ML20137M863

From kanterella
Jump to navigation Jump to search
Pressurized Thermal Shock Evaluations in Accordance W/10CFR50.61 for B&W Owners Group Reactor Pressure Vessels
ML20137M863
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 01/31/1986
From: Lowe A, Skidmore S, Snidow N
BABCOCK & WILCOX CO.
To:
Shared Package
ML20137M841 List:
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR 77-1159656, 77-1159656-00, BAW-1895, NUDOCS 8601290088
Download: ML20137M863 (34)
v • d • e


Text

_ . - . . - . .

R%cAment i f BAW 1895 January 1986

.) 2 . .. _ , ,_ . .._ . [ b 5 $ b AI 1

O O i E E:~'T . ZZC3R Pressurized Thermal Shock Evaluations in Accordance with 10 CFR 50.61 for Babcock & Wilcox Owners Group Reactor Pressure Vessels a m.oono otwa m 3., anom osooo m Babcock & Wilcox i> l'UR .i M4 l)ertix >t! < orrip.triy l

1

BAW-1895 January 1986 PRESSURIZED THERMAL SHOCK EVALUATIONS IN ACCORDANCE WITH 10 CFR S0.61 FOR BABC0CK & WILCOX OWNERS GROUP REACTOR PRESSURE VESSELS l

t by A. L. Lowe, J r. PE S. A. Skidmore N. L. Snidow Prepared for AAW Owners Groun Materials Ca==f ttee Arkansas Power & Light Company Duke Power Company Florida Power Company GPU Nuclear Sacramento Municipal Utility District Toledo Edison Company 4

B&W Document No. 77-1159656-00 Prepared by BABCOCK & WILCOX Nuclear Power Division P.O. Box 10935 Lynchburg, Virginia 24506-0935 Babcock &WHees a McDermotI company

Babcock & Wilcox Nuclear Power Division Lynchburg, Virginia Report Number BAW-1895 January 1986 )

Pressurized Thermal Shock Evaluations In Accordance With 10 CFR 50.61 For Babcock &

Wilcox Owners Groun Reactor Pressure Vessels A. L. Lowe, J r., PE, S. A. Skidmore and N. L. Snidow Key Words: Pressurized Thermal Shock, Screening Criteria, Fracture Toughness, Reference Temperature (Pressurized Thermal Shock),  ;

Reference Temperature (Nil Ductility Temperature), Atypical Wold Metal. Linde 80 Wald Metal. Reactor Pressure Vessel ABSTRACT Pressurized thermal shock evaluations were performed in accordance with 10CFR50.61, Fracture Toughness Requirements for Protection Against Pressur-ized Thermal Shock Events," for the eight Babcock & Wilcox Owners Group 177 FA reactor pressure vessels. The projected values of RT for all the PTS materials in the reactor vessel beltline region are below the screening j criteria at the expiration date of the operating license of all the plants.

The evaluation of the atypical weld metal showed the projected values of l RT also are below the screening criteria at the operating license NDT expiration date for the affected reactor vessels.

I I

i ii Babcock &WHeem ,

a McDermott company

l TABLE OF CONTENTS eaan

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 i
2.

SUMMARY

........................... 2-1 l

l

3. BASIS OF INPUT DATA ..................... 3-1 1

I 3.1 Fluence Estimates . . . . . . . . . . . . . . . . . . . . 3-1 3.2 Chemical Compositions . . . . . . . . . . . . . . . . . . 3-3 3.3 Mater i al P ro pe rt i es . . . . . . . . . . . . . . . . . . . 3-3 1

4. REACTOR VESSEL SPECIFIC CALCULATIONS . . . . . . . . . . . . . 4-1 41 Pressurized Thermal Shock . . . . . . . . . . . . . . . . 4-1
  • . ' .yp ical Wel d Metal . . . . . . . . . . . . . . . . . . .

. 4-2

5. CERTIFICATION ........................ 5-1 i
6. REF E RE NCE S . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 l

APPENDIX i

A. Evaluation Of Atypical Weld Metal .............. A-1 11i Babseek&WWIfees a McDermotI company

LIST OF TABLES J

eno 2-1 Summary Of Pressurized Thermal Shock Evaluations l For B&W Owners Group Reactor Pressuro Vessels ........ 2-3 l

f 3-1 Reactor Pressure Vossol Core Loading Schemos L and Basos For Fluence Estimates ............... 3-4 4-1 Evaluation Of Oconoo Unit 1 Reactor Pressure Vessel In Accordance With Pressurized Thermal Shock Criterion . . . . 4-3 4-2 Evaluation Of Oconee Unit 2 Roactor Pressure Vossel In Accordance With Pressurized Thormal Shock Critorion . . . . 4-4 l

I 4-3 Evaluation Of Oconeo Unit 3 Reactor Pressure Vessel In Accordance With Pressurized Thermal Shock Critorion . . . . 4-5 4-4 Evaluation Of Throo Milo Island Unit 1 Reactor Pressure Vessel l

In Accordance With Pressurized Thormal Shock Critorion . . . . 4-6 4-5 Evaluation Of Crystal River Unit 3 Reactor Pressure Vessel In Accordance With Prossurized Thermal Shock Critorion . . . . 4-7 4-6 Evaluation Of Arkansas Nuclear One-Unit 1 Reactor Pressuro Vossol In Accordance With Pressurized Thormal Shock Critorion ....................... 4-8 4-7 Evaluation Of Rancho Seco Roactor Prossuro Vessel l In Accordanco With Pressurized Thormal Shock Critorion . . . . 4-9  !

l 4-8 Evaluation Of Davis Dosso 1 Reactor Pressure Vossol In Accordance With Pressurized Thormal Shock Critorion . . . . 4-10 )

4-9 Evaluation Of Reactor Pressuro Vossols With Atypical Wold Metal ..................... 4-11 $

iv Babcock &Wilcom  ;

a McDermott company J

k

(

l. INTRODUCTION

(

The Nuclear Regulatory Commission (NRC) amended its regulations for light

( water nuclear power plants, ef fective July 23, 1985,1 to (1) establish a scrooning criterion related to the fracture resistanco of pressurized water reactor (PWR) vessels during pressurized thermal shock (PTS) events; (2) require analyses and schedule for implementation of flux reduction programs that are reasonably practicable to avoid excooding the screening critorion; and (3) require detailed safety evaluations to be performed beforo plant operation beyond the scrooning criterion will be considered.

Those amendments are intended to produce an improvement in the safety of PWR vossols by identifying those corrective actions that may be required to

( provent or mitigato potential PTS events.

Transients and accidents can be postulated to occur in pressurized water reactors (PWRs) that result in severo overcooling (thermal shock) of tho l reactor vossol concurrent with high pressure. In those pressurized thermal shock (PTS) events, rapid cooling of the reactor vessel internal surface causes a temperaturo distribution across the reactor vessel wall.

This temperaturo distribution produces a thermal stress on the reactor vessel with a maximum tensile stress at the insido surface of the vossol.

The magnitudo of the thermal stress varios with the rate of change of temperature and with timo during the transient, and its offect is compounded by coincident pressure stressos.

