ML20154Q100

Adjoint Flux Program for Point Beach,Units 1 & 2

ML20154Q100
Person / Time
Site:	Point Beach
Issue date:	12/31/1984
From:	Shaun Anderson, Balkey K, Lau F, Meyer T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20154Q088	List: ML20154Q083 ML20154Q100
References
REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR TAC-59972, TAC-59973, WCAP-10638, NUDOCS 8603210031
Download: ML20154Q100 (58)

Text

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WESTINGHOUSE PROPRIETARY CLASS 3 WCAP-10638 CUSTOMER DESIGNATED DISTRIBUTION

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ADJOINT FLUX PROGRAM FOR POINT BEACH UNITS 1 AND 2 S. L. Anderson K. R. Balkey WORK PERFORMED FOR WISCONSIN ELEC1RIC POWER COMPANY DECEMBER 1984 Approved:

((A Approved:

7

' f%M F. L. Lau, Manager T.A.Meyer,kanager Radiation and Systems Reactor Vessel Integrity Analysis Programs Although the information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its Licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION NUCLEAR ENERGY SYSTEMS P. O. BOX 355 PITTSBURGH, PENNSYLVANIA 15230 kh k O K

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i EXECUTIVE

SUMMARY

i l

i The Westinghouse Adjoint Flux tool has been used to assess the ef fects that past, present, and projected fuel management strategies have on fast neutron i

exposure levels in the Point Beach Unit 1 and 2 reactor pressure vessels.

In regard to the low leakage fuel management already in place at the Point Beach plants, the plant specific evaluations have demonstrated that, for the

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low leakage case, the average peak fast neutron flux at the O' azimuthal t

i position has been reduced by about 30% at Unit 1 and 25% at Unit 2 relative to that existing prior to implementation of low leakage. Other vessel locations have been impacted to a lesser degree. Since significant deviations from the low leakage scheme already in place will af fect the exposure projections j

beyond the current operating cycle, future loading patterns should be i

evaluated, as they evolve, to determine their potential 1:npact on the various considerations related to the reactor vessel.

I The magnitude of the neutron flux at the surveillance capsule locations and 4

the lead factors relating capsule exposure to maximum vessel exposure have been impacted by the in-place low leakage loading. Capsule withdrawal y

schedules should be adjusted based on the plant specific information given in this report and factoring in future changes in fuel management strategy.

Excellent agreement has also been demonstrated between measured data from

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three withdrawn surveillance capsules and the values calculated using the adjoint flux tool.

The fuel management strategies employed to date essentially have had no impact on the schedule of applicability for the heatup and cooldown curves currently in the Point Beach Unit 1 and 2 technical specifications used for plant operations.

The effect of resulting neutron fluence level changes on reactor vessel l

integrity relative to pressurized thernal shock (PTS) has been evaluated in i

terms of reference nil-ductility transition temperature (RTNOT) data.

Because of current uncertainty in the chemistry values for the beltline region welds of both reactor vessels, a parametric study was employed.

11 For both vessels, the RT values based upon the docketed and NRC estimated NOT weld material chemistry data remain below the proposed NRC screening values l

for PTS using projected fluence exposures through end-of-life (i.e., 32 effective full power years). Caution must be exercised on all future fuel loadings to insure that this situation is maintained. Additional flux reductions beyond those currently in place would be needed to comply with the NRC screening criteria if the weld chemistry values are concluded to be near the upper estimated data or if consideration is given in the future to vessel life extension.

The results provided in this report should enable Wisconsin Electric Power Company to comply with the initial requirements of the forthcoming NRC rule l

for PTS.

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TABLE OF CONTENTS P,Lqe, EXECUTIVE

SUMMARY

i TABLE OF CONTENTS iii LIST OF TABLES iv LIST OF FIGURES viii I.

INTRODUCTION 1

II. NEUTRON EXPOSURE EVALUATION 3

1. Method of Analysis 3

2. Fast Neutron Fluence Results 8

3. Influence of an Energy Dependent Damage Model 34

4. Potential Impact on Excore Neutron Detectors 36

5. Surveillance Capsule Withdrawal Schedules 39

6. Impact on Heatup/Cooldown Curve Schedule of Applicability 40 III. IMPACT ON REACTOR VESSEL INTEGRITY 44

1. Current Licensing Basis for Pressurized Thermal Shock 44

2. Identification and Location of Beltline Region Materials 45

3. Definition of Plant Specific Material Properties 45

4. Status of Reactor Vessel Integrity in Terms of RT NDT

5. Discussion of Results 57 I 'v'. 00nttuS10N5 AND RECOHHENDATIONS 59 V.

REFERENCES 62 VI.

APPENDICES A-1 A. Power Distributions

_1 B. RT Values of Point Beach Unit 1 and 2 Reactor Vessel NDT j_1 Beltline Region Materials

o SECTION I INTRODUCTION The Westinghouse Adjoint Flux Program provides a cost effective tool to assess the effects that past and present core management strategies have had on

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neutron fluence levels in the reactor pressure vessel, along with several other considerations. This report presents the results from application of the adjoint flux program to the Point Beach Unit 1 and 2 reactor vessels for Wisconsin Electric Power Company. Both of these plants have operated for several fuel cycles using non-design basis core management schemes.

The adjoint flux program also provides results that will enable Wisconsin Electric Power Company to comply with the initial requirements of the forthcoming (December 1984) Nuclear Regulatory Commission (NRC) rule for pressurized thermal shock (PTS) [1].

Section II outlines the adjoint methodology, discusses the fuel management schemes that have been used to date, and presents the neutron fluence data for both the design basis and actual core management strategies. The resulting impact of these actual fuel management schemes on the intermediate and power range excore detectors, the lead f actors and removal schedules for the surveillance capsules, and the schedule of applicability for the heat-up and cooldown curves is also given.

The effect of resulting fluence level changes on reactor vessel integrity relative to PTS is presented in Section III.

Following a discussion of the beltline region material property data for the two reactor vessels, reference nil-ductility transition temperature (RTNOT) res d s are gesented in accordance with the NRC proposed PTS rule requirements [1]. Because of uncertainties in the chemistry data for beltline region welds, the RT NDT results are shown via a parametric study to assist utility decision making relative to PTS.

Conclusions and recommendations for utility consideration are given in Section IV, and the references for the report are given in Section V.

o 2

Appendix A shows the power distributions used in the neutron fluence analysis. Appendix 8 provides a tabulation of the input and results for the RT calculation for all beltline region materials for the Point Beach Unit ET 1 and 2 reactor vessels.

i i

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i SECTION II NEUTRON EXPOSURE EVALUATION II.1 METHOD OF ANA!YSIS A plan view of the Point Beach Unit 1 and 2 reactor geometry at the core midplane is shown in Figure II.1-1.

Since the reactor exhibits 1/8th core symmetry only a 0*-45' sector is depicted. Six irradiation capsules attached to the thermal shield are included in the design to constitute the reactor vessel surveillance program. Two capsules are located 180' symmetrically at azimuthal positions of 13', 23', and 33' from the reactor core cardinal axes as shown in Figure 11.1-1.

1 A plan view of a single surveillance capsule attached to the thermal shield is shown in Figure 11.1-2.

The stainless steel specimen container is 1-inch j

square and approximately 63 inches in height. The containers are positioned axially such that the specimens are centered on the core midplane, thus spanning the central 5.25 feet of the 12-foot high reactor core.

In performing the fast neutron exposure evaluations for the reactor geometry shown in Figures 11.1-1 and II.1-2, two sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was utilized to provide baseline data derived from a design basis core power distribution against which cycle by cycle plant specific calculations can be compared. The second set of calculations consisted of a series of adjoint analyses relating the response of interest at several selected locations within the reactor geometry to the power distributions in the reactor core. These adjoint importance functions when combined with cycle specific core power distributions ' yield the plant specific exposure data for each operating fuel cycle.

3

C Ficure II.1-1 Point Beach reactor Geometry PRESSURE VESSEL

-SURVEILLANCE CAPSULE O

13" (CAPSULES V R.

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23 ' CAPSULES T_P, j..

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32 ' ' CAPSULE 5 5 *.

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5 Figure !!,1-2 Surveillance Capsule Geometry s

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The forward transport calculation was carried out in R,0 geometry using the DOT discrete ordinates code [2] and the SAILOR cross-section library [3]. The SAILOR library is a 47 group, ENDF-BIV based data set produced specifically for light water reactor applications.

Anisotropic scattering is treated with 8

expansion of the cross-sections.

3 The design basis core power distribution utilized in the forward analysis was derived from statistical studies of long-term operation of Westinghouse 2-loop plants.

Inherent in the development of this design basis core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery.

Furthermore, for the peripheral fuel assemblies, a 2a uncertainty derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power was used. Since it is unlikely that a single reactor would have a power distribution at the nominal

+2a level for a large number of fuel cycles, the use of this design basis distribution is expected to yield somewhat conservative results.

