ML18054A631

Revised Analysis of Reactor Pressure Vessel Fast Fluence for Palisades Nuclear Plant Through End of Cycle 8.

ML18054A631
Person / Time
Site:	Palisades
Issue date:	03/31/1989
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
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ML18054A630	List: ML18054A628 ML18054A631 ML18054A634
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NUDOCS 8904120232
Download: ML18054A631 (33)

Text

ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 PALISADES REACTOR VESSEL FLUENCE REDUCTION REPORT April 3, 1989 62 Pages OC0389-0107-NL04 e:*;t(>4 :l 2C>2:32 PDR ADOC:K p

ANALYSIS OF THE REACTOR PRESSURE VESSEL FAST FLUENCE FOR THE PALISADES NUCLEAR PLANT THROUGH END OF CYCLE 8 March, 1989

-* Performed by the Reactor Engineering Department Palisades Nuclear Plant Consumers Power Company

• SECTION TABLE OF CONTENTS TITLE PAGE

1. 0 INTRODUCTION 1 2.0

SUMMARY

3 3.0 METHODOLOGY 7

3. 1 OVERVIEW 7 3.2 FUEL MANAGEMENT 7 3.3 GEOMETRY 8 3.4 ~ MATERIAL CROSS SECTIONS 8 3.5 NEUTRON SOURCE 8 3.6 NEUTRON TRANSPORT ANALYSIS 9 3.7 FLUENCE LIMITS 9 3.8 FLUENCE EXTRAPOLATION 10 4.0 RESULTS 20 4.1 FLUENCE DISTRIBUTION 20 4.2 VESSEL LIFETIME 20 4.3 CALCULATION UNCERTAINTY 20 5.0 DISCUSSION 25 5.1 IMPACT OF RESULTS 25 5.2 FUTURE ACTION 25

6.0 REFERENCES

27 7.0 APPENDICES 28 7.1 WESTINGHOUSE FLUENCE ANALYSIS REPORT 7.2 CPCO FLUENCE ANALYSIS PROGRAM MI0887-0055A-OP03 i

LIST OF TABLES TABLE TITLE PAGE 2.1 Summary of Reactor Vessel Fluence Limits 5 2.2 Fast Neutron Flux Reduction Achieved with Cycle 6 8 Loading Pattern

3. 1 Target Fluence Limits for Pressure Vessel Welds and Base Metal Based on 10CFRS0.61 3.2 Target Fluence Limits for Pressure Vessel 12 Welds and Base Metal Based on Regulatory Guide 1.99, Rev 2.

4.1 Inner Vessel Wall Flux and Fluence (E>l.O MeV) -21 4.2 Summary of Fluence Limit Violation Dates 22 with Current Fuel Management to End of Life 4.3 Summary of Fluence Limit Violation Dates 23 with Cycle 8 Fluence Reduction to End of Life

• MI0887-0055A-OP03 ii

LIST OF FIGURES FIGURE TITLE PAGE 3.1 Cycle 8 Peripheral Loading Pattern 13 3.2 Palisades Core Geometry Cycles 1 - 7 14 3.3 Palisades Core Geometry Cycle 8 15 3.4 Plan View of Palisades Wall Capsules 16 3.5 Assembly Radial Power Distribution for Cycles 1 - 7 17 3.6 Assembly Radial Power Distribution for Cycle 8 18 (Preliminary) 3.7 Assembly Radial Power Distribution for Cycle 8 19 (As-Loaded) 4.1 Azimuthal Reactor Vessel Fast Fluence Distribution 24 7.1 Ex-Vessel Dosimeter Locations

• MI0887-0055A-OP03 iii

1

1.0 INTRODUCTION

The final rule concerning fracture toughness requirements for protection against pressurized thermal shock events was published in July, 1985 as 10CFRSO. 61. The rule required a submittal of projected values of the re_ference temperature for pressurized thermal shock (RTPTS) for reactor vessel beltline materials at the vessel inner surf ace as calculated using the method specified in 10CFRS0.61, paragraph b.2. Consumers Power's initial response ,to the rule

[l] indicated that no concern existed for Palisades as the projected RTPTS values were at or within the given screening criteria.

