ML20039F861

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Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for Calvert Cliffs 1 & 2 Reactor Vessels
ML20039F861
Person / Time
Site: Calvert Cliffs, Palisades  
Issue date: 12/31/1981
From:
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML13308A045 List:
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-B, NUDOCS 8201130484
Download: ML20039F861 (44)
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Text

_. _

y ],f AP EN XB bo EVALUATION OF PRESSURIZED THERMAL SH0CK EFFECTS DUE TO SMALL BREAK LOCA'S WITH LOSS OF FEEDWATER

~

FOR THE CALVERT CLIFFS 1 & 2 REACTOR VESSELS Prepared for BALTIMORE GAS AND ELECTRIC COMPANY NUCLEAR M/ER SYSTEMS DIVISION P

POWER Eliiiiil SYSTEMS COMBUSTION ENGl JEERING INC suoi,solgf

I LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF:

A.

MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REX)RT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD; OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF,OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THis REPORT.

i

ABSTRACT This Appendix to CEN-189 provides the plant-specific evaluation of pressurized thermal shock effects due to

-small break LOCA's with extended loss of feedwater for the Calvert Cliffs 1 & 2 reactor vessels.

It is concluded that the crack initiation would not occur for the transients considered for more than 32 effective full power ~ years, which is assumed to represent full plant life.

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CEN-189 Appendix B TABLE OF CONTENTS SECTION TITLE PAGE ABSTRACT Bl.

PURPOSE BI B2.

SCOPE BI B3.

INTRODUCTION B1 84 THERMAL HYDRAULIC ANALYSES BI B5.

FLUENCE DISTRIBUTIONS B5 B6.

MATERIAL PROPERTIES B10 B7.

VESSEL INTEGRITY EVALUATIONS B20 B8.

CONCLUSIONS B39 i

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Bl.0 PURPOSE This Appendix provides the plant-specific evaluation of pressurized thermal shock effects of the SB LOCA + LOFW transients presented in the main body of the CEN-189 report for the Calvert Cliffs 1 & 2 vessels'.

B2.0 SCOPE The scope of this Appendix is limited to the evaluation of the SB LOCA +

LOFW transients presented in CEN-189, as applied to the Calvert Cliffs 1 & 2 reactor vessels.

Other C-E NSSS reactor vessels are reported in separate Appendices.

B

3.0 INTRODUCTION

This Appendix to CEN-189 was prepared by C-E for Baltimore Gas and Electric for their use in responding to Item II.K.2.13 of NUREG-0737 for the Calvert Cliffs 1 & 2 reactor vessels.

This Appendix is intended to be a companion to the CEN-189 report.

The transients evaluated in this Appendix are those reported in Chapter 4.0 of the main report. Chapter B5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter B6 reports the plant-specific material properties and change of properties due to irradiation, based on the methods of Chapter 6.0 of the report.

Chapter B7 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to the material properties discussed in Chapter B6.

I B4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to perfom the plant-specific vessel evaluation reported in this Appendix are those reported in Chapter 4.0 of CEN-189. As discussed in the body of the report, there are several plant parameter conservatisms included in the analyses to develop these transients due to the reference plant approach used which could be eliminated by performing more detailed plant-specific themal-hydraulic system analyses.

Removal of these available conser-vatisms by add"tional analyses was not perfomed due to the favorable conclusion achieved.

B1

B5. Calvert Cliffs Units 1 and 2 Fluence Distribution The fluence distributions for Calvert Cliffs Units 1 and 2 were developed using the methodology described in Chapter 5.

Baltimore Gas and Electric supplied estimates of the cumulative energy output of each unit as of December 31, 1981. These estimates were 4.77 Effective Full Power Years (EFPY) at 2700 Megawatts-thermal (ht) for Unit I and 3.865 EFPY at 2700 Nt for Unit 2. After adjusting for differences in cumulative energy output the flux distributions were assumed to be identical for each unit.

The peak fast neutron flux at the vessel wall is quoted in the report on 2

10 (n/cm -s) as shown on the analysis of the surveillance capsule as 4.7 x 10 Table B5-1. Multiplying this flux by the number of seconds in 4.77 and 3.865 18 (n/cm ) and 5.71 x 10 2

years the year end peak fluence values of 7.05 'x 10 2

(n/cm ) were obtained for Units 1 and 2, respectively. The peak wall fluence data is summarized in Table B5-1.