The adjoint analyses were also carried out using the P cross section 3

approximation from the SAILOR library. Adjoint source locations were chosen at the center of each of the surveillance capsules as well as at positions along the inner diameter of the pressure vessel. Again, these calculations were run in R,0 geometry to provide power distribution importance functions for the exposure parameters of interest.

Having the adjoint importance functions and appropriate core power distributions, the response of interest is calculated as R

I (R,0) F (R,0) R dR de R,0 "

R 0

where:

R,0 = Response of interest ($ (E > 1.0 MeV), dPa, etc.) at R

radius R and azimuthal angle 8.

~~

I (R,0) = Adjoint importance function at radius R and azimuthal angle 0 F (R,0) = Full power fission density at radius R and azimuthal angle 0

7 o

It should be noted that as written in the above equation, the importance function 1(R,0) represents an integral over the fission distribution so that the response of interest can be related directly to the spatial distribution I

of fission density within the reactor core.

Core power distributions for use in the plant specific fluence evaluations for Point Beach Units 1 and 2 were taken from the design of each operating cycle for the two reactors.

The specific power distribution data used in the analysis is provided in Appendix A of this report.

The data listed in Appendix A represents cycle averaged relative assembly powers.

Therefore, the adjoint results are in terms of fuel cycle averaged neutron flux which when multiplied by the fuel cycle length yields the incremental fast neutron fluence.

l The transport methodology, both forward and adjoint, using the SAILOR cross-section library has been benchmarked against the CRNL PCA f acility as well as against the Westinghouse power reactor surveillance capsule data base [4].

The benchmarking studies indicate that the use of SAILOR cross-sections and generic design basis power distributions produces flux levels that tend to be

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conservative by from 7-22%. When plant specific power distributions are used

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with the adjoint importance functions, the benchmarking studies show that fluence predictions are within i 15% of measured values at surveillance capsule locations.

As stated in the preceding paragraph, the use of the generic power distribu-tions along with the P scattering cross-sections produces conservative 3

estimates of pressure vessel exposure.

Prior analysis performed for the Point Beach reactors and reflected in the technical specifications for the two plants (see Section II.6) utilized the generic power distributions with P) scattering cross-sections.

These prior calculations were considered to be best estimate values with an uncertainty of i 25%.

Thus, the prior analysis is more appropriately compared with current calculations using plant specific power distributions than with current design basis calculations that are intended to produce conservative results. Comparisons of this sort will be l

presented on Figures 11.6-1 and 11.6-2 of this report.

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II.2 FAST NEUTRON FLUENCE RESULTS Calculated fast neutron (E >1.0 MeV) exposure results for Point Beach Units 1 and 2 are presented in Tables 11.2-1 through 11.2-14 and in Figures 11.2-1 through 11.2-8.

Data is presented at several azimuthal locations on the inner radius of the pressure vessel as well as at the center of each surveillance capsule.

In Tables II.2-1 through 11.2-4 plant specific neutron flux and fluence levels at 0*,15*, 30*. and 45* on the pressure vessel inner radius are listed for each operating cycle of Point Beach Unit 1.

Also presented are the design basis fluence levels predicted using the generic 2-loop core power distribution at the nominal + 2a level. Similar data for the center of surveillance capsules located at 13*, 23*, and 33* are given in Tables 11.2-5 through 11.2-7, respectively.

In addition to the calculated data given for the surveillance capsule locations, measured fluence data from previously withdrawn surveillance capsules are also presented for comparison with analytical results.

In the case of Unit 1, capsules were removed f rom the 13* location at the end of cycles 1 and 5 as well as from the 33* location following cycle 3.

Plant specific and design basis fast neutron flux and fluence data at the inner radius of the pressure vessel are given in Tables 11.2-8 through 11.2-11 for each operating cycle of Point Beach Unit 2.

As in the case of Unit 1, data are presented for the 0*, 15*, 30*, and 45* azimuthal angles.

Evaluations of plant specific and design basis fluence levels at the three surveillance capsule locations are given in Tables 11.2-12 through 11.2-14.

For Unit 2, surveillance capsules were removed from the 13* position following cycles 1 and-5 and f rom the 23* position af ter the termination of cycle 3.

Oosimetry evaluations f rom these three capsule' withdrawals are also listed in Tables 11.2-12 and 11.2-13.

Several observations regarding the data presented in Tables 11.2-1 through 11.2-14 are worthy of note.

These observations may be sunnarized as follows:

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9 4

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TABLE 11.2-1 F.AST NEUTRON (E > 1.0 Mev) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTHAL ANGLE POINT BEACH UNIT 1 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design (EFPS)

(n/cm2-sec)

Specific Basis I

10 18 I

1 4.63 x 10 4.42 x 10 2.05 x 10 2.23 x 10 10 18 18 2

2.71 x 10 4.42 x 10 3.25 x 10 3.53 x 10 10 18 18 l

3 3.66 x 10 4.03 x 10 4.72 x 10 5.01 x 10 10 18 I9 4

2.18 x 10 4.69 x 10 5.74 x 10 6.06 x 10 I

10 18 18 5

2.43 x 10 4.64 x 10 6.87 x 10 7.23 x 10 I

10 18 IO 6

2.59 x 10 4.20 x 10 7.96 x 10 8.48 x 10 10 18 18 7

2.71 x 10 4.36 x 10 9.14 x 10 9.79 x 10 10 18 I9 8

1.92 x 10 3.44 x 10 9.80 x 10 1.07 x 10 10 I9 9A 1.77 x 10 3.17 x 10 1.04 x 10' 1.16 x 10 I

10 I9 I9 10 1.96 x 10 2.80 x 10 1.09 x 10 1.25 x 10 0

I9 I

11 1.86 x 10 3.49 x 10 1.15 x 10 1.34 x 10 0

Note: Design Basis $ = 4.82 x 10 n/cm - sec I

10 TABLE II.2-2 FAST NEUTRON (E > 1.0 Mev) EXPOSURE AT THE PRESSURE VESSEL INNER RA0!US - 15* AZIMUTHAL ANGLE POINI BEACH UNIT 1 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design (EFPS)

(n/cm2-sec)

Specific Basis 10 18 18 1

4.63 x 10 2.63 x 10 1.22 x 10 1.37 x 10 10 18 18 2

2.71 x 10 2.71 x 10 1.95 x 10 2.17 x 10 3

3.66 x 10 2.49 x 10' 2.86 x 10' 3.25 x 10'O 10 18 4

2.18 x 10 2.86 x 10 3.49 x 10 3.89 x 10 7

10 18 18 5

2.43 x 10 2.89 x 10 4.19 x 10 4.60 x 10 I

10 18 18 6

2.59 x 10 2.62 x 10 4.87 x 10 5.37 x 10 10 IO 7

2.11 x 10 2.65 x 10 5.59 x 10 6.17 x 10 7

10 18 18 8

1.92 x 10 2.11 x 10 5.99 x 10 6.73 x 10 10 18 18 9A 1.77 x 10 2.01 x 10 6.35 x 10 7.26 x 10 10 I0 10 1.96 x 10 1.93 x 10 6.73 x 10'8 7.84 x 10 10 18 18 11 1.86 x 10 2.16 x 10 7.13 x 10 8.38 x 10 10 2

Note: Design Basis + = 2.95 x 10 n/cm -sec

l l

11 O

e IABLE 11.2-3 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE y_ESSEL INNER RADIUS - 30' AZIMUTHAL ANGLE POINT BEACH UNIT 1 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design (EFPS)

(n/cm2-sec)

Specific Basis 10 I#

I#

1 4.63 x 10 1.66 x 10 7.69 x 10 9.12 x 10 7

10 18 18 2

2.71 x 10 1.80 x 10 1.26 x 10 1.45 x 10 10 18 18 3

3.66 x 10 1.66 x 10 1.86 x 10 2.17 x 10 10 18 18 4

2.18 x 10 1.96 x 10 2.29 x 10 2.60 x 10 10 IO 18 5

2.43 x 10 2.01 x 10 2.78 x 10 3.08 x 10 10 18 I0 6

2.59 x 10 1.91 x 10 3.27 x 10 3.59 x 10 10 IO 8

7 2.7I x 10 1.78 x 10 3.76 x 10 4.12 x 10 10 18 IO 8

1.92 x 10 1.53 x 10 4.05 x 10 4.50 x 10 10 0

18 9A 1.7/ x 10 1.50 x 10 4.32 x 10 4.85 x 10 10 1.96 x 10 1.52 x 10 4.61 x 10 5.23 x 10 j

10 18 18 11 1.8b x 10 1.47 x 10 4.89 x 10' 5.60 x 10' 10 Note:

