 T ray AT sys 1 Hydro 3125 10 100.0 70- 70- 0 400- 400-70 70 2 Leak 2250 200 100.0 70- 70- 0 400- 400-70 70 VII-25 l l
Table 4.4 Design Basis Transient - OBE Seismic Press. No.of Rate Temperatures (*F) Event Operation Occur-(psia) rences. ("F/hr) T ac3 Tpaz AT sys 1 OBE Seismic 2250 1440 0 621 653 Old (1) No stratification. Table 4.5 Design Basis Transient - Full Flow Water Slug Press. No.of Rate Temperatures (*F) Event Operation Occur-(psia) rences. (*F/hr) T acs T raz M 3,3 1 Upset 2250 100 Step 563 653 90 2 Heatup/cooldown 2250 500 Step 563 653 90 Table 4.6 Heat Transfer Coefficients for Slug Flow Events Heat Transfer Coefficient (BTU /hr ft?"F) l Top Bottom I 3a 4613 2 3b 2 4613 4a 4631 2 l 4b 2 4631 1 5 1391 2 l 6 2 11080 7 183 2 ., VII-26
1 4.3.2 Past Ooeration System AT
\
In addition to the design transients given in Section 4.3.1, a review of past i operation of Units 2 and 3 was made to identify any system AT in excess of the values included in the design transients. Table 4.7 gives the highest system AT during Unit-2 heatups since 1983 based on Reference 16. Similarly, Table i 4.8 gives the maximum heatup system AT for Unit-3. Extreme transients were included in the fatigue analysis in addition to the design transients. Table 4.7 Unit-2 Maximum System AT l PressurizerPregg. Outage Heatup and Temperature RC. Temp. System AT , Dates Dates (' F) ( F) P (psia) T (, I.) 11/13ld3- 12/13,14/ 2300 656 270 386(2) , 12/20/83 83 06/20/84- 07/24/34 2250 653 268 385I21 ! 07/?6/84 01/13/84- 02/10/84 2310 657 278 379R) 02/16/84 03/15/86- 05/31/86 2250 653 275 378R) 06/12/86 l 10/20/84- 04/03/85 2250 653 280 373(2) > 04/17/85 j Notes: (1) Saturation temperature corr'sponding to pressure (P). i (2) This value of system AT exceeds the value allowed by SCE Operating i Instruction," Plant Startup from Cold Shutdown to Hot Standby," which is 340*F (Reference 38). This limiting value of the system ! AT will not be exceeded in ti.e future during normal operation of j the plant (Reference 40). l f i VII-27 I i
t l i J i I 1 i ! f . Table 4.8 Unit-3 Maximum System AT 4 l. l j PressurizerPrep2s. j Outage Heatup and Temperature ) RCS Temp. System AT I Dates Dates (*F) (* F) P (psia) T (* F) 1 ! 01/16/88- 01/21/88 2270 654 300 354(') l 1 01/26/88 l
- 01/03/87- 03/01/87 2250 653 322 331 l 03/12/87 i
- d 04/15/92- 05/07/92 -
660 333 327 i
~
j 05/10/92
- 01/25/92- 03/21/92 -
660 334 326 d 03/21/92
. 06/11/84- 07/03/84 2270 654 330 324 l l 07/07/84 !
Notes: (1) Saturation temperature corresponding to pressure (P). (2) This value of system AT exceeds the value allowed by SCE Operating 4 Instruction," Plant Startup from Cold Shutdown to Hot Standby," i 4 which is 340*F (Reference 38). This limiting value of the system AT will not be exceeded in the future during normal operation of j the plant (Reference 40). i } i ! i ! i ! l 4 l l ! i
. f 1 !
3 ) ! a j VII-28 i ! i
i i 6 I 4.4 Pipe Wall Temperature Measurements Survey l l Heatup test data were collected at SONGS Unit-2 using surface mounted ' thermocouples (T/C). Figure 4.8 shows the locations of pipe surface temperature measurements (Reference 3). Two thermocouples were mounted at the top (0*) and bottom (180") of the surge line at location 1. Locations 2, 3, 4 > and 5 were fitted with bands of six thermocouples that were 60' apart starting i from the top (0*) . Location 6 had two thermocouples, one on the east and one on the west side of the pipe. Figure 4.9 shows the arrangement of the thermocouples on the pipe outside surface at the different locations. The ' thermocouples used were Omega Type K with Type K wiring. The wiring was run outside of containment to the plant computer, a GOULD 32/9780. All thermocouple data were recorded at two minute intervals by the plant computer. The following plant parameters were also collected at two-minute ; intervals: a) Cold leg temperature, l b) Surge lire temperature, , ! c) Pressurizer pres ,ure and level, d) Reactor coolant pump status. The plant computer produced a hard copy of the recorded data (Reference 17). i Figures 4.10 through 4.15 show typical plots of: a) Pipe top (0 ) temperature, b) Pipe bottom (180*) temperature, c) The difference between the top and bottom temperatures (delta T) at locations 1 through 5, respectively as shown in Figure 4.8. These plots cover the period from 8/7/1988 to 8/14/1988 during heatup. Figure 4.16 shows l the RCS cold leg temperature, the pressurizer temperature, and the system AT during the same period. The system AT during heatup is larger than cooldown. Conservatively, the analysis was based on heatup system AT for both heatup and cooldown transients. l l VII-29
j - , _ w l e _ rz cz ae
'rM e
u e ug c r. P,5 l _b w aL . 6
+
n-i s - t n e m e r u s s r a e
- M e
r u t a r e p m e T n j . j e w . w e p w e i a p - y f i 4s T s,i< h. s o e w s n >
+
/4 s o %
J i
' ' ,s t a
c s ss , s o
' L s 8 s xs 4
@ s
(, ' e r u s
@ i f
g e s ,uO t
- i < , J,l, .li j, iit, 3i.
T"P TOP *
'~ ash Side 10 ohoc} 2 o'cLOCW
--+ +
__ _p _ 5 7 DJ , 4 OClock OOCLock Bolk'" sdtom ' West Sde , Location 1 Locations 2, 3, 4 and 5 Location 6 I 4 4 i Figure 4.9 Thermocouple Installation Orientation on the Surge Line i
. _ _ . _ . . _ _ _ _ _ _ . _ _ _ _ _ - - - - - - - - - - - - - - ^ - - - - - - - - - - - - - - ~ - ~
,og rsou e/r to e/i4 __ ___.___ _ ._
m .._ _ _ _. _ _ _ _ _ _ v-- 3 s y ___ _ _ __ _ - - __ . . . _ _ _ _ - . . . _ . _. C \ g S
,oo _ __ .__ _.__ __.
(v
/4, ..
.p
~
J h . _ . . _ _ w e 3oo __ . _ _ . __._ __ __ __ ___ . . _ _ _ _ _ _ . __ . . _ . _. - . _. . i ggg ._ ._ - - __ _ - - _. --. _ _ - . - - - - - _ . . _. ._ U == =f iou _ _ _ . _ ___ __ .___ __ _ __ _ ____ _ _ _ ___ C ._. _ ..
/
O ~ ~ ~ ~ '~ - - - --
-100 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
0 20 40 60 80 100 120 140 I t>O Ilut OtHS ) l a ti ' lb el Dfi l A l Figure 4.10 Top and Bottom Temperatures and Top-to-Bottom AT at location (1)
. . . _ . _ . . ____...m. . _ _ . _ . _ . . - . . _ , _ _ _ . . _ . _ _ _ . _ _ . . _ - . . _ _ - - _ . . . _ . _ _ . - . _ _ _ _ _ . . _ . . _ _ . . _ _ _ _ _ _ . _ _ . . -
foo ruou s/r to e/i4
.A 4.