In a NRC response to the initial submittal [2], Consumers was requested to provide additional information and was advised that the initial conclusion that the screening criteria were met was incorrect. An incorrect margin term had been incorporated into the calculation of the base metal RTPTS; use of the correct term caused the screening criteria of 270°F to be exceeded. Consumers' response [3] provided the requested information and also o~tlined a fluence reduction program to be initiated with the next fuel cycle (Cycle 8). In addition, the response committed to re-evaluating the current accumulated pressure vessel fast fluence (E>l.O MeV) and providing a report to the NRC describing the expected results from implementation of a fluence reduction program for Cycle 8 and beyond. By letter [4], the NRC accepted our commitment to provide a report to the staff by September 30, 1987. As

-described below, our letters of September 29, 1987 and March 9, 1988 presented justification for delay.

Prior to submittal of Consumers' fluence reduction program report, discussions with the NRC in late 1987 indicated that the reference temperature (RT) corre-lation in 10CFRS0.61 would probably be changed to the correlation used in the draft Regulatory Guide 1.99, Rev 2. It was recommended that Consumers examine the impact of this change on the fluence reduction program under development; a new schedule for submitting the program report was submitted to the NRC in Sep-tember, 1987 [5]. Evaluation of the proposed RT correlation change revealed that the limiting beltline material changed to the vessel axial welds, with more restrictive fluence levels corresponding to the PTS screening criteria.

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2 Based on the results of the evaluation, Consumers elected to redesign the planned Cycle 8 loading pattern to incorporate thrice-burned fuel assemblies with st.ainless steel shielding rods located near the axial weld locations to maximize the attainable fluence reduction via fuel management. A revised mile-stone schedule for submitting the fluence reduction report was provided to the NRC [6]. The information contained herein is intended to address the fluence evaluation and reduction program as developed for Cycle 8. In the November 1988 initial submittal to the NRC [9], it was mentioned that an error had been made in generating the neutron source file used to calculate the vessel wall fluxes.

Later, it was discovered that there was also an error in the cross-section file.

In the present revised report, both of these corrections have been incorporated.

The data base utilized in this study consisted of power distribution informa-tion developed from core monitoring system data and fuel vendor-supplied core simulator models, and plant-specific material compositions. The modeling of the vessel and fluence analysis were performed by Westinghouse (Section 7.1 of this report), using the DOTIIIW discrete ordinates transport code and the SAILOR and BUGLE-80 cross section libraries

• Results presented in this report address the accumulated vessel fluence through

\

the end of Cycle 7 as well as the flux reduction obtained with the Cycle 8 low-leakage loading pattern. Vessel fluence limits based on the 10CFRS0.61 PTS screening criteria and both the 10CFRS0.61 and Regulatory Guide 1.99, Rev 2 RT correlations are calculated based on the vessel material chemistries. Vessel lifetimes are calculated relative to the fluence limits assuming both standard and flux-reduction fuel management. In addition, appendices are included that provide the complete Westinghouse fluence analysis report and the status of Consumers' in-house fluence analysis program.

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3 2.0

SUMMARY

Pressure vessel fluence limits based on the PTS screening criteria of 10CFRS0.61 were calculated using the reference temperature -(RT) correlations of both 10CFR50.61 and Regulatory Guide 1.99, Rev 2 using the vessel chemistries provided in Reference 3. The results are summarized in Table 2.1 and show the dramatic reduction in the vessel weld fluence limits with the use o'f the Regulatory Guide 1.99 RT correlation. The axial welds become the limiting material as opposed to the base metal which was the limiting material under the 10CFR50.61 RT correlation.

A loading pattern was designed for Cycle 8 that provided substantial flux re-duction at the axial weld locations. The associated flux reductions for the primary vessel materials are shown in Table 2.2. Fast flux (E>l,O Mev) reduc-tions of approximately 50% ~ere obtained at the axial weld locations, with modest reductions of about 14% at the circumferential and base metal peak_

locations. The Cycle 8 loading pattern used in this analysis_ was not the final loading pattern used to load the reactor; the final loading pattern predicted radial power distributions indicate slightly higher power levels overall in the peripheral assemblies. The minor differences between the two loading patterns are not expected to drastically alter the final results.