Design Basis & = 1.97 x 10 n/cm2 _ s,c

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12 TABLE 11.2-4 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 45* AZIMUTHAL ANGLE POINT BEACH UNIT 1 Cycle irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant ~

Design

( F.F PS)

(n/cm2-sec)

Specific Basis 10 1

4.63 x 10 1.49 x 10 6.90 x 10'#

8.06 x 10'#

I 2

2.ll x 10 1.60 x 10 1.12 x 10'"

1.28 x 10'"

10 3

3.bb x 10 1.46 x 10 1.86 x 10' 1.91 x 10 7

10 18 4

2.18 x 10 1.76 x 10 2.04 x 10'0 2.29 x 10 7

10 5

2.43 x 10 1.81 x 10 2.48 x 10'"

2.72 x 10'8 10 18 6

2.59 x 10 1.76 x 10 2.94 x 10'O 3.17 x 10 10 1

2.11 x 10 1.b/ x 10 3.36 x 10'"

3.64 x 10' 10 18 18 8

1.92 x 10 1.50 x 10 3.65 x 10 3.97 x 10 10 9A 1.Il x 10 1.47 x 10 3.91 x 10' 4.28 x 10'O 10 IO 10 1.96 x 10 1.55 x 10 4.21 x 10'O 4.62 x 10 10 11 1.86 x 10 1.28 x 10 4.45 x 10'"

4.95 x 10'"

10 2

Note: Design Basis + = 1.74 x 10 n/cm,3,c

13 y,.

TABLE II.2-5

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t'k FAST NEUTRON (E > 1.0 Mev) EXPOSURE AT THE 13* SURVEILLANCE CAPSULE CENTER - POINT BEACH UNIT 1 2

Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm )

No.

Time Flux Plant Design Capsule 2

(EFPS)

(n/cm -sec)

Specific Basis Data I

8 1

4.63 x 10 1.34 x 10 6.20 x 10 6.82 x 10 5.44 x 10 8

2 2.71 x 10 1.36 x 10 9.89 x 10 1.08 x 10 3

3.66 x 10 1.24 x 10" 1.44 x 10 1.62 x 10 4

2.18 x 10 1.44 x 10 1.75 x 10 1.94 x 10 5

2.43 x 10 1.44 x 10" 2.10 x 10 2.30 x 10 2.08 x 10 6

2.59 x 10 1.30 x 10" 2.44 x 10 2.68 x 10 7

2.71 a 10 1.33 x 10' 2.80 x 10 "

3.08 x 10 '

8 1.92 x 10 1.02 x 10" 3.00 x 10 3.36 x 10 I

9A 1.77 x 10 9.65 x 10 3.17 x 10 "

3.62 x 10" 10 10 1.96 x 10 9.13 x 10 3.35 x 10 3.91 x 10" 11 1.86 x 10 1.05 x 10' 3.55 x 10 4.18 x 10 Il 2

Note:

Design Basis $ = 1.47 x 10 n/cm -sec r

m

_.-..--..-,_y

14 TA8LE II.2-6 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 23* SURVEILLANCE CAPSULE CENTER - POINT BEACH UNIT 1 2

Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm )

No.

Time Flux Plant Design Capsule (EFPS)

(n/cm2-sec)

Specific Basis Data 10 18 1

4.63 x 10 7.62 x 10 3.53 x 10 4.11 x 10 10 18 18 2

2.71 x 10 8.15 x 10 5.74 x 10 6.52 x 10 7

10 18 18 3

3.66 x 10 7.50 x 10 8.48 x 10 9.77 x 10 7

10 I9 I9 4

2.18 x 10 8.70 x 10 1.04 x 10 1.17 x 10 10 I

I 5

2.43 x 10 8.92 x 10 1.25 x 10 '

1.39 x 10 '

7 10 I9 I9 6

2.59 x 10 8.27 x 10 1.47 x 10 1.62 x 10 10 I9 I9 7

2.71 x 10 8.00 x 10 1.69 x 10 1.86 x 10 10 I9 I9 8

1.92 x 10 6.66 x 10 1.81 x 10 2.03 x 10 I

10 I9 II 9A 1.77 x 10 6.53 x 10 1.93 x 10 2.18 x 10 I

10 I9 I9 10 1.96 x 10 6.50 x 10 2.06 x 10 2.36 x 10 10 I9 I9 11 1.86 x 10 6.73 x 10 2.18 x 10 2.52 x 10 10 2

Note: Design Basis + = 8.88 x 10 n/cm -sec n

15 TABLE 11.2-7 FAST NEUTRON (E

's 1.0 MeV1 EXPOSURE AT THE 33* SURVEILLANCE CAPSULE CENTER - POINT BEACH UNIT 1 Cycle Irradiation Cycle Avg.

' Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design Capsule (EFPS)

(n/cm2-sec)

Specific Basis Data 10 18 18 1

4.63 x 10 6.77 x 10 3.13 x 10 3.73 x 10 0

18 18 2

2.71 x 10 7.33 x 10 5.12 x 10 5.91 x 10 10 18 8

18 3

3.66 x 10 6.73 x 10 7.58 x 10 8.86 x 10 7.55 x 10 10 18 9

4 2.18 x 10 8.04 x 10 9.34 x 10 1.06 x 10 5

2.43 x 10 8.28 x 10 1.13 x 10 1.26 x 10 '

6 2.59 x 10 7.91 x 10" 1.34 x 10 1.47 x 10 I9 I9 I

7 2.71 x 10 7.26 x 10 1.54 x 10 '

1.68 x 10 '

10 I9 I9 8

1.92 x 10 6.30 x 10 1.66 x 10 1.84 x 10 10 I

I9 9A 1.77 x 10 6.18 x 10 1.77 x 10 '

1.98 x 10 10 I9 10 1.96 x 10 6.31 x 10 1.89 x 10 2.14 x 10 11 1.86 x 10 5.85 x 10 2.00 x 10 2.29 x 10 '

10 2

Note: Design Basis & = 8.05 x 10 n/cm _,ec l

D 9's

16 TABLE II.2-8 FAST NEUTRON (E > 1.0 Mev) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS -

0* AZIMUTHAL ANGLE P_0 INT BEACH UNIT 2 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design (EFPS)

(n/cm2-sec)

Specific Basis 10 i

4.S8 x 10 4.42 x 10 2.02 x iO "

2.21 x 10'"

10 3

2 3 25 x 10 4.47 x 10 3.48 x 10 "

3.77 x 10'd I

I0 3

2.15 x 10 4.38 x 10 4.68 x 10'"

5.10 x 10'"

10 i

2.15 x 10 4.21 x 10 5.84 x 10'"

6.43 x 10'"

7 10 5

2.81 x 10 4.30 x 10 7.05 x 10'"

7.78 x 10'"

0 I0 18

'i 6

2.68 x 10 3.05 x 10 7.87 x 10 9.07 x 10 10 3

/

2./8 x 10 3.35 x 10 8.80 x 10 "

1.04 x 10 10 8

2.65 x 10 3.30 x 10 9.67 x 10'O 1.17 x 10" 10 9

2.43 x 10 3.60 x 10 1.05 x 10

1.29 x 10 10 2.24 x 10 3.14 x 10 1.12 x 10' 1.40 x 10 I

10 2

Note:

Design Basis + = 4.82 x 10 n/cm -sec 1

g.

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17 TA8_LE 11.2-9 FASI NEUTRON (E > 1.0 Mev) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 15* AZIMUTHAL ANGLE POINT BEACH UNIT 2 2

Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm )

No.

T i ne I-l ux l'idnl Design

( E *c PS)

(nA m -sec)-

Specific Basis 0

1 4.58 x 10 2.63 x 10 1.20 x 10'"

1.35 x 10'"

10 18 2

3.25 x 10 2.11 x 10 2.09 x 10'"

2.31 x 10 10 IO 3

2.15 x 10 2.h3 x 10 2.81 x 10'O 3.12 x 10 10 4

2.IS x 10 2.58 x 10 3.52 x 10'"

3.93 x 10'"

I 10 38 18 5

2.81 x 10 2.70 x 10 4.28 x 10 4.76 x 10 I

10 30 18 6

2.68 x 10 2.13 x 10 4.85 x 10 5.55 x 10 10 IO IO 1

2.lH x 10 2.05 x 10 5.42 x 10 6.37 x 10 10 I

8 2.bb x 10 2.04 x 10 5.96 x 10'O 7.15 x 10 "

10 18 I8 9

2.43 x 10 2.16 x 10 6.48 x 10 7.87 x 10 I

10 3

10 2.24 x 10 1.97 x 10 6.92 x 10'O 8.53 x 10 "

0 2

Note:

Design Basis $ = 2.95 x 10 n/cm,s,c

18 TA3LE 11.2-10 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 30* AZIMUTHAL ANGLE POINT BEACH UNIT 2 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design 2

(EFPS)

(n/cm -sec)

