ML20092H874

Nonproprietary Technical Justification for Eliminating RHR Lines Rupture as Structural Design Basis for Comanche Peak Nuclear Power Plant - Unit 2

ML20092H874
Person / Time
Site:	Comanche Peak
Issue date:	12/31/1991
From:	Adamonis D, Schmertz J, Witt F WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20034D235	List: ML20092H864 ML20092H870 ML20092H874
References
TXX-92075, WCAP-13166, NUDOCS 9202210406
Download: ML20092H874 (60)

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W2stinghcuse Preprietary Ciss 3 WCAP 13166 TECHNICAL JUST!FICATION FOR ELIMINATING RESIDUAL HEAT REMOVAL LINES RUPTURE AS TiiE STRUCTURAL DEf!GN BASIS FOR COMANCHE PEAK NUCLEAR POWER PLANT - UNIT 2 December 1991 J. C. Schmertz Y. S. Lee S. A. Swamy Verified: l F.J.$1tt' N

. Structural Mechanics Technology n f1

/ /

Approved: U)// L"<( e*-e--

D. C. Adamonis, Manager Materials, Mechanics and Diagnostic Technology Work Performed Under Shop Order: JTTP 5600 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division

.' P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 a c1991 Westinghouse Electric Corp.

All Rights Reserved WPF1012J/122891:10

4 TABLE Of CON 1CNTS it.CliED 1111c PJtue

1.0 INTRODUCTION

1.1 Background 11 1.2 Scope and Objective 1+1 1.3 References 1-3 ,

1 2.0 OPERATION AND STABILITY Of THE RHR LINE 2.1 Stress Corrosion Cracking 21 .

2.2 Wa;or Hammer 23 2.3 Low Cycle and High Cycle fatigue 2-3

-2.4 Summary Evaluation of RHR Lines for Potential Degradation During Service 24 2.5 References 2-5 a

. 3.0 MATERIAL CHARACTERIZATION

, 3.1 Pipe and Weld Materials 3-1

, 3.2 Material Properties 3-1

. 3.3 References 32 4.0 LOADS FOR FRACTURE HECHANICS ANALYSIS 4.1 Nature of the Loads 41 4.2 Loads for Crack Stability Analysis 4-2 4.3 Loads for leak Rate Evaluation 43 4.4 Loading Conditions for the RHR Lines 4-3 4.4.1 Summary of Loads and Geometry for the RHR Lines 4-3 4.4.2 Governing locations for the RHR Lines 43 5.0 FRACTURE MECHANICS EVALVATION 5.1 Global failure Mechanism 51 5.2 Leak Rate Predictions 5-2 5.2.1 General Considerations 52 5.2.2 Calculation Method 53 WPF1012J/010992:10 iii

TABLEOFCONTENTS(continued)

Section Title Dgs 5.2.3 Leak Rate Calculation 54 5,3 Stability Evaluation 54 5.4 References 55 6.0 ASSESSMENT OF HARGINS FOR THE RHR LINES 61

7.0 CONCLUSION

S 71 APPENDIX A Limit Homent A1 e

'e

?

WF1012J/010992:10 iv

LIST OF TABLES libh lill.2 E190 s

3-1 Room Temperature Mechanical Properties of the 3-3 the RHR Lines Materials (Loop 1) 3-2 Room Temperature Mechanical Properties of the 3-4 RHR Lines Materials (Loop 4) 33 Room Temperature ASME Code Minimum Properties 35 34 Average and Minimum Mechanical Properties including 3-6 Loop 1 and loop 4 at Room Temperature 3-5 Representative Tensile Properties for the RHR Lines 3-7 Appropriate for Without and With Valve Leaking Cases

. 36 Modulus of Elasticity (E) 3,7

. 4-1 Summary of LBB Loads and Stresses of RHR Lines 4-5 Without Valve Leaking 4-2 Summary of LDB Loads and Stresses of RiiR lines 4-5 With Valve Leaking 51 Leakage Flaw Size for the RHR Lines 56 Without Valve Leaking 52 Leakage Flaw Size for the RHR Lines 5-6 With Valve Leaking 53 Sunmary of Critical flaw Size for the RiiR Lines 57 Without Valve Leaking WPf 1012J/010??2:10 V

LIST OF TABLES (continued)

Ishin I'lls E19e r

54 Summary of Critical Flaw Size for th9 RHR Lines 5-7 With Valve Leaking 61 Leakage Flaw Sizes, Critical flaw Sizes and Margins 6-2 for the RHR Lines - Without Valve Leaking 62 Leaktge flaw Sizes, Critical Flaw Sizes and Margins 63 for the RHR Lines - With Valve Leaking 63 LBB Conservatisms 64 e

s 9

e e

i Wof1012J/010992:10 vi

LIST OF FIGURES E19Et lillt PJSf!

3-1 Comanche Peak RHR Layout Loop 1 3-8 32 Con.inche Peak RHR Line Layout Loop 4 39 41 Comanche Peak RHR Line Showing the Governing 46 Locations 5-1 Fully Plastic Stress Cistribution 58 52 Analytical Predictions of Critical Flow Rates 59 of Steam Water Mixtures 3 [ ).c,. Pressure Ratio as a 5-10 Function of L/D

~

5-4 Idealized Pressure Drop Profile through a 5 11 Postulated r. rack 5-5 Loads Acting on the Model at the Governing Location 5-12 56 Critical flaw Size Prediction for RHR Lines 5 13 Node 2030 - Without Valve Leaking - SHAW 5-7 Critical Flaw Siza Prediction for RHR Lines 5-14 Node 2142 - Without Valve Leaking - SAW 58 Critical Flaw Size Prediction for RHR Lines 5 15 Node 2152 - With Valve Leaking - SAW i

, 5-9 Critical flaw Size Prediction for RHR Lines 5-16 Node 4480 - With Valve le . king - SHAW i

f-l Pipe with a Through-Wall Crack in Bending A-2 WPf1012J/010992:10 yii 1 . . _ - . _ - .,_ - - _ - . - ...- - . .