Vessel lifetime based on when the PTS screening criteria- are met were deter--

mined for fuel management schemes with and without flux reduction for Cycle 8 and beyond. The fluence limits were based on the more restrictive Regulatory Guide 1.99, Rev 2 RT correlation and assumed that Cycle 8 concluded in March, 1990, with operation beyond end of Cycle 8 at a capacity factor of 75%. With no flux reduction utilized, the PTS screening criteria would be exceeded at the axial welds in approximately October, 1995; with flux reduction incorporated in Cycle 8 and beyond, the PTS limit would be exceeded at the axial welds again, but not until about March, 2002. Both predicted dates are far short of the assumed nominal plant operating license expiration date of December, 2011.

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4 While the flux reduction obtained in Cycle 8 substantially reduced the axial weld flux levels, the reduction is insufficient to remain within the PTS screening criteria through the minimum plant life (nominal end of operating license). Further flux reduction will be possible through more aggressive low-leakage fuel management in Cycle 9 and beyond, and possibly through reanalysis of Cycle 8 using the final, refined loading pattern model, but the incremental improvement believed possible will probably still be insufficient.

In order to allow plant operation at least until the nominal license expiration date, additional PTS-addressing measures will have to be implemented (eg, Regula-tory Guide 1.154 analysis, vessel annealing, etc).

Results of further flux reduction obtained with more aggressive low-leakage fuel management in Cycle 9 and beyond, including reanalysis utilizing the final Cycle 8 loading pattern model, will be provided in a report to the NRC in August, 1989 .

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5 TABLE 2.1

SUMMARY

OF REACTOR VESSEL FLUENCE LIMITS PTS Screening Fluence Limit (10 19 n/ cm 2)

Material Criterion (°F) 10CFRS0.61 RG 1. 99, R72 Axial Weld 270 7.971 1.634 Circumferential Weld 300 11. 984 3.495 Base Metal 270 4.134 6.046 MI0887-0055A-OP03

IJ 6

TABLE 2.2 FAST NEUTRON FLUX REDUCTION ACHIEVED WITH CYCLE 8 LOADING PATTERN Neutron Flux (n/cm 2 sec, E>l.O MeV)

Material Cycles 1 - 7 Cycle 8 Flux Reduction (%)

Axial Weld oo 4.23 x 1010 2.07 x 1010 -51.1 30° 4.17 x 1010 2.21 x 1010 -47.0 Circumferential Weld 5.62 x 1010 4.81 x 1010 -14.4 Base Metal 5.62 x 1010 4.81 x 1010 -14.4 MI0887-0055A-OP03

7 3.0 METHODOLOGY

3. 1 Overview The pressure vessel fast neutron fluence levels (E>l.O MeV) were calculated utilizing available historical and predictive fuel cycle information. Flux calculations were performed by Westinghouse (See Section 7.1).

The primary analytical model was based on a two dimensional (R,8) discrete

• ordinates code DOTIIIW representation of the Palisades reactor vessel config-uration. The representation includes a model of the core/vessel geometry, the neutron source distribution, and nuclear interactions as represented by cross section data. Two calculations were performed. The first calculation modeled a configuration representative of an average over cycles 1 through 7. The second calculation modeled a Cycle 8 configuration having thrice-burned fuel, assemblies with stainless steel shielding rods near the axial weld locations._

3.2 Fuel Management Palisades followed a standard OUT-IN fueling scheme through Cycle 7. In this scheme, only fresh fuel was placed around the core periphery. This approach results in the maximum overall core neutron leakage and fast flux to the reactor vessel, but minimizes power peaking and generally provides the

. greatest thermal margin.

With the issuance of Regulatory Guide 1.99, Rev 2, it was found that the axial welds would be responsible for limiting the life of the Palisades reactor vessel. It was decided to alter the fuel management strategy to distribute the power away from these critical weld locations for Cycle 8 operation. A low leakage loading pattern was adopted to improve the neutron economy and to reduce the fluence levels at the axial welds. Since the reload fuel for Cycle 8 had already been built, the flexibility to alter the fuel management was limited due to the predetermined new fuel assembly enrichment and quantity. A total of 16 thrice-burned stainless steel shielded assemblies were installed at the MI0887-0055A-OP03

8 core periphery. In addition, 8 twice burned assemblies were placed on the core periphery. The remaining 24 peripheral locations were filled with fresh fuel assemblies (Figure 3.1). With this arrangement, it is anticipated that the reduced power in the peripheral assemblies would reduce the primary source of fast neutrons reaching the reactor vessel axial welds.

