ML20154A907

Pressure Temp Limits for 10 EFPY Toledo Edison Co, Davis-Besse Nuclear Power Station - Unit 1

ML20154A907
Person / Time
Site:	Davis Besse
Issue date:	11/30/1987
From:	Carey R, Collins L, Lowe A BABCOCK & WILCOX CO.
To:
Shared Package
ML20154A885	List: ML20154A881 ML20154A907
References
77-1170844, 77-1170844-00, BAW-2011, TAC-66699, NUDOCS 8805160238
Download: ML20154A907 (16)

Text

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Docket No. 50-346 k

License No. NPF-3 Serial No. 1510 BAW-2011 November 1987 4

PRESSURE-TEMPERATURE LIMITS FOR 10 EFPY THE TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION - UNIT 1 I

s by I

A, L. Lowe, Jr., PE.

t.

R. L. Carey L. L. Collins J. W. Ewing W. A. Pavinich W. E. VanScooter

f K. K. Yoon i

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B&W Document No. 77-1170844-00 o

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BABC0CK & WILCOX is'uclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 I [ ado $k $50bo346.

Babcock & Wilcox P

DCD a McDermott company

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.k-4 CONTENTS Page 1.

INTRO,9UCT ION...........................

1 - 1 2.

DETERMINATION OF REACTOR COOLANT PRESSURE B0UNDARY PRESSURE-TEMPERATURE LIMITS 2-1 3.

DEVELOPMENT OF TECHNICAL SPECIFICATION PRESSURE-TEMPERATURE LIMITS 3-1 4.

CERTIFICATION 4-1

-APPENDIXES A.

Revised. Technical Specifications Pressure-Temperature Operating Limitation............,............ A-1 8.

References..........

.................. B-1 3

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INTRODUCTION This report presents the Davis-Besse Unit I reactor coolant pressure boundary pressure-temperature operating limits for 10 EFPY.

The data used to develop these limitations are based on the analysis of Davis-Besse Nuclear Power Station Unit I reactor vessel surveillance capsule TEl-A as reported in BAW-l 1882.1 The report contains data which supports the development of the pressure-temperature limits for normal operation, both heatup and cooldown,

-l inservice leak and hydrostatic tests and reactor core operation for 21 EFPY.

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To minimize the impact of the effects of neutron radiation induced changes on the operating limitations, the pressure-temperature limitations, as defined in this report, were calculated for 10 EFPY.

These limits are adequate for current operations and are justified by the data obtained from the first three surveillance capsules as presented in BAW-1882.

New pressure-tempera-l ture limitations for additional EFPY operation will be developed before the

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reactor approaches the limits presented in this report.

In addition, the revised technical specifications and pressure-temperature limits as adjusted for Davis-Besse Nuclear Power Station Unit 1 through 10 EFPY are contained in Appendix A.

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DETERMINATION OF REACTOR COOLANT PRESSURE B0UNDARY PRESSURE-TEMPERATURE LIMITS The pressure-temperature limits -of the reactor coolant pressure boundary (RCPB) of Davis-Besse Unit 1 are established in accordance with the require-ment:. of 10CFR50, Appendix G.2 The methods and criteria employed to esta-blish operating pressure and temperature limits are described in topical report BAW-10046A.3 The objective of these limits is to prevent nonductile failure during any normal operating condition, including anticipated opera-tion occurrences and system hydrostatic tests.

The loading conditions of interest include the following:

1.

Normal operations, including heatup and cooldown.

i 2.

Inservice leak and hydrostatic tests.

3.

Reactor core operation.

The major components of the RCPB have been analyzed in accordance with 10CFR50, Appendix G.

The closure head region, the reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of the reactor vessel (and consequently of the RCPB) that regulate the pressure-temperature limits.

Since the closure head region is significantly stressed at relatively low temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first several service periods.

The reactor vessel outlet nozzle also affects the pressure-temperature limit curves of the first several service periods.

This is due to the high local stresses at the inside corner of the nozzle, which can be two to three times the membrane stresses of the shell.

After the first several years of neutron radiation exposure, the RT f the NDT beltline region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature mits of the RCPB.

