ML18096B166

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Rev 1 to Fracture Toughness Analysis for Salem Units 1 & 2 Reactor Pressure Vessels to Protect Against Pressurized Thermal Shock Events,10CFR50.61.
ML18096B166
Person / Time
Site: Salem  PSEG icon.png
Issue date: 11/13/1992
From: Gustems B, Kent R, Rosenfeld E
Public Service Enterprise Group
To:
Shared Package
ML18096B165 List:
References
NFU-060, NFU-060-01, NFU-60, NFU-60-1, NUDOCS 9212220329
Download: ML18096B166 (45)
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Text

NFU-060 Revision 1 September 1, 1992 FRACTURE TOUGHNESS ANALYSIS FOR SALEM UNITS 1 AND 2 REACTOR PRESSURE VESSELS TO PROTECT AGAINST PRESSURIZED THERMAL SHOCK EVENTS I

-j 10CFRS0.61 PStiG 9212220329 921215 PDR __ ADOCK 05000272 P PDR.

'\_

  • FRACTURE TOUGHNESS ANALYSIS FOR SALEM UNITS 1 tfillD 2 REACTOR PRESSURE VESSELS TO PROTECT AGAINST PRESSURIZED THERMAL SHOCK EVENTS 10CFR50.61 Prepared by~ ..~~ Date B: D. Gustems Engineer Reviewed Date R. S. Kent Nuclear Fuels Engineer Nuclear Department Approved b y £ E. s. Rosenfeld w

Manager - Nuclea Fuel Date /(

1

/13/? 2-

  • Nuclear Department
  • COPY NO. - - - - -

\ l NFU-060 Revision 1 September 1, 1992 ABSTRACT Fracture toughness calculations for Salem Units 1 and 2 Reactor Pressure Vessels (RPV) have been performed in response to the Nuclear Regulatory Commission final rule, 50FR29937-29945, on protection against Pressurized Thermal Shock (PTS) events. Salem Units 1 and 2 are owned and operated by Public Service Electric and Gas Company, New Jersey.

The toughness state of the RPV is characterized by an "Adjusted Reference Temperature for Pressurized Thermal Shock" (ARTpts>*

The method specified in a final rule has been used to calculate ARTpts of reactor vessel beltline materials and compared to the screening criterion established by the NRC. Values of ARTpts less than the screening criterion present an acceptably low risk of vessel failure from PTS events.

Two key components required to calculate RTpts are neutron fluence (n/cm2) and the chemical content and mechanical properties of RPV beltline materials. The neutron flux/fluence was calculated using transport code DOT IV.3. To enhance the accuracy of the calculation, the calculated flux was compared to the measured flux from the reactor vessel radiation surveillance program, and a calculation bias was determined. This bias was applied to normalize the calculated flux to the measurement. The chemical content and mechanical properties of the RPV beltline materials were obtained from Westinghouse Electric Corporation and Combustion Engineering .

. A:IOCFR50.BOO

' t NFU-060 Revision 1 September 1, 1992

  • TABLE OF CONTENTS .

SECTION TITLE

1. 0 INTRODUCTION . . . . . * * * . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
2. 0

SUMMARY

OF RESULT.S . . . * . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.0 MATERIAL PROPERTIES OF SALEM 1 AND 2 REACTOR PRESSURE VESSELS .............*......... 5 3.1 OVERVIEW . . . . . . . . . . . . . . * . . . . . . . . . . . . . . . . . . . 5 3.2 MATERIAL DATA FOR SALEM VESSELS ........... 5

4. 0 VESSEL FLUENCE CALCULATION ..................... 19
4. 1 OVERVIEW . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
4. 2 GEOMETRIC MODEL ..*.....**................. 21 4 .*3 FISSION SPECTRUM *......................... 21 4*4 SOURCE GENERATION .................*....... 21 4 *5 CROSS SECTIONS .............*.............. 2 2
4. 6 DOT ANALYSIS ...........*....*......*...... 22 4.7 BENCHMARK OF CALCULATION RESULTS .......... 23 4.8 VESSEL FLUENCE RESULTS .................... 24 4.9 VESSEL FLUENCE CALCULATION UNCERTAINTY .**. 24 5.0 ARTPTS RESULTS FOR SALEM 1 AND 2 VESSELS ........ 32

6.0 CONCLUSION

S . . . . . . . . . . . . . . . . . * . . . . * . . . . . . . . . . . . . 35 7*0 REFERENCES ******************************* ** ***** 3 6 A:IOCFR50.BDG