TABLE B5-1 Peak Vessel Effective Full Peak Fluence Wall Flugnce Full Power Level Power Years Accumulation Rate Unit (n/cm )

(Nt)

(EFPY) r/cm2 per EFPY 1

7.05 x 1018 2700 4.77 1.48 x 1018 2

5.71 x1018 2700 3.865-1.48 x 1018 The azimuthal shape of the fluence distribution was calculated using the SPADRAC code as described in Section 5.2.2.

The reference azimuthal distri-bution was obtained from a D0T-R9 calculation using a Millstone Point-2 model. This distribution was then adjusted using the SPADPAC results to represent an updated Calvert Cliffs Units 1 and 2 azimuthal distribution.

The detailed time averaged power distributions used in the SHADRAC calcu-lations were represented in a nodalized fom as shown in Figures B5-1 and B5-2.

l B2

A time averaged power distribution representing Units 1 and 2 was chosen from a detailed pin power distribution calculated for Unit I during Cycle 5.

This distribution was chosen because the assembly power distribution closely approximated the tine-averaged assently power distributions calculated from initial plant startup to December 31, 1981.

The resulting azimuthal fluence distribution which was applied to Units 1 and 2 is shown in Figure B5-3.

The orientation of the 00 reference point is shown in Figure B5-4.

One eighth core symmetry is assumed.

The axial and radial. fluence distributions in the reactor vessel are obtained for Calvert Cliffs Units 1 and 2 from a DOT-RZ calculation. The resulting axial distribution is shown in Figure B5-5 and the radial dis-tributions are shown in Figure B5-6.

i' The fluence distributions were applied as described in the following sections of this report.

References:

B5-1.

J. S. Perrin et. al., Calvert Cliffs Unit No. I Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule 263, Battelle Columbus Laboratories, September 18, 1980.

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1 APPENDIX B CALVERT CLIFFS UNITS #1 AND #2 B.6 MATERIAL PROPERTIES The methods used to develop and evaluate the materials for the Calvert Cliffs Units di and #2 reactor vessels are described in Section 6.0 in the main body of the report. The chemistry data (nickel, copper, and phosphorus content) and initial (pre-irradiation) toughness properties' of the reactor vessel shell course plates and welds are summarized in Tables B6-1 and B6-2.

In cases where the chemistry exceeded the Regulatory Guide 1.99 prediction limits (0.35% and 0.012% P), those upper limit values were used in the reference te=perature shift calculations.

In cases where the weld metal nickel content was not determined, it was conservatively estimated using information on the type of weld wire used (eg, high MnMo ve. sus MnMcNi wire). For the Calvert Cliffs Unit #1 weld =ents, the weld inspection records and welding certification reports indicated that the longitudinal seam welds could be expected to contain high nickel (greater than 0.30 w/o) since they were fabricated with MnMcNi wire, so the nickel content was conservatively estimated to be 0.99 w/o as indicated in Table B6-1.

For the Calvert Cliffs Unit #2 weldments, nickel content was not determined for weld seam 2-203.

That weld seam could be expected to contain low nickel (less than 0.30 w/o) since it was f abricated with high MnMo wire, so the nickel content was conservatively estimated to be 0.20 w/o as indicated in Table B6-2.

The toughness properties given in Tables B6-1 and B6-2 are the drop weight NDTT (if determined) and the initial reference temperature, RTNDT. For the plate materials, the RTNDT was determined using transversely oriented Charpy impact specimens er by converting longitudinal impact data using Branch Tech-nical Position MTEB 5-2*.

For the weld material, the RTNDT was estimated using the weld qualification test results benchmarked to the surveillance weld for the vessel, as discussed in Section 6.0 and described below.

The individual weld qualificatien test results (three Charpy impact speci-ments tested at +10F) are listed in Tables B6-3 and B6-4.

Each weld for Unit 41 which exhibited an average Charpy energy of 98 ft-lb or greater. (the

  • " Fracture Toughness Requiremen's for Older Plants," U.S. Atomic Energy Commis sio n, Regulatory Standard Review Plan.