SECTICN 1.0 l

', lHTRODUCTION 1.1 packaround The current structural design basis for the residual heat removal (RHR) lines l in Comanche Peak Unit 2 requires postulating non mechanistic circumferential l and longitudinal pipe breaks. This results in additional plant hardware (e.g.

pipe whip restraints and jet shields) which would mitigate the dynamic consequences of the pipe breaks. It is, therefore, highly desirable to be realistic in the postulation of pipe breaks for the RHR lines. Presented in ,

this report are the descriptions of a mechanistic pipe break evaluation method '

and the analytical results that can be used for establishing that a i circumferential type break will not occur within the Residual Heat Removal ,

(RHR) Lines. This methodology used is generally referred to as leak before-break (LBB). The evaluations considering circumferentially oriented flaws cover longitudinal cases.

. The phenomena which could cause potential thermal cycling and stratification

. in the Comanche Peak Unit 1 RHR lines a're described in WCAp 12258 (1 1) using the experience from the plants where cracking was observed due to thermal cycling and stratification in general. The cracking was observed in the RHR line of the Genkai Plant Unit 1 in particular. WCAP 12258 describes thermal loading, structural stability and the crack sizes corresponding to 10 GPM leakage for two cases: These are the case where the first valve from the primary loop is not leaking and the case where this valve is leaking. The RHR lines layout and geometries of Comanche Peak Unit 2 are same as those of Unit

1. The operating transients of both units are identical. Therefore the loadings obtained from the evaluation as described in WCAP 12258 were used in the LBB evaluations presented in this report.

1.2 Scooe and Ob.iectives The general purpose of this investigation is to demonstrate leak before-break (LBB) for the RHR lines. The scope of the report is limited to the high ,

energy portion of the RHR lines (primary loop junction to the first ulve).

WPf1012#010692:10 1-1

l l

The leak before break demonstrations for the RHR lines are considered for two -l cases: 1) when the leakage through the first valve from the primary loop .

would not occur and 2) the leakage through the first valve would occur. The

• first case corresponds to the non stratification case and the second case '

includes the thermal stratification effects due to the valve leakage.

Schematic drawings of the piping systems are shown in Section 3.0. The recommendations and criteria proposed in NUREG 1061 Volume 3 (1 2) are used in j this evaluation. The criterla and the resulting steps of the evaluation procedure can be briefly summarized as follows:

1) Calculate the applied loads. Identify the location at which the highest stress occurs.

2)- Identify the materials and the associated material properties.

3) Postulate a through-wall flaw at the governing location. The site of

the flaw should be large enough so that the leakage is assured of ,

detection with margin using the installed leak detection equipment when ,

the pipe is subjected t'o normal operating loads. A margin of 10 is ,

demonstrated between the calculated leak rate and the leak detection .

capability.

4) Using maximum faulted loads, demonstrate that there is a margin of at least 2 between the leakage size flaw and the critical size flaw.

5) Review the operating history to ascertain that operating experience has indicated no particular susceptibility to failure from the effects of corrosion, water hammer or low and high cycle fatigue.

6) for the materials actually in the clant provide the material properties and-justify that the properties used in the evaluation are representative of the plant specific material. .

7) Demonstrate margin on applied load. ,

l WPF1012J/122891 10 12

_ _ . . _.._____u______. ._ _ _ _ _ _ __

1he flaw stability analyses are performed using the methodology described in SRP 3.6.3 (1 3).

~

The leak rates are calculated for the normal operating condition. The leak rate prediction model used in this evaluation is an [

]'d . The crack opening area required for calculating the leak rates is obtained by subjecting the postulated through wall flaw to normal operating loads (1-4). Surface roughness is accounted for in determining the leak rate through the postulated flaw.

The computer codes used in this evaluation for leak rate and fracture mechanics calculations have been validated (bench marked).

1.3 ReferenCri 1-1 WCAP-12258, Evaluation of Thermal Stratification for the Comanche Peak Unit 1 Residual Heat Removal Lines, April 1989 (Westinghouse Proprietary

. Class 2).

. 1-2 Report of the U.S. Nuclear Regulatory Commission Piping Review Conunittee Evaluation of Potential for Pipe Dreaks, NUREG 1061, Volume 3, November 1984.

1-3 Standard Review Plan; public comments solicited; 3.6.3 Leak Before-Break Evaluation Procedures; federal Register /Vol. 52, No.167/ friday, August 28, 1987/ Notices, pp. 32626 32633.

1-4 NUREG/CR-3464, 1983, "1he Application of fracture Proof Design Methods Using Tearing Instability Theory to Nuclear Piping Postulated Circumferential Through Wall Cracks."

WH iO12J/122891:10 13

SECTION 2.0

^

OPERATION AND STABILITY Of lHE RHR LlHE I

2.1 Stress Corrosion Crackina The Westinghouse reactor coolant system primary loop and connecting Class I lines'have an operating history that demonstrates the inherent operating i stability characteristics of the design. This includes a low susceptibility to cracking failure from the eff ects of corrosion (e.g., intergranular stress j corrosion cracking). This operati: history totals over 450 reactor years, including five plants each having ou r 17 years of operation and 15 other plants each with over 12 years of operation.