Specific Basis 10 1

4.58 x 10 1.66 x 10 7.60 x 10'#

9.02 x 10'#

10 18 18 2

3.25 x 10 1.86 x 10 1.36 x 10 1.54 x 10 3

2.1b x 10 I.H2 x 10 1.8/ x 10' 2.08 x 10' 4

2.15 x 10 1.79 x 10' 2.36 x 10'"

2.63 x 10'"

I 10 18 2.81 x 10 1.81 x 10 2.87 x 10'O 3.18 x 10 10 18 18 6

2.68 x 10 1.83 x 10 3.36 x 10 3.71 x 10 10 i

1 2./8 x 10 1.50 x 10 3.77 x 10' 4.26 x 10' 10 18 18 8

2.65 x 10 1.46 x 10 4.16 x 10 4.78 x 10 10 18 9

2.43 x 10 1.49 x 10 4.52 x 10'0 5.26 x 10 7

10 18 10 2.24 x 10 1.49 x 10 4.85 x 10'O 5.70 x 10 0

2 Note: Design Basis $ = 1.97 x 10 n/cm _3,c

19

.T_A8t E II. 2-11 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE y_ESSEL INNER RADIUS - 45* AZIMUTHAL ANGLE POINT BEACH UNIT 2 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design 2

(EFPS)

(n/cm -sec)

Specific Basis 10 I#

1 4.58 x 10 1.49 x 10 6.82 x 10'#

7.9/ x 10 7

10 I8 2

3.25 x 10 1.72 x 10 1.24 x 10'O 1.36 x 10 10 I

18 3

2.'$ x 10 1.69 x 10 1.71 x 10 1.84 x 10 10 18 18 4

2.15 x 10 1.63 x 10 2.15 x 10 2.32 x 10 10 18 18 5

2.81 x 10 1.51 x 10 2.58 x 10 2.81 x 10 7

10 IO IO 6

2.68 x 10 1.5/ x 10 3.00 x 10 3.27 x 10 7

10 I0 18 7

2.18 x 10 1.49 x 10 3.41 x 10 3.76 x 10 7

10 IO I0 8

2.65 x 10 1.44 x 10 3.80 x 10 4.22 x 10 10 IO 18 9

2.43 x 10 1.47 x 10 4.15 x 10 4.64 x 10 7

10 18 10 2.24 x 10 1.44 x 10 4.47 x 10'O 5.03 x 10 0

2 Note: Design Basis $ = 1.74 x 10 n/cm _,,c

20 TABLE 11.2-12 FAST NEUTRON (E > 1.0 Mev) EXPOSURE AT THE 13* SURVEILLANCE CAPSULE CENTER - POINT 8EACH UNIT 2 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design Capsule l

(EFPS)

(n/cm2-sec)

Specific Basis Data I

II 18 18 18 1

4.58 x 10 1.34 x 10 6.14 x 10 6.73 x 10 6.27 x 10 I

Il II II 2

3.25 x 10 1.36 x 10 1.06 x 10 1.15 x 10 II II II I

3 2.75 x 10 1.32 x 10 1.42 x 10 1.55 x 10 7

II II I9 4

2.75 x 10 1.29 x 10 1,77 x 10 1.96 x 10 7

II II I9 I9 5

2.81 x 10 1.34 x 10 2.15 x 10 2.37 x 10 2.29 x 10 l

6 2.68 x 10 9.98 x 10 2.42 x 10 2.76 x 10 7

10 II I9 7

10 II II i

7 2.78 x 10 9.97 x 10 2.70 x 10 3.17 x 10 I

10 I9 I9 8

2.65 x 10 9.92 x 10 2.96 x 10 3.56 x 10 II I8 II 9

2.43 x 10 1.06 x 10 3.22 x 10 3.92 x 10 I

10 II I9 10 2.24 x 10 9.48 x 10 3.43 x 10 4.25 x 10 10 2

Note: Design Basis $ = 1.47 x 10 n/cm -sec f

e i

21 TABLE II.2-13 4

FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 23* SURVEILLANCE 3

CAPSULE CENTER - POINT 8EACH UNIT 2 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

No.

Time Flux Plant Design Capsule (EFPS)

(n/cm2-sec)

Specific Basis Data 10 I9 18 1

4.58 x 10 7.62 x 10 3.49 x 10 4.07 x 10 10 18 18 2

3.25 x 10 8.22 x 10 6.16 x 10 6.95 x 10 7

10 18 18 I9 3

2.75 x 10 7.98 x 10 8.36 x 10 9.40 x 10 8.61 x 10 7

10 I9 I9 4

2.75 x 10 7.93 x 10 1.05 x 10 1.18 x 10 I

10 I9 I9 5

2.81 x 10 8.28 x 10 1.29 x 10 1.43 x 10 7

10 I9 I9 6

2.68 x 10 7.77 x 10 1.49 x 10 1.67 x 10 10 I9 I9 7

2.78 x 10 6.51 x 10 1.68 x 10 1.92 x 10 7

10 I9 I9 8

2.65 x 10 6.44 x 10 1.85 x 10 2.15 x 10 I

10 I9 I9 9

2.43 x 10 6.60 x 10 2.01 x 10 2.37 x 10 10 I9 10 2.24 x 10 6.41 x 10 2.15 x 10

'2.51 x 10" U

10 2

Note: Design Basis $ = 8.88 x 10 n/cm -sec 4

+r

22 TA8tE II.2-14 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE 33* SURVEILLANCE CAPSULE CENTER - POINT 8EACH UNIT 2 Cycle Irradiation Cycle Avg.

Cumulative Fluence (n/cm2)

N3.

Time Flyx Plant Design Capsule (EFPS)

(n/cm4-sec)

Specific Basis I

10 18 18 1

4.58 x 10 6.77 x 10 3.10 x 10 3.69 x 10 7

10 18 18 2

3.25 x 10 7.68 x 10 5.60 x 10 6.30 x 10 10 18 18 3

2.75 x 10 7.55 x 10 7.67 x 10 8.52 x 10 10 18 I9 4

2.75 x 10 7.39 x 10 9.71 x 10 1.07 x 10 5

2.81 x 10 1.30 x 10 1.18 x 10 "

1.30 x 10 "

10 6

2.68 x 10 7.52 x 10 1.38 x 10 "

1.52 x 10 "

10 7

2.78 x 10 6.19 x 10 1.55 x 10 "

1.74 x 10 "

7 10 8

2.65 x 10 6.01 x 10 1.11 x 10 "

1.95 x 10 "

7 10 9

2.43 x 10 6.10 x 10 1.86 x 10 2.15 x 10 "

I 10 10 2.24 x 10 6.13 x 10 2.00 x 10 2.33 x 10 "

7 10 10 2

Note:

Design Basis + = 8.05 x 10 n/cm -sec

i 23 1.

For both units, calculated plant specific fast neutron (E > 1.0 Mev) l fluer.ce levels at the surveillance capsule center are in excellent f

agreement with measured data. The maximum difference between the plant l

specific calculations and the measurements is less than 7%. Differences I

of this magnitude are well within the uncertainty of the experimental i

results.

j 2.

For Unit 1, low leakage fuel management introduced following cycle 7 has reduced the peak flux on the pressure vessel by about 30%. This reduction i

has been maintained over the last four operating fuel cycles.

3.

For Unit 2, low leakege fuel management introduced following cycle 5 has reduced the peak flux on the pressure vessel by about 25%. This reduction has been maintained over the last 5 operating cycles.

4.

For both of the Point Beach reactors the maximum neutron flux incident on the pressure vessel (0* azimuthal position) during the fuel cycles using l

out-in fuel management (cycles 1-7 for Unit 1 and 1-5 for Unit 2) was, on the average, approximately 10% less than predictions based on the design basis core power distributions.

}

Graphical presentations of the plant specific fast neutron fluence at key l

locations on the pressure vessel as well as at the surveillance capsule center 1

are shown in Figures 11.2-1 through 11-2-4 as a function of full power operating time for Point Beach Units 1 and 2.

For Unit 1, pressure vessel data is presented for the O' location on the circumferential weld as well as j

for the 15* longitudinal weld. For Unit 2 pressure vessel fluence is given l

for the O' position on the beltline circumferential weld.

For both reactors l

corresponding fluence levels at the center of 13*, 23' and 33' surveillance i

capsules are also depicted.

i In regard to Figure II.2-1 thrcugh 11.2-4, the solid portions of the fluence curves are based directly on the plant specific evaluations presented in this i

report. The dashed portions of these curves, however, involve a projection j

into the future. Since both Point Beach Units have been comitted to a l

l consistent form of low leakage fuel management for several cycles, the average neutron flux at the key locations over the low leakage fuel cycles was used i

i

24 p.

9 Figure II.2-1 3

I Maximum Fast Neutron (E >1.0 MeV) Fluence l

ons as a 6

s Function of Full Power Goerating

~ ~ " ~ ' ~ ~

i Time - Unit 1 1

i 4

I

{

r..