3.3 Geometry The Palisades reactor exhibits one-eighth (1/8) core symmetry, thus only a zero to 45 degree sector has been included in the DOT models (Figures 3.2 and 3.3).

Figure 3.2 is typical for Cycles 1 - 7 geometry. Figure 3.3 represents the core modification for Cycle 8, with the replacement of fuel by stainless steel rods in four outer rows of two assemblies. In these models a surveillance capsule at the 20° azimuthal location is also shown. This capsule corresponds to the W-290 capsule (Figure 3.4), which was removed at the end of Cycle 5 and analyzed for measured flux values [7].

3.4 Material Cross Sections The DOT model analysis employed a P expansion of the scattering cross sec-3 tions. The microscopic cross sections used in the analysis were obtained from the SAILOR and BUGLE-80 cross section libraries. Macroscopic cross sections were calculated for each region in the model. Plant specific material composi-tions and the corresponding atomic densities were used for this analysis.

3.5 Neutron Source The neutron source inputs to the DOT calculations were determined from the core power distributions [8]. Detailed pin-by-pin power distributions for the outer assemblies were used. Normalized radial assembly power distributions based on assembly energy generation are shown in Figures 3.5 and 3.6 for Cycles 1 - 7 and Cycle 8, respectively. Figure 3.7 shows the predicted normalized radial power distribution for the actual, as-loaded Cycle 8 loading pattern.

MI0887-0055A-OP03

,._,' l_\

9 3.6 Neutron Transport Analysis

. The spatial distribution of neutron flux in the reactor was calculated using the DOT computer code. DOT program solves the Boltzman transport equation in two-dimensional geometry using the method of discrete ordinates. Third order scattering (P ) and s angular quadratures were used. Fluence levels were 3 8 obtained by multiplying the flux by the effective full power seconds at 2530 MWT.

3.7 Fluence Limits Target fluence limits for pressure vessel welds and base metals are calculated using the 10CFR50.61 correlation for RTPTS and the vessel material PTS screening criteria. The reference temperature correlation is given as:

I+ M +(-10 + 470Cu + 350CuNi)fD*27 Where RTPTS is the adjusted reference temperature for pressurized thermal shock considerations (°F)

I is the initial reference temperature (°F)

M is the margin term (°F)

Cu, Ni are the copper and nickel content (in weight percent), respectively f is the accumulated fluence (E>lMeV) in units of l0 19 n/cm 2 The corresponding fluence limits are determined by solving the RT correlation for the fluence value. The target limits are shown in Table 3.1.

Target fluence limits for pressure vessel welds and base metals are also calculated using Regulatory Guide 1.99, Rev 2 reference temperature correlation and the 10CFR50.61 PTS screening criteria.

MI0887-0055A-OP03

10 The adjusted reference temperature for each material in the belt line is given as:

RT = I+ M +(CF*f(0.28-0.10 log f))

NDT I, M and f have the same meaning as above. CF, the chemistry factor, depends on the contents of copper and nickel in the belt line materials. This factor is provided in Regulatory Guide 1.99, Rev 2. The corresponding fluence limits are determined by solving the RT correlation for the fluence valu~ and are shown in Table 3.2.

3.8 Fluence Extrapolation Based upon the Cycles 1 - 7 averaged fuel management scheme, the estimated end of plant life dates are calculated relative to the fluence limits of 10CFR50.61 and Regulatory Guide 1.99, Rev 2 for the axial welds, circumferential weld and base metal. Similar calculations are performed for the Cycle 8 fuel managem.ent scheme. It is assumed that the Palisades Plant will be operating at 75%

capacity factor after EOC8, which is currently scheduled to end in March, 1990.

The nominal projected end of plant life is estimated by assuming that the* plant would operate 40 years beyond initial criticality under its existing provi-sional license to December, 2011.