Sovero reactor system overcooling ovents with pressurization of the

{

reactor vossol (PTS ovents) are postulated to result from a variety of causos. Those includo systun transients, some of which are initiated by instrumentation and control system malfunctions (including stuck open valvos in either the primary or secondary system), and postul ated accidents such as small break los s-o f-co ol an t accidents, main steam line breaks, and foodwater line breaks. As long as the fracture resistanco of the reactor vossol material is relatively high, thoso events are not expected to causo vossol failuro. Howevor, the fracturo resistance of the roactor vessel material docreases with the intograted exposure to fast neutrons during the life of a nuclear power plant. The rate of decrease is 1-1 Babcock & Wilcom A M(f>f fmott (ompJtly

dependent on the chemical composition of the vessel wall and weld materials. If the fracture resistance of the vessel is reduced suf ficiently by neutron irradiation, severe PTS events could cause small flaws that might exist near the inner surface to propagate into the vessel wal l . The assumed initial flaw might be enlarged into a crack through the vessel wall of suf ficient extent to threaten vessel integrity and, therefore, core cooling capability.

The toughness state of reactor vessel materials can be characterized by a " reference temperature for nil ductility transition" (RTNDT). At normal operating temperatures, vessel materials are quite tough and resistant to crack propagation. As the temperature decreases, the metal =

gradually loses toughness over a temperature range of about 100 0F. RT NDT is a measure of the temperature range at which this toughness transition occurs. Its value depends on the specific material in the vessel wall and the integrated neutron irradiation received by the vessel. These ef fects are determined by destructive tests of material specimens. Correl ations, based on tests of irradiated specimens, have been developed to calculate NOT as a function of neutron fluence for various material the shift in RT compositions. The value of RT at a given time in a vessel's life NDT is used in fracture mechanics calculations to determine whether assumed pre-existing flaws would propagate when the vessel is subjected to overcooling events.

On the basis of studies of severe overcooling events that have occurred, generic calculations of postulated PTS events that could occur, and vessel integrity calculations, the NRC concluded that a value of RT can be NOT selected so that the risk from PTS events for reactor vessels with smaller RT NDT values is acceptable. (The risk of vessels with higher values of RT NDT might also be shown to be acceptable but the demonstration would requiro detailed plant-specific ovaluations and possibly modifications to existing equipment, systems, and proceduros.) The NRC approach to selection of tho RT NDT scrooning critorion is described in detail in SECY-82-465.2 In summary, the approach was to uso a deterministic fracture mechanics algorithm to calculato tho value of RT NDT f r which assumed pro-existing flaws in the reactor vossol would be predicted to initiato (grew doopor 1-2 Babcock & WHees a Mc DermotI company

s s

into the vessel wall) assuming occurrence of one of the severe overcooling L

events that have been experienced. These " critical" values of RT NDT were related to the expected frequency of the experienced severe overcooling events based on a limited data base, consisting of eight events in 350

{

reactor-years. -

The designation RTPTS (Reference temperature for pressurized thermal shock) is the nil ductility temperature of the material as defined by 10CRF50.61, Paragraph (b)(2) for use as a screening criterion. This designation is g

) used to avoid confusion with the RTNOT used to characterize the toughness state of reactor pressure vessel materials.

On the basis of these studies, the NRC concluded that the PWR reactor pressure vessels with conservatively calculated values of RT PTS less than 2700 F for plate and forging material and axial welds, and less than 300 F0 for circumferential welds present an acceptably low risk of vessel failure from PTS events.

The requirements of 10 CFR 50.61 further state the following:

"For each pressurized water nuclear power reactor for which an operating Itcense has been issued, the licensee shall submit projected values of RT (at the inner vessel surface) of reactor vessel PTS beltline materials by giving values from the time of submittal to the I expiration date of the operating Itcense. The assessment must spect fy l the bases for the projection, including the assumptions regarding core loading patterns. This assessment must be submitted by January 23, 1986 and must be updated whenever changes in core loadings, surveillance measurements, or other information indicate a significant change in projected values."

1-3 BabsectsaWIlses a woeemost company

2.

SUMMARY

)

L Table 2-1 provides a summary of the pressurized thermal shock ovaluations as required by 10 CFR 50.61 for each of the B&W Owners Group reactor

(

pressure vessels.

A brief description of the status of each reactor pressure vessel is as follows:

2.1 Oconeo Nuclear Station Unit 1 - (Table 4-1) Projected values of RT PTS i

( for all materials in the reactor vessel beltlino region are below tho l scrooning critoria at the expiration date of operating licenso.

2.2 Oconoo Nuclear Station Unit 2 - (Table 4-2) Projected values of RT PTS for all materials in the reactor vessel boltline region are below the scrooning critoria at the expiration date of operating license.

2.3 Oconoo Nuclear Station Unit 3 - (Tablo 4-3) Projected values of RT PTS for all materials in the reactor vossol boltlino region are below the scrooning critoria at the expiration date of operating licenso.

2.4 Throo Mile Island Unit 1 - (Table 4-4) Projected values of RT PTS II all matorials in the roactor vessel boltline region moot the scrooning critoria at the expiration dato of operating license.

(

2.5 Crystal River Unit 3 - (Tablo 4-5) Projected values of RT f r all PTS matorials in the reactor vossol boltlino region are below the scrooning critoria at the expiration dato of operating liconso.

2.6 Arkansas Nuclear One, Unit 1 - (Tablo 4-6) Projected values of RT PTS for all materials in the reactor vessel beltlino region are below the scrooning critoria at tho expiration dato of operating licenso.

(

2.7 Rancho Soco Unit 1 - (Table 4-7) Projected values of RT PTS for all materials in the reactor vessel boltlino region are below the scrooning

{

critoria at tno expiration dato of operating liconso.

2-1 Babcock & WIfcom A M(Dermott company

~

\

2.8 Davis Besse Unit 1 - (Table 4-8) Projected values of RT f r all PTS materials in the reactor vessel beltline region are below the screening criteria at the expiration date of operating license.

l l

2.9 The evaluatlon of the atypical weld metal for the three reactor pressure vessels required to be evaluated by the procedure established by the NRC showed that the projected values of RTET are below the screening criteria at the expiration date of the operating licenses.

i 2-2 NatteestsaWHess A W DermotI(ompetty

~ _

TABLE 2-1. SumARY OF PRESSURIZED THE184AL SHOCK EVALUATIONS FOR Bar OstERS GR(1P REACTGt PRESSimE VES$ELS Current License RT at Current Calendar Year Plant - - Fwnt ratton Date Ltc E Fwntration Criteria Exceeded # Reference Oconee Eclear Station Unit 1 November 6, 2007 231 vs. 270 2032 Table 4-1 Oconee kclear Station, Unit 2 November 6, 2007 292 vs. 300 2011 Table 4-2 Oconee Nclear Station, Unit 3 November 6, 2007 220 vs. 300 2100 Table 4-3 Three Mile Island Unit 1 May 18, 2006 270 vs. 270 2006 Table 4-4 Crystal River Unit 3 September 25, 2008 267 vs. 300 2031 Table 4-5 Arkansas helear One Unit 1 December 6, 2008 251 vs. 300 2053 Table 4-6 Rancho Seco Unit 1 October 11, 2008 265 vs. 270 2012 Table 4-7 Davis Besse Unit 1 March 24, 2011 217 vs. 300 2126 Table 4-8

  1. Assumes 0.80 ut111zation factor (EFPY/ Calender Year) and no change in future fuel cycles.

e, l E l

2 5D 2

l

3. BASIS OF INPUT DATA The pressurized thermal shock regulations require that the data used to perform the specified calculations must be traceable by including the source of all values included in the assossment. The relationship of the material on which any measurements are made to the actual material i in the reactor pressure vessel must be described. For the fluence values, the assessment must specify the bases for all projections including the assumptions regarding core loading patterns such as standard vs. low-leakage f cores.

