Text

Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCAs W/Loss of Feedwater for Palisades Reactor Vessel
	ML20039F862
Person / Time
Site:	Palisades
Issue date:	12/31/1981
From:	; ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	;
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ML13308A045	List: ML13308A044; ML14092A301; ML17297B168; ML20039F858; ML20039F860; ML20039F861; ML20039F862; ML20039F863; ML20039F864; ML20039F866; ML20039F867; ML20039F868;
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-D, NUDOCS 8201130485
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60' SEEND X D EVALUATION OF PRESSURIZED THERMAL SH0CK EFFECTS DUE TO SMALL BREAK LOCA'S WITH LOSS OF FEEDWATER FOR THE PAllSADES REACTOR VESSEL Prepared for CONSUMERS POWER COMPANY NUCLEAR VER SYSTEMS DIVISION i

' POWER P'si R SYSTEMS Em COMBUSTION ENGINEERING. INC 820113" VSS

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 REPORT, 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.

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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 Palisades reactor vessel.

It is concluded that 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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e CEN-189 Appendix D i

TABLE OF CONTENTS SECTION TITLE PAGE ABSTRACT F

Dl.

PURPOSE D1 D2.

SCOPE D1 i

D3.

INTRODUCTION D1

{

04~

THERMAL HYDRAULIC ANALYSES D1 05.

FLUENCE DISTRIBUTIONS D2 D6

. MATERIAL PROPERTIES D7 07.

VESSEL INTEGRITY EVALUATIONS D13 D8.

CONCLUSIONS D24 i

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Dl.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 Palisades reactor vessel.

02.0 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 Palisades reactor vessel.

~

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

D

3.0 INTRODUCTION

This Appendix to CEN-189 was prepared by C-E for Consumers Power Company for their use in responding to Item II.K.2.13 of NUREG-0737 for the Palisades reactor vessel.

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 D5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter D6 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 D7 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to the material properties discussed in Chapter D6.

D4.0 THERMAL HYDRAULIC ANALYSES The pressure-temperature transients used to perform 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 thermal-hydraulic system analyses.

Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved.

D1 J

D5

. Palisades Fluence Distribution The fluence distribution applied to the Palisades reactor was based on the peak fluence and azimuthal distribution provided by Consumers Power plus radial and axial distributions which were calculated as described in Sections 5.2.3 and 5.2.4.

The peak fluence'was quoted as 4.24 x 1018 n/cm2 as December 31, 1981.

This value assumes an integrated energy output of 4.215 Effective Full Power Years (EFPY) at 2530 Megawatts-thermal (Mwt). The azimuthal fluence distribution as transmitted by Consumers Power is shown in-Table DS-1.

The distribution shown for the interval between 0 and 2 centimeters from the inner surface of the vessel was used.

Figure 05-1 shows a plot of the azimuthal distribution used for this analysis.

The axial and radial distributions were calculated using an RZ-DOT model based on a Millstone Point-Unit 2 design. Adjustments to account;for differences in core lengths were made by using the top of the active core as a reference point. The resulting axial and radial fluence distribu-tions are shown in Figures D5-2 and 05-3, respectively.

t D2

i TABLE D5-1 1 Mev at Palisades Vessel as a Function of l

FLUX **

Angle and Distance from Surface i

FLUX VS. DISTANCE (CM) FROM INTER SURFACE ZONE ANGLE *(REVOLUTIObS)

FROM TO 0-2 24 46 6-8 8-10 1

0 0.0073 0.'619 0.513 0.392 0.285 0.183 2

0.0073 0.0147 0.694 0.5 74 0 437 0.317 0.203 3

0.0147 0.0220 0.825 0.679 0.514 0.371 0.237 4

0.0220 0.0293 0.963 0.786 0.591 0.424 0.269 5

0.0293 0.0354 1.000 0.818 0.615 0.441 0.280 6

0.0354 0.0388 0.9 71 0.796 0.601 0.433 0.276 7

0.0388 0.0402 0.952 0.779 0.591 0.427 0.2 72 8

0.0402 0.0412 0.944 0.771 0.586 0.424 0.270 9

0.0412 0.0422 0.937 0.765 0.581 0.420 0.268 10 0.0422 0.0432 0.927 0.75 7 0.5 76 0.416 0.265 11 0.0432 0.0446 0.910 0.746 0.567 0.410 0.262 12 0.0446 0.0480 0.8 79 0.725 0.550 0.399 0.255 13 0.0480 0.0541 0.847 0.699 0.532 0.385 0.247 14 0.0541 0.0642 0.859 0.710 0.540 0.391 0.250 15 0.0642 0.0743 0.956 0.784 0.592 0.426 0.2 72 16 0.0743 0.0845 0.987 0.809 0.610 0.439 0.279 17 0.0845 0.0946 0.912 0.751 0.569 0.410 0.262 18 0.0946 0.1047 0.817 0.673 0.511 0.369 0.236 19 0.1047 0.1148 0.741 0.611 0.464 0.336 0.215 l