B6-1 B10

c average Charpy energy for the surveillance weld at 10F) was considered to be at least as tough assthe surveillance w' eld; i.e.,

that weld seam RT was NDT

-80F or less. For those' Unit #1 weld qualification test results exhibiting an average Charpy energy less than 98 ft-lbl1tne RTNDT was increased by an amount equivalent to the temperature diIference between the average Charpy energy cransition curve for the surveillance weld and the average Charpy energy for the vessel weld test results. The weld qualification rest results for Unit #2 were compared to tihe average Charpy er.crcy for the Unit #2 cur-veillance weld, 93 ft-lb, in a similar manner.

In effect, the temperature at which 50 f t-lb or better exists was iletemined, and the RT _fvr was established

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N at a temperatu,re 60F below that value.

" Maps" of the cylindrical portion of the Calvert Cliffs Units #1 and

  1. 2 reactor vessels are given in riguies B6-1 and B6-2.

They show the locations of the plates and welds listed in Tables B6-1 and B6-2 and their corresponding values of initial RT (F) located within a rectangle on the Figure.

RT values for the vertical weld sear.s (designated 1-203, 2-203, and 3-203) are shown at a cingle seam buc apply to all three vertical seams in a given chell course. Included in the Figure are the locations of the inlet and outlet nozzles, the core midplane, and the extremities of the active core.

Figures B6-3 and B6-4 are mapq of, ddjusted RT values for g

important locations at the inner surface of the Calvert Clif f s Units #1 and #2 vesscis predicted for December 31, 1981. The predictions are o

based on the test estimate neunen fluence, 0.70 5 x 10 n/cm and 0.571 9

x 10 n/cm ( E >lMev ), (corresponding to 4.77 and13.865 effective full power years at peak flux location on the inside surface of the 1

vessel) for Chits #1 and #2, respectively, the initial PS and

(

ecpper, phosphorus, and nickel contents given in Tsles PG-1 and B6-2, and the normalizd neutron flux profiles given in Secticn B.5.

The values of adjusted RT (initial F"NM plus pre C eted shift) are located in j

rectangles ad'.ecent to the plate and weld designations. The cTg valueJ apply to the inner surface of the vessel in the region indicated by a circle. The circled regions generally represent areas of peak I

neutron flux for a given weld seam or plate.

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~UNIl]/lREACTORVESSELMATERIALSs

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ChemicalContentM)

' Product-Material Drop Weight Initial Fom Identification ' - NDTT-('F_)_

RTNOT ( F)

Nickel Cooper.

Phasphorus, b

0.009 l

Plate:

D-7205-1 10 10a O.57 0.12

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Plate ND-7205 10 10a 0.50 0.12b 0.009 8

b

. Plate

'P-7205-3 10' 10 0.54 0.12 0.012J

-l Plate 0-7206-1 20a 0.55 OJ11 0.011 i i

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Plate D-7206-2

-30

-30a 0.64 0.12 0.011 e

c' Plate D-7206-3

-10 10c 0.64

'O.12 0.011s Plate D-7207-1 0

10a 0.54 0.13 0.010 Plat e.

D-7207-2

-10

-10 a 0.56 0.11 0.009 a

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i Plate

.0-7207-3

. -20a 0.53 0.11 0.003 d

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Weid' 1-203 A,B,&C N/A

-30 0.99 0.35f 0.012 Weld 2-203 A,B,aC N/A

-30 0.99 0.35 0.012f d

8 f

d

, eld 3-203 A,B,8C N/A

-50 0.71 0.20 0.016 i

W i

Weld 203 N/A

-60 0.74 0.35 f 0.012.f d

Weld 9-203

-80

-80C 0.13c 0.24 c 0.014c c

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N/A Not Available a

Deterr.ined using Branch Technical Position MTEB 5-2 b

Esti:nated based on average for similar Calvert Cliffs Unit 1 plates having reported analyses

-c Surveillance proqram data

.d Estimated (see text. and Table B6-3) e Estimated Ni content (high nickel type wire) l f

Regulatory Guide 1.99 upper bound prediction limit 5

4 1

TABLE B6-2 CALVERT CLIFFS UNIT #2 REACTOR VESSEL MATERIALS Product Material Drop Weight Initial Chemical Content (%)

Form Ident i fica t ion _

flDTT (*I')