For the service period for which the limit curves are established, the maxiaum allowable pressure as a function of fluid temperature is obtained' bI Babcock & Wilcox J McOctmott company

through a point-by-point comparison of the limits imposed by the closure head region, the outlet nozzle, and the beltline region.

The maximum allowable pressure is taken to be the lowest of the three calculated pressures.

The limit curves for Davis-Besse Unit I are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the tenth EFPY.

The tenth EFPY was selected because it is estimated that the third surveillance capsule will be withdrawn at the end of the refueling cycle when the estimated capsule fluence corresponds to approximately the T/4 end-of-life value.

The time difference between the withdrawal of the third and the currently available surveillance capsule data i

provides adequate time for re-establishing the operating pressure and temperature limits for the per.iod of operation beyond the third surveillance l

capsule withdrawal.

The unirradiated impact properties were determined for the surveillance beltline region materials in accordance with 10CFR50, Appendixes G and H.

For the other beltline region and RCPB materials for which the measured properties are not available, the unirradiated impact properties and residual elements, as originally established for the beltline region materials, are listed in Table 2-1.

The adjusted reference temperatures are calculated by adding the predicted radiation-induced RT and the unirradiated RT The e

riDT NDT.

predicted RT is calculated using the respective neutron fluence and copper I

f1DT and nickel contents.

Figure 2-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel beltline region wall.

The supporting information for Figure 2-1 are the predicted fluences that have been demonstrated in BAW-18821 to be conservative.

The design curves of

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Regulatory Guide 1.99, Rev. 2,4 were used to predict the radiation-induced RT values as a function of the material's copper and nickel content and ilDT neutron fluence.

The neutron fluences and adjusted RT values of the beltline region ilD7 materials at the end of the tenth full-power year are listed in Table 2-1.

The neutron fluences and adjusted RT values are given for the 1/4T and t1DT 3/4T vessel wall locations (T = wall thickness).

The assumed RT f the t1DT closure head region and the outlet nozzle steel forgings is 60F, in accord-l ance with BAW-10046A.

Babcock & Wilcox J McDermott company

Il' Figure 2-2 shows the reactor vessel's pressure-temperature limit curve for normal heatup.

This figure also shows the core criticality limits as required by 10CFR50, Appendix G.

Figures 2-3 and 2-4 show the vessel's pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively.

All pressure-temperature limit curves are applicable through ten EFPY.

Protection against nonductile failure is ensured by maintaining the coolant pressure below the upper limits of the pressure-temperature limit curves.

The acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve.

The reactor is not permitted to go critical until the pressure-temperature combinations are to the right of the criticality

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limit curve.

To establish the pressure-temperature limits for protection against nonductile failure of the RCPB, the limits presented in Figures 2-2 through 2-4 must be adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor vessel l

I, controlling the limit curves.

This is necessary because the reactor vessel is the most limiting component of the RCPB.

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e-Table 2-1.

Data for Preparation of Pressure-Temperature Limit Curves for Davis-Besse Unit 1 Reactor Vessel -- Applicable Throuah 10 EFPY S.deat see-ledwee ells & _ _

l-Cheetcal af f

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(homestry testeel I

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me. se 1,pe keys.e te.ase.e Ct. se cess ee s les at ene e/c e/o en/a letter alg.I 1/4 'I 31/4 'I stargia f t/4 11/4 aum 2en inwe, & a 2 a.,a te se S

, 9 01887 4.44 8 68 26 ele a

4 14/6 64 64 As J 23 3 tawa, El 2 upper wil

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$6 44 att 244

$A10s, il 2 to et hil 5.6Mle e et S el to one le t

14/s 78 67 es 212 nield asyyer tort e Se.e (10 ft) etW No 9.44137 8.50 0 64 I6B

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M "I 62 35 64/12 524 SI I

ese 2 88 u la upper (ersee team (UD 94tl este tes 9 41 Elf 0.29 8 44 264 me as2 4 us ed needle (6atee ie.e (60058 24 Tes 56 MIS 6.24 4 43 178 e2 42%

88 66 883 139 es 2s2 me ta le=er (nac e se.e (10 titt 247 he

3. 5MI6 8.84 8 64 168

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-n

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Figure 2-1.