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I* It 11 TABLE 3.1 TARGET FLUENCE LIMITS FOR PRESSURE VESSEL WELDS AND BASE METAL BASED ON 10CFR50.61 Chemistry 10CFR50.61 Parameters Fluence Screening Limit Material Cu(w/o) Ni(w/o) I(°F) M(°F) Criterion(°F) (10 19n/ cm 2 )

Axial Weld .19 1.10 -56 59 270 7. 971 Circumferential Weld .20 .97 -56 59 300 11. 984 Base Metal .25 .54 - 5 48 270 4.134 MI0887-0055A-OP03 i

',j

' It 12 TABLE 3.2 TARGET FLUENCE LIMITS FOR PRESSURE VESSEL WELDS AND BASE METAL BASED ON REGULATORY GUIDE 1.99, REV 2 Chemistry. Reg Guide 1.99 Parameters Fluence Screening Limit Material Cu(w/o) Ni(w/o) I(°F) M(°F) Criterion(°F) (lol9n/cm 2 )

Axial Weld .19 1.10 -56 66 270 1.634 Circumferential Weld .20 .97 -56 66 300 3.495 Base Metal .25 .54 - 5 34 270 6.046 Chemistry Factors: Axial Weld - 229.00°F Circumferential Weld - 218.65°F Base Metal - 167.60°F MI0887-0055A-OP03

13r figure 3.1 Cycle 8 Peripheral Loading Pattern Stainless Steel Shield Rods 180° Number of Accumulated

• 2 0

......_41-_ *--i--+---+-+--+--+--+-+-~~--_,_-"_..__..._~3~

90 ° . _

3

--;+------ 2 70 ° 330° Axial Weld Azimuthal Location

I* a' 14 Figure 3.2 Palisades Core Geometry Cycles l - 7 Ill

... J..

J"J" (O""..r.

~*.....

"\ ~

~~

C'/ (0/ ~-~ '&

9>17 ~ *.$' '6'

• ~\

~....... '~

111

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(0/  ;.--o> *o

• 9'. :;_)

\ ...-". 9'

...-". ...9._,,

Ill

..90 *.9.

• ./_ '.9

<9 Ill

~*....

" Fuel H

..... IH.I 1w.1 111.1 ..... ..... .....

I Al! 1119 t Cftl 1n. I ..... .....

Ml0887-0055A-<lP03

15 Figure 3.3 Palisades Core Geometry Cycle 8 Ill Ill Ill IH Ill SS

" Rod Regions II

"

• 71 111.1 ..... 111.1 .....

110 IUI 1u.1 I Cll I

..... 111.1 .....

1+-~~~....._..L-~~+-~~~+-~.....t...-U...._~...1...L-..._~_..1L--+--~..J.1..~f-l-~~--1--~~--+~~~-4~..l.-~.-I 1***I "'*I HI0887-0055A-OP03

t6 Figure 3.4 Plan View of Palisades Watt Capsules W1U W-80

~lentad A-IO W&ll Rector V euel w-100 W1U W-110 Cont Shroud w111----

w-2t0 Wiii

---

__:~~~~~-.-~~~~~~----------Wml w-2tO w-210

,..,. \liew MI0887-0055A-OP0l

' I l7

• Figure 3.5 Assembly Radial Power Distribution for Cycles 1 - 7 o*

I I

I 0.892 0.874 0.695 l . 138 l. 14 7 l .152 l .133 0.956 0.611

o. 981 1.074

1. 144 1.145 1.153 1.11 & l . 127 l .002 0.923 1. 059 0.972 1.032 1.142

l. l 09 1 . 122 l . 116 1. 127 l.015 1.171 0.938 l. 175 1.103 l .144

1. 108 1.005 1. 141 1.069 0.972 1.118 1.047 1 .074 1.126 0.937 0.96S 0.892 Normalized Power 1.138 Axial Peaking Factor for Outer Assemblies MI0887-0055A-OP03

18 Figure 3.6 Assembly Radial Power Distribution for Cycle 8 (Preliminary) o*

I I

I 0.356 0. 756 ' 0.641

1. 084 l. 135 I 1. 145 I

l

• l 01 1. 1061 1. 133 0.576 0. 221

1. l 04 1. 108 : 1. 139 1. 114 1. 12a 1.206

1. 098 1.259 1.008 1. 165 0.907 ,45*

1. 1OJ 1. 103 1

• 155 1.164 /

/

/

l. 219 1. 043 1.299 1. 076 1. 239

1. 115 1.166 1.009 1.286 1. 005 1.133 .