In 1978, the United States Nuclear Regulatory Commission (USNRC) formed the second Pipe Crack Study Group. (The first Pipe Crack Study Group established in-1975 addressed cracking in boiling water reactors only.) One of the

. objectives of the second Pipe Crack Study Group (PCSG) was to include a review -

of the potential for stress corrosion cracking in Pressurized Water Reactors

. (PWR's). The results of the study performed by the PCSG were presented in NUREG 0531 (Reference 2-1) entitled " Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants." In that report the PCSG stated:

"The PCSG has determined that the potential for stress corrosion cracking in PWR primary system piping is extremely low because the' ingredients that produce IGSCC are not all present. The use of ,

hydrazine additives and a hydrogen overpressure limit the oxygen in the coolant to very low levels. Other impurities that might cause stress corrosion cracking, such as halides or caustic, are also rigidly controlled. Only for brief periods during reactor shutdown when the coolant is expued to the air and during the subsequent startup are conditions even marginally capable of producing stress corrosion cracking in the primary systems of PWRs. Operating experience in PWRs o' supports this determination. To date, no stress corrosion cracking has been reported in the primary piping or safe ends of any PWR."

W 101*t J/122891:10 21

1 During 1979, several instances of cracking in PWR feedwater piping led to the '

establishment of the third PCSG. The investigations of the PCSG reported in

lillREG 0691 (Reference 2-2) further confirmed that no occurrences of IGSCC have been reported for PWR primary coolant systems.

As stated above, for the Westinghouse plants there is no history of cracking failure in the reactor coolant system loop ar connecting Class I piping. The discussion below further qualifies the PCSG's findings, for stress corrosion cracking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, susceptible material, and a corrosive environment. Since some residual stresses and some degree of material susceptibility exist in any stainless steel piping, the potential for stress corrosion is minimized by properly selecting a material immune to SCC as well as preventing the occurrence of a corrosive environment.

The material specifications consider compatibility with the system's operating environment (both internal and external) as well as other material in the system, applicable ASME Code rules, fracture toughness, welding, fabrication, ,

and procsssing, ,

The elements of a water environment known to increase the susceptibility of .

austenitic stainless steel to stress corrosion are: oxygen, fluorides, chinrides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g.,

sulfides, sulfites, and thionates). Strict pipe cleaning standards prior to i operation and careful control of water chemistry during plant operation are used to prevent the occurrence of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally. During flushes j and preoperational testing, water chemistry is controlled in accordance with j written specifications. Requirements on chlorides, fluorides, conductivity, and pH are included in the acceptance criteria for the piping.

During plant operation, the reactor coolant water chemistry is monitored and maintained within very specific limits. Contaminant concentrations are kept -

below the thresholds known to be conducive to stress corrosion cracking with the major water chemistry control standards being included in the plant .

operating procedures as a condition for plant operation, for example, during WPF1012J/12?891:10 p 22 l

I

normal power operation, oxygen concentration in the RCS and conneeting Class I lines is expected to be in the ppb car.ge by controlling charging flow chem-istry and maintaining hydrogen in the reactor ccol+nt at specified concentra.

tions. llalogen concentrations are also stringently centrolled by maintaining concentrations of chlorides and fluorldes within the specified limits. This is assured by controlling charging flow chtmistry. Thus during plant opera-tion, the likelihood of stress corrosion cracking is minimized.

2.2 Ritter HAmmtt Overall, there is a low potential for water hammer in the RCS and connecting RHR lino since they are designed and operated to preclude the voiding condition in normally filled lines, lho RCS and connecting RHR line including piping and components, are designed for normal, upset, emergency, and faulted condition transients. The design requirements are conservative relative to both the number of transients and their severity. Relief valve actuation and the associated hydraulic transients following valve opening are considered n the system design. Other valve and pump actuations are relatively slow transients with no significant e'fect on the system dynamic loads. To ensure

. dynamic system stability, reactor coolant parameters are stringently

, controlled. Temperature during normal operation is maintained within a narrow range by control rod position; pressure is controlled by pressurizer heiters and pressurizer spray also within a narrow range for steady state conditions.

The flow characteristics of the system remain constant during a fuel cycle because the only governing parameters, namely system resistance and the reactor coolant pump characteristics are controlled in the design process.=

Additionally Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system. preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping and connected RHR lines are such that no significant water hammer can occur.

i Tf1012J/010f 92:10 2-3

2.3 Low Cyclednd HidLhd.LfAWRtt i

Low cycle fatigue considerations are accounted for in the design of the piping system through the fatigue usage factor ovaluation to show compliance with the rules of Section 111 of the ASMC Code.

Pump vibrations during operation would result in high cycle fatigue loads in the piping system. During operation, an alarm signals the exceedence of the RC pump shaft vibration limits, field measurementt have been made on the reactor coolant loop piping of a number of plants during hot functinnc1 testing. Stresses in the elbow below the RC pump have been found to be very small, between 2 and 3 ksi at the highest. Recent field measurements on typical PWR plants indicate vibration amplitudes less than 1 ksi. When translated to the connecting RHR lines, these stresses would be even lower, well below the fatigue endurance iimit for the RHR line material and would result in an applied stress intensity factor below the threshold for fatigue crack growth.

2.4 Summary Evaluation of RHR Line_for Potential Degradation Durina Seivice ,

In the Westinghouse PWR design only one incident of service cracking has been .

identified in the RHR piping. In that specific case the cracking was attributed to thermal cycling resulting from valve malfunction. The thermal ,

cycling caused an initial surface flaw to grow and reveal its presence by leakage. The leakage was detected by the plant leak detection systems, the plant was shutdown and necessary repairs were completed, in addition only one incident of wall thinning has been identified in RHR lines of Westinghouse PWR design. However, this is of no concern in the present application as described later in this section. Sources of such degradation are mitigated by the design, construction, inspection, and operation of the RHR lines, ,

Based on a review of references 2-3 through 2 6 cr.ly one incident of water hammer has bee,n reported in a PWR RHR system. This incident was a result of .

incorrect valve line up preceding a pump start. The only damage sustained was to several pipe supports. Therefore it is concluded that water hammer in the i WH101?J/010672:10 2-4

RHR system is unlikely to affect piping integrity or to cause pipe system degradation.