R_TNDT("FJ Ni ck_e l_

CoJger Phosphorus b

Plate D-8905-1

-20

-20a 0.52 0.14 0.009 b

Plate D-8905-2

-20

-20a 0.52 0.14 0.009 b

Plate D-8905-3

-20

-20a 0.61 0.14 0.011 8

Plate D-8906-1

-10 10 0.56 0.15 0.006 a

Plate D-8906-2 10 10 0.56 0.11 0.007 8

Plate D-8906-3

-10 5

0.55 0.14 0.005 a

Plate D-8907-1

-10

-8 0.60 0.15 0.005 c

c Plate D-8907-2 10 20 0.66 0.14 0.005 a

Plate D-8907-3

-20

-10 0.74 0.11 0.006 d

Weld 1-203 A,B,&C N/A

-50 0.71 "

0.20 0.016 d

Weld 2-203 A,B,8C N/A

-50 0.20 0.12 0.018 d

Wald 3-203 A 8,8C N/A

-80 0.18 0.23 0.013 d

Weld 8-203 N/A

-80 0.18 0.30 0.01 3 c

c c

c Weld 9-203

-60

-60c 0.04 0.20 0.016 w

il/A Not Available a

Determined using Branch Technical Position MTEB 5-2 b

Estimated based on average for similar Calvert Cliffs Unit 2 plates having reported analyses c

Surveillance program data d

Estimated (see text and Table B6-4) e Estimated Ni content (low nickel type wire)

TABLE B6-3 CALVERT CLIFFS UNIT,11 REACTOR VESSEL WELD SEAM TOUGllNESS DATA d

Charpy Qualification Test Results Average Energy Estimated Weld Seam at 10*F (ft-lb) at 10 F (ft-lb)

RTNDT ( F) 1-203 A/C 35, 50, 43 44.3

-30 62, 47, 62 57.0

-50 2-203 A/C 35, 50, 43 44.3

-30 62, 47, 62 57.0

-50 3-203 A/C 63, 47, 62 57.0

-50 62, 59, 60 60.3

-50 co E

30, 74, 73 75.7

-50 3-203 102, 102, 103 104.0

-6d3 a

9-203 151, 121, 123 131.7

-30 b

c Surveillance Weld 62, 85 97.5

-30 a Neld seam 9-203 fabricated with sace heat of wire (33A277) and lot of flux (Linde 0091 lot 3922) as surveillance weld, so same RTNDT b Test results at 0*F c Actual RT DT based on drop weiqht and' Charpy test data N

d Estimated using the method described in the text e Conservatively adjusted to higher RTNDT to account for difference from weld seam 9-203 qualification test results

TABLE B6-4 CALVERT CLIFFS UNIT #2-REACTOR VESSEL WELD SEAf1 TOUGilNESS DATA Charpy Qualification Test Results Average Energy Estimatede Weld Seam at 10 F'(f t-lb) at 10*F (ft-lb)

RTNDT (*F) 1-203 A/C 62, 59, 60 60.3

-50 30, 74, 73 75.7

-50 2-203 A/C 66, 75, 73 73.0

-50 a

3-203 A/C 151, 121, 123 131.7

-30 a

8-203 151, 121, 123 1 31.7

-30 9-203 101, 103, 107 105.3

-00 c

d y'

Surveillance Weld 70.5, 78.5, 39.5 93.0

-60 a Same wire heat and flux lot as Calvert Cliffs Unit il surveillance material, so same RTflDT b Weld seam 9-203 fabricated with same heat of wire (10137) and lot of flux (Linde 0091 lot'3995) as surveillance weld, so same RTNDT-c Test results at 0*F d Actual RTNDT based on drop weight and Charpy test data e Estimated using the method described in the text 9

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J B.7.0 Calvert Cliffs 1 & 2 Vessel Integrity The fracture mechanics analysis is performed using the plant specific properties of the Calvert Cliffs 1 & 2 vessels. The attenuation of the peak fluence value is considered in three dimensions (r, z, 0),

and the superposition of the fluence profile and the weld geometry map is used in calculating the predicted RT value at all points in NDT the vessel as a function of Effective Full Power Years (EFPY). This information is used in locating the points in the vessels having the highest RTNDT at each of the three axial sections of interest:

1) middle of core, z 136.35 in.

=

2) top of core, z 66.

in.

=

3) above core, z 40.5 in.

=

i whe're z is the axial distance below the centerline of the nozzle.

From the predicted RTND. values, the material toughness properties' Kgg and K, are determined from the calculated temperatures for the SBLOCA g

+ LOFW transients using the method described in Section 7.6.