1.200 0.997 1.100 1.166 1.128

. ~

0.931 0.356 Normalized Power 1.084 Axial Peaking Factor for Outer Assemblies HI0887-0055A-OP03

19

• Figure 3.7 Assembly RAdial Power Distribution for Cycle 8 (As Loaded) a*

I I

I

.363 .764 .629 1.126 1.163 1.068 .599 .226 45*

1.210 1.194 1. 002 1. 223 .905 /

/

1.218 1.034 1.263 1.040 1. 208 1.017 1.259 1.016 1.126 1.217 1.010 1.123 1.182 1.144

.947

20 4.0 RESULTS 4.1 Fluence Distribution The azimuthal distribution of inner vessel wall fluence is given in Figure 4.1.

For cycles 1 through 8, the maximum peak flux/fluence occurs at the nominal 16° location on the vessel wall. Fluence levels at the two axial weld locations 0° and 30° are marked by vertical lines. Fluence limits corresponding to axial welds, circumferential welds and base metal are shown as horizontal lines.

These limits are calculated per Reg Guide 1.99, Rev 2. Table 4.1 provides the flux/fluence levels for 0°, 16° and 30° azimuthal locations. In this table, other parameters used for the calculations of the fluence from the flux (Sec 7.1) are also provided.

4.2 Vessel Lifetime Table 4.2 provides the fluence limit violation dates assuming the standard fuel management scheme, typical of cycles 1 through 7, for Cycle 8 and beyond .

Calculated dates are given both for the 10CFRS0.61 PTS screening criteria and the Reg Guide 1.99, Rev 2 RT correlations (based upon the known vessel material chemistries). Table 4.3 is presented for the flux-reduction fuel management scheme assuming the Cycle 8 loading pattern, with thrice-burned fuel assemblies with stainless steel shielding rods at the selected core locations (Figure 3.1), for Cycle 8 and beyond.

4.3 Calculation Uncertainty Fuel loading patterns for Cycles 1 through 7 are very similar to each other.

Measured flux for the W-290 capsule at the EOC 5 is 6.73 X 1010 n/cm 2 -sec [8].

This compares well with the calculated flux at the center of W-290 capsule of 7.47 X 10 1 0 n/cm 2 -sec (Section 7.1, Table 3 - 7) for Cycles 1 - 7. The calcula-ted flux is about 11% higher than the measured value.

A number of factors contribute to the uncertainty in the projected peak fast

• fluence at the reactor vessel wall. These factors are due to the conversion of measured activity data to the flux, material composition and neutron cross sections, power distribution and cycle-by-cycle variation in the lead factors.

An uncertainty of +/-25% is estimated in the calculated vessel wall fluence.

MI0887-0055A-OP03

' (

21 TABLE 4.1 INNER VESSEL WALL FLUX AND FLUENCE eE>l.O MeV)

Azimuthal CYCLES 1 - 7 CYCLE 8 CYCLES 1 - 8 Location Average Flux Total Fluence Average Flux Total* Fluence ex 10 10 n/cm 2 -sec) ex 10 19 n/cm 2 ) ex 10 10 n/cm 2 -sec) ex L0 19 n/cm 2 )

oo Axial Weld 4.23 0. 951 2.07 1. 014 16° Base Metal 5.62 1. 264 4.81 1.411 30° Axial Weld 4.17 0.938 2.21 1.006 NOTES: 1. The rated full power is 2530 MWT.

2. Flux values are computed at 0.125°, 16.44° and 30° for the nominal angles of 0°, 16° and 30° respectively.

3. EFPY at the EOC 7 is 7.13.

4. EFPY at the EOC 8 is the estimated value of 8.10

• MI0887-0055A-OP03

TABLE 4.2

SUMMARY

OF FLUENCE LIMIT VIOLATION DATES WITH CURRENT FUEL MANAGEMENT TO END OF LIFE Material Fluence Levels ( 10 19 n I cm 2) 10CFR50.61 Reg Guide 1.99, R/2 EOC7 (8788) EOC 8 (3790~ Increment Time to Limit Date of Limit Time to Limit Date of Limit (Approx.) (Approx.)