Predicted Fast Neutron Fluence at Various Locations Through Reactor Vessel Wall for 10 EfPY - Davis-Besse Unit 1 6.0 5.63x1018n/cm2

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el

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is 3.18x1018n/cm2 s

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1.59x1038n/cm2 8

yes el Wall T/2 Locatior.

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7.47x1018n/cin2 ae Vessel Wall 3T/4 Location 0.0 i

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9 10 l B

EFPY

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De

.k k e=

?O M

f i

=

v Figure 2-2.

Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation -

lleattm. Applicable for First 10 EFPY - Davis-Besse Unit 1 2400 Assumed RI

" I NDI' 2200 Settline Region 1/%I 147 A

473 70 ll J

8eltline Region 3/4I 107 B

473 85 Closure 6 ead Region 60 C

48 3 115 Outline Noizie 60 0

625 175 2000 t

625 230 l

1005 240 G

1305 270 1800 u

2250 325 c,;

i 1305 310 1

2250 365

- 1600 g

the acceptance pressure-temperature combinations are below and to the right of the limit curve (s). the limit curves de hot include the a

U pressure differential between the point of system pressure f

1400 measurement and the pressure on the reactor vessel region controlling the limit curve, nor do they include any additional

+f margin of safety for possible lastrument error.

G i

I 1200 8

o g

1000 Applicable for lleatup F

g Rates up to 50f/h 800 m

0 Oe ci 600 D

g Cality 0

e 400 A

B C

An 2O

!' ax-200

s.

!k O

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j S0 100 150 200 250 300 350 400 450 m

Reactor Vessel Coolant Temperature, F

p

-m-%

g p,

A Figure 2-3.

Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation -

Cooldown. Applicable for First 10 EFPY - Davis-Besse Unit 1 2400 Assumed ITNDg, I Folat Pressure, psig leep., I 2200 - settilne segion 1/%I 147 A

307 70 Beltline Region 3/%I 107 3

470 120 Closure IIcad Region 60 C

625 165 2000 - outline morate 60 0

625 19 5 t

852 200 f

1H3 255 G

2250 300 1800 o

the acceptable pressure-temperature combinations are below "x

and the the right of the limit curve (s). the Ilmit curves c

do a t include the prnsen dif funtial betvun the point 4

1600 - of systes pressure measurement and the pressure on the u

3 reactor.essel region controlling the limit curve, nor do they include any additional margin of safety for F

g 1400 possible instrument error.

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1200 8

o T

1000 Applicable for Cooldown U

Rates up to 100F/h to 270F I

and then 50F/h 800 E

u S

M 600 C

D m

E'd 400 B

-no On A

Cooldown Rates fh 200 U to 50F/hr Up to 100F/ilr

• t j=

0 t

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I ag 50 100 150 200 250 300 350 400 M

Reactor Vessel Coolant Temperature. F t

R.

G Figure 2-4.

Reactor Vessel Pressure-Temperature Heatup and Cooldown Limits for Inservice Leak and livdrostatic Tests for the First 10 EFPY - Davis-Besse Unit 1 2600 Pressure, psig Iemp., I Assumed RI n

ngl, G

00

,,3,,,,,,,gi,, 3jg, 3g7 g3g 79 Beltline Region 3/41 107 8

524 90 2200 closure stead Region 60 C

625 120 Outline Norrie 60 D

625 205

[

1151 220 g" 2000 I

1910 28 0 G

2500 310 p

[

1800 the acceptance pressure-temperature combinations are below and the 3

the right of the limit curve (s). the limit curves do not, include the pressure differential between the point of system pressure E

1600 - measurement and the pressure on the reactor vessel region controlling the limit curve, nor do they include any M

1400 - additional margin of safety for possible lastrument sirer.

2 fo O

o 1200 Applicable for lleatup and

[

E h

1000 Cooldown Rates up to o

50 F/h (<50F in any 1/2-h 800 Period) g t:

600

=

U C

400 0

A R.

200 ct 5$

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g$g 50 100 150 200 250 300 350 400 p

Reactor Vessel Coolant Temperature, F o -

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3.