The following describes the sources for all data used to evaluate the B&W Owners Group reactor pressure vessels.

3.1 Fluence Estimmten The integrated reactor vessel surveillance program as described in BAW-1543A, Rovision 23 provides a general description of the interrolationt. hip of the fluence values of the participating reactor pressure vessel. This relationship is the basis for cross referencing of the fluence analysis for reactor vessels without active surveillance dosimetry to those reactor vessels with dosimeters. A summary of this information is presented in Table 3-1.

The maximum, or peak, fluence values for each material at the various locations on the inside surface of each reactor vessel at January 1,1986, license expiration, and 32 EFPY are listed in Tables 4-1 through 4-8.

These values were obtained by first determining the maximum fluence on the r inside surface of the reactor vessel and then multiplying by appropriate azimuthal a'nd axial factors. The maximum fluence values with the exceptions noted below were determined as described in the surveillanco capsulo reports listed in Table 3-1.

( The general analytical method uses fluence values obtained from transport theory calculations through the latest fuel cycle included in a 00T codo analysir. These calculations are normalized to the most recont measured dostmotor results obtained from the corresponding surveillance capsulos.

3-1 Mah a McDermott tonyetty

Fluence values are extended beyond this value by assuming that the flux above 1.0 MeV at the reactor vessel is proportional to the flux above 1.85 MeV in the baf fle region at the edge of the core. This method is described in detail in BAW-1485.4 Baffle flux values for the completed fuel cycles woro used explicitly and the flux value for the last cycle was used to extrapolate futuro cycles.

Exceptions to this approach were mado for Throo Mile Island Unit 1 (TMI-1) and Rancho Soco. The last availablo DOT calculation for TMI-1 is for Cycle

1. The referenced report (BAW-1439) oxtrapolates on the basis of Cycle 1 only. In the analysis used in this evaluation, Cycles 2 through 5 were included using the ratio of PDQ baffle flux values to that in Cycle 1.

Cyclo 6 was assumod to be the first low-loakago core with a baffle flux ratio to Cycle 5 of 0.82. Cycle 7 was assumed to be a second low-loakago core with a baf fle flux ratio to Cycle 5 of 0.72. Those ratios are based on PDQ calculations for similar cores. Cycles 8 and beyond were assumed to be very-low-loakago cores. The ratio of flux values at specific locations for very-low-loakage to low-loakago cores were obtained from a recently completed vossol fluence reduction study.5 This ratio is 0.665 at zero degrees from a major axis (azimuthal angle for critical wold) and was used to adjust TMI-1 Cycle 7 to obtain the fluenco values for Cycles 8 and beyond.

The general method described abovo based on extrapolating fluence values using baf fle flux ratios was used to obtain the fluence for Rancho Soco through Cyclo 7. Cycles 8 and beyond were assumed to be very-low-loakago coros. The ratio of flux at 14 degroos from a major axis (azimuthal anglo for most critical wold) for very-low-loakago to low-loakago cores was obtained from the same fluence reduction study used for TMI-1. A value of 0.661 was obtained at the 14 degroo location and was used to adjust Rancho Soco Cyclo 7 to obtain the fluence for Cycles 8 and beyond.

Spatial factors for specific wold locations within the reactor pressure vessel woro obtained in che following way. Axial factors woro obtainod from BAW-1405 oxcept thoso for Davis Bosso (which was not includod in DAW-1485). The values for Davis Bosso woro obtained by comparing wold 3-2 Embcock & Wilcox A M(Defmott (ompJny

1 locations to those for Oconee Unit 1. Azimuthal factors were obtained irom the most r ont DOT calculation for each reactor vessel except for TMI-1, for whicle the azimuthal factors were obtained from the DOT calculation for Oconoo Unit 1 Cycles 3 through 7. It is believed that those values are more representative of future fuel cycles since DOT calculations are not available for TMI-1 beyond Cycle 1.

3.2. Chemical Comoositions l The bases of the chemical composition of the materials in the beltline 6

region of the reactor vessels is BAW-1820 which is supplemented by the data and information in BAW-17997 (Nonproprietary version of BAW-1500P O ).

3.3 Material Prooerties The bases of the material proporties which represent actual measured proporties of the bolt 11no region materials is BAW-1820. In the cases where the NRC regulations did not provido a generic initial value of RT NOT f r oither SA-533 Grado B plato, or SA-508 Class 2 forging material, the statistical average value of these materials was calculated using the data base presented in BAW-10046P.9 Those values are as follows:

Plate Material, SA-533, Grado B = +1F Forging Material, SA-508, Class 2 = +3F L

3-3 Babcock & Wilcox a McDermott company

t TABLE 3-1. REACTOR PRESSURE VESSEL CORE LOADING SCHEMES AND BASES FOR FLtENCE ESTIMATES J ANUARY 1,1986 Current Conversion To Plant Name Core loading Fuel Cvele LL_ Fuel Cvele Bases of Fluence Estimates Oconee Nuclear Station, Unit 1 Low-Leakage 9 6 Capsule OCl-A (BM-1837)10 Oconee Nuclear Station, Unit 2 Low-Leakage 8 5 Capsule OC2-A (BAW-1699)I1 Oconee Nuclear Station, Unit 3 Low-Leakage 9 6 Capsule OC3-B (BM-1697)12 Three Mile Island Unit 1 Standard 5 6,8* Capsule THIl-E (BAW-1439)I Crystal River Unit 3 Low-Leakage 6 4 Capsule CR3-C (BAW-1898)I#

Y Arkansas Nuclear One, Unit 1 Low-Leakage 7 4 Capsule AN1-A (BAW-1836)15 a

Rancho Seco Low-Leakage 7 4,8* Capsule RSI-D (BAW-1792)16 Davis Besse Low-Leakage 5 5 Capsule TEl-A (BAW-1882)17

  • Assumed Conversion to Very Low-Leakage Fuel Cycle R

5R

?R

!=

iO IE 2=

4OM

4. REACTOR VESSEL SPECIFIC CALCULATIONS 4.1 Pressurized Thermal Shock For the purpose of comparison with the PTS criterion, the value of RT PTS for each of the reactor pressure vessel materials must be calculated as described in the following paragraphs. The calculation must be made for each weld, plate, and forging in the reactor vessel beltline. For each material, the RT is the lower of the results given by Equations 1 and PTS
2. Equation 1 was applicable to all the materials in B&W Owners Group reactor pressure vessels.

0 Equation 1: RT PTS =I+M+[-10+470Cu+350CuNi3f 0

.270 Equation 2: RT PTS =I+M+283f .194

a. "I" means the initial reference temperature of the unirradiated mate-rial measured as defined in the ASME B&PV Code Section III, Paragraph NB-2331. If a measured value is not available, the following generic mean value must be used: 0 0 F for weld made with Lindo 80 flux.
b. "M" means the margin to be added to cover uncertainties in the i

values of initial RT NDT, copper and nickel content, fluence and the calculational procedures. In Equation 1, M=48 F if a measured value of I was used and M=59 F if the generic mean value of I was used.

i

c. "Cu" and "N1" mean the best estimate weight porcent of copper and

! nicko1 in the material.