NFU-060 Revision 1 September 1, 1992

  • LIST OF FIGURES FIGURES TITLE 3 .1.1 REACTOR PRESSURE VESSEL ASSEMBLY . . . . . . . . . . . . . . . 7 3 .1. 2 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE SALEM UNIT N0.1 REACTOR VESSEL . . . . . * . . . * . . . . . . . . . . . . . * . . . . . . G *
  • 8 3 .1. 3 IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIAL FOR THE SALEM UNIT NO. 2 REACTOR VESSEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.2.1 ARTPTS VERSUS FLUENCE FOR SALEM 1 RPV WELDS ..*.* 13 3.2.2 ARTPTS VERSUS FLUENCE FOR SALEM 2 RPV WELDS ..... 14 3.2.3 ARTPTS VERSUS FLUENCE FOR SALEM 1 RPV BASE METAL B2402 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.2.4 ARTPTS VERSUS FLUENCE FOR SALEM 1 RPV BASE METAL B2403 ......*.*..****.***.....*...... 16 3.2.5 ARTPTS VERSUS FLUENCE FOR SALEM 2 RPV BASE METAL B4 712 ***..........*.....*..*..**..*. 1 7 3.2.6 ARTPTS VERSUS FLUENCE FOR SALEM 2 RPV BASE METAL B4 713 . . . . . . . . * . . . . . . . . . . . . . . . . . . . . . . 18 4 .1.1 FLOWCHART FOR CALCULATING NEUTRON FLUENCE ....*. 27 4.7.1 ARRANGEMENT OF SURVEILLANCE CAPSULES IN THE REACTOR VESSEL ...*.****..........*.*..** 28 A:IOCFR.50.BOO

NFU-060 Revision 1 September 1, 1992 LIST OF TABLES TABLE NO. TITLE 3.2.1 RPV BELTLINE REGION WELD CHEMISTRY FOR SALEM UNITS 1 AND 2 ***.*****...*..******** 10 3.2.2 RPV BELTLINE REGION WELD MECHANICAL PROPERTIES FOR SALEM UNITS 1 AND 2 ***.....**.** 11 3.2.3 RPV BELTLINE REGION PLATE MATERIAL CHEMICAL AND MECHANICAL PROPERTIES FOR SALEM UNITS 1 AND 2 ********.*.......*****.. 12 4.7.1 BENCHMARK RESULTS OF NEUTRON FLUX .**..*.****.*. 29 4.8.1 REACTOR PRESSURE VESSEL FLUENCE FOR SALEM 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 4.8.2 REACTOR PRESSURE VESSEL FLUENCE FOR SALEM 2 * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 5.0.1 ARTPTS FOR SALEM 1 RPV BELTLINE REGION MATERIALS *..**..*....***..**........*********** 3 3 5.0.2 ARTPTS FOR SALEM 2 RPV BELTLINE REGION MATERIALS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4 A:IOCFRS0.800

NFU-060 Revision 1 September 1, 1992

  • LIST OF TABLES TABLE NO. TITLE 3.2.1 RPV BELTLINE REGION WELD CHEMISTRY FOR SALEM UNITS 1 AND 2 ............*.***..*... 10 3.2.2 RPV BELTLINE REGION WELD MECHANICAL PROPERTIES FOR SALEM UNITS 1 AND 2 ..*..*....*.. 11 3.2.3 RPV BELTLINE REGION PLATE MATERIAL CHEMICAL AND MECHANICAL PROPERTIES FOR SALEM UNITS 1 AND 2 ****.**.**************** 12 4.7.1 BENCHMARK RESULTS OF NEUTRON FLUX ..........*... 29 4.8.1 REACTOR PRESSURE VESSEL FLUENCE FOR SALEM 1 . . . . . . . . * . . . * . . . . . . * . * . . . . . * . *. e **** 30 4.8.2 REACTOR PRESSURE VESSEL FLUENCE 1

FOR SALEM 2 . * ********************************** 31 5.0.1 RTPTS FOR SALEM 1 RPV BELTLJ~ 1EGION MATERIALS . . . . . * . * . . . . . . * * . * * .. . . * . . . * . ..*. ti **** f9 33 5.0.2 RTPTS FOR SALEM 2 RPV BELTLINE REGION MATERIALS . . . . . . . . . * . . . . . . . . . . " . . . . . . . . . . . . . . . . . 3 4

  • A:IOCFRSO.BDG

~. NFU-060 Revision 1 September 1, 1992

1. 0 INTRODUCTION This report presents the RPV fracture toughness calculations for Salem Units 1 and 2 which have been performed in response to the NRC final rule on protection against PTS events (Reference 1). Salem Units 1 and 2 are Westinghouse, 4 loop PWR's with thermal rated power of 3411 MW.

Salem 1 started operation in 1976 and is currently operating in Cycle 11. Salem 2 started operation in 1981 and is currently operating in Cycle 7. The current cycles employ the low leakage loading pattern which are designed to reduce the neutron flux leakage from the core.

Two key inputs are required to calculate the fracture toughness of the reactor pressure vessel which is characterized by the quantity ARTpts (Adjusted Reference

~emperature for Eressurized ~hermal ~hock). These are neutron Fluence and the chemical/mechanical properties of RPV beltline region materials. The neutron Fluence calculations have been performed by PSE&G using the industry accepted transport theory method. The fluence methodology was recommended by Babcock & Wilcox Company (B&W). The chemical/mechanical data of beltline region welds were obtained from Combustion Engineering, who fabricated the Salem pressure vessels. The beltline region plate data were obtained from Westinghouse Electric Corporation.

Section 2.0 summarizes the results of ARTpts and vessel life. Section 3.0 documents the material properties of Salem vessels. The vessel fluence calculation methodology A:1ocFR.SO.soo Page 1

..