Wall thinning by crosion and erosion-corrosion ef f ect s "ill not occur in the RHR lines due to the low velocity, typically less than 10 f t/sec and the material, austenitit stainless steel, which is highly resistant to these degradation mechanisms. Per NUREG 0691 (2-2), a study of pipe cracking in PWR piping, only two incidents of wall thinning in stainless steel pipe were reported as noted earlier. One incident was related to the PHR system.

However, this occurred in the pump recirculation path which has higher flow velocity and is more susceptible to other contributing factors such as cavitation, than the RHR piping near the primary loop. Therefore, wall thinning is not a significant concern in the portion of the system being addressed in this evaluation, flow stratification, where low flow conditions permit cold and hot aater to separate into distinct layers, can cause significant thermal fatigue loadings.

lhis was an important issue in PWR feedwater piping where temperature differences of 300'f were not uncommon under certain operational conditions,

. Stratification is believed to be important where low flow conditions and a temperature differential exist. This is not an issue in the RHR line, where typically there is no flow during normal plant operation. During RHR operation the flow causes sufficient mixing to eliminate stratification.

The maximum normal operating temperature of the RHR piping is about 618'f.

This is well below the temperature which would cause any creep damage in stainless steel piping.

2.5 Reference 1 21 Investigation and Evaluation of Stress-Corrosion Cracking in Diping of light Water Reactor Plants, NUREG.0531, U.S. Nuclear Regulatory

/ Commission, february 1979.

Wrf101?J/010602:10 2-5

l I

2-2 Investigation and Evaluation of Cracking incidents in Piping in - t Pressurized Water Reactors, NUREG 0691. U.S. Nuclear Regulatory Commission, September 1980.  :

I 23 Utter, R. A., et. al., " Evaluation of Water Hammer Events in Light Water Reactor Plants," NUREG/CR 2781, published July 1982.

24 " Report of the U.S. Nuclear Regulatory Commission Piping Review Committee Evaluation of Other Dynamic Loads and Load Combinations,"  :

NUREG 1061 Volume 4 Published December 1984 25 Chapman, R. L.. et, al., " Compilation of Data Concerning Known and Suspected Water _ Hammer Events in Nuclear Power Plants, CY 1969 May 1981," NUREG/CR-2059, Published April 1982.

26 *Evaltation of Weter llammer Occurrence in Nuclear Power Plants,"

NUREG 0929 Revision 1, Published March 1984.

I i e

9 P

i t

WPF1012At N42:10 26

l SECTION 3.0 MATERIAL CHARACl[RilA110N 3.1 hpt._A1Wfjd Ma t er t3h The pipe materials of the RHR lines for the Comanche Peak Unit 2 are A376/Tp316. A403/WP316, A312/IP304 and A182/TP316. This portion of the RHR lines does not include any cast pipe or cast fittings. The welding processes used were GTAW, SMAW and SAW. Weld locations are identified in figures 3 1 and 3 2. The two RHR lines of Comanche Peak Unit 2 are connected to loops 1 and 4.

In the following section the tensile properties of the materials are presented for use in the leak before break analyses.

3.2 B31grial Prenerties The room temperature mechanical properties of the two RHR lines materials were obtained from the Certified Materials Test Reports and are given in Tables 31

. and 3 2. The room temperature ASME Code Section til minimum properties (3-1)

. are given in Table 3-3. It is seen that the measured properties well exceed those of the Code. As mentioned in Section 1.2, LBB evaluations are performed for two cases, without and with leaking conditions in the first valve. The material properties for these evaluations are found separately for each case, since temperatures at the locations to analyze are different for each case.

The minimum and average tensile properties at room temperature are given in Table 3-4.

The minimum and average tensile properties were calculated by using the ratio of the ASME Code Section 111 properties at the temperatures of interest alluded to above. The modulus of elasticity values were established at the temperatures of interest from the ASME Code Section 111 (lable 3-6). In the leak before-break evaluation, the representative minimum properties (yield stress and ultimate strength) at the given temperature are used for the flaw stability evaluations and the representative average properties are used in the leak rate predictions. These properties are summarized in Table 3-S.

i WPf1012J/010Y92:10 3-1

3.3 Referentti '

31 ASME Boller and Pressure Vessel Code Section 111. Division 1, Appendices July 1. 1989. ,

1 I

4 t

h I

m -

WPF101?J/122891:10 3-2

TABLE 3 1 l

, ROOH TEMPERATURE MECHANICAL PROPERTIES OF THE RHR LINES MATERIALS (LOOP 1 - RHR LINE) 10 Heat / Mat'1/ Ultimate Yield Elong. R/A Serial No. Type Strength Strength

_ A' CW2991 2 1 SA376/316 88.500 47,100 56 N/A 6 B C8456 SA403/WP316 83,600 40,300 56 N/A I C CW2991-2-1 SA376/316 88.500 47,100 56 N/A 0 L6CUH SA403/WP316 88,400 45,200 50 N/A E CW2991-2 1 SA376/316 88,500 47,100 56 N/A F CW1281-5 SA376/316 88,500 48,600 50 N/A G 04722 SA403/WP316 82,200 44,500 54 N/A H CW1281-5 SA376/316 88,500 47,100 56 N/A i

1 49189 SA182/TP316 79,500 43,000 62 77.5 J L4328 SA376/316 79.900 38.400 58.3 68.4

, _, K CW3020 4 SA376/316 83.100 49,300 49 N/A

, L F90746 SA312/TP304 89.100 49,400 55 N/A

• shown in Figures 3-1 and 3-2 b: N/A means not available WP51012J/010992:10 33

TABLE 3-2 ROOM TEMPERAlVRE HECHANICAL PROPLRTIES Of THE RHR LINES HATERIALS (LOOP 4 - RHR I.INE)