DEVELOPMENT OF TECHNICAL SPECIFICAT?0N PRESSURE-TEMPERATURE LIMITS i-The pressure-temperature litaits established for the technical specification were determined for selected heatup and cooldown rates by comparing the individual uncorrected pressure-temperature curves for the nozzle, beltline, and closure head over the. operating temperature range of the reactor vessel (see Section 2).

The limiting pressure (minimum) at each temperature was selected as the basis for developing the maximum pressure for setting the actual operating limitations.

Differential pressure corrections were then applied to the resulting limiting curves to account for the pressure cif-4 ferential between the analyzed regions of the reactor vessel and the system pressure sensor on the reactor coolant system hot leg.

The resulting corrected data points were plotited to obtain a bounding techni-(

cal specification curve for normal operations.

Also, heatup' and cooldown curves at various rates (OF!hr) over the various operating temperature ranges

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were combined into composita, bounding operating limit curves.

The resulting changes to the applicable technical specification section and the revised pressure-temperature curves for Davis-Besse are shown in Appendix A.

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Babcock & Wilcox aAROerme company

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CERTIFICATION The pressure-temperature operating limits for Davis-Besse Unit I reactor pressure vessel were calculated using approved procedures and established methods and techniques in accordance with the requirements of 10CFR50, Appendix G.

6 RE l3N'.'87 M

I A. L. Lowe, Jr., P.E. [/

Date Project Technical Manager This report has been reviewed for cal content and accuracy.

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0' WN AloV l6,l4 t7 L. 8. Gross, P.E.(m'aterial properties)Date '

" j" M&SA' Unit U %~, %

Ww. rt.1997 K.K.Yoop,P.E.(fractureanalysis)Date M&SA Unit /

l Verification of independent review.

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A d. lk /98 7 A. O. McKim, Maniger Date M&SA Unit This report has been approved for release.

/(f/ 7l'$ l J/ F. Walters

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PF

/ ogram Manager l

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APPENDIX A Revised Technical Specifications I

Pressure-Temperature Operating Limitations i

I Intentionally Omitted I'

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s APPENDIX B References F

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t B-1 Babcock & Wilcox a McDermott Co.mpany

l y in pvN e

1.

A.

L. Lowe, Jr., et al.,. Analyses of Capsule TEl-A, The Toledo Edison

Company, Davis-Besse Nuclear Power Station Unit 1,

Reactor Vessel Materials Surveillance Program, BAW-1882, Babcock & Wilcox, Lynchburg, Virginia, September, 1985.

2.

Code of Federal Regulation, Title 10, Part 50, Fracture Toughness Requirements for Light-Water Nuclear Power Reactors, Appendix G, Fracture Toughness Requirements, Federal Register, Vol. 48, No.104, May 17,1983.

3.

H. S. Palme, et -al., Methods of Compliance With Fracture Toughness an.

Operational Requirements of Appendix G to 10CFR50, BAW-10046A, Rev. 1, g g Babcock & Wilcox, Lynchburg, Virginia, July 1977, 1--

4.

U.S. Nuclear Regulatory Commission, Radiation Embrittlement of Reactor Vessel Material, Draft Regulatory Guide 1.99, Revision 2, Dated February 10, 1986.

5.

A.

L.

Lowe, Jr.,

et al.,

Pressurized Thermal Shock Evaluations in Accordance With 10CFR50.61 for Babcock & Wilcox Owners Group Reactor Pressure Vessesl, BAW-1895, Babcock & Wilcox, Lynchburg,

Virginia, January 1986.

6.

A.

S.

Heller and A.

L.

Lowe, Jr.,

Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds, BAW-1803, Babcock & Wilcox, Lynchburg, Virginia, January 1984.

7.

J. D. Aadland, Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information, BAW-1820, Babcock

& Wilcox, Lynchburg, Virginia, December 1984.

8.

K. E. Moote and A. S. Heller, BAW 177-FA Reactor Vessel Beltline Keld Chemistry Study, BAW-1799, Babcock & Wilcox, Lynchburg, Virginia, July 1983.

9.

A. L. Lowe, Jr., et al., Integrated Reactor Vessel Material Surveiilance Program (Addendum), BAW-1543A. Rev.

2.

Addendum 1, Babcock & Wilcox, Lynchburg, Virginia, November 1987.

0' B.aucock & Wilcox J McDermott Company