m _ __._ __ __ .__ ____ _ _ _ . _ _ ( g _ _ _._ _ _ _ _ _. _ e o s 400 -
--]
g j-
-}\ :- -- - -- --- - - -
4 :> ), r V x Iv x 5 r2 3oo h __ _ u 200 -- - - - -
/
.- M ioo ____ ._
_._ f _\ AC }_ . _ . _s j s ArJA%] \ O 20 40 60 80 100 120 140 I t>O IIML (IIRS ) Ioe not DO I A l
^
figure 4.11 Top and Bottom Temperatures and Top-to-Bottom AT at Location (2)
. . . . . . . _ . _ . _ _ _ _ _ _ _ . . _ . _ - > _ _ , _ . _ _ _ _ _ . . _ . _ _ . _ _ .____.____.,_._,_____.__m.._____._ _ . ___
700 .f ROM 8/710 8/14
.oo _ _ _ _ _ . ,
, FA .' J b N
^
u _ __ _ _ _ __ ___ _ _ _ _ . _
,f_. ~-_I_ _
t 0 e 4 _ _ _ _ 1_ _ ._ tf 9_ f *_ _ g ._. _ ( g y_ n (< W f y " 200 - g
, A l /
~ >
300 .._ _
- y,,- ' _
N e f'] g. -
-~t '
O O 20 40 60
] \ -. A 80 100 120 140 160 itut (Hets )
TOP HOI DELIA I t I t figure 4.12 Top and Bottom Temperatures and Top-to-Bottom AT at Location (3) _ . _ . . _ _ . . . - _ . . - _ . _ . . . . _ _ _ . . . _ . . _ . . ~ _ _ , _ . . . . . . . . . . . . . _ _ _ . _ . . . _ . . - _ . ,
i i } 4 l l i l
! l !
j v
- ' {ld -l l ! h e e
kIk ! l l l l
/
l !a\ i g ! l / , .s l i I ? \,J % { %fM l ( ,
! il 11 l I(l _
ls I l4 ! 4 i
<T K l d 3 A i
; l C \\ l lr . ;G a
,=
l 3 1 71
'>l
!f l)
I 7 2 Ig E I l l5l 5 i I dl (l ILI i E ! I I h kl b 8 l 1 I I im m '* - O O
~
t ! I I l \\ j : l i
=
e ! l 4 I k h k 8
. k 8
~ 8 7
L. j U C30) 3knivw3en31 E k vil- 3 5 1 s . . _ - _ - - , , _ _, . _ _ _ , _ . _ _ _ _ _ . - _ .
700 FNOW 8/7 IO 8/14 s l .__ 600 p fy-W , n -- - - - - - - - - - - g _ _ _ _ k _ _ _ -- t - 1 _ w o ,oo __ .__ ._ _ __ _ _ __ __ _ _ _ _. _. w y 3 A) -s - -- y ^ 6 p E y T U $ 200 d
~
Gl l ) } f\r t 3
\ "
,00 _ _
s N O - - - - " - - 0 20 40 60 80 100 120 140 160 ilWE (HRS ) IOP BOI DEllA i Figure 4.17 Top and Bottom Temperatures and Top-to-Bottom AT at Location (5)
-__._..__m _ . . _ _ _ . _ _ _ _ . _ _ . . _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ - - _ _ . _ ___ _____ __ ..- - m _ - , - _ . , - . .i--_ e- -
700 FROM 8/7 TO 8/I4 m - - y) ooo _ C 500 -- - - - - - . - - w 8 - - - g 400 - - - - - - - _ _._
~ m g g ::: x.C ~ . _ . _-.
_ r ---- (,J w 300 - 4 % 200 i= A , 100 0 20 40 60 1 80 100 120 140 160 isMt (HOURS SHI M S C010 i f G 3 SYSIE M Di t lA 1 Figure 4.15 Cold Leg and Pressurizer Temperatures and System AT During Plant Heatup (from 8/7 to 8/14/1988, SONGS Unit-2) f
4 I I 4.5 Pressurizer Liauid Sample Line The pressurizer liquid sample line was evaluated for stratified flow , conditions in the pressurizer surge line. All loading of this line are the i same as in the previous evaluation (original design report) except for the thermal movement at the branch point. The branch point thermal movements were calculated under thermal stratification conditions in the surne line. , s l l 1 l l VII-38
~1 l
5 METHODOLOGY l l 5.1 Introduction I i Thermal stratification occurs when the temperature of a fluid, inside a i horizontal pipe, varies from top to bottom resulting in temperature gradients in the pipe wall. This phenomenon often exists in the pressurizer surge line 1 for the following reasons. . l (a) The hotter pressurizer water has lower density than the colder RCS hot i leg water. Thus, during insurges and outsurges, the hotter fluid tends to flow on top of the colder water, (b) The flow rates during insurges and outsurges are, in most cases, i relatively small. Small flow rates can result in thermal ! stratification (Reference 39). l Test data, recorded during plant heatup (Reference 17), confirm the occurrence l of thermal stratification in the pressurizer surge line, and show that the top l of the horizontal section of the pipe is at considerably higher temperature j than the bottom. 5.2 Structural Analysis of the Pressurizer Surge line Structurai analysis of the pressurizer surge line at SONGS 2 and 3 was performed. The finite element program ANSYS was used to calculate the resulting thermal stresses in the pipe, the rigid support loads, the pipe displacement, and the nozzle loads. 5.2.1 Calculation of the Temperature Profile in the Pipe Stratified flow in the pipe generates temperature gradients in the pipe wall that change nonlinearly in the vertical direction, as shown by the test results (References 3 and 17) described in Section 4.4. The actual temperature profile was based on the test data collected at SONGS Unit 3 in 1988 using surface mounted thermocouples at different locations of l the pressurizer surge line, as explained in Section 4.4 of this summary report. The ratio of the system AT to the actual pipe wall AT was calculated at maximum top-to-bottom AT from the test data. This ratio was assumed constant for all values of ATsrs (see Section 3 of this summary report). Only linear temperature profiles can be modeled using ANSYS. The ratio between the linear temperature profile and the equivalent nonlinear pipe wall l VII-39
l l l temperature was calculated equal to 0.95. This ratio was based on equivalent ' pipe bending moment between two models: the first is a three-dimensional model with nonlinear temperature distribution, and the second is a straight pipe , model with linear temperature distribution. : Both ratios calculated above were applied to the system AT to obtain the equivalent ATun,,,. This AT y n,,,. was used in the ANSYS model. 1 l i i l l l 5.2.2 Thermal Stratification Analysis An ANSYS model of the pressurizer surge line was generated using element type STIF20 for straight pipe segments, and STIF60 for curved pipe segments. The ! top-to-bottom temperature differential (AT was obtained from the system l AT by applying the two ratios explained in N ,s)ection 5.2.1 above. 3 The system l for Units 2 and 3. Only l AT the was based section horizontal on the maximum of the pipe past wasoperation AT, *d stratified. considere ) The ANSYS finite element model of the pressurizer surge line is based on the geometry and support configuration shown in Figure 4.2. Figure 5.1 shows a computer plot of the finite element model including the ANSYS node numbers. A sample of the results obtained is shown in Figure 5.2, which shows the displaced configuration of the line due to thermal stratification. j Results of the surge line analysis were obtained in the form of nodal forces and moments, reaction force at the rigid support, and displacements at all locations of snubbers, spring hangers, and branch connections. Calculated forces and moments were used to calculate ASME Code stresses and demonstrate compliance with the Code requirements. Rigid support load was compared with
. design load, and displacements at the locations of snubbers and spring hangers were compared with allowable ranges.
5.2.3 Response of the Surae Line to Thermal and Mechanical Unit loads l l To simplify the analysis, unit loads based on the design loading conditions l including thermal expansion, thermal stratification and nozzle movements were ; i applied independently. Results from the unit loads were combined, after i applying appropriate scaling factors, using the principle of superposition to calculate the results for the actual loads. In the structural analysis of the pipe, internal pressure was applied, and the elbow stiffness includes the , effect of pressure in accordance with ASME Code Subsection NB-3650. VII-40
?r e nun'zer 69 Cille kh 66 65 je '13 S.3 f !
2 b7 A Q2
/
~ N 86
.1
\g3 yy Nob Le<j - d5 d4 N24 4 N .N2 5 43',.4t'49 5 x 42' Q6 Q7 40 *'Al' q8 39.
N19 .w'p.
% R;3g oppor b Figure 5.i Computer Plot of the AN3ys rinite Element Model
l Press uriZM Side i 1 1 i l kk ., t si l i N
,s
.Y
/
I
' .s . ,
D
' ~ ~ .. .
/
~~..'