NFU-060 Revision 1 September 1, 1992

  • is described in Section 4.0. The ARTpts results are presented in Section 5.0, followed by conclusions which are discussed in Section 6.0 *
  • A:IOCFRSO.BDG Page 2

~*

I . v ~ ,

NFU-060 Revision 1 September 1, 1992 2.0

SUMMARY

OF RESULTS The PTS rule requires the licensee to submit projected values of ARTpts at the inner vessel surf ace of reactor beltline materials by giving values from the time of submittal (August 31, 1992) to the expiration date of the operating license. The operating license of Salem 1 will expire midnight of August 13, 2016 and the operating license of Salem 2 will expire midnight of April 18, 2020. The key results are as follows:

1. For Salem Unit 1, the limiting beltline region material is plate #B2402-l. The limiting beltline region material is defined as that material which reaches the ARTpts screening criterion first, thereby limiting the vessel life. The ARTpts of plate B2402-1 is projected to be 211.4°F by August 31, 1992 and 255°F by August 13, 2016. These ARTpts values are below the screening criterion of 270°F.
2. For Salem Unit 2, the limiting beltline region materials are weld numbers 2-442B and C. The ARTpts of these welds are projected to be 142.9°F by August 31, 1992 and 219°F by April 18, 2020. These ARTpts values are below the screening criterion of 270°F.
3. The ARTpts projections are based upon the assumption that the future Salem Unit 1 cycles will be very similar in design to the low leakage loading pattern of cycle 8. Therefore, fluence accumulated during cycle 8 operation was used to estimate f luence for the future cycles. Similarly for Salem Unit 2, the projections are based upon the assumption that the low leakage design of cycle 5 will be continued into the future.

A:IOCFRSO.BOO Page 3

NFU-060 Revision 1 September 1, 1992

4. For Salem Unit 1, the screening criterion for ARTpts for the limiting plate is projected to be reached at the end of 45.4 effective full power years (EFPY) of operations.
5. The limiting RPV welds for Salem Unit 2 are projected to reach the screening criterion after a total plant operation of 86.4 EFPY *
  • A:IOCFR50.BDG Page 4

NFU-060 Revision 1 September 1, 1992 3.0 MATERIAL PROPERTIES OF SALEM 1 AND 2 REACTOR PRESSURE VESSELS 3.1 Overview A good set of RPV material data is very important in accurately assessing the RTpts* The chemical and mechanical properties of the Salem weld data, used in this study, were the result of a thorough investigation performed by Combustion Engineering and our recent response to GL 92-01. (Reference 16).

Figure 3.1.1 shows the assembly of the Salem Reactor Pressure Vessel. The Salem vessels were fabricated by Combustion Engineering. The Reactor Vessel beltline materials that directly surround the effective height of the core and the adjacent region that are predicted to experience sufficient neutron irradiation damage consist of the intermediate and lower shell course and their associated welds. Figures 3.1.2 and 3.1.3 identify and schematically locate the Reactor Vessel plates and welds in the beltline region for Salem units 1 and 2 respectively (Reference 5).

3.2 Material Data for Salem Vessels The chemical and mechanical properties of the beltline region welds and plates of Salem Units 1 and 2 which are necessary to calculate the fracture toughness of the vessel are tabulated in Tables 3.2.1, 3.2.2 and 3.2.3. The fracture toughness state of the vessel is characterized by the quantity ARTpts (i.e., Adjusted Bef erence ~emperature for £ressurized ~hermal ~hock) .

2 are used. The equation is:

A:IOCFRSO.BDG Paqe 5

NFU-060 Revision 1 September 1, 1992

  • ART = I + M + (CF) f (0.28 - O.lLog (f) )

where I = Initial Reference ~emperature of unirradiated material for Nil ~uctility ~ransition; RTNDT M = Margin to be added to cover uncertainties in the initial RTNDT, copper and nickel content, f luence and the calculations procedures CF= Chemistry factor, a function of the copper and nickel content from tables 1 and 2 in 10CFR50. 61.

f = Best estimate neutron fluence in units of

  • 1019 n/cm2 (for energies greater than or equal to 1.0 Mev) at the clad-base metal interface on the inside surface of the vessel at the material location which receives the highest f luence for the period of service in question.

Using the above equation and Salem specific material data, ARTpts is calculated and plotted as a function of vessel f luence for all the welds and plates in the beltline region. Figures 3.2.1 through 3.2.7 represent such a family of ARTpts versus f luence curves ranging through the vessel lifetime for Salem Units 1 and 2.

These plots enable the determination of RTpts knowing the actual vessel f luence (which is calculated in Section 4.0) during different times of plant operation *

  • A:IOCFR.50.BDG Page 6
  • FIGURE 3. 1. l REAC?OR PRESSURE VESSEL ASSEMBLY Top Hm Cen* OWc Top Hem!

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am &int ~Km rar.-.

  • Page 7

PIGORB 3 .1. 2 IDBRTIPICATIOR ARD LOCATIOR OP BBLTLIHB RBGIOB MATERIAL POR TBB SALEM UNIT NO. 1 REACTOR VESSEL Nozzle Shell 8-042 2-0428---- 90° Intermediate Shell 180°

-042C B2402-2*

CORE 270° 9-042 B2403-3-042 Lower Shell 3-042C B2403-3 270°10-042 Relative to the midplane of the active core, the circumferential weld 8-042 is about 91.5" above the core cidplane and circ.

welds 9-042 and 10-042 are 16.7" and 123.2n below the core rnidplane respectively.