ID Heat / Mat'l/ Ultimate Yield Elong. R/A Serial No. Type _ Strength _ Strength _ _ _ _ _ . , , _ _ _

A' CW2991 2-1 SA376/TP316 88.500 47,100 56 N/A b 8 C8456 SA403/WP316 83,600 40.300 56 N/A C CW2991 2-1 SA376/TP316 88,500 47,100 56 N/A D L6CWH SA403/WP316 86,800 45,300 53 N/A E CW2991 2 1 SA376/TP316 88,500 41,100 56 N/A f 1281 14 SA376/TP316 86,500 42,350 -56 N/A G 0 5530 SA403/WP316 77,400 35,600 59.7 N/A H CW1281 2 SA376/IP316 85.500 48,900 43 N/A 1 52111 SA403/WP316 80,000 48,000 64 76.5 J L4328 SA376/TP316 79,900 38,400 58.3 68.4 K 3055-1 4 SA376/TP316 85,000 43,100 56 N/A L 39745 SA312/TP304 87,100 46,900 56 N/A ,

• shown in figures 3-1 and 3-2 6: N/A means not available

• i WPf1012J/010992:10 3-4

TABLE 3 3 Room Temperature ASME Code Minimum Properties Material Yield Stress Ultimate Strenoth (P5i) (psi)

A376/TP316 30,000 75,000 A403/WP316 30,000 75,000 _

1 6

0 6

m L

P O

4 wnesans22e91:io 35

TABLE 3-4 AVERAGE AND HINIMUM HECHANICAL PROPERTIES .

INCLUDING LOOPS 1 AND 4 AT ROOM TEMPERATURE E

Material Average

• Hinimum*

Yteld SA376 45,625 38,400-Stress /TP316 SA403 42,743 35,600

/WP316 Ultimate SA376 86,279 79,900 Strength /TP316 SA403 83,143 77,400

/WP316 _

• From Tables 3-1 and 3-2.

WF1012J/122891:10 3-6

l TABLE 3-5 .

1

, [. . REPRESENTATIVE TENSILE PROPERTIES FOR RHR LINES APPROPRIATE FOR WITHOUT AND WITH VALVE LEAKING CASES l 3:

i Temperature Minimum Averag,e Minimum Remark i ('F) Vield' Yield Ultimate (psi) (psi) Strength *

(Psi)

SA376- 617 20,647 24,5's2 61,293 w/o leaking

/TP316 5M03 617 21,178 25,427 63,326 w/o leaking l

fWPal6 b

l- SA376 530 22,506 26,741 61,702 w/ leaking

/TP316 SA403 585* 21,880 26,270 63,482 w/ leaking

/WP316 ,

TABLE 3-6

, MODULUS OF ELASTICITY (E)

Temperature E_(ksi)

+ 617'F 25,215 530'F 25,650 585'F 25,375

, w b; From Tables 3-1 and 3-2

Governing Locati m % @ 4480 as discussed later in section 4.4.2

• Governing Locat'en Nec 2152 as discussed later in section 4.4.2 i

f I

WPF1012J/010992:10 3-7

' Pipe'--12" Schedule 140 Minimum Wall Thickness: 1.0125 in.  :

FW - Field Weld -

SW Shop Weld LOOP l FW -

N [)

'# /

wg J

QN'i w

--( D n @ w -

~

,, Fw ev .

' FV W l ,- N

h W '

( -l M l , )

1

/ FW . wk ,e FW L ()

Figure 3-1 ,

Comanche Peak RHR Lines Layout - Loop 1 wro,6autzintito 3-8 M'i,ei

l

.- , ~

4 Pipe . 12" Schedule 140

., Minimum Wall Thickness: 1.0125 iU.

fW - Field pf3]q

-SW - 3 hop Weld ,'

s .

@x y , 'yW LOOP 4 ,

c 4 ..

- W F .

• N '

W.

L

@', xW m g

- N N' y r

-Figure 3-2

• . Comanche Peak RHR Lines Layout - Loop 4 wro9m/121791s to 39

SECTION 4.0 LOADS FOR FRACTURE MECHANICS ANALYSIS 4.1 Nature of the loads Under normal operating conditions, the RHR lines are subjected to axial and bending 1sads which arise from deadweight, pressure, and thermal expansion.

Under faulted conditions, the loads caused by Safe Shutdown Earthquake (SSE) arc superimposed on these normal operating loads.

Figures 3-1 and 3-2 show schematic layouts of the RHR lines for the Comanche Peak Unit 2 and identify the weld locations. Diameters and minimum wall thicknesses are shown in Figures 4-1 with governing locations.

The stresses due to these axial loads and bending moments were calculated by the following equation:

o = { + j' (4-1) where, o - stress F - axial load M = bending moment A = metal cross-sectional area 2 - section modulus The bending moments for the desired loading combinations were calculated by the following equation:

N, = Uv 2 gj) o.s (4-2) where, WPF1012J/122891:10 4-1

M = bending moment for required loading B

My = Y component of bending moment -

M = Z component of bending moment Z

The axial load and bending moments for crack stability analysis and leak rate predictions are calculated by the methods to be explained in Sections 4.2 and 4,3 which follow.

4.2 Loads for Crack Stability Analysis The faulted loads for the crack stability analysis by the absolute sum method are as follows:

F = lF DW l+lF TH l +lFP p i+lF SSEl I4~3) l (My )Dh + I NY THI + IN Y SSE My = l (4-4)

I (NZ)Dd + I NZ THI+IN Z SSEl (4-5)

M "

Z DW = Deadweight .

TH = Applicable thermal load (normal or stratified) .

P = Load due to internal pressure .