Critical crack depth diagrams are constructed from the applied K vs crack depth y

curves and the calculated material toughness curves. By performing the same fracture mechanics analysis a number of times for increasing plant life (EPFY) the integrity of the Calvert Cliffs 1 & 2 vessels for the SBLOCA + LOFW transient is evaluated.

B.7.1.1 Summary of Physics and Materials Data Input to Fracture Mechanics Analysis A detailed survey was performed on the combined fluence and material properties maps of the Calvert Cliffs 1 vessel to detennine the most critical locations in terms of radiation embrittlement. The properties are considered independently at the three axial sections.

At each section, the combination of fluence and materials data were evaluated for a large number of points around the circumference. The adjusted RT values at the inner vessel radius were compared, and the location NOT with the highest RT value was used in the fracture mechanics analysis.

NDT At the mid-core level the location of highest RT ccurs in the NDT weld material at an azimuthal angle of 270 degrees. The fluence factor at this location is1.0 of the peak fluence in the vessel.

B20

The materials data at this point are as follows:

PCT.

Ni

.99

=

.35 PCT.

Cu

=

.012 PCT.

P

=

-20%

Initial RT

=

riDT 19 At the 12/31/81 level of 4.8 EFPY, and peak fluence of.709 x 10 n/cm2 (E > 1 MeV), this corresponds to a point fluence of.709 x 1019 2

0 n/cm and an adjusted surface RT value of 244 F.

riDT At the top of core leve, the location of highest RT occurs in flDT the weld material at an azimuthal angle of 270 degrees. The fluence j

factor at this -location in the vessel is.32 of the peak fluence. The materials data at this point are as follows:

.99 PCT.

Ni

=

.35 PCT.

Cu

=

.012 PCT.

P

=

-20%

Initial RT

=

NDT 19 At the 12/31/81 level of 4.8 EFPY, and peak fluence of.710 x 10 2

19 n/cm (E)1 MeV), this corresponds to a point fluence of.226 x 10 2

n/cm and an adjusted surface RT value of 137%.

t1DT At the above core level (about halfway between the top of core and the inlet nozzle), the location of highest RT occurs in the weld material NDT at an azimuthal angle of 90 degrees. The fluence factor at this point is.003 of the peak fluence in the vessel.

The materials data for this point are as follows:

.99 PCT.

Ni

=

.35 PCT.

Cu

=

.012

PCT, P

=

- 0%

Initial RT

=

riDT L

B21

19 At the level Of 4.8 EFPY, and peak fluence of.710 x 10 n/cm2 (E >l MeV), this corresponds to a point fluence of.002 x 1019 2

n/cm and an adjusted surface RT value of -4 F.

NDT This represents the materials information available at the time of the analysis. Lower initial weld metal RT values were iiDT subsequently justified by additional testing. The use of the present values therefore provides a conservative evaluation of vessel integrity.

B.7.1.2 Sumary of Physics and Paterials Data Input to Fracture Mechanics Analysis A detailed survey was performed on the combined fluence and material properties maps of the Calvert Cliffs 2 vessel to determine the most critical locations in tenns of radiation embrittlement. The properties are onsidered independently at the three axial sections.

At each section, the combination of fluence and materials data were evaluated for a large number of points around the circumference.

The adjusted RT values at the inner vessel radius were compared, NDT and the location with the highest RT value was used in the fracture NDT mechanics analysis.

At the mid-core level, the location of highest RT occurs in the NDT plate material at an azimuthal angle of 90 degrees. The fluence factor at this location is 1.0 of the peak fluence in the vessel.

The materials data at this point are as follows:

PCT.

Ni

.56

=

.15 PCT.

Cu

=

.006 PCT.

P

=

10 F Initial RT

=

NDT 19 At the level of 3.9 EFPY, and peak fluence of.578 x 10 19 n/cm (E >l MeV), this corresponds to a point fluence of.554 x 10 2

n/cm and an adjusted surface RT value of 94 F.

NDT 822

At the top _of core level the location of highest RT occurs in NDT the plate material at an azimuthal angle of 90 degrees. The fluence factor at this location in the vessel is.32 of the peak fluence..The materials data at this point are as follows:

PCT.

Ni

.56

=

PCT.

Cu

.15

=

.006 PCT.