3 i

i 2

i i

e i

l i

l i

[_. _

i, i

i 8

20 10 i

e c

9 I

6 i

I I

j l

I

{

i i

i I

l m

i j

gs j

-4 G

,I'...

~

0*

v g

cg3 I

u.

E i

15*

4 h2 s

U 5

=

i e

p.

s-i' T

l

/

l s,

/r 19 M.,

10

/ !

= _. _ _.

8

/

i G

9 es I /

1,'

I l$

~

~

~~

f l

'f

_s i

i

l

,, l

/

/

i

~~

~

f 4

i I

s 1

I i

j l

f

f...

...~..

i I

i i

i 2

+

t l

,I 1

i l

i i

i 6

e t

10 i

U 5

10 15 20 25 30 35 Operating Time (EFPY)

25 p

i 7

Figure 11.2-2 s

7 1

Maximum Fast Neutron (E>1.0 MeV) Fluence i

i at the Center of the Surveillance 6

Capsules as a Function of Full Power 5

i Operating Time - Unit 1 l..

4 i

i I

i 3

I 1

......_..[_...._.

. __._..[..

i.

l t

4 1

l b.

l I

g i

Capsule I

20 i

Location om 10 3

s 13*

e 9

i I

i a

6 r

l l

7 l

N 23" Et t

l I

33"

/

1 cs t.

.s.

......-. (;.,e

.l s

v4 l -

e g

_ _..s<

[ _

4.c_.

wa

?

j j

l l

~

r e

e t..._...

.o.

i j

i f/

u

,,82

/4 i

I l

i I

}

l'

/

k 1019

[

i i

,y 7g r

n

/,//

I i

4, g

/,l l

i.

g I

es 5

=....... _. _

i i

i 3

i

'I

~

l

,t i

f s

s l

2:

I i

e i

l l

i

}

I8' I

i 10 u

5 10 15 20 25 30 35'

~ ~ ~ ~ ~ ~

~ ~ ' ' ~ ~ ~

l Operating Time (EFPY)

26 i,,

3 i

L_.

Figure 11.2-3 Maximum Fast Neutron (E>1.0 MeV) Fluence i

I 6

at the Beltline Weld Location as a Function of Full Power Operating 5

Time - Unit 2 l

4 i

9 3

i j

i

.--r*'----------

i 2

1 i

l I

i q

8 20 10 3

t i

9 i

e 8

i i

7

-eu 6 l

8 8

l i

j 15 l

i.-

e4 I

v e

3 U

-3 8

w i

i

.s.

I i

e

.. t.

o 1

e J

t 2

=

t

=

j l

l 4

/

s I

j

/

l U

i

/

/

l 1

I9 -

10 i

_/3 i

3 g,

j g

,l u._..__

7 2=

1,

i s.

3l w

i

~

'C i

4

)_..

j i

i 3

i 2

t I

l l

10 i

j l

18 i

_i_...

0 5

10 15 20 25 30 35 Operating Time (EFPY)

1 l

s 27 e

4 9

Figure II.2-4 i

4 q

l Maximum Fast Neutron (E>1.0 MeV) Fluence i

7 at the Center of the Surveillance 6

i 1

Capsules as a Function of Full Power Operating Time - Unit 2 5

i.

s i

3 l

I___..._.

l l

e Capsule Location 8

20 l

l e 10 s 13e e

I i

e 9

f I

t' 8

7

=*

-.. _ 23*

~

3

,,4 56

_ _. _3 3 a.._-- -

i r

_e 5 l

.p.

...-p, ua c

c G,

i

./

_-,, #*s Y d'

e t

- 3 8

g

}

,s

/ /

m oe l

l I

l f

19 10 f

/

I i

g r

g l l-J'

?

it.

,t IU

_a__....__....

f l

M W

j i

I i

i i-l l

I 3 i 7

I l

i 2

I 4

i 1

I-I 1,

l e

i 10 l

l 10 i

0 5

10 15 20 25 30 35 Operating Time (EFPY)

2.

28 for all temporal projections.

In particular, the neutron flux average over j

cycles 8 through 11 was used to project future fluence levels for Unit 1,

]

while the neutron flux average over cycles 6 through 10 was employed for Unit 2.

4 It should be noted that implementation of a more severe low leakage pattern f

would act to reduce the projections of fluence at key locations. On the other hand, relaxation of the current low leakage patterns or a return to out-in fuel management would increase those projections.

In any event it would be j

prudent to update the fluence analysis as the design of each future fuel cycle j

evolves.

I In Figures 11.2-5 and II.2-6, the azimuthal variation of maximum fast neutron f

(E > 1.0 MeV) fluence at the inner radius of the pressure vessel is presented i

as a function of azimuthal angle for Units 1 and 2, respectively. Data are presented for both current and projected end-of-life conditions.

In Figure 11.2-7, the relative radial variation of fast neutron flux and fluence within the pressure vessel wall is presented. Similar data showing the relative

)

axial variation of fast neutron flux 'and fluence over the beltline region of i

the pressure vessel is shown in Figure 11.2-8.

A three-dimensional description of the fast neutron exposure of the pressure vessel wall can be constructed using the data given in Figure 11.2-5 through 11.2-8 along with l

the relation i

1 l

$(R,0,Z) = $(9) F(R) G(Z) i where: $ (R,0,Z)

Fast neutron fluence at location R, 0, Z within

=

i the pressure vessel wall j

$ (9)

Fast neutron fluence at azimuthal location 9 on

=

j the pressure vessel inner radius from Figure 11.2-5 or 11.2-6 l

l F (R)

Relative fast neutron flux at depth R into the

=

pressure vessel from Figure 11.2-7 l

G (Z)

Relative fast neutron flux at' axial position Z from

=

Figure 11.2-8

29 i

i.-

9 s

Figure 11.2-5

~~'

7 i

Maximum Current and Projected EOL Fast i

Neutron (E>1.0 MeV) Fluence at the i

e Pressure Vessel Inner Radius as a Function of Azimuthal Angle - Unit 1 i

i j

4 l

j l.

3 i

1

_...__.__l

_.._..___..l

___...i I

i i

l

.I I

8,r 20 l

10 m

9 I

e, 7

i 4

d 6

5 i

m g

I l

s4 I

I y.

n.

l

.s 03 c

w,

...L_

._,-E.

I s

l g2 x

Q 4

5 6

r E

t I_.

32 EFPY l

i z

n l

1

! 10 '3 i

=;

J,,-

i ii I

s y

4 t

N

Z

9.0 EFPY i

4 i

i 5

_.L.

i

=. _.

l

?

r 1

l 1

t 1

l i

i 2,

l' i

t 18 ll l

I l

10 0

10 zu Ju eu du

- ~ ~ - ' ' - --

Azimuthal Angle (deg.)

4 4

30 Figure II.2-6 Maximum Current and Projected EOL Fast Neutron (E>1.0 MeV) Fluence at the

"-~

Pressure Vessel Inner Radius as a 5

Function of Azimuthal Angle - Unit 2 2

i i

l I

i 3

j 2

1 l

i

.I 8

20 I

l

• 10 f

I e

e, I

9 1

I 6

i I

g;- 5

[

E t

u i

%4 I

g i*

4 i

e u,3 m

N e

i o

m u

N l

g2 w

w; s I

u l

~

J - - 3 2 E F P Y i

l t

O f

• l 10 1.;,

9i i

i 2,.

.5

__ _... \\__ -

7 t

i a_

i

\\!

l

~

\\

l I

E i

y, i

N 9.2 EFPY i _. _

_. r i

t

}

S l

i I

i i

i

)

I l

10 10 9

0 10 20 30 40 50 Azimuthal Angle (deg.)

31 Ic i

9

~ - ~ ~ ~

3 Figure II.2-7 7

I Relative Radial Distribution of Fast Neutron (E>1.0 MeV) Flux and Fluence 6

Within the Pressure Vessel - Units 1 & 2 2

s

~

r--

j 4

l l

j i

3 i

i e

1 l

I i

2 l

1 i

i i

f1.0 f

l e

9 X

- ~~

e 3

a X

h7 I\\

l

$^

I

\\

l

(

=

.1/ 4.. T

[

7 6 4 t

j e

c g

l, -

t wj i

i l

,s I

l 1

25 t.

.._g f g. ;

f, l

I l

i i

i r

e i,

5 1

g t

g s

I I

.M I

3/4 T l

• • 0.1 i

to j;

9 y

s j

g p:

,3 i

??

t s

n V4 t

i s$

I l'

1 i

i 3L

-.l_....

