ML20140A671

Pressurized Thermal Shock Evaluation Per 10CFR50.61 for Reactor Vessels in Byron & Braidwood Units 1 & 2

ML20140A671
Person / Time
Site:	Byron, Braidwood, Zion, 05000000
Issue date:	01/13/1986
From:	Shawn Campbell, Lowe A, Moore K BABCOCK & WILCOX CO.
To:
Shared Package
ML20140A658	List: ML20140A653 ML20140A671 ML20140A682
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR 77-1159832, 77-1159832-00, NUDOCS 8601230298
Download: ML20140A671 (18)

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l PRESSURIZED THERMAL SHOCK EVALUATION IN ACCORDANCE WITH 10CFR50.61 FOR THE REACTOR VESSELS IN RYRON UNITS 1 & 2 AND BRAIDWOOD UNITS 1 & 2 January 13, 1986 B&W Contract #583-7497 Task 001

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Prepared by Babcock A Wilcox Company Nuclear Power Olvision P. O. Box 10935 Lynchburg, Virginia 24506 0935 B6 Mah i

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REPORT DOCUMENTATION i

DOCUMENT IDENTIFIER: 77-1159832-00

{ TITLE: PRESSURIZED THERMAL SHOCK EVALVATION IN ACCORDANCE WITH 10CFR50.61 FOR THE REACTOR VESSELS IN <

8YRON UNITS 1 & 2 AND BRAIDWOOD UNITS 1 & 2 l

DATE: JANUARY 13. 1986 i

'l 4 CUSTOMER: Com0NWEALTH EDISON 1

! CUSTOMER ORDER NO.: P. O. #303654 '

B&W CONTRACT NO.: 583-7497. TASK 001 f

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4 TABLE OF CONTENTS I Page

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

2. BACKGROUND , . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

3. INPUT DATA . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1. Ma te r i a l s Da ta . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2. heutron Fluence Estimates ................. 3-2 4 RT CALCULATIONS . . . . . . . . . . . . . . . . . . . . . . . . 4-1 PTS

5. CONCLUSIONS ........................... 5-1

6. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 List of Tables Table

1. Evaluation of Byron 1 Reactor Pressure Vessel in Accordance with Pressurized Thermal Shock Criteria . . . . . . . . . . . . . 3-3

2. Evaluation of Byron 2 Reactor Pressure Vessel in Accordance with Pressurized Thermal Shock Criteria . . . . . . . . . . . . . 3-4

3. Evaluation of Braidwood 1 Reactor Pressure Vessel in Accordance with Pressurized Thermal Shock Criteria ....... 3-5 4 Evaluation of Braidwood 2 Reactor Pressure Vessel in Accordance with Pressurized Thermal Shock Criteria ....... 3-6

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i Pressurized Thermal Shock Evaluations in Accordance with 10CFR50.61 for the i Reactor Vessels in Consnonwealth Edison Company's l Byron Units 1 & 2 and Braidwood Units 1 & 2 l

ABSTRACT I Pressurized thermal shock evaluations were performed in accordance with i

10CFR50.61, Fracture Toughness Requirements for Protection Against Pressurized i ThermalShockEvents"forthereactorvessel(RV)beltlineregionmaterials

, in the Consnonwealth Edison Company's Byron Units 1 & 2 and traidwood Units 1 &

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2. The projected values of RT PTS for all these materials are below the screening criteria for fast neutron fluences projected to 32 effective full power years (2.80 x 10 I's E > 1 MeV).

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1. INTRODUCTION The Nuclear Regulatory Comission's pressurized thermal shock (PTS) rules for pressurized water reactors (PWRs) are contained in 10CFR50.611 . This document requires that licensees submit projected values of reference temperature for <

each of the reactor vessel beltline materials. Thesevalues.(RTPTS),as determined for the Comonwealth Edison Byron Units 1 and 2 and Braidwood Units 1 and 2, are presented in this report. It also contains PTS background information, a description of the reactor vessel beltline materials, the source of the materials, neutron fluence data, and a review of the calculational methods employed. .