Page s

PIGORB 3.1.3

  • IDERTIPICATIOH ARD LOCATION OP BBLTLIRB REGIOR MATERIAL POR TBB SALEM UNIT HO. 2 REACTOR VESSBL Nozzle Shell 8-442 90° B4712-2 2-442A Intermediate 180° .

Shell B4712-l CORE 84712-3 270° 9-442 90° 3-442A - - -

0°~-t--t-~~---i~~~-i~~l80° 3-4428 ..

Lower Shell 84713-1 3-442----

270°10-442 Relative to the oidplane of the active core, the circumferential weld 8-442 is about 92.1" above the core midplane and circ. welds 9-442 and 10-442 are 16.7" and 123.2" below the core ~idplane respectively.

Page 9

NFU-060 Revision 1 September 1, 1992 TABLE 3.2.1 RPV BELTLINE REGION WELD CHEMISTRY (1) FOR SALEM UNITS 1 AND 2 Weld Weld Wire Cu Ni Basis Seam Heat/Flux Lot .llil.sll .llil.sll Salem 1 8-042 12420/3708 0.22 1.02 MML (3)Analysis 2-042 34B009/3692 0.18 1.00 (2) H. B. Robinson A/C & 39Bl96/3692 head weld and Ni-200 Wire Salem 1 surv.

weld analyses 9-042 13253/3791 0.25 0.72 D.C. Cook and Salem 2 surv.

weld analyses 3-042 34B009/3708 0.19 1. 00 ( 2) H. B. Robinson A/C + Ni-200 wire head weld analysis Salem 2 8-442 20291/3854 0.28 0.74 Cooper Sta.

& IP2809/3854 surv. weld (20291), Bare Wire analysis and Reg. Guide 1.99 (IP2809) 2-442 13253/3833 0.23 0.73 D. c. Cook, A/C &20291/3833 Salem 2, Cooper Sta. Surv. weld analyses 9-442 90099/3977 0.175 0.20(2) MML(3) Analyses 3-442 21935/3889 0.20 0.86 MML Ananlyses A/C & 12008/3889 Notes: (1) source is reference #2 (2) Estimated values (3) C-E's Metallurgical and Materials ~aboratory (MML)

Page 10

TA 3.2.2

  • **1

. .

RPV BELTLINE REGION WELD MECHANICAL PROPERTIES (1) FOR SALEM UNITS 1 AND 2 Initial RTNDT( 2 ) Margin(3) To be Added To Weld Weld Wire Weld Flux Cover Uncertainties C°F>

Unit Seam Heat/Flux Lot .en ~

Salem 1 8-042 12420/3708 -56 1092 65.51 2-042 348009/3692 -56 1092 65.51 A/C 398196/3692 -56 1092 Ni-200 Wire 9-042 13253/3791 -56 1092 65.51 3-042 348009/3708 -56 1092 65.51 A/C + Ni-200 Wire Id Pl Salem 2

~ 8-442 20291/3854 -56 1092 65.51 Cl)

IP2809/3854

!-.&

!-.& 2-442 13253/3833 -40 1092 65.51 A/C 20291/3833 9-442 90099/3977 -56 0091 65.51 3-442 21935/3889 -56 1092 65.51 A/C 12008/3889

( 1) Source is Reference #2.

( 2) Generic RT NOT for C-E's SAW Weld; u =17 (3) Per NRC PTS Final Rule A:BRIANfAB.BOO

Plate No. Chemical Composition (3)

T 3.2.3 RPV BELTLINE REGION PLATE MATERIAL CHEMICAL AND MECHANICAL PROPERTIES FOR SALEM UNITS 1 AND 2 Mechanical Properties (1) Margin (2) To Be Added

  • To Cover Uncertainties (2)

Cu (w/o) Ni (w/o) Initial RTNDT(°F)

Salem 1 82402-1 0.24 0.52 45 34 B2402-2 0.24 0.50 -5 34 B2402-3 0.22 a.so -23 34 B2403-1 0.19 0.48 10 34 B2403-2 0.19 0.49 26 34 B2403-3 0.19 0.48 30 34 Id Salem 2 Pl t..Q CD B4712-1 0.13 0.56 0 34

~ B4712-2 0.14 0.60 12 34 N

B4712-3 0.11 0.57 10 34 B4713-1 0.12 0.60 8 34 B4713-2 0.12 0.57 8 34 B4713-3 0.12 0.58 10 34 Notes: ( 1) Based on measured data from transverse charpy specimens.

(2) Per NRC PTS Final Rule.

(3) Measured data.

A:BRIANTAB.BOO

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NFU-060 Revision 1 September 1, 1992 4.0 VESSEL FLUENCE CALCULATION 4.1 Overview A key input parameter in the calculation of RPV fracture toughness is the neutron f luence (n/cm2) at the vessel clad-base metal interface. This section describes the vessel f luence calculation which was performed for every Salem cycle completed to date including the current cycles, to determine the current cumulative fluence and to project future vessel fluence.

Salem Units 1 and 2 are both 4 loop, 193 fuel assembly Westinghouse design PWR's, rated for a thermal power of 3411 MW. Salem 1 is presently operating in cycle 11 and Salem 2 is in the seventh cycle of operation. Both the current cycles employ the low leakage loading pattern designed to significantly reduce the vessel flux at key weld locations.