20 0.1148 0.1250 0.702 0.5 79 0.440 0.318 0.204 e

1 1

I the 450 axis and 0.125 is at the principal axis.

Zero revolutions is at Octant symmetry assumed.

- Relative to peak flux.

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f-APPENDIX D PALISADES

{

D.6 MATERIAL PROPE.< TIES The methods used to develop and evaluate the materials for the Palisades reactor vessel 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 Table D6-1.

The copper content for weld seams 1-112 and 2-112 was estimated using the highest measured value fer the-other weld seams.

j t

In cases where the weld metal nickel content was not determined, it was conservatively estimated using information on the type of wire (eg, high MnMo versus MnMoNi wire) or the weld process (inclusion of Ni-200 wire duringL

~ weld deposition). For the Palisades weldments, the weld inspection records -

4 and welding certification reports indicated that all the welds could be ex-pected to contain high nickel (greater than 0.30 w/o), so the nickel content was conservatively' estimated to be 0.99 w/o as indicated in Table D6-1.

The toughness properties given in Table D6-1 are 'the drop weight NDTT (if determined) and the initial reference temperature, RTNDT.

For the plate materials, the RT was determined using transversely oriented Charpy impact NDT specimens or by converting longitudinal impact data using Branch Technical Position MTEB 5-2*.

For the weld material, the RT was estimated using the NDT weld qualification test results benc tmarked to the Fort Calhoun surveillance weld.

(RTNDT was not determined for the Palisades surveillance weldment, so data from a similar weldment were used instead.)

The methodology used is discussed in Section 6.0 and described below.

i-The individual weld qualification test results (three Charpy' impact specimens tested at +10F) are listed in Table D6-2.

Each weld which l-exhibited an average Charpy energy of 57 ft-lb or greater (the average I

Charpy energy for the benchmark weld at lOF) was considered to be at least as tough as the benchmark weld i.e., that weld seam RTNDT ***

I

=* " Fracture Toughness Requirements for Older Plants,"

U.S. Atomic Energy Commission, Regulatory Standard Review Plan.

%9

-50F or less. For those weld qualification test results exhibiting an -

average Charpy energy less than 57 ft-lb, the RT as increased by an NDT amount equivalent to the temperature difference between the average Charpy energy transition curve for the benchmark weld and the average Charpy energy for the vessel weld test results. In effect, the temperature at which 50 ft-lb or better exists was determined, and the RT was established at a temperature 60F below that value.

NDT A " map" of the cylindrical portion of the Palisades reactor vessel is given in Figure D6-1.

It shows the locations of the plates and welds listed in Table D6-1 and their corresponding values of initial RTNDT (F) loccted within a rectangle on the Figure.

RT NDT for the vertical weld seams (designated 1-112, 2-112, and 3-112 ) are shown at a single seam but apply to all three vertical seams in a given shell course. Included in the Figure are the locations of the inlet and cutlet nozzles, the core midplane, and the extremities of the active core.

Figure D6-2 is a map of adjusted RT "E # ""

NDT locations at the inner surface of the ' Palisades vessel predicted for December 31, 1981. The predictions are based on the best estimate 19 neutron-fluence, 0.424 x 10 n/cri ( E>lMeV ), (corresponding to 4.215 effective full power years at peak flux location on the inside surface of the reactor vessel), the initial RT and copper, phosphorus, and NDT nickel contents given in Table DS-1, and the normalizd neutron flux profiles given in Section D.5.

The values of adjusted RTNDT (1 itial i

RT plus predicted shift) are located in rectangles adjacent to the pg7 plate and weld designations. The RT Values apply to the inner NDT surface of the vessel in the region indicated by a circle. The circled regions generally represent areas of peak neutron flux for a given weld seam or plate.

D6-2 08

TABLE 06. PALISADES. REACTOR VESSEL MATERIALS.