L

d. "f" means the best estimate ncutron fluence, in units of 10 19 n/cm2 l

(E greator than or equal to 1 MeV), at the clad-base metal interface 1

on the insido surface of the vossol at the location where the material in question roccives the highest fluence for the period of service considered.

f

\

The results of the reactor pressure vossol specific PTS calculations using Equation 1 and the data sources described in Section 3 which moot the L

4-1 Embcock & Wilcox A McDermott company

requirements as described in 10CFR50.61 are included in Tables 4.1 through l 4.8. '

l

[ 4.2 Atvofcal Wald Metal The NRC letter of December 12, 1979 transmitting the staff review of i BAW-10144A,18 " Evaluation of the Atypical Weldment," stated in part:

" ... conclude that in calculating pressure-temperature operating limits for these vessels, the properties of atypical material should be considered."

l l The effect of pressurized thermal shock on reactor pressure vessel integrity I

was not an issue at the time. The NRC letter further stated: "... the probability that atypical weld metal was used in fabricating the subject vessel is very low." However, any evaluation of reactor vessel integrity involving the mechanistic approach using fracture mechanics dictates that the properties of the atypical weld metal be evaluated for those locations which contain weld metal designated as WF-70.

The NRC evaluation of the atypical weld metal determined that the material was uniquely different from other materials used in the fabrication of reactor pressure vessels. The limited irradiation data then available indicated that the material would also exhibit a uniquely dif ferent response to neutron irradiation. Therefore, the NRC evaluation prescribed t a procedure for evaluating the response of the atypical weld metal to I neutron radiation damage. This procedure is based on Regulatory Guide 1.99, Revision 1,19 however, the procedure is adopted by Regulatory Guide 1.99, Revision 220 for those cases where actual irradiation results are available for a given material. The available surveillance data for the atypical weld metal and the evaluation of change in RT NDT resulting from

]

neutron radiation are presented in Appendix A.

i An evaluation of the atypical weld metal in the three reactor pressure vessels which have weld metal WF-70 at the inside surface in the beltline region (i.e. Crystal River 3, Three Mile Island Unit 1, Rancho Seco) was performed for the neutron exposure periods designated by 10CFR50.61 using the formulation specified for the atypical wel d metal . The results of f these evaluations are presented in Table 4-9.

4-2 Babcock M 8com a McDermott company

e r

_tm

,e

_E 3 01 4 68 23 A 0241 91 1 T'e R e 2 4 3 1 2 2 5 1 1 1 1 1 1 1 2W10 0 2 2 3 3 2222 n

d e e t t t a L t

s N e Y O t P o F 8 4 80 0 0 0 1 A 03 3 8 00 _

I R

C E 2 4 3 1 3 3 5 1 1 1 1 1 1 1 2W1 2 9 832222 344 _

E T )

2 3 -

I 8 5

M 2 m

e n e t t K tc n r C s/ S e e 000000 00A 000000 07 7 7 7 7 O

H E n T et P r i c r 7 7 7 7 7 7 0 2222 223 3W3 0

2222 2 S t9 SC s1 _

L eE _

A B0 o _

( r N

I 2 t E 2 e m

0 0 8 0 8 00 0 0 0 0 0 80 1 1 1 1 1 1 1 1 A 1 1 1 1 1 1 H 3 t E E E E E E E E E E E E E E E T >:

a /

e n e s

550 00 1 70 0 901 7 0 5 3.W 0557 0 4 00 05 1 1 D t n n e

F  : r c 7 ae .

e i T f t c L ri n R

t er e u

S tC l Y S i F P 80 80 0 0 80 0 0 0 0 0 8 F 1 1 1 1 1 1 1 1 A 1 1 1 1 1 1 E rg e E E E E E E EE E E E E E E E R

P Cn i f

c a 2 3

98222 1 0 1 20 1 1 1 1 0

0. W 2 5 0 0. ?. 7 1 0 7 990 gn r H ne u ie S T

I nr e W ec eS d

t 0

0 s 0 7 0 0 0 0 07 0 0 0 0 0 0 8 E r I n 1 1 1 1 1 1 1 1 1 A 1 1 1 1 1 1 E E E E E E E E E E E E E C ct n N Sa o 7 7 000 0 0 7. W E 001 e. 0 0 32223 3 A J 5233 3352 D te 1 R ac O n C Ye n

C Pu F A Fl .

f e 9 90 99 00 00 0 0 0 080 EF e r 55555555 55555 5 5 N r n o o e M I f t

. . t L 34 e e t t E n u S a T * * * * *

  • S t lc N ] l ) ) ))

. E s a t 3 1 1 1 1 1 nC n + + + + + + 000000000 V 7 o l([{( (

0 C _I E 0 _Nn R 2 T RI U

S S 6 E ot R r /e P e sh 5000 3 3 4 1 0 0 0 4 5 5 5 b , tc 0 50 58050 0558555 R m6 t e nN 00000000000000 0 O e4 el T v te t t

C o s s r

_ A N : e i ee a

E R t Dm ee eo s 0 52 01 1 56 5 1 1 999 9 1 1 1 1 1 1 223222222

r CC 0O000 000O00O000 1 ee tt T ai . o I Dr d t N C 2 a t M e U n e e

C d n

og .

E in w . 8 i T 0 2 L ti K an 00 * " "

  • 5 3 A A V""***"**

S 0 re S 8 A

- C ie O pr xc n e

F ES 21 1 1 259 5 3 3 00 O t t r 4 - - - - - 3 2 0 7 0 3 2 N

eo st i e

t b e 5 7 8 5 001 2 55 0 7 0 0 0 1 1 2 1 91 1 1 1 0 4 4 4 O

I n

es r

c s H N ee o

R I

1 2 2 0 0 - - - - - - - - -

MGC3 C 3 C 2 C2ASASFWS A FWS A ASS A AS T

A cr i a D e 3

- 3 U L e t e hp L Y l A g r ] ) a V nr t e

]S1 1 5 9 ]

r e

E ia e 0 0 3] 5 s td M 0 5) 0] ) l r an n 1 . .0 10 5 5 5 e

. re i o ( 000 100 1 11 00 001 01 0 P 1 el t r

- pa a d1 1 [l o c e p 4 OC t L o

t t t

teIl nd

[0((

o t lWe del edl del d e e W .ldo dleW o e F

I . . e n s e teh S S sW u

W .