The method used in the calculation of vessel f luence is conservative and consistent with the current industry method for such analysis. A discrete ordinate s 0 transport analysis was performed to determine the fast neutron flux using the DOT IV.3 code. Additionally, the "calculated" flux values were compared with the "measured" values from the analyses of withdrawn surveillance capsules for equivalent irradiation periods, to determine the calculation bias. This bias was then factored into the calculation of vessel fluence. This vessel fluence methodology was recommended by Babcock & Wilcox Company (Reference 6).

A:IOCFR50.BDG Page 19

NFU-060 Revision 1 September 1, 1992 The process of calculating vessel f luence can be divided into nine key tasks. The first task is concerned with the development of the geometric model which provides an adequate description of the physical system. The second task is to develop cycle representative combined (U235 + Pu239) fission spectrum. The third task is to calculate the fuel assembly multipliers to account for pin power deviations between measured and design powers. The "as-designed" pin power distribution are obtained from PDQ calculations. Combination of the first three tasks with some' additional input enables the calculation of a

\

source term which is both energy dependent and spatially distributed and represents the fuel cycle of interest. This represents the fourth task. The fifth task constitutes generation of macroscopic cross sections. The sixth task involves combing the geometric model, the source term and cross sections plus some additional input for the DOT calculation and obtain neutron fl~x. The seventh task involves the determination of axial shape factors which are applied

,_to the two dimensional flux to account for the leakage along the third dimension (Z-axis). The eighth task involves benchmarking the calculated flux values against corresponding measured data. Finally, the ninth task involves summing the results for individual cycles to obtain cumulative vessel fluences and predict future vessel fluence.

Figure 4.1.l is a flowchart which illustrates the above calculation steps. A brief description of these steps is provided below.

A:IOCFRSO.BOO Paqe 20

NFU-060 Revision 1 September 1, 1992 4.2 Geometric Model The geometric model (Reference 7) describes the physical system in two dimensions; radially (R) and azimuthally (theta). Radially, the model extends from the core center to approximately 10 cm into the concrete primary shield and describes the core, baffle, barrel, thermal shield, the surveillance capsule, pressure vessel and its liner, mirror insulation, concrete and water in between the cylindrical vessel, components. Azimuthally, the model assumes eight core symmetry and extends from O to 45 degrees. The (R-0) model was divided into 82 radial intervals and 45 angular intervals. Such a mesh description was determined to be adequate based on neutronic consideration and computer limitations.

4.3 Fission Spectrum A combined {U235 + Pu239) fission spectrum (Reference

7) for every fuel cycle was developed using the pure, normalized Watt spectrum. of U-235 and Pu-239 and modifying it with the integrated fission rate from each isotope. This method allows the inclusion of spectral hardening effects due to plutonium build up. The Watt spectra were based upon ENDF/B-V recommended spectral forms and constant9. The integrated fission rate was obtained from PDQ calculations at a representative statepoint during the fuel cycle.

4.4. Source Generation The relative pin power distribution from PDQ (References 7, 8) was used to calculate the DOT source term. The PDQ pin power distribution were obtained at a number of statepoints during fuel cycle life. Since the PDQ pin powers represent the "as-designed" power A:JOCFR50.BDG Page 21

NFU-060 Revision 1 September 1, 1992 distribution, they were adjusted to reflect reality by

'using appropriate multiplication factors. These factors were derived based on the comparison of end-of-cycle measured versus designed assembly burnups.

The SORREL code (Reference 9) was used to calculate the DOT source term. This code, time averages the pin power distributions and calculates energy dependent and spatially distributed source term based upon the input fission spectrum and geometric model.

4.5 Cross Sections A 27 neutron group macroscopic cross section set was generated for input to DOT (Reference 10) using the BUGLE 80 master cross section library (Reference 11) and the GIP code (Reference 12). The 27 neutron groups covered the energy range from 17 MeV to 0.06 MeV.

These groups were considered sufficient since neutron flux greater than or equal to 1.0 MeV is desired.

4.6 DOT Analysis The discrete ordinates transport analysis was performed using the DOT-IV.3 code (Reference 13). This analysis was performed for all the past and current Salem cycles (Reference 14). All the 2 dimensional (R-8) DOT executions used P 3 expansion of the scattering cross section and an sa angular quadrature. DOT executions were performed in the forward mode. The fluxes calculated by DOT using the (R-8) geometry implicitly contains a flat axial power shape. To factor the axial power shape into the results, an axial factor was obtained. A cycle average factor was calculated considering the cycle average power distribution of the fuel assemblies nearest to the vessel target point of A:IOCFRSO.BDG Page 22

NFU-060 Revision 1 September 1, 1992

  • interest. In each case the axial factor was computed as the ratio of the flux at the target due to the actual axial shape, to that due to a flat axial shape.

4.7 Benchmark of Calculation Results To date, three surveillance capsules have been removed and analyzed from Salem 1: capsule "T" was removed at the end of cycle 1 (Reference 15), capsule "Y". was removed at the end of cycle 5 (Reference 3) and capsule "Z" was removed at the end of cycle 7 (Reference 21).