P

=

10 4 Initial RT

=

NDT 19 At the 12/31/81 level of 3.9 EFPY, and peak fluence of.578 x 10 19 n/cm ' (E > 1 MeV), this corresponds to a point fluence of.184 x 10 n/cm and an adjusted surface RT valueof57%.

NDT At the above core level (about halfway between the top of core and the inlet nezzle), the location of highest RT occurs in the weld NDT material at an azimuthal angle of 90 degrees. The fluence factor at this point is.003 of the peak fluence in the vessel. The materials data for this point are :ss follows:

.710 PCT.

Ni

=

.200

PCT, Cu

=

.015 PCT.

P

=

-20D Initial RT

=

NDT 19 At the 12/31/81 level of 3.9 EFPY, and peak fluence of.578 x 10 19 n/cm (E > 1 MeV), this corresponds to a point fluence of.002 x 10 2

n/cm and an adjusted surface RT value of M F.

NDT This represents the best materials information that was available at the time of the analysis.

l.ower initial weld metal RT values a re NOT subsequently justified by additional analysis.

The use of the present values therefore provides a conservative evaluation of vessel integrity.

[

B23

B.7.2.1 Results of Fracture Mechanics Analysis for SBLCCA + LOFW %

Open PORV's (Case 4)

The stress analysis for this case is presented in Section 7.8.1 of the report. The fracture mechanics analyses were performed for this case using the Calvert Cliffs 1 vessel properties and predicted fluence levels up to the, assumed end-of-life condition of 32 EFPY. The critical crack depth diagram at the mid-core level of the vessel for 32 EFPY is given in Figure B.7-1.

For times greater than 65 minutes in the transient, K is calculated to exceed g

the initiation toughness, KIC, for a range of initial flaw sizes.

vs. time shown in Figure 7.14 of th'e However, from the plot of Kg report it is seen that warm-prestressing would occur after 10 minutes is continually decreasing.

in the transient, beyond which time Kg Thus, no crack initiation would occur under these circumstances. The upper shelf toughness line indicates the flaw depths for which Kg=

200 ksiN. This represents the upper limit of applicability for linear elastic fracture mechanics. A ductile failure mechanism would be expected for crack sizes above this limit. The fact that warm-prestressing would preclude crack initiation prevents initially small flaws from extending into that region.

The critical crack depth diagram for the top of core level at 32 EFPY is shown in Figur9 B.7-2.

For this case, also, initial flaws within a certain range of depth are calculated to exceed the level of initiation toughness after 65 minutes in the transient.

From the plot of K vs.

g time for the top of core leve1 in Figure 7.15 it is seen that warm-prestressing occurs after 10 minutes in the transient. Thus, no crack initiation would occur under these conditions at the top of core level of the vessel.

Figure B.7-3 shows the critical crack depth diagram at the above core level of the vessel for 32 EFPY.

It is apparent frem this figure that the calculated stress intensities are below both the initiation and arrest toughness levels, thus there is no potential for brittle crack initiaticn in the vessel above the top of the core for this transient. This is because of the relatively low fluences at this height on the vessel wall.

j B24

B.7.2.2 Results of Fracture Mechanics Analysis for SBLOCA + LOFW ---+

Open PORV's (Case 4)

The stress analysis for this case is presented in Section 7.8.1.

The fracture mechanics analyses were performed using the Calvert Cliffs 2 vessel properties and considering flunece levels up to the end-of-life condition of 32 EFPY. The critical crack depth diagram at the mid-core level of the vessel is shown in Figure B.7-4 for 32 EFPY.

A small region is apparent at 100 minutes in the transient for stress intensities greater than the arrest toughness of the material, however, no initiation region is indicated for these conditions.

The critical crack depth diagrams for the top of core and above the core levels are given in Figures B.7-5 and B.7-6, respectively, for 32 EFPY.

These figures indicate that neither the initiation nor the arrest toughness is exceeded during this transient at these levels of the core throughout the plant life. This is due to the relatively low fluences and corresponding low shifts in the RT values in these areas of the vessel.

NDT These results indicate that no loss of vessel integrity would occur as a result of this type of SBLOCA + LOFW transient during the expected life of the Calvert Cliffs 2 vessel of 32 EFPY.

B.7.3.1 Results of Fracture Mechanics Analysis for SBLOCA + LOFW Restoration of Feedwater (Case 5)

The stress analysis for this transient is presented in Section 7,8.2 of the report.