W t. W WW t s

B 12 et e . . . . . e A W ne e e r s mer m.tmot t t t T rR t t t t t t t a at t t t C eu e e n1 e g tee t t e

e z i e o e e ih r

. t tri Ci r L n oonon s

t e s n ei el od N eE S hh ss teCC r rm r r r r r r r e

C . L L L r rm r r r s

e u

R t t

e e e e e e e e edl e e e e e st p p s st p pd wt p s w l

e o a n p p a e n p pi o n p a o V B lI uU L LI U U M L I U L L '

,W fi Dq3ygM

- seio:B32J

- ijfj1[i

TABLE 4.7. EVALUATION OF OCONEE UNIT 2 REACTOR PRESSURE VESSEL IN AcmRDANCE WITH PREteJ1Rf 7FD THEM4AL SHOCK mITERION

1. Operating License Expiration Date: November 6, 2007 3. EFPY at Screening Criteria: 29
2. Calendar Years to Screening Criteria: 25 4. Fluence at Screening Criteria: 1.1E19 n/cm 2 Metertet Desertatten Dies t eet Cenetente for PTS Reacter Vesset Heat Ceaseeftlene e/o NT Calculottene. F Inalde surface Fluence. n/nes screening 8etttlee Renten Locetton Bhaber Twee Cooper Nicket I et NT Mernip 1 Jan 1900 32 EFPY Licanes Eastro Criterie _ cateut eted RT__'as_t ro 32 EFPY L11e_npe d sA50s, Ct2 0.00 0.76 l+31' 59 2.4E18 9.1E18 7.1 E10 270 98 94 Laser wante sett moo 7 Upper Shott A4F183
  • 0.04 0.75 +10 48 3.2E10 1.2E10 9.4E10 270 79 77
  • 0.02 0.08 +20 48 3.2E18 1.2E10 0.4E10 270 73 73 Leser Shett AAW184 Upper C1rce. Weld 11005) WF-154 ASA/Linde 80 0.31 0.58 0 50 2.4E18 9.1E18 7.1E10 300 254 241
  • 0.35 0.87 0 50 3.2E18 1.2E10 9.4E10 300 300 292 Middle Circa. Weld (1005) WF-95
  • 0.31 0.50 0 58 1.SE18 8.7E18 5.3E18 300 111 100 (mer Circia. Weld (1005) WF-112 "vetwee setteeted per peregraph 3.3 g

s b

.I sr eg 58r ED l

I$

2 a8M

r 6

l l

"8 n'; E88EEi8 i :

5 3> 3 S

! 3-E 333aR*E

= 2=k r _

w i oc o lnu l U .N i

u o

-5

.s F .I L g we  :

" "* E!E !!!IIII e 8a g a x.

~

Y c

^E - a =

2 2:25*.!

=ggggh:

g R E. ) [. 4J44J 4

- -e B e' 4 3

$ "u w um l g og - ggggg,g oc g ggggg>=

5 58 cu s  ;;;;; 4 83 6

a 3

: : : : <k u=

..e..

e. e g 44444 4

>- o I' O. 3 m-z

. . ,. "##2 **

d m* 8

> -  !. 3E5

  • Ea,: ass ====

y 8 3

w m e

- #2 sl -

6 G @

n, -

5 5 l '  : ess==

  • ddddddd W

y

e. zE I

m

- S.h -

Sh 848;;88 . .....

M Qb z C o

2 O I E p

,m g a_

u g

$$ 9 we au MQ l I' . . }: . . .

g c

@O 5 c"  !- lasE;e,I 3

Q u m

- 8

= a

.-liini sa 4

> c6  %

w =

7 u

2; ec 1

m .- . .

4 &d 2 Ea..as *

:- : 1

. .  : .32, Q "n 2 -ili a,,. 1

,_ t= 's  :  :

2- .

lee.g ee 1

.a -

3;8:ss: =..= -

a: - ss: 8 1  : 'E l l e astrs a

- r f 4-5 Babcock &WHeos a McDermott company

e

  • l l

l l

.a

~

s a s. .s. = .= $ .e xk E k e 3 = 3 .

e s !

w

1 8

y 3

s**R.e*sI'>****

>m m i i

e -

3 t:: e. ..e.e .e e <e<e m.m.

W r

w ou C

u s

~

8 g ". ; j 55555555.E.5 saeaesegwews w .

L

. 1 E to 1  ::::::: a

" U'S m -

j =====s==<=

44444~'44)dh4 mc ,

c@ 6

-. S c6 8 83 a 8 t.

i

~ ========<=.:

W 99894s.4924);

2

..8 .

a CL 3 m m- .

N , sasssssssssa

> m ,.

. . .i l

m: r i l m m

!! ' :. s. :. s. :

jd_j E113E.......

g afI 1 ~

W

$ i i e N esasssssssas l

l -

2_f didid&4&&a&&

l -  :

- m I 5 i x .48".;.99.;9999 8

m

.. u ee g l

gr . .}

s g w ci s 2 .4 -

l N C *** '

X O CD w

U. c" i s,. o c x ~o 2 . .e.

i g

tb we

: 8&aAa.

een ~ ~,-s5 eo  : n Haan 4 46As; (

. a .

8

- c. .

- I

  • v 2 1

- - i n 3 3' g2 w

m c6

- 5 1 .s E.k i e

g si j -lgggggg*ds:sd 1 t

i 4 5* 3 35333*3 .

i s.

ou  : . a=sassa s w

s .g.....  :

g "8 3

. Eo II::: -

R"

"'  : =sessa222:sereegg! u aa . .

55 E i gij::1:: }

s iBE':*.aseR2s'* r i, 4-6 N AMb8 a McDermott company

l ll l, llll)Jllll l i e

r

_t

,o m

_E e 50 1 9 803 05 8 2 0A 7 8t B32 23 1 8 N

T R e r

1 1 1 1 1 2W2 i 222 O d g I e ji R t e L E t T u I c Y l P R ) a F 0 9 9 0 3 3 A 4 8 8 8 1 C 2 4

C E 2

2 8 8 08 1 1 1 1 1 2W21 1 8 03 3 3 222 K 3 C y O t m H sc S S E/ n e n t t L t n r '

A s9 S ee 00 00 90A 0 0 0 0 0 T et 7 7 7 7 7 0 0 N e1 P r i c r 22 22 2 3W3 0 32227 7 7 F BE S C E ( 2 H

T 21 3 e D > : r E a f o 8 8 80 8 8 8 8 0 8 8 7 i s 1 1 1 1 1 1 A 1 1 1 1 1 I  : r t E E E E E E E E E E E E R

t ae it a

c e

/ s 5

2. P. 2 25.W208 57 7 7 7 5 7 4 0 5 S 60 S ri n n S er e

. c E tC e t R i c L n

P rg e Cni u Y

H gn F l

P 0 0 0 8 8 8 8 8 88 8 T F 1 1 1 1 1 1 A 1 1 1 1 1 I ne e E c

E E E E 38 E E E E EE E 4 991 W i e nr a 2 6. B. 8 3. W S.

f 3 7 0 990 7 S 580 8 E ec r u

C eS S N r A ct e D Sa d 8 i 9 R s 9 8 8 8 8 0 8 9 8 8 0 0 O te n 1 1 1 1 1 1 1 A 1 1 1 1 1 I E E E E E E E E E E E C ac n 5 0 0 0 0 5. W 21 0 11 18871 C n o A Ye J 1 2 2 221 Pu 1 N Fl I EF L F e E . . .

t a 9 88 8 8 90 9 8 99 9 S 34 e r 5 4 4 4 4 5 5 5 5 555 S r o

n o E

V f te t M

e e t

E n t u R a t t

e T R

)

U s a 3 000 5 S 8 nC S 0 C o te te +2+ 2- +1 4 0000000 E 0 R 2 P T

_ , R R 5 O 2 T a t C r / ke a 2 4 4 881 3 9 9 55 5 A e 75555885 5555 E b . ic R m n N 000000000000 t o e5 e 3 t4 r t.

p t T e tas e o e r

I S : h c p 0 0 02 280 51 9 9 9 N a C s o ao 1 221 1 2 1 3 3 222 U i CC 0 00000O0 00O0

r R ee E tt 0 V ai 8 28 r I Dr t e R e nC CG d n

e og r , . i L T s3 L A in 0 3 * *

  • V***"**

T ti 5 5 A A S S an S $ A Y re R ie C pr n xc t o

1 21 2 9 -

1 0

F ES t r 4 - - - - 5 9 4 8 O e e 9 4 4 7 7 7 8 05 0 5 N

eo trc t b a

4 4 4 4 1 1 7 1 8 1 1 J 3 3 3 3 - - - - - - -

st e e r4 4 4 4 A F F F F F A O n s H N e

A C C CC S W W W W W S I es D 3

T A

cr ia t 3

e h U Le i p L Y r ) e A g e 1 1) r 0

V E

nr ia N t

o 0 4 80 ) e r

td n o

. . 10 1) 11) 1) ) p e

an t D. 0. 01 001 0101010 00

. re t e

I I [0'( 1 [ [ [ r e

S el c d p

- pa o '

4 OC t L t d d lo dl dl ld dl d e t le leW o e e e e E

s s o n

s e W W . WWWW t s

sat s....