For Salem 2 three surveillance capsules have been removed: capsule "T" at the end of cycle 1 (Reference 4), capsule "U" at the end of cycle 3 (Reference 22) and capsule "X" at the end of cycle 6. All six surveillance capsules were located at 40° azimuthal location as shown in Figure 4.7.1. The calculations were benchmarked against the four measured state-

\ points, and the results are presented in Table 4.7.1.

As discussed in WCAP-11554 (Salem 2 Capsule "U" results), Westinghouse has made improvements to capsule measurement results by including the use of a c factor to account for the effect of power distribution changes on neutron monitor response. The same method is used in WCAP-11955 for Salem 1 capsule "Z". The measured versus predicted results for the last capsule (U and Z) for each unit appear to be significantly better than the results for previous capsules. It was therefore decided that for the purpose of generating a CALIB factor for PSE&G f luence calculations only the most recent capsule result for each Unit would be used.

DOT analyses to predict these capsule results (for use in deriving CALIB factors) were not explicitly performed for this report. Instead, measurement A:IOCFRSO.BDG Page 23

NFU-060 Revision 1 September 1, 1992 results for the inner vessel surface at 45 Deg were inferred from the Westinghouse reports for each unit for the lastest capsule and were used to calculate a CALIB factor for each Unit.

Based on the results of the above calculations a CALIB factor of 0.99 was chosen as a reasonable but conservative value to be used for both Salem Units for fluence predictions and projections in this report (Reference 20) . These results provide a measurable degree of confidence in the calculation methodology.

Results from Unit 2 capsule "X" were not available at the time this report was prepared and will be submitted if the results are significantly affected when the DOT analysis has been completed.

4.8 Vessel Fluence Results Tables 4.8.1 and 4.8.2 present the Salem 1 and 2 maximum vessel fluence results for all the beltline region welds and plates at different times of vessel life. The vessel fluence is calculated on the inside surface of the clad-base metal interface. The projected fluences for Salem 1 were based on the cycle 8 low leakage design and Salem 2 fluence were projected based on cycle 5. The tables also lists the maximum fluence permissible for the beltline region materials and the total effective full power years necessary to accumulate the maximum f luence before exceeding the PTS screening criterion.

4.9 Vessel Fluence Calculation Uncertaintv The calculation of the ARTpts requires the use of best estimate fluence value, since the "margin" parameter in A:IOCFR50.BDG Page 24

NFU-060 Revision 1 September 1, 1992

  • the ARTpts equations already account for the uncertainty involved in the fluence calculation.

Therefore, it is not necessary to apply additional calculation uncertainty to the vessel fluence results.

The following discussion only serves to highlight the key uncertainty contributing factors in such a calculation.

1. Capsule Flux/Fluence Uncertainty Since the calculation model is benchmarked against measured dosimeter data from the capsule, the uncertainty associated with these measurement must be included. The factors to be considered are:
a. The uncertainty associated with the measurement of dosimeter activities.
b. Calculation Procedure used to obtain the calculated dosimeter responses (i.e.

"measured" flux) introduces uncertainty due to such items as cross sections, material compositions, yields, source term, etc.

c. In the calculation of capsule flux, nominal position of the capsule holder tube and capsule specimens are assumed. Where as the "measured" flux derived from the dosimeter activity measurement correspond to the "as-built" position of the surveillance capsule *
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NFU-060 Revision 1 September 1, 1992

2. Vessel Flux/Fluence Uncertainty The calculated flux/f luence at the vessel location has two uncertainty contributing factors:
a. The calculation method is benchmarked against the capsule values by a measured-to-calculated normalization factor. This factor is applied to the vessel f lux/f luence and thus the vessel flux accounts for the capsule uncertainty described in (1) above.
b. The normalization factor obtained from the capsule is extrapolated to the vessel as it is assumed to apply at the vessel location.
3. Fluence Prediction The uncertainty in the prediction of f luences to end-of-life includes the above two terms plus the estimated uncertainty in the prediction method used. This term includes the uncertainty associated with the choice of a particular cycle as the equilibrium cycle for extrapolation.

The PSE&G vessel flux calculation method closely reflects the standard industry approach for performing such a calculation. The method used by PSE&G was recommended by Babcock & Wilcox Company. Therefore, the to~al uncertainty associated with this calculation is estimated to be similar to B&W's estimate, i.e.,

varying between+/- 20% to+/- 30% (Reference 17).

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FIGURE 4.1.1

'rf FLOWCHART FOR CALCULATING NEUTRON FLUENCE J

PDQ-07 Provide!t Relative Pouer Di!ttribution I

<R-8)

Geometric Model SORREL.