' .,' ,f
, /
~,
e Figure 5.2 Computer Plot of the Stratified Surge Line Showing the Bowing Effect l Vil-42
l l The following unit loads were applied to the finite element model of the pressurizer surge line described in Section 5.2: (a) RCS Hot Leo Surce Nozzle Movement Anchor displacements applied at the hot leg end of the pressurizer surge line. These displacements represent the movement of the , RCS hot leg surge nozzle due to thermal expansion. l
\
(b) Pressurizer Surce Nozzle Movement l Anchor displacements applied at the pressurizer end. These ! displacements represent the movement of the pressurizer surge nozzle i due to thermal expansion. l (c) Normal Thermal Expansion l Uniform temperature rise of 10 F, representing thermal expansion without stratification. l (d) Thermal Stratification Top-to-bottom stratification of 10 F in the horizontal run of surge line, representing thermal stratification conditions. ] The unit load cases described above are shown schematically in Figures 5.3 and l 5.4. 5.2.4 Response of the Surce Nozzles to Unit Loads The nozzles are connected to the vertical piping section and do not experience thermal stratification. However, they are impacted by the bending moments and forces induced by stratification in the horizontal sections of the pipe. Unit thermal and mechanical loads were applied independently, and the results-were multiplied by appropriate factors and combined to calculate the response to actual loads, based on the principle of superposition. The models were generated using ANSYS. These models are shown in Figures 5.5 and 5.8 for the RCS hot leg surge nozzle and the pressurizer surge nozzle, respectively. ANSYS thermal analysis two-dimensional element type STIF55, and the structural element type STIF42 were used to generate the models of the surge nozzles. Structural element type STIF25 was used when asymmetric loading was applied. Vll-43
AVn PWz AU wzzag AWg N C i ss , 4 s sh T=70 F
?% day aw s T=70 F gy AT=0*F T=O*F RCS Hot leg Surge Nozzle Displacement Pressurizer Surge Nozzle Displacement i
9 Figure 5.3 Unit Load Cases for Fatigue Analysis - RCS and Pressurizer Surge i Nozzle Displacements i
Pn1 PR2 unswo muzun Y Hab
$ ] N'*3 4 .4 a T.8(fF T7 F AT=0 AT-lO F Unifonn Pipe Wall Temperature of 80'F Thermal Gradient in the Pipe Wall (top-to-bottom AT-10*F)
(10*F above reference temperature) b J i Figure 5.4 Unit load Cases for Fatigue Analysis - Uniform and Stratified Flow Conditions i
t The unit load cases comprise the following: i (a) Unit Thermal Shock Thermal transient analysis was performed to calculate the temper'ture time history due to a thermal shock on the inside surfact of the pipe. Conservatively, the outside surface of the model was assumed to be adiabatic (perfectly insulated) with no heat transfer to the surroundings. Results of this computer run are stored on a data file generated by the thermal transient analysis. A second run uses the data stored to calculate the time history of r membrane, bending, peak, and total stresses due to the thermal gradients generated by the thermal shock. Sample results are shown ' in Figure 5.5, which shows the temperature time history at the inside and outside nodes of section 1330-01284 (see Figure 5.5). ' The corresponding stress components are shown in Figure 5.7. (b) Internal Pressure Each two-dimensional axisymmetric model was subjected to internal !' pressure of 2250 psi (operating pressure) on the inside surface of the pipe. Aipropriate end load was applied on the end of the model representing the cap load. l (c) Unit Force Perpendicular to Pipe Axis A unit shearing force was applied at the pipe side of the model ; perpendicular to the pipe axis. ! (d) Unit Force in the Axial Direction of the Pipe A unit force was applied on the side of the pipe in the direction of l the pipe axis. l (e) Unit Bendina Moment I A unit moment was applied at the pipe side of the nozzle model ! perpendicular to the nozzle axis . This moment has a bending effect on the nozzle. ; (f) Unit Twistina Moment A unit moment was applied at the pipe side of the nozzle model in the direction of the nozzle axis. This moment has a twisting (torsional) effect on the nozzle. VII-46
- i A similar analysis was performed on the pressurizer surge nozzle to calculate its unit load case, which are similar to the unit load cases of the RCS hot leg surge nozzle. Figure 5.8 shows a computer plot of the finite element model used for the pressurizer surge nozzle. This model includes the pressurizer head, the nozzle safe end and part of the surge line piping. The nozzle geometry is based on Figure 4.6. Sample results of the thermal analysis are shown in Figure 5.9, which shows the temperature time history response of the inside and outside nodes of section 1422-1419 (see Figure 5.8) to a thermal shock. The corresponding stress components are shown in Figure ,
i 5.10. As in the case of the surge line, results of the unit load cases described above were combined by superposition in the fatigue analysis of the surge nozzle. A more detailed description of the fatigue analysis can be found in Section 5.3. ; i d 4 a VII-47
El
/ PIPS
==
M _ _Sec bien 13 3 o -129 4
- _9ee kan 1209-1105
%Ce Gnd
_ joe kon 327-8073 __ _ Seckten lo 5 -930 _ _ Secb*on 63-916
- Mozzle Bo dy
= ; ' , ,n',
=: g4' 3
=tm- .
g water Gap % e p' 4 q Adiabahic %ndary Sleeve - -; ., s s 3 ..._ s+I.de 'CrtrMMb - Y Of y , i ; L I,6
.u.a m w nt w H.E Leq Z ,X l;'C "' ' nn: t--+
J
/
Figure 5.5 Finite Element Model of the Hot leg Surge Nozzle Vi t - 48
I e so 11 - 1 e. 11 l - co t%
^
T il E o n tm
- e9 w -
is N
~ 8 8 l l f :v
- ~9 e l IE 8 11 l l
- s .
e 1 I erv So + - o o l l ] ~5 5 a, l l . Yo E a ; ct _\ l l
- 1. e
-- y u i ; ,
a s l3 l 0
- l - ; l
'? z c l ll\ l I ! I a l
< =
l\s- l _ l i I I e l l l n
'Mj l
- j =
j L l I O O
' E
_ _ _ . _ :- C - o s
] ! 1 .T l ) l ' - ~
m e n. c r. w #. n - o o o o o o o o o *o (sovosnow2) - (J 030) 36nive3eW3. l Vii- 49
_ _., - ---- - - - ~ l I l I e t I j - s
.j l
. ,. i 1
. 5 $ wb"
, j 2
. ,. .i .
.. . i l 4 l
i / //! .' I 8
- l
-( i y s
t l < is ' E E 1/ i 2
-i l V -
; ~
I ,/ s
< ~ ,
I g i
, a ;
e m
, 4 i )$ g m
!g!
d
.,. i lf
] i i !
! l;- *
'x c
# fF i % ; C O 4
i 6 ! l
. a I //
f f i e' gl
) - u O g
l i / ! // /
;a 8!
e I
,, M g -
i i ! j - i
,,/ sf
- .: e i I,
/ ,
?
l o i i f / I j!
, < x i 1o u .
l4 . # ^ 4 -. g O
. . [-
,1 w 4 A
/.
j . I f =
, 4
,! I o 9
s
- h. ,.y . &
w - i .
- : : e
- i
! Si :
cs y ,
*l 2 ,/ / / ! ! l -
=
,' ' / s 3 e-6
~ - I .
le 3 e 4; I ;/ fi
-! i
.w :
+
a g
~!
i I/ , a i 4 { fj
=
i
, /
=
u - !i
/ .
i
< ,e >
- , 3. s m
[ i
/ i t'
.O
+ g
/ e , . ,
%(\ e 1 b
b l g M i I 7 , i l 2
)
i [ 4 g _+ i
- w 1 N.
i i I f ,
!' O w -l .
1 -
- w w
,, y v 1 j i %
j 3 1
. .,' I
[ e h E 9' x' 9
*' 2 9 O
-. r. _ **
.l J rse esa.ogo O5d) 5 53=.5 4 Vil- 50 e .
-w. f-- t- w-;--- p-., e, w % - ep -- v
-t y=. p--s, y r--
I i J 1 Pipe $ . -Se e bcon is'e i - l 'f 4 'f i 9,c k ,o i422-14t3 { se Se En b
- seeb.,n l 248 1180 i
tso Ie 80 by
.fgx
/
Wder 6ajo f
, T hR: -
/
'D +
' % rurix e r Head i
% rhern,aI \ s 9Ieeve Q '
~,
*p.
\
#" fi\
w\.
'S Y a\
. \r, . . . .\ Ihbk u nu
)
4 1 Figure 5.8 Finite Element Model of the Pressurizer Surge Nozzle Vll - 51
ll 2 si
~
II g l 3 3 9 E C
== N y .3 g $
M ll _2 8 o l Ta ! c l
!$ g
- l eo b u * .$
f l - l l l l 05 [" e4 l l e ta
*N '-
3
!!\
i ,i l I i l i i ' s2 ice l l l } l
- if l l l l Ml l E = l' a ?
l l l- Nw : N o e 8, t [ R *
? @ h C D w m n . o
- C o o o o o o o .O l
(se esac.1) (J 030) 3e nai=3en31 - VII- 5 2 i I
l l l l i l 1
-2 l i
I m 1 [ , i I e I '
. i l l ,
; 7 N
! 3 I #
U
! f, -
/- } ? t 5
) i j i l i
- o I '
\ l ,
I l
) .
v i l. !