L . . et . . e B 12 W a e at t t t e

t t t cr reircc ig tg ig a t t t t t t t t t A r P ts e T e e n n n s ah e e he e i iC t e s n ee N S mS SCC e C L oL oo t L e e

r r r r r r r el F t e e e e e a edt r r r r s p p w w e pe w p ep es e e t

w e

te e p p o o p pt o p p a V g L U U L L U U N L UU L

  • ban RaKW > er3EOM 8 502$=8a E

_ ll 1 llljl

llllll N

O _d I 37 3 7 3 01 07 0 R T_'e R e 7 3 2 3 3 0 5 01 2 1 1 1 1 1 2 2 2 E n T d e c e

I t t '

t s L O t e

t Y K t e F P

3 4 0 4 0 0 4 0 01 C C E 7 4 24 3 0 0 0 2 3 O 1 1 1 1 22 22 H 2 3

S

)

L 0 A 6 g N 2 a e I m t n r t

E tc S e e 00 000 00 000 H s/ T et P r 7 7 7 7 7 0 0 0 7 7 2222 2 3 3 3 22 T En c r t

D F

7 t9 s1 eE S C e

f B0 r s

t

(

2 t

na 0 0 00 0 0 8 0 0 0 S 2 f 1 1 1 1 1 1 1 1 1 1 E E E E E E E E E E S

3 : 5000 05 0003 E >a m e

/ s R i 0 0 0 00 004 50

r n n P e ae e. ic H it ri c L T n I er e u

W tC l f i

rg F P F

00 0 0 0 0 0 0 00 E e E 1 1 1 1 1 1 1 1 1 1 E E E E E E E E E E C Cni c 41 1 1 1 41 200 N a 2 A gn f r

3 01 1 1 1 0 1 07 0 D ne u R ie S nr e m

C A

ec eS r

ct d

i I

s 0 n

0 1

0 n

8 0 0 8 0 0 0 8 0 0 1 1 1 1 1 1 1 1 1 1 E E E E E E E E E E N Sa o 1 00001 0000 I J 22222221 1 2 te L ac 1 E n S Ye S Pu F e E Fl t V EF . a S 0 0 00 0 0 0 0 0 e r S 4 4 4 4 5 5 5 5 5 r n o E e M

. . fo t R

t 34 e o t

S t S n e c lu T

  • E t l R -

R e a nC t

)3 0 00

+ 1 1 1 000000 P e e [ - - -

t C f G t n

T C

A E

8 0

0 st I

~

R 2 / e a k 0 5 2 5 23 0 4 55 1 ,

. ic 7 655505555 6 n N 000000 0000 T t e I r et N e tct i U b e s r

- m e e e p K

O e7 c6 e

D :

h o C n CC a o o 37 57 5 4 1 500 01 1 1 1 23222 0000000000 -

R a 0 A i 3

E  : r 2 f ee e t e C r d e

r tt e 0, n ai w i M Dr T 03 0 3 * * * / * * *

  • L C 5 6 A n

S A A S A og SS A S in N ti n A an e 1 K re t r 1 221 1 -

3 - - - - 2 2 0 0

R ie t e e 1 04 04 0 1 7 0 0 A pr i t b - 2 1 2 1 1 1 1 1 1 xc r c e e a 6 1 1 1 1 - F- AF

- - F -

F O

N ES eo st D

t s H H e

lY A 22G2F W WS WW 3 3

O n e h p

I es i r e T cr e c A i a t e

0 e

U L

A V

E Le g

nr ia Y M t

t s

e' e

) ]$) ] 1 5 0 5 5 5 0 0 0 00 0 1 0 0 0

[

1 l 111 1 [(

P r

e r

e td c p an e d d d

. re t L e

l dlel o W e e a t dl d e

el e a t 6 n e W.WWW s

- pa et . e . .t . e 4 OC V ae t t eerueml it t

t F r R t t 1 t t t e eg e o e o e a e r n n s t . . t e S h h h h t Cr t f o o e B c n e S E S sC 12 CL L A oi e r r r e o l r r r r r w T R t t ta e ee e edl e e e t r p p s s pd s p s e e a p p a a pi e p a V S N U U L L U M L UL h

C. e eg*e8M e Ee.O23C.

SsI*

I l  ! ll I

i

,f ,l e

r

_i

_a

_'ls oed 1 t 51 53 4 5 A T d i oee2 4 2 ee0

, R a 1 t i 222 1 1 2W d .n e

t t c

L la N u O c I

R 2

m l

e m 52 e4 4 e4 e3 3 eA E

T /

c C E 2

4 2 oee32ts30 1 1 t 22 t o72W 3

I n R

C 8 g

1 K E n a C 9 r O e tn. ei oooooooooooA H

S 7 7 f P r

. t c ir 7 7 7 7 7 D co7 22222 3 3 3 222W7 7 2  : sC L a A i t

a  : r o l ae r E i t t 4 ri s a eeeeeeeeeee 1

T er E 1 1 t l t 1 t t t 1 t E E E E E E E E E E f tC 41 1 1 1 41 4 o. e e. A D

F i

rg /

= s n

e n

57 7 77 57 5 eeeW 7 Cn i ,

e c

f gn e i R c L t

S ne n e

i e u eesessssees S nr mfe.eeseeeseeeeW l t 1 i t i i i i t i i E ec F E E E E E E E E f E R

P eS r e c

1 1 1 1 e. 1 e. 7 s. o. A H

ct S a f a

r a

T u I te S W ac e n d s e

E Ye is o neeeeesesse C Pu n t i t i i i t i 1 i i 1 A N Fl I n

E f E f E E E E E E E A EF a 1 o. e. o. e.1 e. 1 o.77 W O

s J

1 22222a22222 m 34 y F n

. in eeeeeoeeeeee N s r d 4 sd sns6sSSS I r n o n L

fo io t E a S e l t

S n u 6 . c T E 0 t l R V 0 e a nC o oon

t. t as 2eooooooo 2 o - - -

E C f RL 1 7

_i a l

S 1 Nr S

_ E r R e et P b / e o6 mk c

eooeeeee3 3 e3 R t2 e.c. c. s. s. s. s. s.e e. 5. e.

O c mni s oaccaooooooc T O: t.c t C a i A i i t E  : r m t e ee r

R ee h s e s2 2 ooe1 e3 3 s3 tt C a n P. 3 2 2 2 3 2 O ai e e i. 1 1 t. t.