Product Material Drop Wei ht Initial Chemical Content-(%)

9 Form Identification NDTT ( F)_

RTND_T_( " F,)_

Nickel Copper Phosphorus 0

Plate D-3802-1 10 20 0.49 0.25 0.012 b

Plate D-3802-2 0

30d 0.48 0.25 0.015 b

Plate D-3802-3 10 20 0.55 0.25 0.011.

Plate D-3803-1

-10

-Sc 0.53 0.25.

0.011 c

Plate D-3803-2

-30

-30 0.50 0.25 0.011 8

Plate D-3803-3

-30 Oa 0.48 0.25 0.013 b

Plate

'D-3804-1

-30 0

0.45 0.25 0.017 8

8 Plate'

.0-3804-2

-40

-30 0.50 0.25 0.018 b

a Plate D-3804-3

-30

-25 0.54 0.25 '

0.010 d

Weld 1-112 A,B,&C N/A

-25 1.20 0.28" O.021 d

Weld 2-112 A,B,&C N/A

-25 1.20 0.28 0.021 d

I Weld 3-112 A,B,&C N/A

-45 0.99 0.28 0.018 d

I Weld 8-112 N/A

-45 0.99 0.28 0.013 d

c Weld 9-112 N/A 25 1.27c 0.22' O.011 8

N/A Not Available Determined using Branch Technical Position HTEB 5-2 a

b Estimated based on average for Palisades plates having reportedanalyses c

Surveillance program data d-Estimated (see text and Table D6-2)

Estimated (highest measured.value for other weld seams) e f

Estimated Ni content (high nickel type wire or weld process)

_.. ~

TABLE D6-2 4

PALISADES

. REACTOR VESSEL WELD SEAT 1 TOUGHNESS DATA.

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

RTNDT (*F)

.1-112 A/C 35, 39, 48 40.7

-25 2-112 A/C 35, 39, 48 40.7

-25 j

3-112 A/C 46, 56, 59 53.7

-45 3-112 46, 56, 59 53.7

-45 9-112 35, 4S,'42 41.7

-25 71, 57, 42 56.7

-45 c3 Benchmark Held" 51, 55 '

57.0

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a ' Benchmark Weld - Fort Calhoun surveillance weld b Test results at 0"F c. Actual I!TNDT based on drop weight and Charpy test results d Estimated using the method described in the text L

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D.7.0 Palisades Vessel Integrity The fracture mechanics analysis is performed using the plant specific properties of the Palisades vessel. 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 the vessel as a function of f1DT Effective Full Power Years (EFPY). This information is used in locating the points in the vessel having the highest RT at each of the three NDT axial sections of interest:

1) middle of core, z 138.63 in.

=

2) top of core, z 72 in.

=

3) above-core, z 43.5

=

where z is the axial distance below the centerline of the nozzle.

From the predicted RT values, the material toughness properties K and f4DT IC i

K are determined from the calculated temperatures for the SBLOCA + LOFW la transients using the method described in Section 7.6.

Critical crack vs crack depth curves and depth diagrams are constructed from the applied Kg the calculated material toughness curves. By performing the same fracture mechanics analysis a number of times for increasing plant life (EFPY) the integrity of the Palisades vessel for the SBLOCA + LOFW transient is evaluated.

0.7.1 Sunnary of Physics and Materials Data Input to Fracture Mechanics Analysis A detailed survey was performed on the combined fluence and material properties 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 riDT at the inner vessel radius were compared, and the location with the highest RT value was used in the fracture mechanics analysis.

flDT At the mid-core level,the location of highest RT occurs in the weld f1DT material at an azimuthal angle of 30 degrees. The fluence factor at this location is.93 of the peak fluence in the vessel.

013

The materials data at this point are as folicws:

1.2 PCT.

Ni

=

.28 PCT.

Cu

=

.021 PCT.

P

=

-20%

Initial RT

=

NDT 19 At the 12/31/S1 level of 4.2 EFPY, and peak fluence of.422 x 10 19 n/cm2 (E >l MeV), this corresponds to a point fluence of.392 x-10 2

n/cm and an adjusted surface RT value of 1719.

f4DT At the top of core level the location of highest RT ccurs in the NDT weld material _ at an az.imuthal angle of 30 degrees. The fluence factor at this location in the vessel is.29 of the peak fluence. The materials data at this point are as follows:

1.2 PCT.

Ni

=

.28 PCT.