SSE = SSE loading including seismic anchor motion The faulted loads for the crack stability analysis by the algebraic sum method l are as follows:

F =

lFDW_+ FTH + Fpl +lF SSE l (4-6)

My =

l(M y)DW + (My )THi +IN Y SSEI (4-7)

M = l (4-8)

Z l(M z)DW + (M Z )THI+lMZ SSE where the subscripts are the same as indicated above.

In this analysis, the absolute sum method was used.

I l

WPF1012J/122891:10 4-2

. _ _ _> - . _ _ _ .~ __ _ -._ _ _ . .

4.3_ Loads for leak Rate Evaluation 4

The normal operating loads for leak rate predictions were calculated by the algebraic sum method as follows:

F -

FDW + FTH + FP p (4-9)

N -

Y (My )DW + (My)TH (4 10)

M Z

(MZ )DW + INZ)TH (4'II)

The parameters and subscripts are the same as those explained in Section 4.1.

4.4 Loadina Conditions for the RHR Lines The normal operating loadings for the RHR line are pressure (P), deadweight (DW) and normal operating thermal expansion. The faulted loadings consist of normal operating loads plus Safe Shutdown Earthquake (SSE) loads including the Seismic Anchor Motion.

L To ev.luate the effect of thermal cycling, the loads resulting from thermal cycling were substituted for normal thermal expansion loads as applicable.

4.4.1. Summary of loads and Geometry for the RHR Lines The load combinations were evaluated at various weld locations. Normal loads and faulted loads were determined using the algebraic and absolute sum method l respectively.

4.4.2 Governina locations for the RHR lines Figures 3-1 and 3 2 show schematic layouts of the RHR lines for Comanche Peak Unit 2 and identify the weld locations. The diameter and minimum wall thickness is shown in Figure 4-1 with the governing locations, which are in the RHR line connected to loop 4.

WPF1012J/010992:10 4-3

l l

All the welds at Comanche Peak RHR lines are fabricated using the GTAW, SMAW, -

and SAW procedures. The following governing locations were established based -

on the magnitude of total faulted stress for without and with valve leaking conditions. In all cases, the initial root passes are GTAW followed by either

~~

a SAW weld or a SMAW weld for the remaining, so that only the SAW and SMAW procedures are analyzed for in this report. As will be shown in Section S,0, the SAW and SMAW analyses are more conservative than the GTAW analyses.

Crack stability calculations are performed for these two welding procedures and twice the crack length for the 10 GPM leak rate is demonstrated to be stable for SMAW and SAW.

The governing locations without the valve leaking condition are found to be:

SA376/TP316 Node 2030 (SMAW)

SA403/WP403 Node 2142 (SAW)

The governing locations with the valve leeking condition are found to be : ,

SA376/TP316 Node 4480 (SMAW) ,

SA403/WP316 Node 21S2 (SAW) ,

The loads and stresses at these governing locations for the normal and the faulted loading combinations for the without and with valve leaking conditions are shown in Table 4-1 and 4-2 respectively. Figure 4-1 shews the governing locations.

l e

l j WPF1012J/010992:10 l 4-4 l

l

TABLE 4-1

, SUMHARY OF LBB LOADS AND STRESSES OF RHR LINES WITHOUT VALVE LEAKING Node Case Axial Axial Bending Bending Total Force Stress Moment Stress Stress (1 b)__1p_si) _{in-lb) (psi) (psil_

2030 Normal 186,921 5,007 618,468 0,087 11,094 Faulted 209,853 5,621 1,327,137 13.062 18,683-2142 Normal 181,562 4,863 829,339 8,162 13,025 Faulted 222,184 5,951 1,315,809 12,950 18,901 TABLE 4-2

SUMMARY

OF Lb8 LOADS AND STRESS OF RHR LINES WITH VALVE LEAXING a,c.e

. I

~ l

\

l e

WPF1012J/122891:10 4-5

Pipe - 12" Schedule 140 Minimum Wall Thickness: '

1.0125 in. W FW Field Weld SW - Shop Weld -

2030 (8) s N w w LOOP 4 , W

.I f

,W fY 2142

{

2l52 N SW

. FV 4480 y

W' .

b y o GOVERNING LOCATIONS WITHOUT VALVE LEAKING -

g 5 GOVERNING LOCATIONS WITH

• 7 VALVE LEAKING fy \

L g SW FW ' g Figure 4-1 ,

Comanche Peak RHR Lines Showing the Governing Locations WPF0948J/121791:10 4-6

SECTION 5.0 FRACTURE MECHANICS EVALVATION 5.1 ~ Elobal Failure Mechanism Determination of the conditions which lead to failure in stainless- steel should be done with plastic fracture methodology because of the large amount of deformation accompanying fracture. One method for predicting the failure I

of ductile material is .the [ ]*' method, based on traditional 1 plastic limit _ load concepts, but accounting for [ ]and  ;

taking into account the presence of a flaw. The flawed component is predicted to fail when the remaining net section reaches a stress level at which a plastic hinge is formed. The stress level at which this occurs is termed as the flow stress. [

,ja.c.e This methodology has been shown to be applicable to ductile piping through a large number of experiments and is used here to predict the critical flaw size *

, in the RHR lines. The failure criterion has been obtained by requiring

. ' equilibrium of the section containing the flaw (Figure 5-1) when loads are applied. ~ The detailed development is provided in Appendix A for a through-wall circumferential flaw in a pipe section with internal pressure, axial force, and imposed bending moments.- The limit moment for such a pipe is given by:

( )*' (5-1) where:

I-i g

]a,c.e WPF1012J/010692:10

, 5-1 L

~ - --

[

]'d d (5 2) }

The analytical model described above accurately accounts for the internal pressure as well as imposed axial force as they affect the limit moment. Good agreement was found betwsen the analytical predictions and the experimental results (reference 5-1). Flaw stability evaluations, using this analytical model, are presented in section 5.3.