' a i ,
i . }lI-i . e g-
- e o
I i j . ' g 1
.t- x u
\
o s f & il 9 4 "e i / , i
,/ l j _s i
- I r } i 4* g EZ Rl /I , [i {j 3 5
- i i
z *
/ :
k I'
/ I ,i s; i i il _~ 5 s t
t E u l l 1 \ /,e \ l \, o 3 a o A / ' N &
- I/
\yf i
l1i
- O e:
L
/ 3 - -
a e t i e
~
m l , i a
'/ \ 3
\ T. -
N 8 s A _ s
- , .c ,-
- I t
2, k 3
~ 8
~ 8_ 8 S o (sowesmowi) ,
U5c 553eAS Vil-53
5.2.5 Intearal Attachment Analysis Snubbers are connected to the surge line by means of dummy pipes. Similarly, , the rigid support load is transmitted by means of rectangular lugs, as shown in Section 4 of this attachment. Unit thermal and mechanical loads were applied independently, and the results were multiplied by appropriate factors and combined to calculate the response to actual loads, based on the principle of superposition. The three-dimensional models were generated using ANSYS. Three-dimensional models of the rectangular lug and the dummy pipe were used due to the absence of axial symmetry and to capture the details of the welds. ANSYS thermal analysis three-dimensional element type STIF70, and the structural element type STIF45 were used to generate the models of the integral attachments. Figures 5.11 , and 5.12 show a computer plot of the dummy pipe model, and figures 5.14 and ' 5.15 show a sample of the thermal analysis results obtained for the transient shown in Figure 5.13. Similarly, Figure 5.16 shows a computer plot of the rectangular lug model, and Figures 5.17 and 5.18 show a sample of the thermal analysis results obtained for the a thermal shock transient. t Critical sections were selected for stress calculations in each model based on material and geometric discontinuities. The following unit thermal and mechanical loads were considered: (a) Unit Thermal Shock Thermal transient analysis was performed to calculate the temperature time history due to a thermal shock on the inside surface of the pipe. Conservatively, the outside surface of the model was assumed to be adiabatic (perfectly insulated) with no heat . transfer to the surroundings. Results of this computer run are stored on a data file generated by the thermal transient analysis. A second run uses the data stored on FILE to calculate the time history of membrane, bending, peak, and total stresses due to the thermal gradients generated by the thermal shock. l (b) Internal Pressure Each three-dimensional (solid) model was subjected to internal pressure (on the inside surface of the pipe). Appropriate end load was applied on the end of the model representing the cap load. VII-54
l t i i i (c) Unit Force Perpendicular to Pipe Axis i' A unit force was applied on the integral attachment, rigid support lug or snubber dummy stub, representing the actual operating l loads. (d) Unit Force in the Axial Direction of the Pir,e i A unit force was applied on the end of the pipe in the direction of the pipe axis.
- Figures 5.11 and 5.12 show two computer plots of the three-dimensional finite l element model of the dummy pipe. Figure 5.11 also shows the critical ,
! sections, representing geometric discontinuities, selected for stress I calculations. Figure 5.13 shows how the thermal- shcck was applied to the l dummy stub model. Sample results of the thermal analysis are shown in Figure ! 5.14, which shows the temperature time history response of some inside nodes in the hot layer and cold layer to a thermal shock. Stress components acting on section 5-1883 (see Figure 5.11) are shown in Figure 5.15. , l l Similarly, Figure 5.16 shows a computer plot of the three-dimensional model . used to generate the base load cases for the rigid support rectangular lug. t l The figure also shows the critical sections selected for the calculation of l stress components. Sample results of the thermal analysis are shown in Figure j 5.17, which shows the temperature time history response of the inside and outside nodes of section 717-362 (see Figure 5.16) to a thermal shock. Figure 5.18 shows the stress components time history acting on section 719-403. l l I l Vll-55
l d I
- %~-
t an *' srAo , e if $ piP e -
*M I =Q T%nbe g]
'I>e rny S M kf f % tail A W: %. . ,
. ~,.,, a r .
1y 'g; \ , s
,cQf -
s s s' s I. --
-- -,/o q %xm,Y , ,
,.\
--Z .
4 s ,
. ~,. .,
Y. . . ' "' ,/\- '
't
, h -@' -
g\ l' ,
- r. i y : $N'Mr '.
' r N- '
f.+ ,
' ( 4W'ud ir\ i
'N(Nr-t'.;-(?
\ Md s
$' %.:eg)? s ;
t??.$c \/ Wi l l _ vy A / j ; s ' ' ', - 1 M' / - f i j' ! 9ee behil A , R- , s,/ )
,w \
.. \Y l W
\.iVYp'/p V?
i P,'pe.
# 1 l Q % v"*. '
Figure 5.11 Finite Element Model of the Dunny Pipe Showin Selected for Stress Calculations (Front View)g Sections i V il -~5 6
4 4 l l
}
I I 4 J i [ 4
. c.'.. '
,s"
- : 1
..+. -::hk . ~ .- .?
gg
%W.<~.~4..n.x+.
' '~
T ' , % ,,, l t s 7 )
%,~ . = , -- . s9 .,.9., .a l
F.% ~.-1 k, %A .Wn. - s,p.W*,, . j/ L. . . .- Weld t y% _.s # .< g - g %. ' f, 'r W: 4 Nr- .-. _ vs yip
- e .,.m 0n ,s , ' '-
C'a.
*li M' . . . .
>k<,N*.
f Y.^ j Q.g* .
;' '< /.'~ - %,
s > N' " I
*i.,*,,.<4,' ,#,- g
'n[h '
gSh5.k Nci ,. i 2 % ?-4 Y E % %. d g'g' ,?'sq'.i'A\ ' Q2'5'[h ,Ys 9 lf IL 'h' ' ti f'}, d \ Mt ' I,. ./ ' :
.-{, ' ~ ('v .
,0 3', [.3 % -lii ?~ ' V\
, ' }1
5 l-l.y yv .
, f, h y/l /.j I .' l h -
; y.. A, ' /
- n s; f , !
' i sj ,.Aj'
,lj RW kM Q
I l: i l- 'l !,!i N. f& u+j . l N ; }l % 'l-~ l I
! ~~
l
/ N r-j l % ..r**
u,. Figure 5 12 Fintte Element Mode} of the Dusesy ptp, (Rearyiew) I Vil-5 7
Iipe Ajdskelic
%mn dery h=5096
/ Mob Ar [1 1 F 10 o
< 7 '7"y
/
\ NN\x
/,,_
3 CD
}\ / Co hl {
// y er-l
~ h = G00 8T0l p,,,e l / n kr fl' *F
%mmmy5%)o Therma / Fansienf 1
i Figure 5.13 Unit Thermal Shock (100*F) at Dummy Pipe Location i
,Il; I. 1 Iil
_. n a 0 / 0 s 4
- a d n
- w a U -
- . g n e o O
, +- .
w i t
- c a e 7 0 S
r _ - , 0 e 3
- y 2 a
- 7 l
/-
s 2 t l o
- e H
- W a n
- t
) n o
- S o 0
0 ND w r y a _2 O o
/ I c
L t sn
- / (
t s e s io Hi ec t
- I t
e me
/ _
n iS T _ i r ee
~
/ o o
N. o ry ua 2 v tL g- .i 8 a w rd w
/ w el po
_ - /
- mC e
y Ta g
=
5
^ 4 1
- ,. ' n n ' - 5 o
l-
, \
n '
- i
_ e _ n. i r e u g W 0 w o o o o 0 0 0 o e. i l 0 1 s t s -w n 3 2 1 4 F n
. o C aw w t ?, 1M i
<_ TUe l ;
1 l I I i I I e i 1 g r ' s j = 5-5o
, l
- i , -
. ,1 '
L o i : l/l ' i i I/ li ,
,2 n I
i
! !/
// ll !- .
< z
/
l ; : il li I=N .- i i , l 1 !
/o + jI ,
'/
i - g 8 k.
/ i.
I i i : // /I i g .