4 l

i i

l t

1 I

l j

l 2

l 1

i l

l I

i l

e l

1 l

i l

l I

~

• ~~-

0 2

4 6

8 10 12 14 16 18 20 Depth into the Pressure Ve55el (cm)

I

s s

O uN y,

[",.lptrE **a

,h.% f,jtj,.,.,3

,gi 4

p, J

l, m c,m

,Nem e

l.

iy

A

. r.r

.,i"6 fe.

qI' e{

1,mn y3:9h h;:

=m a" c.m >x m

<a,,,n*o3 o% 7 u

_e a

e.

,k p

v 2gcn,03. rve ;3 2m< mm cx a3Q

c.,,3n2

,i

, r gp L ;:

+.,.

3*

i,.

,8 r+e2$.eamhn,n2c=omug,N e

ia wc,= <am CQ,.

g

'o.so

.l

, 3:

a**

j.

g3O l,

?l

..,p.-

i.t,.

g.

ii t

g.

U

.l.I..

4 lp o.

,.i

'l 1js

}.

.l

, r 6

. "s.

1 pl,.tI iI.

-'[

Ig-

[.,. J, %...

'.. g,.

t.

l

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h'l-

..gi.

1..-

p 9tgl

,,,i

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l e

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ll.;t tge.

.g gs

t 4

1jf j.

);,v.,

s,.

'\\",

' 4 i,7' r p- ' _

r -

h r-

. fr

.+

T.

0

p. jg-. h r t'

!,in h,

i..

2 r,.

-:Tg

i h,.

tl'.

m j!, W,. -.'c. t;tt. ',.. )r I

, i i 'L 6

h.i cr.;t6

,,g-(i!,

iI,,

!3;71 e.

.;..t g./.

..l*

1

Iii, i ;

' 1 i -

. a, t' c [c

[ ej 97i re - -

6

. hv l.

i p

4 1.

g".,.

t' rr. 4 t.;;

l-

., hn - r r

-. i P r..,e,

-r a

- tr l

oi l

n p j,.

o ;rj r lj.ly. l'rl 9fl,-g n j,+.

[.l*

r,

_.,. e

e i']:

t'!,

' mIt, f

...l.

8 j

.!lg

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,. e

, -. 7ljl j.

...1, I',l, s#,,

l l

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3

'.6+

6,,:. -

f

~.... i,j i

,'+ j t r

l,-

T; i,'.e*l

,t

.q(

i

.t,

.I l

,i,.

.f.

. ' t fI e'

il.

l.

l

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c{

,. 4 n

o~F}..

. g it l

I,

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u 1!

l l *,

• t'J.

}

t e

,, i.

,t, i -

j.

v.

.r ' r i

r.

t 7

a

=,,

..l s

e

. r,

..R c.

s.

p- ".

[.j

+

i.

i.,n ii.

s, m

p c

u a

j,..

e

r.,..

s.

t.

e 1

a.

i 1

,i...

c-cv

.i o

~

9, 1.,;..,

i, t,

Tl,.

4 l,

.4

,,t c

m.c

-l t

p4, 1..

l.

r..

n.

l.

a u.

i.

l i

l, o o#

c.

im i.

,a_

a i ::,. -

..j; o

4 pn

..,.a s.

4.,,.

E M

.l l

r.

.1 u.

4, y..7 qc.,

%l

.@3,,,.

cl

.i r

, _'e g,:

ll.

i.

i

. j,j; i rf..

9 l,,

. o..j i'

i,.,

.!l p. j. '..

I

.(

,,.. p i.

p q{, c..

t J..

,o..

I,.

}

o oo#

-lj.

ti.;

, {,.

a.

1,

[4; i.

.,:p.

t

. a. 8 oo "8

o

.oO Nao 2C@

(

o~.w C 2na ; 3 no,= 2 o. *e2e en3 aQ

33 1

Analysis has shown that the radial and axial variations within the vessel wall are relatively insensitive to the implementation of low leakage fuel management schemes. Thus, the above relationship provides a vehicle for a reasonable evaluation of fluence gradients within the vessel wall.

i d

=.

i 34 i

i 11.3 INFLUENCE OF AN ENERGY DEPENDENT DAMAGE MODEL The use of fast neutron fluence (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water i

reactor applications has traditionally been accepted for development of damage trend curves as well as for implementation of trend curve data to assess i

vessel condition.

In recent years, however, it has been suggested that an i

exposure model that accounts for dif ferences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure j

vessel wall.

I Because of this potential shif t away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853 4

" Analysis and Interpretation of Light Water Reactor Surveillance Results",

recommends reporting calculated displacements per iron atom (dPa) along with fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dPa function to be used for this evaluation is specified in ASTM j

Standard Practice E693 " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dPa)."

f For the Point Beach Units 1 and 2 pressure vessels, iron atom displacement j

rates at each surveillance capsule location and at positions within the vessel wall have been calculated. The analysis has indicated that for a given location the ratio of dPa/$(E > 1.0 MeV) is insensitive to changing core I

power distributions. That is, while implementation of low leakage loading patterns significantly impact the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum at a given

)

location are of second order. The dPa/$(E > 1.0 MeV) ratios calculated for key locations in the Point Beach reactor geometry are given in Table II.3-1.

l The data 1[ Table 11.3-1 may be used in conjunction with the fast neutron fluence data provided in Section 11.2 to develop distributions of dPa within the surveillance capsules and the reactor pressure vessel.

=...

a 35 TABLE 11.3-1 dPa/$(E > 1.0 MeV) RATIOS FOR THE POINT BEACH

~

UN115 1 AND 2 REACTOR GEDMETRY LOCATION (Pa/efE > 1.0 Mev) 13* CAPSULE 1.87 x 10-21 1

23* CAPSULE 1.77 x 10-21 f

33' CAPSULE 1.79 x 10-21

-21 VESSLL INNER RADIUS 1.66 x 10 VLSSEL 1/4 THICKNESS 1.80 x 10-21 j

-21 VLSSEL 3/4 THICKNESS 2.50 x 10 N01L: RATIOS ARE IN UNITS OF i

[ DISPLACEMENTS PER ATOM]/[n/cm )

1

,s 4

t '

36 11.4 POTENTIAL IMPACT ON EXCORE NEUTRON DETECTORS 1

Whenever fuel management strategy changes in a way that significantly alters the power distribution around the core periphery, the potential exists to change output signals from both the power range detectors located at the 45' azimuthal position and the intermediate range detectors located at the 0*

azimuth. As 3 result recalibration of these detectors is necessary to assure proper core power correlations.

In addition, changing azimuthal distributions from cycle to cycle can affect the overlap setpoints between the intermediate and power range chambers.

From an operational viewpoint it is helpful to have an a priori knowledge of the impact of a new fuel cycle design on the output of these chambers.

The data in Figures 11.4-1 and 11.4-2 are provided as an aid in the evaluation of the potential changes in detector response. The data as presented represents the response contribution by fuel assembly to the detector of interest. When multiplied by the relative assembly powers for a given fuel cycle design and summed over the applicable assemblies, the relative response of the detector is obtained for that particular fuel management scheme.

Evaluation for two successive fuel cycles will yield the percentage change in detector response from cycle to cycle.

The data in Figures 11.4-1 and 11.4-2 are applicable to both of the Point Beach reactors.

37 Figure II.4-1 Relative Response of Intermediate Range Excore Neutron Detectors per Unit Assembly Power 0

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11.5 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES l

As discussed in Section 11.2, plant specific fluence evaluations for the center of surveillance capsules located at 13*, 23*, and 33' were presented in Figures II.2-2 and 11.2-4 for Point Beach Units 1 and 2, respectively. The data presented on those curves represent the best available information upon which to base the future withdrawal schedules for capsules remaining in the Point Beach reactors.

In the past, withdrawal schedules have been based on the assumption of a constant exposure rate at the surveillance capsule center and a constant lead factor relating capsule exposure to maximum vessel exposure. With the widespread implementation of low leakage fuel management neither of these assumptions can be assumed to be universally valid.

It becomes prudent, therefore, to utilize the actual anticipated capsule exposure in conjunction with appropriate materials properties data to establish capsule withdrawal dates that will provide experimental information that is of most benefit to the utility.

In evaluating future withdrawal schedules, it must be remembered that the fluence projections shown in Figures 11.2-2 and II.2-4 assume continued operation with the low leakage fuel management scheme currently in place. The validity of this assumption should be verified as each new fuel cycle evolves and if significant changes occur withdrawal schedules should be adjusted accordingly.

\\

O 40 II.6 IMPACT ON HEATUP/COOLDOWN CURVE SCHEDULE OF APPLICABILITY Since the fuel management schemes that have been used to date can impact the fluence pertinent to the schedule of applicability of the heatup and cooldown c'urves, this effect is studied in this section.

Heatup and cooldown curves were generated for the Point Beach Unit 1 and 2 reactor vessels in 1977, and the results were reported in WCAP-8743 [20] and WCAP-8738 [21], respectively. The bases for these curves relative to material properties, fluence data, and irradiation damage trend curves are also provided in these reports.