5.2 Leak Rate Predictions Fracture mechanics analysis shows in general that postulated through-wall cracks in the RHR lines would remain stable and do not cause a gross failure of this component. However, if such a through-wall crack did exist, it would be desirable to detect the leakage such that the plant could be brought to a safe shutdown condition. The purpose of this section is to discuss the me' hod t which will be used to predict the flow through such a postulated crack and present the leak rate calculation results for through-wall circumferential .

cracks. .

5.2.1 General Considerations -

The flow of hot pressurized water through an opening to a lower back pressure (causing choking) is taken into account. For long channels where the ratio of the channel length, L, to hydraulic diameter, D g, (L/D )g is greater than [ ]*dd both [ ]*dd must be considered. In this situation the flow can be described as being single-phase through the channel until the local pressure equals the saturation pressure of the fluid.

At this point, the flow begins to flash and choking occurs. Pressure losses due to momentum changes will dominate for [ ]*dd. However, for large L/D H values, the friction pressure drop will become importsat and must be considered along with the momentum losses due to flashing.

WPF1012J/010692:10 5-2

t 5.2.2 Calculat_ingl Method In using the [

)..c..

The flow rate through a crack was calculat:d in the following manner. Figure 5-2 from reference 5-2 was used to estimate the critical pressure, Pc, for the primary icop enthalpy condition and an assumed flow. Once Pc wa. found for a _

given mass flov, the [ ]

was found from figure 5 3 taken from reference 5 2. For all cases considered, since [ ]d Therefore, this method will yield the two-phase pressure drop due to momentum effects as illustrated in figure 5-4. Now using the assumed flow rate, G, the frictional pressure drop can be calculated using AP, = [ ] d (5-3) where the friction factor f is determined using the [ ] The

. crack relative roughness, c, was obtained from fatigue crack data on stainless

. steel samples. The relative roughness value used in these calculations was [

] RMd.

The frictional pressure drop using Equation 5-3 is then calculated for the assumed flow and added to the [

j*** to obtain the total pressure drop from the system under consideration to the atmosphere. Thus, Absolute Pressure - 14.7 = [ ] 5-4) for a given assumed flow G. If the right-hand side of equation 5-4 does not agree with the pressure difference between the piping under consideration and the atmosphere, then the procedure is repeated until equation 5-4 is satisfied to within an acceptable tolerance and this results in the flow value through the crack.

WPF1012J/010692:10 5-3

L 5.2.3 Leak Rate Calculations .

1 .

Leak rate calculations were performed as a function of postulated through-wall -

crack length for the critical locations previously identified. The crack -

opening area was estimated using the method of reference 5-3 and the leak rates were calculated using the calculational inethods described above. The leak rates were calculated using the normal operating loads at the governing locations identified in section 4.0. The crack lengths yielding a leak rate of-10 gpm (10 times the leak detection capability of 1.0 gpm) are shown in Table 5-1 for RHR lines without leaking at the first valve from the primary loop and in Table 5-2 for the RHR lines with leaking at the first valve from the primary loop .

The capability of each of the pressure boundary leak detection systems are given in Tatle 5.2-9 of the Comanche Peak FSAR.

5.3 Stability Evaluation A typical segment of the pipe (at the governing location) under maximum loads of axial force F and bending moment M is schematically illustrated as shown in [

figure 5-5. In order to calculate the critical flaw size, plots of the limit ,

moment vei 3 crack length were generated as shown in figures 5-6 to 5-9 (as recommended in reference 5-4) discussed below. The critical flaw size corresponds-to the intersection of this curve and the maximum load line. In this evaluction the critical flaw size was calculated using the lower bound base metal tensile properties established in section 3.0.

As discussed earlier, the welds at the locations of interest (i.e. the governing locations) are SMAW and SAW. Therefore, "Z" factor corrections for SMAW and SAW welds were applied (references 5-5 and 5-6) as follows:

Z - 1.15 [1 + 0.013 (0.0. - 4)) (for SMAW)

_ (5-5) i Z = 1.30 [1 + 0.010 (0.D. - 4)] (for SAW) (5-6) ,

where 0.D. is the outer diameter in inches. Substituting 0.0. - 14.00 inches, l the Z factor was calculated to be 1.30 for SMAW and 1.43 for SAW. The Z WPF1012J/010992:10 5-4 l

factor for_GTAW is 1.0. The applied loads were increased by the Z factors.

The shop welding (SW) in the RHR lines is SAW and the field welding (FW) is a combination of GTAW and SMAW. Therefore the critical flaw lengths of the RHR lines are obtained using SMAW for FW and SAW for SW and leak-before-break (LBB) is demonstrated for the two welding procedures (SPAW and SAW). The critical flaw size without the valve leaking is given in Table 5-3 and Table 5-4 is the critical flaw size under the valve leaking condition. The plots of limit load versus crack length were generated as shown in Figures 5-6 and 5-7 for the condition of without leaking at the valve and in Figures 5-8 and 5-9 for the condition of with leaking at the valve.

5.4 References 5-1 Kanninen, M. F. et al., " Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks" EPRI NP-192, September 1976.

5-2 [

, p.c.e 5-3 Tada, H., "The Effects of Shell Corrections on Stress Intensity Factors and the Crack Opening Area of Circumferential and a Longitudinal Through-Crack in a Pipe," Section 11-1, NUREG/CR-3464, September 1983.

5-4 NRC letter from M. A. Miller to Georgia Power Company, J. P. O'Reilly, dated September 9, 1987.

5-5 ASME Code Section XI, Winter 1985 Addendum, Article IWB-3640.