Cu

=

.021 PCT.

P

=

. -20 F Initial RT

=

tIDT 19 At the 12/31/81 level of 4.2 EFPY, and peak fluence of.422 x 10 19 n/cm (E > 1 MeV), this corresponds to a point fluence of.124 x'10 2

n/cm and an adjusted surface RT value of 88 F.

NOT At the above core level (about halfway between the top of core and the inlet nozzle), the location of highest RT occurs in the plate NDT material at an azimuthal angle of 123 degrees. The fluence factor at this point is.002 of the peak fluence in the vessel. The materials data for this point are as follows:

.48 PCT.

Ni

=

.25 PCT.

Cu

=

.015 PCT.

P

=

U 30 F Initial RT

=

NDT Dl4

19 At the 12/31/81 level of 4.2 EFPY, and peak fluence of.422 x 10 2

19 n/cm (E >l MeV), this corresponds to a point fluence of.001 x 10 2

n/cm and an adjusted surface RT value of 37 F.

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

D.7.2 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 of the report. The fracture mechanics analyses were performed for this case using the Palisades 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 D.7-1.

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

IC, r a range of initial flaw sizes. rHowever, from the plot of K vs time shown in g

Figure 7.14 of the report it is seen that warm-prestressing would occur after 10 minutes in the transient, beyond which time K is g

continually decreasing. Thus, no crack initiation would occur under these circumstances. The upper shelf-toughness line indicates the 200 ksiTin.'. This represents the upper flaw depths for which K

=

g 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 precludes crack initiatior.

prevents initially small flaws from extending into that range.

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

For this case, also, initial flaws within a certain range of depth are calculated to exceed the level of initiation toughness after 85 minutes in the transient.

From the plot of K vs g

time for the top of core level 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.

D15

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

It is apparent from this figure that the

-calculated stress intensities are below both the initiation and arrest toughness levels, thus there is no potential for brittle crack initiation 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.

D.7.3 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 Palisades 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 D.7-4.

The calculated stress intensity values exceed the arrest toughness after 72 minutes, and a small initiation region is apparent at 100 minutes in the transient. The fac t that warm-prestressing occurs for this transient after 78 minutes, as shown in the plot of K vs time in Figure 7.17 of the report, indicates g

that crack initiation would not occur under these conditions. The 200ksilfEE.representstheupper upper shelf toughness line for K

=

y limit of applicability of LEFM. A ductile failure mechanism would 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 0.7-5.

Similarly, the diagram for the above the core level of the vessel at 32 EFPY is shown in Figure 0.7-6.

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.

1 016

s D.7.4 Conclusion

~

These results demonstrate that the integrity of the' Palisades vessel would be assured throughout the assumed plant life fer the SBLOCA + LOFW transient with recovery of feedwater, and for the SBLOCA + LOR:

transient where the PORV's are opened.

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'023

/, -

D 8.0 CONCLUS!ONS This Appendix to CEN-189 provide's the results of analytical evaluations Is of pressurized thermal shock effects on the Palisades reactor vessel

/

for cases of 5 SBLOCA + LOFW, in response to the requirements of Item

/

II.K.2.13 of NUREG-0737. Two different scenarios were chosen for eval-e'

/J

,uation based on remedial actions to prevent inadequate core cooling:

ii f

1.

SBLOCA + LOFW + p0RV's opened after 10 minutes i

2.

SBLOCA + LOFW + Aux FW reinstated af ter 30 minutes Thermal-hydraulic system transient calculations were performed on a reference-plant basis, as reported in CEN-189 with the parameter variations over th'd range representing all operating plants. Four different' cases wercianalyzed for each of the two different scenarios defined above, for a total of eight cases. The most challenging of each of tr.e,two different scenarios.was a'.alyzed using linear elastic fracture mechanics methods to determine the critical crack tip stress intensity values for comparison to plant specific materials properties

. b,'

at various times in plant life? The effect of the warm prestress

a y

'f phenomenon is identified where applicable for each transient, and fj i

credited where appropriate, i

}

'II' In this Appendix, the re ults of plant specific neutron fluence pro-

+

[

file calculations are superimposed on plant specific material proper-ties to define vessel capability versus plant life. The results of the generic LEFM analyses were evaluated using the plant specific material properties.

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

024

COMBUSTION ENGINEERING, INC.

.

ML20039F862

Text

3.0 INTRODUCTION

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