Axial Weld oo .951 1.080 .133/EFPY >50 yrs 5.56 yrs 10/1995 30° .938 1.066 .132/EFPY >50 yrs 5.73 yrs 11/1995 Circumferential Weld 1.264 1.436

• 177 /EFPY >50 yrs 15.51 yrs 9/2005 Base Metal 1.264 1.436 .177 /EFPY 20.3 yrs 7/2010 34.73 yrs 11/2024 All times and dates based on 75% capacity factor after EOC8 (3/90).

[\)

[\)

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TARLE 4.3

SUMMARY

OF FLUENCE LIMIT VIOLATION DATES WITH CYCLE 8 FLUENCE REDUCTION TO END OF LIFE Material Fluence Levels (10 19 n/cm 2 ) 10CFR50.61 Reg Guide 1.99, R/2 EOC7 (8 788) EOC 8 (3790~ Increment Time to Limit Date of Limit Time to Limit Date of Limit (Approx.) (Approx.)

Axial Weld oo .951 1.014 .065/EFPY >50 yrs 12. 72 yrs 12/2002 30° .938 1.006 .070/EFPY >50 yrs 11.96 yrs 3/2002 Circumferential Weld 1.264 1. 411 .152/EFPY >SO yrs 18.28 yrs 6/2008 Base Metal 1.264 1. 411 .152/EFPY 23.89 yrs 2/2014 40.65 yrs 11/2030 All times and dates based on 75% capacity factor after EOC8 (3/90).

[\)

w MI0887-0055A-OP03

FIG 4.1 AZIMUTHAL ACTOR VESSEL FAST FLUENCE DISTRIBUTION 24 s.so..:

REG GUIDE 1 .99 REV 2 LIMifn 6-046 E+l9 BASE i~Ei AL o.00-: ..

5.JO-w s.oo-.

0 -4. so -.

w I \.

~. 00-:

J N

~ 3.so - REG GUIDE 1.99 REV 2 LlMJia 3.495 E+19 CIRCUMFERENTIAL ~ELD

• ~

<..)

z:

3.00..;

0)

+ 2.so-w

-..J w

u 2.00 -

z w 1 AXIAL ';ELD

J

_J 1.soi LL.

~

1 .oo] F.OC 8 o.so ' 0 AND JO DEGREES ARE AXIAL WELD LOCATIONS F.OC 7 l

~

o.oo 0 5 10 15 20 25 30 35 '40 AZIMUTHAL ANGLE CDEGREES)

• , t "

25 5.0 DISCUSSION 5.1 Impact of Results The utilization of the Reg Guide 1.99, Rev 2 reference temperature (RT) correla-tion with the current 10CFR50.61 PTS screening criteria substantially reduces the allowable fluence at which the screening criteria are met. In addition to lowering the allowable fluence limits, the limiting material was ~elocated to the axial welds as opposed to the base metal.

The flux reduction obtained with the Cycle 8 loading pattern of about 50% at the axial welds and about 14% at the circumferential weld/base metal peak locations delays exceeding the PTS screening criteria to about March, 2002, as opposed to October, 1995 if no additional flux reduction measures were taken.

The flux reduction is insufficient, however, to allow operation of the plant within the PTS screening criteria until the minimum expected plant life, corresponding to the expiration of the current operating license. in_ D.ecember,,

2011.

In order to maximize vessel lifetime, further measures must be taken in the areas of greater flux reduction, Reg Guide 1.154 analysis to properly define the real PTS screening criteria, and possible vessel annealing/shielding actions to reduce the accumulated vessel embrittlement.

5.2 Future Action The most straightforward method of reducing the vessel fast flux levels is reduction of the source itself, which has been initially addressed 'with the incorporation of low-leakage fuel management and stainless steel shield rods in Cycle 8. Two more refinements of the source reduction technique involve reanalyzing the Cycle 8 flux levels using the final loading pattern and utiliz-ing more aggressive low-leakage fuel management in Cycle 9. Reanalyzing Cycle 8 is not expected to provide a substantial improvement in the vessel fluxes; utilizing more burned fuel on the periphery for Cycle 9 could result in flux levels lower than Cycle 8 on the order of several percent at the axial weld locations, which would possibly add one to two years to the date at which the screening criteria are exceeded. Both of these issues will be addressed in the fluence program report to be provided to the NRC in mid-1989.