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2. BACKGROUND i

The Nuclear Regulatory Commission (NRC) amended its regulations for light 1

water nuclear power plants, effective July 23, 1985 to (1) estabitsh a screening criterion related to the fracture resistance of pressurized water reactor (PWR)vesselsduringpressurizedthermalshock(PTS) events;

(2) require analyses and schedule for implementation of flux reduction prcgrams that are reasonably practicable to avoid exceeding the screening criterion; and (3) require detailed safety evaluations to be performed before l j plant operation beyond the screening criterion will be considered. These I amendments are intended to produce an improvement in the safety of PWR vessels.

by identifying those corrective actions that may be required to prevent or , I mitigate potential PTS events.

,i l Transients and accidents can be postulated to occur in pressurized water reactors (PWR$) that result in severe overcooling (thermal shock) of the I reactor vessel concurrent with high pressure. In these pressurized thermal shock (PTS) events, rapid tooling of the reactor vessel internal surface causes a temperature distribution across the reactor vessel wall. This j j temperature distribution produces a thermal stress on the reactor vessel with

a maximum tensile stress at the inside surface of the vessel. The magnitude of the thermal stress varies with the rate of change in temperature and with time during the transient, and its effect is compounded by coincident pressure stresses.

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Severe reactor system overcooling events with pressurization of the reactor (

vessel (PTS events) are postulated to result from a variety of causes. These i

include system transients, some of which are initiated by instrumentation and l control system malfunctions (including stuck open valves in either the primary I or secondary system), and postulated accidents such as small break loss-of coolant accidents, main steam line breaks, and feedwater line breaks.

As long as the fracture resistance of the reactor vessel material is

relatively high, these events are not expected to cause vessel failure.

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sl However, the fracture resistance of the reactor vessel material decreases with the integrated exposure to fast neutrons during the life of a nuclear power plant. The rate of decrease depends on the chemical composition of'the vessel wall and weld materials. If the fracture resistance of the vessel is reduced s sufficiently by neutron irradiation, severe PTS events could cause small flaws

% that might exist near the inner surface to propagate into the vessel wall.

The assumed initial flaw might be enlarged into a crack through the vessel wall of sufficient extent to threaten vessel integrity and, therefore, core cooling capability.

The toughness state of reactor vessel materials can be characterized by a

" reference temperaturet 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*F. RT is a measure of the NDT 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 effects are determined by destructive tests of material specimens. Correlations, based on tests of irradiated specimens, have been' developed to calculate the shift in RT as a NDT function of neutron fluence for various material compositions. The value of RT at a given time in a vessel's life is used in fracture mechanics NDT calculations to determine whether assumed pre-existing flaws would propagate when the vessel is sub'jected to overcooling events.

On the basis of the 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 NDT 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 require detailed plant-specific evaluations and possibly modifications to existing equipment, systems, and procedures.) The NRC approach to selection of the RT NDT screening criterion is described in detail in SECY-82-465.2 In summary, the approach was to use a deterministic fracture mechanics algorithm to j calculate the value of RT NDT for which assumed pre-existing flaws in the l

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l reactor vessel would be predicted to initiate (grow deeper into the vessel wall) assuming occurrence of one of the severe overcooling events that have been experienced. These " critical" values of RT were related to the NDT 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 10CFR50.61, <

Paragraph (b)(2) for use as a screening criterion. This designation is used to avoid confusion with the RTNDT 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 less than 270*F for PTS plate and forging material and axial welds, and less than 300*F for circumferential welds present an acceptably low risk of vessel failure from PTS events.

The requirements of 10CFR50.61 further state the following:

"For each pressurized water nuclear power reactor for which an operating license has been issued, the licensee shall submit projected values of RTPTS (at the inner vessel surface) of the reactor vessel beltline materials by giving values from the time of submittal to the expiration date of the operating license. The assessment must specify 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."

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3. INPUl DATA The pressurized thermal shock regulations require that the data used to perform the specified calculations be traceable by including the source of all values included in the assessment. The relationship of the material on which ,

any measurements are made to the actual material in the reactor vessel (RV) 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 cores.

The following describes the sources for all data used to evaluate the R.V.

beltline materials in the Byron and Braidwood units. .

3.1. Materials Data The R.V. beltline materials of all four of the Byron and Braidwood units met 2

the requirements of Appendix G of 10CFR50 . This included the use of materials with prescrioed levels of copper and fracture toughness properties.