.. . s 1- 1 i -
/ : il /l -
34 5 I i l
- -e .
i
++
il j7l i y . e
$3= g i I
/
l
./ /
I o= ~,
.I
'2 e
j i ll l /! !/ ! 11 ,
, , , o - 58 o-i I i / -
I ll / l 11 (a l I/ , ill /l i i- c i %.E I / 7 -
! s!/ 2/ l +
w I !" ! EE a i, , y I
=
m
/ / 'IL .^ i
_,.O
=
e
. j 2 3 I
En i .l l, _ E D D @ D 3 D M N N - - C a S c i.*C 5 f" O wl)
.Sc) 553eA5 Vil-60
se.c . 69 s cff6 gg,ygg 3gy Lw3 g ((hh ,k Weld g's $f5bbE?" Y h
,k ?
M)[-Q%.&,-Ys
'*^,' /
.It o'3 l <.~~~' \ y \ ^ ~ 719
\.-
r,
,,. g s'
s '\ s
\
,L E; .. '.l '.
~ - ' ..
\ Es , i - '
s [Wh Tg / I
- .7 V ',tx ' s' N, -b '\ -
s 4 , 9 ,;. b 3 6 2 '.
, ,,j-(T, ',; A,,
g - e . . .
. \ \
l
\
\\
, .' , \, \
). 'g veds. Po"Y ' '
s '\
\' .\ \. -\ h. / /
\ '\ ; '?
\.
l
\\
/ \ },. ' P(pe 4 %d Figure 5.16 Computer Plot of the Finite Element Model of the Rectangular Lug Showing Sections Selected for Stress Calculations Vil- 61
iio _ _ _ _ . _ _ _ - . _ . _ __. , _ 100 - - - - - -
,, ==
2 - n- -
., .g 90 -
4 -' - p Y
/ .
l 1 so / f - Q o w 70 - ---l--
-- J _'--. .-- ---. .
o.- J -
/
'y s,n -- - - - -. -- -- ._ . - -
i
- e. -
(* ,,o _i L_. _ __. ___ ___ .-_ _. w i i y $ an .. - . _ _ _ _ ._ cn F " N 30 , , 4 20' -i :-- - -- 10 -, , O - 0 02 04 06 08 1 1.2 1.4 1.6 1.8 (Thousands) j TIME (SECONDS) n $N%8DI t*X)E (717) - + OUTSaDE NODE (362) 1 i Figure 5.17 Temperature Time History on Section 717-362
33 __ 3o q _ _ _ . . _ . . _ _ _ _ _ _ __ _ _ . . . _ _ _ _ _ .. . _ _ 23 g _ x x ___ _ _. _ . _ . .. . N, ..
,.._. t ,
N* ~ , p ;; 20 ~~
'N -A.
.. g , %
- m. '
A. % m3 I IJ O 'N
=
- iS -. s- --A i
.A - -
A - -- - - ~' ~m y_. _ __ 10 --
',[ --
'Ns '-
~ -
3 _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ p o u -u- u---u - u - - --- -u - -- -- u - - ---- si j o 20 40 i.o seus c.1(ONOS) i ui u a lie n us u e ist N **"A* 11 At lesiAl Figure 5.18 Stress Components' Tine History in the Full Penetration Weld (Section 719-403) a. _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ . _ _ _ _ _m _ - . _ _ . _ _ -, -__ . . _ _ ____.___ _.--_ _ _ . . _ _ , _ _ -
5.3 Surae Line Evaluation \Fatique Analysis Evaluation of the pressurizer surge line, including fatigue analysis, was performed in response to the requirements of Bulletin 88-11, Item 1-d. This i evaluation also includes the pressurizer surge nozzle, the RCS hot leg surge nozzle and integral attachments. It is recommended by Bulletin 88-11, the analysis was performed per a recent ASME Code edition. Accordingly, the 1986 Code Edition was used throughout the fatigue analysis. In addition to the ASME Code, Code Cases N-122-1 (for the rigid support rectangular lug) and N-391-1 (for the dummy pipe) were used to evaluate the surge line at the locations of the integral attachments. A summary of the methodology of this evaluation is presented in this section , as follows: (a) Applicable ASME Code Analysis Criteria (Section 5.3.1) A summary of the applicable ASME Code design equations is included in Section 5.3.1. (b) Pipino ASME Code Evaluation (Section 5.3.2) 1 The methodology for ASME Code evaluation of the surge line piping is explained in Section 5.3.2. (c) Nozzle Evaluation (Section 5.3.3) Section 5.3.3 includes the methodology and the criteria for the evaluation of the RCS hot leg surge nozzle and the pressurizer surge j nozzle. (d) Ricid Support Lua (Section 5.3.4(a)) The analysis methodology of the rigid support rectangular lug is explained in Section 5.3.4(a). Both the ASME Code and Code Case .' N-122-1 were used in the analysis. (e) Dummy Pipe Evaluation (Section 5.3.4(b)) ! The analysis methodology of the dummy pipe welded to the surge line (used as part of the snubber support assembly) is explained in Section 5.3.4(b). Both the ASME Code and Code Case N-391-1 were used in the analysis. (f) Thermal Ratchet Check (Section 5.3.5) Thermal ratchet check is required by the Code when the primary plus secondary stress intensity allowable is exceeded. The methodology of this check is explained in Section 5.3.5. (g) Thermal Stripina Evaluation (Section 5.3.6) Thermal striping analysis methodology is outlined in Section 5.3.6. l 1 l VII-64 l
i i l
~
l 5.3.1 Applicable ASME Code Analysis Criteria i The following ASME Code vessel and piping rules were evaluated for the effects : of all specified loads, including thermal stratification, on nozzles and integral attachments: NB-3200 - Design by Analysis (used for nozzles) , NB-3221 (Design Conditions) Thermal stratification does not impact design conditions. NB-3222 Level A Service Limits NB-3222.1 Primary membrane and bending stress limits (not impacted by ; thermal stratification). i NB-3222.2, 3, 4 Primary plus secondary, and primary plus secondary plus peak ! stress intensity. These conditions are impacted by thermal , stratification. l Pt+ P3 + Q s 3 S, ! U (usage factor) 5 1.0 (includes Level B and Test conditions) If the above limit is not satisfied, then K, should be calculated for fatigue evaluation per NB-3228.5 based on the value of (P + t P, + Q) shown above. Also, apply the rules of NB-3228.5 below, NB-3228.5 Pt+ P3+ Qs3S, excluding thermal bending stress. ! U s 1.0 (includes Test conditions). ; 9 NB-3222.5 Thermal stress ratchet check. This condition is impacted by
- thermal stratification.
NB-3223 (Level B service limits), NB-3224 (Level C service limits), NB-3225 (Level D service limits) and NB-3226 (Test limits) are not impacted by thermal < stratification, and the original analysis is valid. It should be noted that Test condition has been included in the fatigue evaluation. ' NB-3600 - Piping Desian i NB-3652 (Design Conditions) Thermal stratification does not impact atsign conditions. NB-3653 Level A Service Limits NB-3653.1 Primary plus secondary stress intensity range limit (Equation 10). This is impacted by thermal stratification. If Equation (10) limit is exceeded, then VII-65
. _ _ _ , . . _ l
4 i i i I I a l NB-3653.6 sis: ! Simplifiedelastic-plasticanaly/U Equation (12) stress s 3 S ; Equation (13) stress s 3 S, 3 K, should be calculated for fatigue evaluation. NB-3653.7 Thermal stress ratchet check. , l NB-3653.2, 3, 4, 5 i Primary plus secondary plus peak stress intensity range. These l conditions are impacted by thermal stratification. ; i U s 1.0 (includes Level B and Test conditions). i NB-3654 (Level B service limits), NB-3655 (Level C service limits), NB-3656 (Level D service limits) and NB-3657 (Test limits) are not impacted by thermal , stratification, and the original analysis is valid. It should be noted that l Test condition has been included in the fatigue evaluation. ] l Load Combinations l Condition Local Combination Design Pressure, Deadweight , I Level A Pressure, Deadweight j (primary stress only) j Level B Pressure, Deadweight, OBE (primary stress only) Level A and B Pressure, Deadweight, OBE, (primary + secondary) Thermal including stratification Level C Emergency conditions not specified (primary stress only) Level D Pressure, Deadweight, SSE, LOCA l ASME Code stress and fatigue evaluations were performed at 18 piping locations, 5 welded attachment locations, and 2 nozzle locations (see Section 2 for a summary of the results). Note (1): Locations where Equation (12) limit is exceeded are enveloped by the shakedown analysis in Reference 6 (see Section VI, page VI-1), since the calculated stress in this analysis is less than the stress calculated in that reference. VII-66
5.3.2 Piping ASME Code Evaluation j Surge line piping evaluation was performed per the rules of Subsection NB-3650 of the ASME Code. The following Code equations were included in the evaluation:
- a. Code Equation 10 stress (S,) is given by For non-0BE transients:
4 S ,= C1 +C Mj + C3 E,3la,T,-a3 T3] s 3 S, For OBE transient 4 PDv Sn =C2 M + [A3 +C Eadln,T,-a3 T3 l s 3 S,, 2t o+C 2 [D i 3 j [A] = maximum of (2C2 Most D/21) or (C (M i 2 + Most) D/2I) i
- b. Code Equation 12 stress (S,) is given by:
D 3, = s 3 S, i C fM* 2 i i See also note (1) on page VII-66.