MI0887-0055A-OP03

26 Since the planned flux reduction measures do not appear to fully solve the vessel fluence issue relative to PTS, investigations have been initiated to possibly implement a Reg Guide 1.154 evaluation to more realistically define the vessel PTS temperature limits. The schedule for performing the evaluation will be more fully defined once the results from the 1989 fluence reduction report are available. Also, discussions have been held with NSSS vendors on the possibility of performing vessel annea~ing and critical material area shielding.

In order to benchmark vessel fluence calculations, an upgraded vessel dosimetry program has been initiated to supplement the existing surveillance capsule program. An ex-vessel dosimetry program was developed by Westinghouse and hardware installation occurred during the. end of Cycle 7 refueling outage. The dosimetry installed will provide detailed azimuthal and axial mapping of the 270-360 degree vessel quadrant, with gradient chains installed in the other three quadrants to provide accurate axial and cross-quadrant mapping. It is intended to change out the dosimetry each cycle to provide accurate cycle-specific results. The dosimetry will provide measured data for use in.

vessel wall and vessel support fluence evaluations. In addition to the ex-vessel program, Combustion Engineering was contracted to fabricate and install a replacement in-vessel dosimetry capsule to be inserted into the W-290 capsule holder vacated following Cycle 5. Installation attempts were unsuccess-ful during the end of Cycle 7 refueling outage; the capsule will be installed during the next refueling, however. When install~d, this capsule will proyide an excellent through-wall correlation with the ex-vessel dosimetry installed ih the same quadrant.

In addition to implementing the supplemental dosimetry program, efforts have been underway to develop in-house capability to perform vessel fluence analyses using the industry-standard DOT methodology. Initial training in model development and analysis methodology were obtained from Combustion Engineering; in-house capabilities will allow convenient and economical evaluation of vessel fluence accumulation on a cycle-by-cycle basis without depending on outside vendors. Further details on the supplemental dosimetry program and the in-house calculational Erogram development are included in Appendix 7. 2.

MI0887-0055A-OP03

27

6.0 REFERENCES

1. Letter from K W Berry (CPCo) to NRC, "Response to 10 CFR SO. 61 - Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," January 23, 1986.

2. NRC letter from T V Wambach to K W Berry, "Pressurized Thermal Shock Rule (PTS), 10 CFR 50.61, Response for Palisades Plant," May 6, 1986.

3. Letter from K W Berry to NRC, "Response to Request for Additional Informa-tion - Pressurized Thermal Shock (PTS) Rule 10 CFR 50.61," August 7, 1986.

4. NRC Letter from A C Thadani to K W Berry, "Palisades Plant - Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, 10 CFR 50.61," December 5, 1986.

5. Letter from R W Smedley (CPCo) to NRC, "Compliance with Pressurized Thermal Shock Rule 10CFR50.61 - Additional Information," September 29, 1987.

6. Letter from R W Smedley (CPCo) to NRC "Revised Schedule for Compliance with Pressurized Thermal Shock Rule - Draft Regulatory Guide 1.99 Revision 2, and 10CFR50. 61," March 9, 1988.

7. WCAP - 10637, Analysis of Capsules T-330 and W-290 from the Palisades.

Reactor Vessel Radiation Surveillance Program, M K Kunka and C A Cheney, September, 1984.

8. Engineering Analysis Package for PTS study, Reactor Engineering Department, Palisades Plant (1987-88).

9. Letter from R W Smedley (CPCo) to NRC "Compliance with Pressurized Thermal Shock Rule 10 CFR 50.61 and Regulatory Guide 1.99, Revision 2 -

Fluence Reduction Status". Attachment (Preliminary), November 30, 1988.

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28 7.0 APPENDICES 7.1 Westinghouse Fluence Analysis Report Included in this section is the entire Westinghouse fluence analysis report performed in March 1989,- with revised neutron source and cross section data.

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