The chemical compositions and reference temperature data shown in Tables 1-4 were obtained from the Quality Assurance records available at The Babcock &

Wilcox Company, the manufacturer of these vessels. Either SA 508 C1 2 mod. or SA 508 C1 3 forgings were used in the beltline.of these plants. The test data were obtained from coupons of the actual forgings in accordance with Section III, Article NB-2000 of the 1971 Edition of the ASME Code and the following Addenda:

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Byron I All Addenda through Summer 1972 Byron II, All Addenda through Braidwood I, II Summer 1973 l

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The test data shown in these tables for the beltline welds were obtained from weld metal qualification test samples which also met the requirements of Section III, Article NB-2000 of the ASME Code. As can be seen, measured values of RT NDT and copper and nickel concentrations were available for each of the beltline materials.

3.2. Neutron Fluence Estimates The peak fast neutron flux at the inside surface of each reactor vessel is 2.77 x 10 10 nyt as presented in the FSAR for each plant. The estimated peak neutron fluence is 2.77 x 10 10 x 32 effective full power years or 2.80 x 10 19 n/cm2 (E > 1 MeV). The value of 32 EFPY is based on assumed 40-year licensed operating period and 80% full power operation during this period. This fluence was applied to the upper and lower shell forgings and the circumferential weld joining these shells. .

The projected peak neutron fluences at other R.V. beltline locations were based on the following:

e The relationship for relative axial variation of fast neutron flux and fluence within the pressure vessel wall (E > 1 MeV) for Zion Units 1

&2,3 e The relative positions of the beltline materials with respect to the core midplane. These positions are virtually identical for all four reactor vessels.

As shown in Tables 1-4, this value was applied to this weld and the nozzle belt forging. .

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Table 1. Evaluation of Byron 1 Reactnr Pressure Vessel in Acenrdance with Pressurrired Thermal Shock Criteria Material Description Chemical Constants for PTS Calculated Reactor Vessel Heat Composition u/o RT pg$ Calculations. F Inside Surface Fluence, n/cm2 Screening RTpy3, F Beltline Region Location Number Type Copper Nickel Initial RT NOT Margin 32 EFPY Criteria, F 32 EFPY tower Morale Belt 123J218 SA 508 C1 2 mod. 05 .72 +20 48 6.30ElB 270 91 Upper Shell $P-5933 SA 508 C1 2 mod. .05 .73 +40 48 2.80E19 270 123 w tower Shell SP-5951 SA 508 C1 2 mod. .04 .64 +10 48 2.80El9 210 81 Upper Circumferential Weld WF501 ASA/Linde 80 028 .63 +10 48 6.30E18 300 66 Middle Circumferentiel Weld WF336 ASA/Linde 80 .031 46 -30 48 2.80E19 300 31 Lower Circumferential Weld WF472 ASA/Linde 80 .23 .57 +10 48 < E17 300 --

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Table 2. Evaluation of Byron 2 Reactor Pressure Vessel in Accordante with Pressurrired Thermal Shock Criteria Material Description Chemical Constants for Inside Surface Pis Calculated Reactor Vessel Heat Composition, w/o RT py$ Calculations, F Fluence, n/cm2 Screening RTp ;$, F Beltline Region Location Number Type Copper Nickel Initial RT MDT Margin 32 [FPY Criteria, F 32 EIPY SA 508 C1 2 mod. .05 .74 +10 48 6.30E18 270 81 lower Nortle Belt 4P-6107 Upper Shell 49D329) y,g 49C297), 5A 508 C1 3 .01 .70 -20 48 2.80E19 270 30

• Law n S k11 -l-1 SA 508 C1 3 .05 .73 -20 48 2.80E19 270 63 tipper Circumferential Weld WF562 ASA/Linde 80 .03 .65 +40 48 6.30E18 300 98 ASA/Linde 80 053 .62 +10 48 2.80E 19 300 93 Middle Circumferential Weld WF447 ASA/Linde 80 .18 .54. +40 48 < E17 300 --