- c. Code Equation 13 stress is given by:
For non-0BE transient pairs l D M"D+ C(E,3ln,T,-a3 T3 l s 3 S, l l C1 P o f +C2 l For OBE transients CP 3 o +C2 ##
"D+ C[E,3la,T,-a3T3 l s 3 S, where M ost = resultant OBE moment, in.-lb VIl-67 '
2 i i Mo w = maximum deadweight moment, in.-lb P, = pressure range for the transient pair, psi { C1 and C, = secondary stress indices for the- specific component ! under investigation (NB-3680) l
\
= nominal outside diameter of pipe, in. (NB-3683.1) 4 Do 4 t = nominal wall thickness of pipe, in. (NB-3683.1)
D = diameter of interest (either inside or outside _ diameter), in. 1 = moment of inertia of pipe, in." e, (a3 ) = coefficient of thermal expansion on side a(b) of a gross-structural discontinuity, at room temperature, 1/*F (NB-3653.1) T, (T3 ) = range of average temperature on side a(b) of gross j structural discontinuity or material discontinuity, 'F. For generally cylindrical shapes, the averaging of.T (NB-3652.3) shall be over' a distance of Nd,t, for T, and Nd 33t for T 3 E,3 = average modulus of elasticity of the two sides of'a a gross structural discontinuity or material disccatinuity at room temperature, psi ASME (Appendix I Table I-6.0)- C'3
=
valus in Table NB-3681(a)-1 _ M, = maximum thermal bending moment calculated as follows: T"-7 0 1 T"-7 0 Mj = 653-7 0 (Case momen t) + 6 50-70 (Case 2 moment) + 0.5(T y m+ T ) -7 0 (Case 3 momen t) + 80-70 0.95 (Case 4 momemt) 10 where Traz and Tut = pressurizer temperature and the hot let temperature, respectively. Also, Case 1 through Case 4 refer to the four base load cases described in Section 5.2 of this summary report. _ A factor of 0.95 appears in the last term represents the ratio between the maximum stratification temperature in the pipe and the system temperature difference (AT3rs = Traz - Tut) . VII-68
l l i
- d. Primary + secondary + peak stress The primary plus secondary plus peak stress is given by l
KC 1 1 + [ C] +K3 0,,, ( ) ,,x The [C] term appears in the transients with axisymmetric thermal shock, namely, transients number 3a, 3b, 4a, 4b, 5, 6, 7, 27 and 28 (see Section i 4.3 of this report). It is given by the maximum of the following: l I M,D + l KCI Mwp ) + A T,,x o a,, , or - 2 2 y n I K3C2 (0.1493
+ ) + A T,,x (2 0ix+ oj,)
i where Mm = resultant operating basis earthquake (0BE) moment (used only for OBE transients) C,C.K 1 2 and K 1 2
= stress indices !
o,,^ and o,,,^ = peak and steady-state stresses for a 1*F axisymmetric i thermal shock , 1 AT,,, = maximum temperature difference during the transient The scale factor.of 0.1493 is the value of the global bending stress when the local bending reaches a peak of 1.0 (normalized value) l M, = maximum thermal bending (calculated above) As illustrated in Figure 5.19, for all piping locations, the global l stress and local stress will build up at different rates. The local stress is due to the skin effects of the thermal shock, and the global l stress due to the thermal bending moments caused by thermal stratification. Figure 5.19 shows that there are two possibilities for the stress range depending on the relative magnitudes of the local and global stresses. This explains the [C] term above. For transients other than those listed above, stratified flow conditions l are assumed. Same equations are used except that AT,y is the maximum j and o 1 top-to-bottom temperature steady-state stresses for a difference, and o,h,ermal 1*F stratified t ,s hock.are the peak and l l VII-69
E i s 1 j c or Q l Stratified - '
- Not Stratif'ed I
Global Stress Gg G = Steady-State Global Stress max = Maximum Local Stress max ss = Steady-State Local Stress i G and max have the same signs 0.1493 G - i
/ Local Stress ss I
I tirne 1 I I I I First Possible Combination A = max + 0.1493.G I B B = ss + G C=B-A 1 D=0 A I Range = B - D = lssi + IGl IC l l D time l I Second Possible Combination I A l l l Bi Range = A - C = (2. max - ss) - 0.7014.G
< 12 max - ssi + 0.1493 G l D A
tiNHS i I l l lC : o i I i l l Figure 5.19 Typical Transient Thermal Stress Response j for Piping Locations J 1 VII-70
l l It should be noted that the AT,,, term was not used for transients 10a and 10b because these transients do not have a thermal shock. Fatigue evaluation was performed per the rules of NB-3653.6, which can be summarized as follows: 1 S, = 0.5+( '"*""'* )
- X,* (peaks tress)
E,n,y.,3, where S, = alternating stress E,,1,3 , c m , = Young's modulus used in Appendix I of the ASME Code in the fatigue curve E,n,y,,, = Young's modulus used in the analysis and K, is calculated as follows: i l
- 5, 5 35, : l K, = 1. 0 !
1
- 35, 5 Sn5 3mS, :
(1~"I #" K*=1.0+ n (m-1) ( 3S, -1) Sn a 3ms, l K, =1/n The parameters m and n are obtained from Table NB-3228.3(b)-1. l VII-71 l l
t l l 5.3.3 Nozzle Evaluation l Two-dimensional axisymmetric finite element models of the RCS hot leg surge nozzle and the pressurizer surge nozzle were generated, and several base load cases were analyzed as explained in Section 5.2.4 of this summary report. These base load cases include a thermal shock, mechanical loadings including , internal pressure. Results of the thermal base load case was used to l calculate the secondary stress index C ,2 and mechanical base load cases were used to calculate C at the most critical location of the nozzle assembly (the 3 nozzle safe end). The nozzles were evaluated per the rules of NB-3200 (design by analysis) by calculating the maximum primary plus secondary stress intensity. This stress intensity was calculated as the sum of two terms as follows: Pt+ P3+ Q = (P t + P3 + Q)7 + (P + P3 + Q), t (1) where (see also ASME Section NB-3200), l Pt = local membrane stress, P3 = bending stress, l Q = self-equilibrating stress necessary to satisfy continuity of structure. The first term on the right hand side of Equation (1) is the thermal term given by: (Pt+Pp O) 7 = [B] + o, ,( ) The [B] term appears in the transients with axisymmetric thermal shock, namely, transients number 3a, 3b, 4a, 4b, 5, 6, 7, 27 and 28 (see Section 4.3 of this summary report). It is given by the maximum of the following: l 1 ( 2 O_ - O,,) h T, or l l C M, + O_hT where 0,, and Q,, are the maximum and steady-state secondary stresses due to a l'F axisy:c.etric thermal shock. Similarly, o is the maximum inside diameter stress in a straight pipe due to a li*/hr linear heatup, and the term (dT/dt),, is the maximum heatup or cooldown rate of the transient in question. VII-72
1 The mechanical term on the right hand side of Equation (1) is given by: ( Pt+ Ps+ O) u= C1 P o +Cy Mag where Mm = operating basis earthquake (OBE) moment P, = pressure range for the transient pair C and C, = secondary stress indices 1 D, = pipe outside diameter t = pipe wall thickness D = diameter of interest (either inside or outside diameter) I = moment of inertia of the pipe cross section The limit for (P t
+ P, + Q) is 35, per NB-3222.2. If this limit is exceeded, j then simplified elastic-plastic analysis is invoked per NB-3228.5, and a !'