Calculate!t DOT Source Normalization J - Term and other DOT -

Factor Array I

- Input for PDQ Pouer Distribution

<U+Pu>

Fission Spectrum

'

DOT IV.J GIP BUGL~-80 Calculates Neutron - Calculates t Master Fl u x (n/cm 2 -sec> Macroscopic X-Sec-X-Sections tian Library Benchmark Factors .I

- Synthesize 30 Neutron Fl U)( calculation Axial Shape Factor for -

30 Synthesis I Cycle Effective Cycle Neutron Fluence - Fuel Power Seconds Paqe 27

FIGURE 4.7.1 ARRAtlG:C:HENT OF SURVILL.1'\.NCE CAPSULES IN THE REACTOR PRESSURE VESSEL

  • SALEH UNIT l(l) y

.,,

SAL:C:Il mnT 2 ( 2)

I ES: (1) See Reference ~3 (2) See Reference #4 Page 28

NFU-060 Revision 1 September 1, 1992 TABLE 4.7.1 BENCHMARK RESULTS OF NEUTRON FLUX NEUTRON FLUX Cn/cm2--sec) [MEAS-CALC] x 100 MEASURED CALCULATED CALC ]

(%)

Salem Unit 1, End of Cycle 1, Capsule "T" 7.48 + 10 6.84 + 10 9.0 Salem Unit 1, End of Cycle 5,

  • Capsule "Y" 8.36 + 10 7.03 + 10 19.0 Salem Unit 2, End of Cycle 1 Capsule"T 6.91 + 10 5.97 + 10 16.0 NUETRON FLUENCE Cn/cm2.~

Based on:

Salem Unit 1, End of Cycle 7 Capsule "Z" 3.87 + 18 3.98 + 18 -2.7 Based on:

Salem Unit 2, End of Cycle 3, 1.78 + 18 1.78 + 18 -1.0 Capsule"U"

  • At vessel inner surface at 45° A:IOCFR50.BDG Page 29

NFU-060 Revision 1 September 1, 1992 TABLE 4.8.1 REACTOR PRESSURE VESSEL FLUENCE FOR SALEM 1 (1)

Weld No. or Azimuthal Current Fluence; Projected Fluence Maximum Plate No. Location @ up to 8/17/92 at end of fluence to which Fluence c1019n/cm2) Operating License reach PTS is Calculated (lol 9n/ cm2) screening (Degrees) criterion c1019n/cm2) 8-042 45 .32 .83 2 .4 .

2-042A/B 30 .40 1.10 2.2 2-043C 0 .21 .63 2.2 9-042 45 .51 1.35 8.0 3-042A/B 15 .31 1.07 2.0 3-042C 45 .51 1.35 2.0 B2402-1 45 .54 1.38 2.0 B2402-2 45 .54 1.38 11. 6 B2402-3 45 .54 1.38 25 B2403-1 45 .51 1.35 25 B2403-2 45 .51 1.35 25 B2403-3 45 .51 1.35 25 Note: (1) See References 14 and 17 A:IOCFRSO.BDG Page 30

NFU-060 Revision 1 September 1, 1992 TABLE 4.8.2 REACTOR PRESSURE VESSEL FLUENCE FOR SALEM 2Cl)

Weld No. or Azimuthal Current Projected Maximum Plate No. Location @ Fluence; up Fluence at fluence to which Fluence to 5/08/92 end of reach PTS is Calculated (lo19n/cm2) Operating screening (Degrees) License criterion c1019n/crn2) ( 1019n/crn2) 8-442 45 .14 .42 4.74 2-442A 0 .14 .68 4.13 2-442B/C 30 .24 1. 05 2.89 9-442 45 .33 1.38 25.0 3-442A/C 30 .24 1. 07 2.93 3-442/B 0 .14 .67 2.9 B4712-1 45 .33 1. 38 25 B4712-2 45 .33 1. 38 25 B4712-3 45 .33 1.38 25 B4713-1 45 .33 1.41 25 B4713-2 45 .33 1.41 25 B4713-3 45 .33 1. 41 25 Note: (1) See References 14 and 17 A:IOCFRSO.BDG Page 31

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NFU-060 Revision 1

  • September 1, 1992 5.0 ARTpTs RESULTS FOR SALEM 1 AND 2 VESSELS This section describes the ARTPTS {Adjusted Reference Temperature for Pressurized Thermal Shock) calculation for the reactor pressure vessel beltline region materials {Reference 17).

ARTpTs equation which is specified in the PTS final rule has been used. This equation is described in Section 3.2. The material properties given in Tables 3.2.1 through 3.2.3 and neutron fluence values presented in Tables 4.8.1 and 4.8.2 were substituted in the ARTPTS equation. Table 5.0.l and 5.0.2 present the results for Salem 1 and 2 RPV beltline region materials

  • respectively
  • A:IOC:FR50.BDG Page 32

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NFU-060 Revision 1 September 1, 1992

  • TABLE 5.0.l ARTpTs FOR SALEM UNIT 1 RPV BELTLINE REGION MATERIALS(l)

Weld # or Azimuthal Current Projected ARTPTS Plate # Location @ ARTPTS up to ARTPTS @ end Screening which ARTPTS 8/17/92 (OF) of Operating Criterion is calculated License ( ° F) (OF)

(Degrees) 8-042 45 171 231 300 2-042A/B 30 169 229 270 2-042C 0 134 196 270 9-042 45 169 222 300 3-042A[B 15 159 234 270 3-042C 45 188 248 270 B2402-1 45 212 255 270 B2402-2 45 159 201 270 B2402-3 45 133 192 270 B2403-l 45 149 184 270 B2403-2 45 166 201 270 B2403-3 45 169 204 270 Note: (1) See Reference 17 Page 33 A:IOCFRSO.BDG L ___ _

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NFU-060 Revision 1 September 1, 1992