In short, the curves for the Unit 1 reactor vessel were generated using the material properties of the limiting longitudinal weldment in the lower shell plate, i.e., 0.20% Cu and an initial RTNOT = 30*F, in combination with the peak 0* azimuthal position fluence data. Since a longitudinal flaw was postulated in the analysis, these assumptions also bounded the case of a longitudinal flaw assumed to exist in the circumferential weld at the peak azimuthal fluence position.

The curves for the Unit 2 reactor vessel were developed using the material properties of the circumferential weldment between the intermediate and lower forging, i.e.,

O.25% Cu and an initial RTNDT = 30*F, also in combination with the peak 0*

azimuthal position fluence data. A longitudinal flaw was also assumed in the Unit 2 analysis.

The Westinghouse trend curves were used to determine the shift in RT as a result of neutron fluence accumulation.

The curves were NDT originally developed at 10 EFPY for Unit 1 and 5 EFPY for Unit 2 and are currently shif ted to the design basis fluence at 14 EFPY in the technical operating specifications for both units (22].

Figures 11.6-1 and 11.6-2 respectively show the actual and projected fluence values for the Point Beach Unit 1 and 2 vessels against the fluence data per-tinent to the generation of the curves, i.e., the original design basis flu-ence curves.at the 1/4 and 3/4 vessel wall thickness locations.

Keeping all assumptions equal, the fuel management schemes used to date have had a minimal impact on the 14 EFPY schedule of applicability for the heatup and cooldown curves for both reactor vessels. That is, the design basis fluer$ce associated with the 14 EFPY value is equivalent to about 13.5 EFPY and 14-EFPY for the actual and projected fluences for Point Beach Units 1 and 2, respectively.

41 O

Figure 11.6-1 Fast Neutron (E>1.0 MeV) Fluence at the 1/4 Thickness and 3/4 Thickness Locations for Heatup/Cooldown Analysis -

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43 The NRC is currently developing a new irradiation damage trend curve that may be required in the generation of heatup and cooldown curves in the future.

In addition, uncertainty currently exists in defining the pertinent material chemistry for the Point Beach Unit 1 and 2 reactor vessel beltline region weldments (see Section III.3). Since potential changes exist in the bases for the heatup and cooldown curves and the impact of fuel management strategies has been minimal to date, it may be prudent to not make any changes at this time to the technical specifications relative to the pressure-temperature limits.

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59 SECTION IV CONCLUSIONS AND RECOMMENDATIONS Detailed fast neutron exposure evaluations using plant specific cycle by cycle core power distributions and state-of-the-art neutron transport methodology have been completed for the Point Beach Units 1 and 2 pressure vessels and surveillance capsules.

Explicit calculations were performed for the first eleven operating cycles of Unit 1 and for the first 10 cycles of Unit 2.

For both units, projection of the fast neutron exposure beyond the current operating cycle was based on continued implementation of low leakage fuel management similar to that employed during cycles 8 through 11 for Unit 1 and cycles 5 through 10 for Unit 2.

In regard to the low leakage fuel management already in place at the Point Beach Units, the plant specific evaluations have demonstrated that for the low leakage case the average fast neutron flux at the 0* azimuthal position has been reduced by about 30% at Unit 1 and 25% at Unit 2 relative to that existing prior to implementation of low leakage.

In particular, the following data applies at the 0* location.

$ (n/cm2-Sec)

Unit 1 Unit 2 10 10 Out-In Patterr5 4.37 x 10 4.37 x 10 Low Leakage Pattern 3.14 x 10 3.25 x 10 This location represents the maximum fast neutron flux incident on the reactor pressure vessel. At other locations on the vessel, as well as at the surveillance capsules, the impact of low leakage will differ from the data presented above.

It should be noted that the reductions of 30% for Unit 1 and 25% for Unit 2 are in the neutron flux at the vessel wall. Since neutron fluence is an

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0 60 integral effect these reduction factors can be realized only if the low leakage flux levels are maintained for extended periods of time. Therefore, to achieve maximum benefit low leakage implementation should take place as j

early as possible and remain in place for as long as possible.

It should be further emphasized that significant deviations from the low leakage scheme already in place will affect the exposure projections beyond the current operating cycle. A move toward a more severe form of low leakage (lower relative power on the periphery) would tend to reduce the projection.

On the other hand, a relaxation of the loading pattern toward higher relative power on the core periphery would increase the projections beyond those reported. As each future fuel cycle evolves, the loading patterns should be analyzed to determine their potential impact on vessel and capsule exposure.

Implementation of low leakage loading also affects the magnitude of the neutron flux at the surveillance capsule locations as well as the lead factors relating capsule exposure to maximum vessel exposure. Therefore, the assumptions of constant lead factors and capsule exposure rates over plant lifetime are not necessarily valid. Data depicting actual capsule exposure as a function of full power operating time has been presented in this report for both Point Beach Units.

It is recomended that capsule withdrawal schedules

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be adjusted based on this plant specific information.

Again, the potential l

impact of future changes in fuel management strategy should also be factored into any rescheduling of capsule withdrawals.

The fast neutron fluence values from the plant specific calculations have been compared directly with measured fluence levels derived from neutron dosimetry contained in three surveillance capsules withdrawn from each of the Point Beach Units.

For Unit 1, the ratio of calculated to measured fluence values ranges from 0.94 to 1.05 for the three capsule data points.

The corresponding range for the U' it 2 ratios is 0.94 to 0.98 for the three capsules removed n

from that reactor. This excellent agreement between calculation and l

measurement supports the use of this analytical approach to perform plant 4

specific evaluations for the Point Beach reactors.

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o 61 The fuel mangement strategies employed to date essentially have had no impact on the schedule of applicability, i.e., 14 EFPY, for the heatup and cooldown pressure-temperature limit curves in the Point Beach Unit 1 and 2 technical specifications used for plant operation.

It is recommended, however, that no action be taken at this time to modify the curves since the NRC is currently revising the irradiation damage trend curves used for developing.heatup and cooldown curves and uncertainty still exists for the material chemistry of the governing weld locations.

Relative to the pressurized thermal shock issue, the impact of current and projected fuel management strategies on the RTNDT ** "'

Beach Unit 1 and 2 reactor vessel beltline region materials was studied. A parametric study using docketed, NRC estimated, and upper estimated values was also employed to evaluate the effects of the current uncertainty in the weld material chemistry data.

For both reactor vessels, the RT values based upon Ge docke W and E NDT estimated weld material chemistry data remain below the proposed NRC screening values for PTS using the projected fluence exposure through 32 EFPY.

Caution must be exercised on all future fuel loadings to insure that this situation is

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maintained.

In the event that the weld material chemistry values are concluded to be near the upper estimates, significant flux reductions beyond those currently in-place would be needed for the RT va hes of bon reactor vessel to stay NDT within the NRC screening criteria. Additional flux reductions would also be required even with the NRC estimated weld material chemistries if consideration is given in the future to vessel life extension.

The results provided in this report should enable Wisconsin Electric Power Company to comply with the initial requirements of the forthcoming NRC rule for PTS.

62 SECTION V REFERENCES 1.

" Analysis of Potential; Pressurized Thermal Shock Events," Nuclear Regulatory Commission,10 CFR Part 50, Proposed Rule, Federa~l Register, Vol. 49, No. 26, February 8,1984, pp. 4498-4502.

2.

Soltesz, R. G., Disney, R. K., Jedruch, J. and Ziegler, S. L., " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation Vol. 5 - Two Dimensional, Discrete Ordinates Transport Technique," WANL-PR(LL)034, Vol. 5, August 1970.

3.

" Sailor RSIC Data Library Collection DLC-76." Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P, r ss Section Library for Light Water 3

Reactors.

Benchmark Testing of Westinghouse Neutron Transport Analysis Methodology -

4.

to be published.

5.

" Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," Regulatory Guide 1.99 - Revision 1. U.S. Nuclear Regulatory Commission, Washington, April 1977.

6.

Lukens Steel Company Test Certificate No. RM12965-NS, January 3, 1966 for Babcock and Wilcox Company.

7.

Draft WCAP, " Analysis of Capsule T From Point Beach Unit 1 Reactor Vessel Surveillance Capsule Test Program", to be published.

8.

Lukens Steel Company Test Certificate No. RM61766-BB, January 20,1966 for Babcock & Wilcox Company.

Wisconsin Electric Power Company letter C. W. Fay to H. R. Denton (Nuclear 9.

Regulatory Commission) " Reactor Vessel Response to Thermal Shock," January 15, 1982.

W 63

10. Babcock & Wilcox Qualification Test Plate No. SA-775, February 7,1966.

11. NRC Policy Issue

" Pressurized Thermal Shock," SECY-82-465, November 23, 1982.

12. 8AW-1799, " Reactor Vessel Beltline Weld Chemistry Study", July 1983.