penalty factor is calculated for use in the fatigue analysis. Furthermore, a thermal stress ratchet per NB-3222.5 is to be performed. The primary plus secondary plus peak stress evaluation is performed using the methodology summarized in Section 5.3.2 of this calculation. However, since both nozzles are located on vertical segments of the surge line, they are not subjected to stratified flow conditions. Accordingly, the AT term, associated with stratified transients, was not used for the nozzles. VII-73
l- . l i I l l 5.3.4 Integral Attachments l (a) Riqid Support Lua The surge line piping located at the rigid support rectangular lug was l evaluated similar to other piping locations except that the Code equations l have some additional terms. Specifically, Equation 10 of NB-3653 becomes: ! I Sn =C1 P,D, + C 2 Dpy +Sn2 g s 3 S, i where the term Sg is defined in Code Case N-122-1 (Reference 36) as follows: 1 L l S" = Ay5+ Z'"'+ #"+ 2 L L, + 2L#*L*
+M' l yt Zyy 1 2 y l
where W, M , uM , 3Q2Q and M t r are loads on the lug (see Figure 5.20). However, l the only non-zero component is W acting on the lug: l l l (T##'.-7 0 ) l W=F1 ( Tp,z-7 0 ) + F2 ( Tut -7 0 ) + F3 10 l Istratification 4p 4y oss a 39 l where F through F, are obtained from the base load cases (see Section 5.2), 3 and F m is the lug force during an OBE seismic event. The C7 factor is calculated in accordance with the Code Case. If Equation 10 stress limit is exceeded, then equations 12 and 13 are evaluated, and a thermal stress ratchet check is performed. In this case, ! additional stress limits per Code Case N-122-1 are checked: f Sn*] s 2.ssy where S *] - W" The load component W is calculated using the same expression as W. l VII-74 l i
,4 - 5. , l 1 1 ( l i i i i i I i t i 1 } l 1 i i j i I
$ 0 e 2Lg4 ' , JL 02 lI
! 2ci l le I[- M o o: Y aj % "' 7 "N I Og i ( Q i 1 4 i i i d 1 i I a 1 4 { rigure 5.20 Rectangular Weld Attachment Load Components j (Code Case N-122-1) I l VII-75
i I The peak stress is calculated per NB-3600 with the following term added: ; i Sg=K i Sg +K i Ea lT - 7 T, l where Ki = 1.3 for ground full penetration welds per Code Case N-122-1, the j second term represents the temperature difference between the lug and the pipe wall. F (b) Dummy Pipe The surge line piping located at the dummy pipe was evaluated similar to other piping locations except that the Code equations have some additional terms. Specifically, Equation 10 of NB-3653 becomes: S ,= C3 +C2 +Se s; 3 S,, 2t 2J where the term S u is defined in Code Case N-391-1 (Reference 37) as follows: CW w Sg = +1. 7 Ea l T 7 -T,l l where load components are defined in Figure 5.21 (W is the only non-zero ; component). The average local temperatures in the dummy pipe and the pipe ! wall are represented bt T yand T,, respectively. The index C,, is obtained l from Code Case N-391-1. Also A is the cross sectional area of the dummy 1 pipe. The value of the load (W) is non-zero for OBE transients, and zero for all other transients. Its value was obtained from OBE seismic analysis of the surge line.
- If Equation 10 is exceeded, then Equations 12 and 13 are evaluated. It should l be noted that S u is not included in either of these two equations.
l The peak stress is calculated per NB-3600 with the following term added: S p7 =K 7 Sn where K7 = 1.8 for full penetration welds on the outside diameter, and K7 = 1.0 for full penetration welds on the inside diameter per Code Case N-391-1, the second term represents the temperature difference between the dummy pipe and the pipe wall. VII-76
i i . i . i i 1 d l, f } } j i d 1 1 1 i Nunphe
\ "L 4
'q I AN [bn cF as
- l f w- s r A
+
- j ,
I l v,~hl ~
"sN L ur b 4 5 l I i Og og Run Wee l
( 4 1 , 1 i 1 i i 1 3 , Figure 5.21 Circular Weld Attachment Load Components 1, (Code Case N-391-1) 4 l VII-77
t 5.3.5 Thermal Ratchet Check [ ASME Code equation results for the surge line piping, summarized in Section 2 ! of this attachment, show that the stress limit of Equation 10 of NB-3653 was l exceeded at some locations. Similarly, the stress limit of NB-3222.2 was exceeded at both the RCS hot leg and the pressurizer surge nozzles. Accordingly, as required by the ASME Code, a thermal ratchet stress check was performed as follows: ; According to the rules of NB-3653.7, the range of AT cannot exceed 3 ' that calculated as follows: y 'S,. A T, range s 0 ,g ca i where i y' = 3.33, 2.00, 1.20, and 0.80 for x=0.3, 0.5, 0.7, and 0.8, j respectively i x = (PDo /2t)(1/S,) P = maximum pressure for the set of conditions under consideration, psi j C, = 1.1 for ferritic material
= 1.3 for austenitic material j l
l l E = Young's modulus, psi e = coefficient of thermal expansion, 1/ F S, = yield strength, psi, taken at the average fluid l temperature of the transient l The above check was performed at the piping locations at which Equation 10 was exceeded. Similar check was performed per NB-3222.5 on the surge nozzles to l ensure that the Code requirements are met. l. VII-78
l l 1 i i 5.3.6 Thermal Stripina Evaluation ) Thermal striping in the surge line was characterized, in the CE0G report on l thermal stratification in the surge line (Reference 3), with a maximum fluid ; AT of 140*F and 28'F at oscillation frequency of 1 Hz and 0.25 Hz, ' respectively. The maximum alternating stress range resulted from a AT of l 140'F at oscillation frequency of 0.25 Hz with heat transfer coefficient of 3,500 BTU /hr ft 2 F and the following fluid properties: thermal conductivity = 9.81 BTU /hr ft "F, thermal diffusivity = 0.16 f t2 /hr. A thermal striping loading spectrum was developed, and is summarized in Table 5.1. This spectrum is based on the design basis transients given in Section 4 of this summary report. l Table 5.1 Thermal Striping Loading Spectrum ; Service Condition Number of Cycles ATs ,s ("F) Normal 300 340 1500 250 : 1600 200 l 2000 150 ! l 175420 90 l 4580 89 ! 34080 86 17640 80 t 200 73 I 30000 60 2000000 32 Upset 30 250 40 80 5 46 195 32 Each cycle in Table 5.1 initiates a cyclic striping process with initial fluid temperature fluctuation listed in Table 5.1. The amplitude of the fluid temperature fluctuation is assumed to decay with time. The decay correlation was developed, and is documented, in the Westinghouse Owner's Group (WOG) report on thermal stratification in the pressurizer surge line (Reference 34). The magnitude of stress variation due to thermal cycling is given by (Reference 39): l _ y 1AT,,uis G alt " 2 Vll-79
where, E is Young's modulus, v is Poisson's ratio, e is the coefficient of thermal expansion, q is a frequency correction factor AND ATage is the amplitude of the fluid temperature cycling near the pipe wall. Based on Reference 2 methodology, AT nge is calculated from AT ,3 as follows: 3 A Tf ,uis =
)ATsrs (3
The frequency correction factor, q, was calculated using the methodology of Reference 37 as follows:
=
71 , s/2 +20+202
~
( . where k is thermal conductivity, h is the heat transfer coefficient between the fluid and the pipe wall, u is the fluid oscillation frequency and E is the thermal diffusivity. Finally, the accumulative fatigue usage factor due to thermal striping was calculated by dividing AT gm into 5*F increments, then calculating the number of striping cycles, at frequency u, consistent with Figure 5.1 in each of the 5 F increments. For each temperature increment, the alternating stress and the number of striping cycles are paired and the corresponding partial fatigue usage factor is calculated using the ASME Code fatigue curve. The fatigue usage factor was then obtained by summation of the partial factors. VII-80
5.4 Licuid Sample Line The piping analysis computer program ME-101 is used to perform the Deadweight, Thermal and Seismic Inertia analyses. The mathematical model was generated based on the piping design condition. Code Case N411 is utilized by using i seismic response spectra curves with damping values of 2% - 5% (extrapolated from 1% damping curves) for Operating Basis Earthquake (CBE). Design Basis Earthquake (DBE) is conservatively taken 2 x OBE. i The piping stresses are checked in accordance with the rules of ASME Boiler and Pressure Vessel Code, Section III, 1974 through 1974 Summer Addenda, which is the design code. The pipe support loads were calculated, and the final evaluation of design loads is performed separately in the pipe support calculations. Thermal plus DBE seismic displacements are reviewed for excessive displacement (>1"), and valve accelerations are evaluated and checked against the allowable acceleration.
)
i l VII-81
i i l 6 REFERENCES
- 1. Calculation number M-1201-015-AA, Revision 2," Pressurizer Surge Line 1201-015-12"-A-EEO."
- 2. Calculation number M-DSC-046, Revision 4," Pressurizer Surge Line."