5-6 Standard Review Plan; Public Comment Solicited; 3.6.3 Leak-Before-Break Evaluation Procedures; Federal Register /Vol. 52, No. 167/ Friday, August 28, 1987/ Notices, pp. 32626-32633.

WPF1012J/122891:10 5-5

TABLE 5-1 -

LEAKAGE FLAW SIZE FOR THE RHR LINES -

WITHOUT VALVE LEAKING aAe TABLE 5-2 LEAKAGE FLAW SIZE FOR THE RHR LINES WITH VALVE LEAKING

- - a,c.e

(

e P Y

l t

- WPF1012J/122891:10 5-6

TABLE 5 3

SUMMARY

OF CRITICAL FLAW SIZE FOR THE RHR LINES -

WITHOUT VALVE LEAKING a,c.e TABLE 5-4

SUMMARY

OF CRITICAL FLAW SIZE FOR THE RHR LINES -

WITH VALVE LEAKING e

4 ese ,

t WPF1012J/122891:10 5-7

1 i

.! 1 l

a Cie t

l-Figure 5-1 .

Fully Plastic Stress Distribution 4

)

- WF0948J/120991:10 5-8

)

i l

1 4

~

- 4, ,e M

~

n E

i 4 5 8

s 5

. u 1

~ -

STAGNATION ENTHALPY (102 Stu/lb)

Figure 5-2 Analytical Predictions of Critical flow Rates of Steam-Water Mixtures wro94w120991:io 5-9

l

~ a.c.e

'e b

Y e

4 m

W I

w f -

a 0

5 ,

LENGTH / DIAMETER RATIO IL/D)

Figure 5-3

[ ] Pressure Ratio as a Function of L/D .

WPF1012J/122891:10 5-10

]

l l

e 4,:.s

~

t-a.:,e

'/ /

o 4

_p = ==-

l l

? Figure 5-4 Idealized Pressure Drop Profile Through a Postulated Crack wro94au120991 oo 5-11

• b n -

I r,r 7..

=

~

og EL C" ,,

.25 i 4.M 2 ;E

.3

• • N

'l u Ec  !

l t2

! l l l I

I I

I I I I

^ I -

__ _J __t

• I l -

I I

l l I I

I I

I 1 w&

Figure 5-5 .

Loads Acting on the Model at the Governing Location wrtwan120991 s to 5-12

, ~. . o .

a,c,e Y

G Figure 5-6 Critical flaw Size Prediction for RHR Line Node 2030 - Without Valve Leaking - SMAW s _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - _ _ _ -

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

- i a,c.e v.

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- Figure 5-7 Critical Flaw Size Prediction for RHR Line  ;

Node 2142 - Without Valve leaking - SAW  ;

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I Figure 5-8 Critical flaw Size Prediction for RHR Line Node 2152 - With Valve Leaking - SAW

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E Figure 5-9 Critical Flaw Size Prediction for PJiR Line Node 4480 - With Valve Leaking - SMAW

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-SECTION 6.0 ASSESSMENT OF MARGINS FOR THE RHR LINES o

"- In the preceding sections, the leak rate calculations, fracture mechanics analysis were performed. Margins at the critical locations are summarized below:

In Section 5.3 using the IWB-3640 approach (i.e. Z factor approach), the critical flaw sizes at the governing locations are calculated, in Section 5.2 the crack lengths yielding a leak rate of 10 gpm (10 times the leak detection capability of 1.0 gpm) for the critical locations are calculated.

The leakage size flaws, the instability flaws, and margins are shown in Tables 6-1 and 6 2 without and with valve leakage, respectively. Both the margins on flaw size and the margins on loads are seen to be mat.

In this evaluation, the leak-before break methodology is applied conservatively. The conservatisms used in the evaluation are summarized in Table 6-3.

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TABLE 6-1 LEAKAGE FLAW SIZES, CRITICAL FLAW SIZES AND MARGINS FOR RHR LINES- WITHOUT VALVE LEAKING

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TABLE 6-2

[ LEAKAGE FLAW SIZES, CRITICAL FLAW SIZES AND MARGINS FOR RHR LINES - VALVE LEAKING

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TABLE 6 3 LBB CONSERVATISMS c factor of 10 on Leak Rate o factor of 2 on Leakage flaw for all cases o Algebraic Sum of Loads for Leakage o Absolute Sum of Loads for Stability o Average Material Properties for Leakage o Minimum Material Properties for Stability e

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WPf1012J/122891:10 6-4

SECTION 7.0 l

6 CONCLUSIONS o

This report justifies the elimination of RHR lines pipe breaks as the structural design basis for the Comanche Peak Unit 2 nuclear plant as follows:

a. Stress corrosion cracking is precluded by use of fracture resistant materials in the piping system and controls on reactor coolant chemistry, temperature, pressure, and flow during nornial operation.

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b. Water hammer should not occur in the RCS piping (primary loop and the attached class 1 RHR lines) because of system design, testing, and operational considerations.

c. The effects of low and high cycle fatigue 'n the integrity of the RHR lines were evaluated and shown acceptable. The effects of thermal stratification were evaluated separately and shown acceptable.  :

( d. Ample margin exists between the leak rates of small stable flaws and the capability of the Comanche Peak plant's reactor coolant system pressure boundary leakage detection system,

e. ' Ample margin exists between the small stable flaw sizes of item d and the critical flaw size,

f. With respect to stability of the critical flaw, ample margins exist between the maximum postulated loads and the plant specific maximum faulted loads.

The -leakage size flaws will be stable because of the ample margins and will j leak at a detectable rate which will assure a safe plant shutdown.

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l WPF1012J/122891:10 7-1

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

Based on the above, it is concluded that RiiR lines pipe breaks should not be .

considered in the structural design basis of the Comanche Peak Unit 2 nuclear 4 plant.

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WF1012J/122891:10 7-2

APPENDlX A

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