13. Nuclear Regulatory Commission letter Steven A. Varga to J. W. Williams, Jr.

(Florida Power and Light Company), " Evaluation of Reactor Vessel Materials Data For Turkey Point Plant Units 3 and 4 Reactor Vessels," April 26, 1984.

14. NUREG-0800- U.S. NRC Standard Review Plan, Branch Technical Position 5-2.

Revision 1, July 1981.

15. Bethlehem Steel Corporation Test Report No. 911, July 15,1968 for Babcock

& Wilcox Company.

16. WCAP-7712. " Wisconsin Michigan Power Co. and the Wisconsin Electric Power Ca. Point Beach Unit No. 2 Reactor Vessel Radiation Surveillance Program",

June 1971.

17. Bethlehem Steel Corporation Test Report No. 917, July 18,1968 for Babcock

& Wilcox Company.

18. Babcock & Wilcox Qualification Test Plate No. SA-1484, October 14, 1968.

19. Emergency Response Guidelines - Revision 1 Westinghouse Owners Group, September 1,1983.

l

20. WCAP-8743, "Heatup and Cooldown Limit Curves for the Wisconsin Electric Power Company and the Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit No.1," January 1977.

21. WCAP-8738, "Heatup and Cooldown Limit Curves for the Wisconsin Electric Power Company and the Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit No.

2," January 1977.

l

64

22. " Point Beach Nuclear Plant Units 1 and 2 Technical Specifications";

Figures 15.3.1-1 and 15.3.1-2 for Unit 1, Amendment 53, Aug. 28,1981 and Figures 15.3.1-3 and 15.3.1-4 for Unit 2, Amendment 57,.luly 10,1981.

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A-1 APPENDIX A POWER DISTRIBUTIONS Core power distributions used in the plant specific fast neutron exposure analysis of the Point Beach Units 1 and 2 pressure vessels were derived from the following fuel cycle design reports:

Fuel Cycle Unit 1 Unit 2 1

WCAP-7430 WCAP-7430 2

WCAP-8120 WCAP-8418 R1 3

WCAP-8325 WCAP-8759 4

WCAP-8652 WCAP-8934 5

WCAP-8861 WCAP-9275 6

WCAP-9131 WCAP-9493 7

WCAP-9368 WCAP-9667 8

WCAP-9548 WCAP-9846 9

WCAP-9841 WCAP-10048 10 WCAP-9974 WCAP-10278

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11 WCAP-10191 A schematic diagram of the core configuration applicable to Point Beach Units 1 and 2 is shown in Figure A.1-1.

Cycle averaged relative assembly powers for each operating fuel cycle of Point Beach Units 1 and 2 are listed in Tables A.1-1 and A.1-2, respectively.

On Figure A.1-1 and in Tables A.1-1 and A.1-2 an identification number is assigned to each fuel assembly location; and three regions consisting of subsets of fuel assemblies are defined.

In performir.g the adjoint i

evaluations, the relative power in assemblies comprising Region 3 has been adjusted to account for known biases in the prediction of power in the In order to rebal' nce the core power for each fuel peripheral assemblies.

a cycle, the relative power in assemblies comprising Region 2 has been 1

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13 14 15 11 16 17 18 19 20 Region 1 Assemblies 12 - 21 Region 2 Assemblies 6 - 11 21 Region.3 Assemblies 1 - 5

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A-3 TABLE A.1-1 CORE POWER DISTRIBUTIONS USED IN THE PLANT SPECIFIC FLUENCE ANALYSIS i

POINT 8EACH UNIT 1 Fuel Cycle Ascembly 1

2 3

4 5

6 7

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9 10 11 1

0.83 0.83 0.75 0.91 0.87 0.80 0.84 0.72 0.65 0.48 0.73 2

0.69 0.68 0.62 0.73 0.74 0.65 0.67 0.43 0.40 0.38 0.43 3

0.86 1.00 0.93 1.05 1.08 1.00 0.98 0.96 1.01 1.01 1.07 4

0.63 0.67 0.62 0.77 0.79 0.74 0.69 0.48 0.46 0.45 0.45 5

0.70 0.74 0.67 0.83 0.86 0.84 0.72 0.78 0.77 0.85 0.52 6

1.08 1.06 1.08 1.06 0.91 1.08 1.18 1.15 0.97 1.12 1.27 7

1.12 1.22 1.05 1.05 1.13 0.95 0.95 1.15 1.10 1.20 1.15 t

8 1.12 1.09 1.22 1.06 1.07 1.22 1.07 1.05 1.31 1.04 0.98

(

9 1.05 1.16 1.14 1.02 1.06 1.14 1.21 1.23 1.03-1.13 1.30 1

10 0.95 1.08 0.93 1.09 1.15 1.16 0.94 1.16 1.14 1.22 1.19 11 1.06 1.20 1.12 1.05 0.98 1.15 1.18 1.05 1.22 1.04 0.93 12 1.14 1.16 1.20 1.02 1.23 0.96 0.96 1.16 1.00 1.12 1.01 13 1.15 1.03 1.14 1.03 0.99 1.15 1.12 0.96 1.18 1.03 1.03 14 1.14 1.10 1.18 1.22 0.90 0.90 1.19 1.22 1.04 1.27 1.26 15 1.11 1.01 1.12 1.02 0.93 1.21 1.08 1.07 1.22 1.05 1.05 16 1.14 1.04 1.10 0.99 1.21 1.02 1.18 1.14 1.03 1.18 1.08 17 1.14 0.95 1.13 0.92 1.02 0.91 0.99 1.09 1.20 1.08 1.19 18 1.14 0.94 1.13 1.20 1.18 1.05 1.22 1.19 1.05 1.25 1.06 19 1.01 0.96 1.01 1.06 0.98 1.05 0.93 1.01 1.17 1.12 1.11 20 1.12 1.09 0.94 1.00 1.06 1.16 1.14 1.19 1.18 1.18 1.11 21 1.09 1.05 1.15 0.95 0.94 0.93 0.85 0.87 1.01 0.89 0.91 i

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A-4 TABLE A. -2 CORE POWER DISTRIBUTIONS USED IN THE PLANT SPECIFIC FLUENCE ANALYSIS POINT BEACH UNIT 2 Fuel Cycle Assembly 1

2 3

4 5

6 7

8 9

10 1

0.83 0.83 0.86 0.81 0.80 0.56 0.71 0.70 0.77 0.67 2

0.69 0.68 0.67 0.64 0.68 0.43 0.41 0.41 0.44 0.38 3

0.86 0.97 0.94 0.95 1.03 1.03 0.96 0.99 0.98 0.92 4

0.63 0.71 0.71 0.69 0.71 0.74 0.46 0.43 0.44 0.48 5

0.70 0.82 0.82 0.75 0.58 0.60 0.80 0.78 0.78 0.74 6

1.08 1.22 1.18 1.18 0.85 0.84 1.03 0.92 1.18 1.09 7

1.12 1.07 0.88 0.96 1.15 1.13 1.17 1.16 1.20 1.12 8

1.12 1.18 1.17 1.09 0.91 1.23 1.20 1.09 1.08 1.26 9

1.05 1.07 0.94 1.22 1.20 1.01 1.06 1.24 1.24 1.16 10 0.95 1.16 1.07 1.10 1.05 1.18 1.16 1.11 1.16 1.08 11 1.06 1.12 1.23 1.02 1.20 1.05 1.02 0.97 0.98 1.25 12 1.14 1.16 0.98 1.21 0.99 1.03 1.02 1.04 1.02 1.11 13 1.15 0.98 1.17 0.99 1.16 1.15 1.20 1.23 1.01 1.16 14 1.14 0.98 1.03 1.20 1.08 1.00 1.06 1.08 1.03 1.07 15 1.11 0.99 1.15 1.01 1.14 1.24 1.25 1.23 1.24 1.14 16 1.14 1.24 1.00 0.93 1.17 1.06 1.04 1.08 0.90 1.05 17 1.14 0.96 1.18 0.99 1.08 1.20 1.19 1.24 1.19 1.23 18 1.14 1.07 1.16 1.21 1.13 1.03 1.10 1.08 1.15 1.08 19 1.01 0.98 1.09 1.20 1.00 1.19 1.17 1.20 1.19 1.12 20 1.12 1.05 1.01 1.02 1.13 1.11 1.08 1.05 1.05 1.06 21 1.09 0.92 0.96 1.07 0.84 0.96 0.91 0.96 0.92 0.91

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A-5 maintained at the cycle average value and the relative power in assemblies comprising Region 1 has been uniformly adjusted to provide an average power over the entire core of 1.00.

In each of the adjoint evaluations, within assembly spatial gradients have been superimposed on the average assembly power levels.

For the'~ peripheral assemblies (Region 3), these spatial gradients also include adjustments to account for analytical deficiencies that tend to occur near the boundaries of the core region.

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