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Table 3. Evaluation of Rraidwood 1 Resctnr Pressure Vessel in Accordance with Pressurrired Thermal shock Criteria Material Description Chemical Constants for Inside Surface PTS Calculated Reactor Vessel Heat Composition, w/o RT pys Calculations. F Fluence, n/cm2 Screening RTPT5' I Beltline Region location Mr Type Copper Nicle1 Initial RT NOT Margin 32 EFPV Criteria, F 32 [FPY Lower Nortle Belt SP-7016 SA 508 C1 2 mod. .04 .71 +10 48 6.30F18 270 75 Upper Shell 49C344) 490383)~g~g SA 508 C1 3 .05 .73 -30 48 2.80fl9 270 53 b Lower Shell 490867)"I~I 49C813) SA508C13 .03 73 -20 48 2.80E19 270 44 Upper Circumferential Weld WF645 ASA/Linde 80 .033 .50 -30 48 6.30E18 300 28 Middle Circumferential Weld WF562 ASA/Linde 80 .03 .65 +40 48 2.80E19 300 102 Lower Circumferential Weld WF653 ASA/Linde 80 .19 .56 -40 48

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Table 4 Evaluation of Braidwood 2 Reactor Pressure Vessel in Accordance with Pressurztred Thermal. Shuk Criteria Material Description Chemical Constants for Inside Surface PTS Calculated Reactor vessel Heat Composition, w/o Rip;$ Calculations. F Fluence, n/ce? Screening RTPT5' I Peltline Region location Number Type Copper Niciel Initial RT PDT Margin 32 EFPY Criteria, F 32 EFPY tower Nozzle Belt SP-7056 SA 508 C1 2 mod. .04 .90 +30 48 6.30E18 270 97 Upper Shell 1 SA 508 C1 3 03 .71 -30 48 2.80El9 270 to l9 33 c'n 10"'I

• II 5 g jl.1 SA 508 Cl 3 .06 .75 -30 48 2.80C19 270 63 Upper Circumferential Weld WF645 ASA/Linde 80 033 .50 -30 48 6.30Ela 300 78 Middle Circumferential Weld WF562 ASA/Linde 80 .03 .65 +40 48 2.80E19 300 102 Lower Circumferential Weld WFC% ASA/Linde 80 038 .60 -16 48 < EI7 300 --

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4. RT A M ATIONS PTS For the purpose of comparison with the PTS criterion, the value of RT PTS IO#

each of the reactor vessel beltline 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 PTS is the lower of the results given by Equations 1 and 2. Equation 1 was applicable to the Deltline materials in the four reactor vessels in the Byron and Braidwood plants.

0 Equation 1: RT PTS =I+M+[-10+470Cu+350CuNi]f.270 Equation 2: RTPTS = I+M+283f .

a. "I" means the initial reference temperature of the unirradiated material 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 j riust be used: 0'F for weld made with Linde 80 flux (as stated in Part 3, i measured values were available for all materials).

l b. "M" means the margin to be added to cover uncertainties in the values of initial RTNDT, copper and nickel content, fluence and the calculational procedures. In Equation 1, M=48'F if a measured value of I was used and l M=59'F if the generic mean value of I was used. (Since measured values were available, M=48'F was employed in these calculations.)

l l c. "Cu" and "Ni" mean the best estimate weight percent of copper and nickel d

(- in the material.

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d. "f" means the best estimate neutron fluence, in units of 10I9n/cm2 (E greater than or equal to 1 MeV), at the clad-base metal interface on the inside surface of the vessel at the location where the material' in question receives the highest fluence for the period of service 1

considered.

The results of the reactor vessel specific PTS calculations using Equation 1 i and the data sources described in Part 3 are included in Tables 1 through 4. -

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5. CONCLUSIONS The Byron Units 1 and 2 and Braidwood Units 1 and 2 reactor vessel beltline I materials met the requirements of Appendix G to 10CFR50. The projected 40-year RT PTS values for these materials are well under the PTS screening <

criteria. All of the calculated RT P values were 5 123 at the estimated peak neutron fluence of 2.80x10 19 n/cm 2(TS E > 1 MeV).

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. 6. REFERENCES 1

1. 10CFR50.61, " Analysis of Potential Pressurized Thermal Shock Events," ,

j July 23, 1985.

! 2. Appendix G,10CFR Part 50, " Fracture Toughness Requirements," March 1

! 1973.

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3. S.L. Anderson, Plant Specific Neutron Fluence Evaluation for Zion Units 1 and 2," WCAP-10902, August 1985.

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