C Dr CG ooooooooocoo E C S n o og i s O in ts H ti .

t er e C an e , e, n N re T v

e3 u A

R ie pr xc e3 = = =

sn s a W======

s F ES es A O

N eo st n

a O n i r 1 2 1 2 3 - - - - 3 4 3 I es t s e e2eoo3 s3 eeoe T cr i t t eesr r 2 t 2 e2rt A

U ia Le r

c e i s H s as 4 cooo - - - - - - -

vazzzr rF F F rF F c c cc uuW nW aW L Y e A g D V nr t s

E ia td an fre 1ssl re t a

3 7 7 2 7 el pa m i is] t ]

- n sosss . .

OC e 4 t t

o o o a o o. o.

ot ooo 1 it r o F e c

1([t(t 1

[

. . a t

a e d ad ae B 12 t t t t.

t. mst t. n t. t t A e i.e s T s n s e ei W .umuuu W no m =

r n t

r t a t u c =it ts i c r c sesg t t t _

e. a t. t. t.h t. e i r n p a n t

c n ai et nsesmc ah e

icia oe o t

e cLLL L r r r r R i c.

p pc. n

c. e. c. s p e ..

e e r. es _

t.

e p p o. pt a p op w L uutL unL uL L L

= o I@ a$8M e En -

1a

( - ll

TARtF 4-8. EVALUATION OF DAVIS RESSE 1 REACTOR PRESStRE VESSEL IN ACmRDANCE WITH PRESSURf 7FD T}iEMAL SHOCK CRITERION

1. Operating License Expiration Date: March 24, 2011 3. EFPY at Screening Criteria >32 (Best Est.2Ill)
2. Calendar Years to Screening Criteria: 140 4. Fluence at Screening Criteria: 5.8E19 n/cm -

Materiet Descristian Osamicet Canetente for PTS asetter vesset Meet Conseettiens e/o R g Calculattana. F Inside Surface Ft- - m/am I Screentog J testated RT-getttene Anator Locetton etaber Twee Ceaser Ilt skal InttInt RT Mornin t.Jon 1900 32 Effy License Eastre Criterta 32 EFPY Lt cense'k!s_t re Norste Sett Ace-903 SA60s. C12 0.04 0.68 60 de 9.0E17 2.7E10 2.1E18 270 111 110 upper Shett AEJ 233 O.04 0.77 +20 48 2.4E18 1.7E19 1.3E10 270 91 SS Lamer shalt

  • 0.02 0.01 48 2.4E18 270 104 104 BCC 241 +50 1.7E10 1.3E19 upper Cf rem. Wold lI.O. 95] WF-232 ASVLinde 80 0.18 0.84 0 58 3.8E17 2.7E10 2.1E18 300 140 136 Upper Circa. Weld (0,0. 915] WF-233
  • 0.29 0.88 0 SS WA WA WA WA WA WA
  • 0.24 0.83 40 2.4E18 Middle Circus. Weld (10D5) WF-182-1 +2 1.7E10 1.3E19 300 230 217
  • 0.18 0.64 0 300 90 Laser Ct esian. Weld (1.0.125) WF-232 SS 1.3E18 9.SE10 7.3E10 82

" 0.29 0.00 0 Laser Ctrcum. Weld (0.0. des) WF-233 68 WA WA WA WA WA WA 4k e

N O

I S. I i

sr 2

$gN

, ,o' 2

8  !

=

- - _ _ _ - ~ _ -

~ _ . - -

- _ -- w TAFu F 4.9. EVALUATION OF REACTOR PRESSURE VESSEL S WITH ATYPICAL WEL D ETAL Inside Surface Fluence, n/cm 2

  • Reactor Vessel Calculated RT Plant Beltline Region Location 32 EFPY License Exofre 32 EFPY LicenseYxofre Crystal River Middle Circum. Weld (1005) 9.6E18 7.2E18 263 252 Three Mile Island Unit 1 Upper Circum. Wold (100%) 7.5E18 5.3E18 253 242 l Rancho Seco Lower Longit. Weld (I.D. 73%) 8.8E18 6.9E18 259 251 Intial Value RT = 90F l Radiation Induck Shift = Per Appendix A p Margin for Uncertainties = 48F C

n 5D E$

IiN

l 1

(

5. CERTIFICATION This report is an accurate description of the pressurized thermal shock evaluations of the Babcock & Wilcox Owners Group reactor pressure vessels in accordance with 10 CFR 50.61.

A8 7

/W E E /d ub'n/ffd A. T! L' owe,"J r. , P.g/ Date Project Technical Manager W - in lO, lf[$b N. L. Snidow  ;/ Date Performance Analysis This report has been reviewed and is an accurate description of the pressurized thermal shock evaluations in accordance with 10 CFR 50.61.

()

, E >~Nl i L'. . Gloss, P.E. '

V Date Ch istry, Materials &

Structural Analysis I

l .10./9S4 C_ G. F. Malan // Date Performance Analysis 5-1 h M8com a McDermott company

6. REFERENCES
1. U.S. Code of Federal Regul atio ns. Title 10, Energy, Part 50. Section 61a. " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," First Published - Federal Register, Vol . 50, No. 141, J ul y 23, 1985.
2. U.S. Nuclear Regulatory Commission Staff Evaluation of Pressurized Thermal Shock, SECY-82-465, November 1982.
3. A. L. Lowe, J r. , et al., Integrated Reactor Vessel Material Surveil-lance Program, BAW-1543A. Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, May 1985.
4. C. L. Whitmarsh, Pressure Vessel Fluence Analysis for 177-FA Reactors, BAW-1485, Babcock & Wilcox, Lynchburg, Virginia, June 1978.
5. J. R. Rodes, Vessel Fluence Reduction Fuel Cycle Study, BAW-1884, Babcock & Wilcox, Lynchburg, Virginia, December 1985.
6. J. D. Aadland, Babcock & Wilcox Owners' Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information, BAW-1870, Babcock & Wilcox, Lynchburg, Virginia, December 1984.
7. K. E. Moore and A. S. Heller, B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study, BAW-1799, Babcock & Wilcox, Lynchburg, Virginia, July 1983.
8. K. E. Moore and A. S. Heller, Chemistry of 177-FA B&W Owners Group Reactor Yessel Beltline Welds, BAW-1500P, Babcock & Wilcox, Lynchburg, Virginia, September 1978.
9. H. S. Palme, H. W. Behnke, and W. J. Keyworth, Methods of Compliance With Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G, BAW-10046P. Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, March 1976.

6-1 BakockM8com a McDermott company

(

(

10. A. L. Lowe, Jr., et al . , Analysis of Capsule OCl-A From Duke Power  ;

, Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program, Revision 1, BAW-1837, Babcock & Wilcox, Lynchburg, Virginia, August  !

1984.

1

11. A. L. Lowe, J r. , et al., Analysis of Capsule OCII-A From Duke Power Company Oconee Nuclear Station, Unit 2, Reactor Vessel Material (

i Surveillance Program, BAW-1699, Babcock 7 Wilcox, Lynchburg, Virginia, Decerber 1981.

12. A. L. Lowe, J r., ej;A, Analysis of Capsule OCIII-A From Duke Power i

Company Oconee Nuclear Station, Unit 3, Reactor Vessel Materials Surveillance Program, BAW-1697, Babcock & Wilcox, Lynchburg, Virginia, October 1981. )

j 13. A. L. Lowe, J r., et al. , Analysis of Capsule TMI-1E From Metropolitan s Edison Company Three Mile Island Nuclear Station - Unit 1, Reactor f {

Vessel Materials Surveillance Program, BAW-1439, Babcock & Wilcox, Lynchburg, Virginia, January 1977.