  • TABLE 5.0.2 ARTPTS FOR SALEM UNIT 2 RPV BELTLINE REGION MATERIALS(l)

Weld # or Azimuthal Current Projected ARTPTS Plate # Location @ ARTPTS up to ARTPTS @ end Screening which ARTPTS 5/8/92 (oF) of Operating Criterion is calculated License (°F) (oF)

(Degrees) 8-442 45 111 167 300 2-442A 0 117 180 270 2-442B/C 30 142 219 270 9-442 45 76 113 300 3-442A/C 30 134 216 270 3-442B 0 108 190 270 B4712-1 45 96 132 270 B4712-2 45 116 155 270 B4712-3 45 95 124 270 B4713-1 45 100 133 270 B4713-2 45 99 132 270 B4713-3 45 101 134 270 Note: (1) See Reference 17 A:IOCFR50.BDG Page 34

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NFU-060 Revision 1 September 1, 1992

6.0 CONCLUSION

S The fracture toughness calculations for Salem Units 1 and 2 were performed in response to the NRC final rule on protection against the pressurized thermal shock events. The methods specified by the NRC were employed in this study. The ARTpts results described in this study are characterized by the fact that the best available material data for Salem vessels have been used, along with standard industry accepted method for calculating vessel fluence. The main conclusions that can be drawn based on the results of this study are as follows:

1. The ARTpts of the Salem Units 1 & 2 vessels remain below the PTS screening criterion until the current expiration date of the operating license.
2. The limiting material which allows Salem 1 to reach the screening criterion is vessel plate
  1. B2402-1. This is caused by the high initial RTNDT of the plate material combined with a high copper content.
3. The low leakage loading patterns have significantly reduced the vessel flux at critical vessel locations. Flux reductions of approximately 50% relative to the conventional out-in designs hav~ been achieved for both the Salem units .
  • A:IOCFR50.BDG Page 35

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NFU-060 Revision 1 September 1, 1992

7.0 REFERENCES

1. Analysis of Potential Pressurized Thermal Shock Events, Nuclear Regulatory Commission Final Rule, 56FR22304 May 15, 1992.
2. Salem Units 1 and 2 Reactor Vessel Weld Data, Combustion Engineering, Inc., Design Input File
  1. TOl.5-020, November 1985.
3. Analysis of Capsule Y from the Salem Unit 1 Reactor Vessel Radiation Surveillance Program, Westinghouse Electric Corporation, WCAP 10694, December 1984.
4. Analysis of Capsule T from the Salem Unit 2 Reactor Vessel Radiation Surveillance Program, Westinghouse Electric Corporation, WCAP 10492, March 1984.
5. PSE&G Reactor Vessel Surveillance Program Data for Salem Units 1 and 2, Memo from H. Trenka (PSE&G) to E. Rosenfeld (PSE&G), NFUI 85-456, October 1985.
6. Fluence Analysis Procedure, Babcock & Wilcox company, NFUI 85-219, March 1985.
7. SORREL:DOT Input Generation for all Salem Units 1 and 2 Cycles, PSE&G Calculation File #DOl.6-302, December 1985.
8. ARMP Maintenance Manual, PSE&G, NFU-0046, May 1985.

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NFU-060 Revision 1 September 1, 1992

9. SORREL Computer Code, PSE&G Request for Code Modification File #D06.6-0152, August 1985.
10. GIP Cross Sections for Salem Unit 1 and 2 cycles

'-

for PTS Analysis, PSE&G Calculation File #DOl.6-297, December 1985.

11. BUGLE-SO, RSIC Data Library Collection, Oak Ridge National Laboratory, NFUI 85-010, Design Input File #DOl.5-205, January 1985.
12. GIP Computer Code, PSE&G Request for code Modification File #D06.6-0156, June 1985.
13. DOT IV.3 Computer Code, PSE&G Request for Code Modification File #D-6.6-164, December 1985.
14. DOT Execution for Salem Units 1 and 2 Cycles, PSE&G Calculation File #DOl.6-755, October 4, 1991.
15. Analysis of Capsule T from Salem Unit 1 Reactor Vessel Radiation Surveillance Program, Westinghouse Electric Corporati~n, WCAP 9678, February 1980.
16. Response to Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity, Salem Generating Station unit Nos. 1 and 2, June 1992.
17. Vessel Fluence Data, Letter from Dr.L. Hassler (B&W) to Q. Dahodwala (PSE&G), NFUI 85-309, July 1985.

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NFU-060 Revision 1 September 1, 1992

18. Salem Unit 1 and 2 Vessel Fluence Update, PSE&G Calculation File DOl.6-770, December 12, 1991.
19. NFU-060 Support Calculations, PSE&G Calculation File DOl.6-845, September 21, 1992.
20. Salem 10CFR50.61 PTS Assessment for 1989, PSE&G Calculation File DOl.6-524, May 18, 1989.
21. Analysis of Capsule Z from Salem Unit 1 Reactor Vessel Radiation Surveillance Progra, Westinghouse Electric Corporation, WCAP 11955, September 1988.
22. Analysis of capsule U from Salem Unit 2 Reactor Vessel Radiation Surveillance Progra, Westinghouse Electric Corporation, WCAP 11554, September 1987.

A:JOCFRSO.BDG Page 38

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