Fracture mechanics analyses were performed using the Cal-vert Cliffs 1 vessel properties with various levels of accumulated fluence up to the assumed end-of-life condition of 32 EFPY. The critical crack depth diagram at the mid-core level of the vessel for 32 EFPY is given in Figure B.7-7 The calculated stress intensity values exceed the arrest toughness after 70 minutes, and an initiation region is apparent at 92 minutes in the transient. The fact that warm-prestressing occurs for this transient after 78 minutes, as shcwn in the plot of K vs time in Figure 7.17 of the report, indicates y

that crack initiation would not occur under these conditions. The 200 ksiTinlrepresentsthe upper shelf toughness lira for K

=

B25

~

upper limit of applicability of LEFM. A ductile failure mechanism 0 91d be expected for crack sizes above this limit, In this case, warm-prestressing prevents initially small flaws from extending into that range.

The critical crack depth diagram for the top of core level at 32 EFPY is given in Figure B.7-8.

Similarly, the diagram for the above the core level of the vessel at 32 EFPY is shown in Figure B.7-9.

Both of these figures indicate that the initiation toughness level is not exceeded at these locations in the vessel throughout the expected plant life for this transient loading condition.

B.7.3.2 Results of Fracture Mechanics Analysis for SBLOCA + LOFW Restoration of Feedwater (Case 5)

The stress analysis for Case 5 is discussed in Section 7.8.2, Using the results of the stress analysis, the fracture mechanics analyses wera performed using the Calvert Cliffs 2 vessel properties and calculated fluence levels up to the assumed end-of-life condition of 32 EFPY. The critical crack depth diagram considering the stresses and properties at the mid-core level of the vessel for 32 EFPY is shown in Figure B.7-10.

A similar diagram for the top of core level is given in Figure B.7-11, and Figure B.7-12 shows the results at the level of the vessel above the core, Each of these three figures shows the same result that the calculated stress intensities due to this transient do not exceed the initiation or arrest toughness levels at any point in the Calvert Cliffs 2 vessel throughout the expected plant life of 32 EFPY, thus no loss of vessel integrity would result from this type of SBLOCA + LOFW transient.

B.7.4 Conclusion These results demonstrate that the integrity of the Calvert Cliffs 3 & 2 vessels would be assured throughout the assumed plant life for the SBLOCA+LOFW' transient where the PORV's are opened, and for the SBLOCA + LOPA transient with recovery of feedwater.

i B26

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-B38

B8.0 CONCLOSIONS lhis Aooendix to CEN-189 provides the results 'of analytical evaluations of pressurized thermal shock effects on thetCalvert Cliffs 1 a 2 reactor vessels for cases of a SBLOCA + LOP.4, in' responic to'the requirements of Item II.K.2.13 of NUREG-0737. Two different scenarios were chosen for

~

evaluation based on.remcdial'actjons to preyent initdequate core coolinc:

g,p y

1.

SBLOCA + LOFW + PORV's opened after 10 minutes s-2.

5BLOCA + LOFW + Abx, FW reinstated after 30 minutes y

4 '.

N *

/

f g.

Thermal-hydraulic system trangit-nt Jalck atio(ts were performed on a l

reference-plantbasis,"asrehofted(nCEN-139withtheparameter variations over the range representing al,l. Operatf_'ngjp1 ants.'

Four different cases were a'nalyzed for jach o(thelcwo differen,t scenarios

'i '

defined above, for a' total of pight; cases. Thehos,t~challe~nging of each of the two differentescena?ios was antlyzed tsing lirear eld? tic fracture machanics methods to determine th'e critic'a1.[ crack tip stress 3

s intensity vylues for comparisen to plant specific naterials ' properties o

at various UmesinLjl' ant life. The effect of the warm p' rest'ress phenomenooxis,identifie'd4here applicable for each;t'raMent, and credited where appropriate.

~-

J

/..

In(this' Apoendix, the results of plant specific neut'rorf fluence Ipro-file calculations'are superimposed on plari specific material proper-n

~

ties to defihe ve'ssel cap'aMiity versus plant life. The results of the generic LEFM' analyses were evaluated using'the pla'nt specific s.

w material propertiei.

It is concluded that. crack initiation would not

's occur due to the SBLOCA '+ LCFW Ucnsients considered :for more than 132 ef'fective full power yearsjf operation, whi~ch is assumed. to s

represent-full plant life.

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