- 3. Combustion Engineering Owners Group (CE0G) Report Number CEN-NPSD 546-P, Revision 1-P," Pressurizer Surge Line Flow Stratification Evaluation," ABB l Combustion Engineering Nuclear Power, December 1991.
- 4. Drawing number S2-1201-ML-015, Revision 3, Line S2-1201-ML-015-12"-A-EE0 l from the Hot Leg to the Pressurizer. l l
S. SONGS Units 2&3 Updated Final Safety Analysis Report (FSAR).
- 6. ASME Boiler and Pressure Vessel Code, Section Ill, Division I, 1986.
- 7. Design Specification DS-1201, Revision 8," Design Specifications for Reactor Coolant System Piping Components."
- 8. Nuclear Systems Materials Handbook, Design Data, Oak Ridge National i Laboratory, 1986.
- 9. Report number 5023-919-74-0, Revision 1, Figure 7.5," Internal Details, i San Onofre III 96" 1.D. Pressurizer," surge nozzle thermal sleeve details.
- 10. Drawing number 5023-923-23-3, Revision 6," Nozzle Details, San Onofre 11 Piping", RCS hot leg surge nozzle details. j
- 11. Drawing number 5023-919-83-01, Revision 3," Nozzle Details, San Onofre III l 96" 1.D. Pressurizer," pressurizer surge nozzle.
- 12. Wolf, H.," Heat Transfer," Harper & Row Publishers, New York,1983.
- 13. "ANSYS Engineering Analysis System User's Manual," Swanson Analysis Systems Inc., Issued: May 1, 1989.
- 14. Report number S023-923-49-3," Project Specification for Reactor Coolant i
Pipe and Fittings for San Onofre Unit 2," Specification number 1370-PE-140, Revision 07, December 1, 1979.
- 15. Drawing number S023-919-45-2, Revision 5,"96 inch ID Pressurizer."
- 16. Calculation number N-0220-027, Revision 0," Unit 2&3 Maximum Pressurizer & i RCS Hot Leg Temperature Difference During Heatup."
- 17. Document number C92083157616, Test Data, Pipe Wall Temperature l Measurements During Unit-3 Heatup.
Vll-82
6 REFERENCES - cont.
- 18. Drawing number 5023-923-5-5, Revision 4, Piping Assembly, San Onofre II, RCS Hot Leg Geometry.
- 19. Document number S023-919-3-6," Project Specification for A Pressurizer Assembly for San Onofre Unit 2 & 3," Combustion Engineering Specification Number 01370-PE-130, Revision 07, September 20, 1979.
- 20. Document number C760930G0218, Pressurizer Design Report, Combustion Engineering Report number CENC-1275," Analytical Report for Southern ,
California Edison, San Onofre Unit No 2 Pressurizer," September 1976. ' l l 21. Report number 90002,"Line Designation List-Masterfile," Revision 74.
- 22. Drawing number S2-1201-ML-001, Revision 8 RCS Hot Leg from Reactor Vessel to Steam Generator.
! 23. Calculation number DC-3606, Revision 0," SONGS-1 Pressurizer Surge Line Structurol Analysis." l l 24. Document number 5023-919-3-6, Specification number 01370-PE-130, Revision 07," Project Specification for a Pressurizer Assembly for San Onofre Unit l Nos. II & III," Combustion Engineering Power Systems, 9-10-79. '
- 25. Young, W. C. "Roark's Formulas f or Stress & Strain," Sixth Edition, McGraw-Hill Book Company, 1989,
- 26. Pipe Support Drawings
- 1. 52-RC-015-H-019, Revision 1, DCN 1, 2, 3, 4.
- 2. 52-RC-015-H-016, Revision 1, DCN 1, 2.
- 3. S2-RC-015-H-017, Revision 1, DCN 1, 2. 3, 4, 5. ;
- 4. S2-RC-015-H-018, Revision 1.
I
- 5. 52-RC-015-H-020, Revision 0, DCN 1, 2, 3, 4.
l
- 6. 52-RC-015-H-021, Revision 0, DCN 1, 2, 3.
- 7. S2-RC-015-H-903, Revision 0, DCN 1, 2, 3, 4, 5, 6.
- 8. S3-RC-015-H-019, Revision 2, DCN 1, 2.
- 9. 53-RC-015-H-016, Revision 0, DCN 1, 2, 3, 4, 5.
- 10. 53-RC-015-H-017, Revision 1, DCN 1, 2, 3, 4, 5, 6, 7, 8, 9.
- 11. 53-RC-015-H-018, Revision 2.
- 12. 53-RC-015-H-020, Revision 0, DCN 1, 2, 3, 4, 5.
Vll-83
i l I i i 6 REFERENCES - cont. ;
- 13. 53-RC-015-H-021, Revision 0, DCN 1, 2.
- 14. S3-RC-015-H-903, Revision 0, DCN 1, 2, 3, 4, 5.
- 27. Grinnell Pipe Hangers, Catalog PH-92, Grinnell Corporation, Exeter NH 03833.
~
- 28. Calculation number DC-3606," SONGS-1 Pressurizer Surge Line Analysis."
- 29. "The Heat Transfer Problem Solver," Staff of Research and Education Association, Piscataway, NJ 08854.
- 30. Kreith F.," Principles of Heat Transfer," 3rd. Edition, Intext Press, Inc.,
1973. ;
- 31. Letter from J. D. Edmundson of ABB Combustion Engineering to J. Santander i of SCE dated July 22, 1992.
Subject:
SCE Pressurizer Surge Line l Evaluation.
- 32. Calculation number P-450-1.50, Revision 12, rigid pipe support ;
I calculation, Unit 2.
- 33. Calculation number P-450-1.150, Revision 8, rigid pipe support calculation, Unit 3.
- 34. Westinghouse report number WCAP-12639," Westinghouse Owners Croup, l Pressurizer Surge Line Thermal Stratification Generic Detailed Analysis,"
l Program MUHP-1091, June 1990. l 35. Document number 90004, Revision 53," Piping Material Classifications," l Pipe Support Material for Dummy Stubs, Plates, and Base Plates.
- 36. ASME Section III Code Case N-122-1," Procedure for Evaluation of the Design of Rectangular Cross Section Attachments on Class I Piping."
- 37. ASME Section III Code Case N-391-1," Procedure for Evaluation of the Design !
of Hollow Circular Section Welded Attachments on Class 1 Piping."
- 38. SCE document S023-5-1.3, Revision 13, Operating Instruction," Plant Startup from Cold Shutdown to Hot Standby."
- 39. SCE document number 1814-AR801-M0004-0,"ASME Code Evaluation of Pressurizer Surge Line Including Thermal Stratification for San Onofre Nuclear Generating Station Units 2 and 3."
- 40. Letter from T. Yackle (SCE) to V. Fisher (SCE) dated October 25, 1993.
Subject:
Maximum Delta-Temperature Between the Pressurizer and Hot Leg, Response to NRC Bulletin 88-11, SONGS 2 and 3. Vll-84
l l 7 NOMENCLATURE A = area, in 2. l e = coefficient of thermal expansion, in/in "F. S = coefficient of volumetric thermal expansion,1/*F. j C = specific heat, BTV/lb, *F. I D, = inside diameter, in. Do = outside diameter, in. E = Young's modulus, psi. F = force, Ib. I F,n = shearing force, lb. i F,, Fy and F, = components of force (F) in the direction of x, y and ! z, respectively, Ib. , g = gravitational acceleration, ft/sec , z h = heat transfer coefficient, BTU /hr ft 2 *F. I = moment of Inertia, in'. l
\
2 k = thermal conductivity, BTU /hr ft *F. M = moment, in-lb. M3 = bending moment, in-lb. M1 = torsional moment, in-lb. M,, My and M, = components of moment (M) in the direction of x, y and z, respectively, in-lb. M cy = resultant operating basis earthquake moment, in-lb v = Poisson's ratio. P = pressure, psi. P,. = Prandtl number. r = radius, in. . I VII-85
7 NOMENCLATURE - cont. R, = Reynolds number. i Ri = Richardson number. l p = density, lb,/in'. l \ S, = stress intensity, ksi. Ax = thermal displacement in the x-direction, in. Ay = thermal displacement in the y-direction, in. Az = thermal displacement in the z-direction, in. t = wall thickness, in. T = temperature, "F. AT = temperature differential, F. Au = displacement in the x-direction, in. Av = displacement in the y-direction, in. Aw = displacement in the z-direction, in. V = velocity, ft/sec. i i l i l l l Vll-86}}