14. A. L. Lowe, J r. , et al., Analyses of Capsule CR3-C Florida Power Corporation Crystal River Unit 3, Reactor Vessel Materials Surveillance Program, BAW-1898, Babcock & Wilcox, Lynchburg, Virginia, January 1986.
15. A. L. Lowe, Jr., et al ., Analysis of Capsule AN1-A From Arkansas Powe:-

& Light Company Arkansas Nuclear One - Unit 1, Reactor Vessel Materials i Surveillance Program, BAW-1836, Babcock & Wilcox, Lynchburg, Virginia, l July 1984.

i

16. A. L. Lowe, Jr., et al., Analyses of Capsule RSl-D Sacramento Municipal i utility District Rancho Seco Unit 1, Reactor Vessel Materials Surveil-lance Program, BAW-1792, Babcock & Wilcox, Lynchburg, Vi rginia, October 1983.

1

(

l t

l 6-2 Babcock &WIIcom )

a McDermott company /

l

17. A. L. Lowe, J r. , et al . , Analyses of Capsule TEl-A The Toledo Edison Company Davis-Besse Nuclear Power Station Unit 1, Reactor Vessel i

j Materials Surveillance Program, BAW-18fl2, Babcock & Wilcox, Lynchburg, Virginia, September 1985.

18. K. E. Moore, et al ., Evaluation of the Atypical Weldment, BAW-10144A, l Babcock & Wilcox, Lynchburg, Virginia, February 1980.

I j 19. U.S. Nuclear Regulatory Commission, Effect of Residual Elements on I Predicted Radiation Damage to Reactor Vessels, Regulatory Guide 1.99.

Revision 1, April 1977.

I

20. U.S. Nuclear Regulatory Commission, Radiation Damage to Reactor Vessel

! Material, Draft Regulatorv Guide 1.99. Revision 2, August 14, 1985.

) .

21. A. L. Lowe, J r. , et al., Analyses of Capsule CR3-B Florida Power Corporation Crystal River Unit 3, Reactor Vessel Materials Surveillance f

Program, BAW-1679. Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, June {

! 1982. {

(

22. A. L. Lowe, J r. , et al., Analyses of Capsule CR3-D Florida Power Corporation Crystal River Unit 3, Reactor Vessel Materials Surveillance '

Program, BAW-1899, Babcock & Wilcox, Lynchburg, Virginia, To Be Published.

l l

1

( l I

l t

6-3 hock &Mcom a McDermott company

I i

APPENDIX A Evaluation Of Atypical Weld Metal A-1 Babcock &Wilcox a McDermott company

1. Introduction During 1978, B&W initiated work with the B&W Owners Group on a program for evaluating the material properties of "early vintage" 177-fuel assembly reactor vessel welds. One of the work phases in this program had the obj ective of characterizing the chemistry of reactor vessel beltline welds. Extensive chemical analyses of the archive sources of reactor vessel welds have been performed as part of this work. Two samples of test  !

(

weldments made for the Crystal River 3 reactor vessel surveillance program were part of the weld metal archives subjected to chemical analysis. The j results of these anal yses, performed by the Mt. Vernon Works Quality Assurance Laboratory, indicatod that one of these samples had atypical i concentrations of nickel and silicon, while the concentrations of the other i elements were in the normal range for MnMoNi:Linde 80 submerged-arc weldments. The other sample had the nominal chemistry. The atypical weld ,

was made with weld wire designated by Heat Number 72105. This heat of weld wire was used in the fabrication of 12 reactor vessels.

2. Regulatorv Position To resolve the atypical weld issue, B&W conducted an extensive investigation of records, metallographic examinations, chemical analyses, and fracture mechanics tests on both unf rradiated and irradiated atypical weld material.

The results of this study are presented in BAW-10144A.

Charpy V-notch tests were performed on both unirradiated and irradiated material. The irradiated specimens were irradiated in the Crystal River 3 reactor vessel. Dynamic and static fracture toughness tests were conducted on one inch thick compact tension specimens at room temperature. Although the dropweight NDT is -20 0F, the results of the Charpy tests show that 50 ft-lbs of energy is absorbed at 150 F, therefore the unirradiated value of RT is 900 F (150 minus 60). Using RTNDT equal to 90 0 F, the toughness NOT I

properties obtained from the fracture mechanics tests, KIc (static) and kid k curve (dynamic), were found to be conservative (i.e., lie above) to the K in ASME B&PV Code,Section XI and the K curve in ASME B&PV Code, Section IR III respectively. Using an RT of -20 F (the dropweight NDT), the NOT fracture mechanics data fall within the scatter of the data of normal A-2 N M3888 a McDermott company

_ _ _ _ _ _ _ . _ _ . . _ _ _ . . . _ I

/

s material used to obtain the K k and K IR curves. This indicates that the f RTNDT value of 90 F is conservative.

The effect of irradiation on the mechanical properties of atypical material

{

have been evaluated, using the test results obtained from the Crystal River 3 surveillance specimens. These specimens were subjected to a fluence of 1.17 x 10 18 n/cm2 . This fluence produced an increase in RT of 35 F.

NDT

( The NRC staf f concluded that the probability that atypical weld metal was used in fabricating any of tho vessels is very low. However, they also f concluded that in calculating pressure-temperature operating limits for these vessels, the properties of atypical material should be considered.

It was determined, "... that an initial value of RT f 900 F was a very

{ NDT conservative value. The increase in RTNDT due to irradiation should be l based on the measured value of 35 0 F at a fluence of 1.1 x 10 18 n/cm2 and the damage prediction slopes in Regulatory Guide 1.99."18

3. Evaluation of Atvofcal Weld Metal Since the NRC staff evaluation of the atypical weld metal was completed, two events have occurred which affact any evaluation of the neutron radiation damage response of the atypical weld metal. The first is the revision of Regulatory Guide 1.99 to reflect a better understanding of radiation damage of reactor vessel materials. This warrants a change from the original approach of evaluating the atypical weld metal based on the slope of the damage prediction curves. The same basic technical relation-ship is defined in the revised Regulatory Guide but based on a statistical evaluation of available reactor surveillance capsule data. The second event is the testing and evaluation of two reactor vessel material surveillance capsules containing atypical weld metal. These data, added to the previously reported capsule data, provide a basis for evaluation of the atypical weld metal using the procedure described in Regulatory Guide 1.99, Revision 2.

l The atypical weld metal data from the three available capsules from Crystal River Unit 3 reactor vessel surveillance program are presented in Table A-1.

A-3 Babcock &WHcom a McDermott company

Table A-1. Irradiated Atvolcal Wald Metal Data Capsule Fluence. n/cm2 Reference 4RI@Taf 21 CR3-B 28 1.17E18 BAW-1679 ]

~

1 CR3-C 122 6.56E18 BAW-1898 22 CR3-D 119 7.50E18 BAW-1899 The equation developed for the atypical weld metal using the data in ]

Table A-1 and the procedure described in Regulatory Guide 1.99, Draft J Revision 2, Section C.2 is as follows:

ARTNDT(Surface) = 125.8f(0.28-0.10logf)

(

l i

l l

l I

I i,

i l

A-4 Bh &WWkom a McDermott company

Retrieved from "https://kanterella.com/w/index.php?title=ML20137M863&oldid=2831755"