ML20135E809

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Proposed Tech Specs Revising Svc Period & Associated Figures IAW TS 3.1.2.7,3.1.2.8 & 10CFR50,App G
ML20135E809
Person / Time
Site: Arkansas Nuclear 
Issue date: 11/26/1996
From:
ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20135E801 List:
References
NUDOCS 9612120073
Download: ML20135E809 (41)
v • d • e


Text

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PROPOSED TECIfNICAL SPECIFICATION CHANGES i-i i

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I 9612120073 961126 Pon ApocK 05000313 PDR

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1 3.1.2.7 Prior to ranching thirty two effsetiva full powsr yacro of l

operation, Figures 3.1.2-1, 3.1.2-2 and 3.1.2-3 shall be updated for the next service period in accordance with 10CFR50, Appendix G.

The service period shall be of sufficient duration l

to permit the scheduled evaluation of a portion of the surveillance data scheduled in ccordance with the latest revision of Topical Report BAW-1543 (5). The highest predit td adjusted reference temperature of all the beltline region naterials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.8.

The provisions of Specification 3.0.3 are not applicable.

3.1.2.8 The updated proposed technical specifications referred to in 3.1.2.7 shall be submitted for NRC review at least 90 days prior to the end of the service period.

l 3.1.2.9 With the exception of ASME Section XI testing and when the core flood tank is depressurized, during a plant cooldown the core flood tank discharge valves shall be closed and the circuit breakers for the motor operators opened before depressurizing the reactor coolant system below 600 psig.

3.1.2.10 With the exception of ASME Section XI testing, fill and vent of the reactor coolant system, emergency RCS nakeup and to allow maintenance of the valves, when the reactor coolant temperature is less than 262*F, the High Pressure Injection motor operated l

valves shall be closed with their opening control circuits for the motor operators disabled.

3.1.2.11 The plant shall not be operated in a water solid condition when the RCS pressure boundary is intact except as allowed by Emergency Operating Procedures and during System Hydrotest.

l Amendment No. M,84,96,-144,140, 18a 464,461, l

l

m

. ~..

BASES All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.(N)

These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table-4-8 of the FSAR.

The umximum unit heatup and cooldown rates satisfy stress limits for cyclic operation. (2)

The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100'F satisfies stress levels for temperatures below the DTT. (3)

The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10CFR50.

Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in FTI Document 77-1258569-01 (4). The limiting weld l

material is being irradiated as part of the B&W Owners Group Integrated Reactor Vessel Material Surveillance Program and the identification and locations of the capsules containing the limiting weld material is discussed in the latest revision to B&W report, BAW-1543. (5)

The chemical composition of the limiting weld material is reported in the B&W Report, BAW-2121P (6).

The effect of neutron irradiation on the RTNDT of the limiting weld material is reported in FTI Calculations 32-1245917-00 and 32-1257716-00 (1).

Figures 3.1.2-1, 3.1.2-2, and 3.1.2-3 present the pressure-temperature limit curves for hydrostatic test, normal heatup, and nornal cooldown respectively.

The limit curves are applicable through the thirty second effective full power l

l year of operation. The pressure limit is also adjusted for the pressure i

differential between the point of system pressure measurement and the limiting component for all allowed operating reactor coolant pump combinations.

l The pressure-temperature limit lines shown on Figure 3.1.2-2 for reactor criticality and on Figure 3.1.2-1 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice hydrostatic testing.

The actual shift in RTNDT of the beltline region material will be established i

periodically during operation by removing and evaluating, in accordance with Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core regicn.

The spray temperature difference restriction based on a stress analysis of the spray line nozzle is imposed to naintain the thermal stresses at the pressurizer spray line nozzle below the design limit.

Temperature requirements for the steam generator correspond with the measured NDTT for the shell.

I l

t i

2 I

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Amendment No. M,M,W,M,M4, 19 I

Tha haatup and cooldown rates stated in this specification are intandad as tha umximum changes in temperature in one direction in a one hour period. The r

actual temperature linear ramp rate may exceed the stated' limits for a time

-period provided that the umximum total temperature difference does not exceed i

the limit and that a temperature hold is observed to prevent the total i

temperature difference from exceeding the limit for the one hour period.

Specification 3.1.2.9 is a ecrire that the core flood tanks are not the source for pressurizing the reactor ecolant system when in cold shutdown.

Specification 3.1.2.10 is to ensure that high pressure injection is not the l

source of pressurizing the reactor coolant system when in cold shutdown. The f

LTOP enable temperature has been calculated in accordance with Code Case N-514.

Instrument error is not included in the reactor coolant temperature of 262'F.

f Specification 3.1.2.11 is to ensure that the reactor coolant system is not 7

operated in a manner which would allow overpressurization due to a temperature transient.

e i

REFERENCES j

k (1)

FSAR, Section 4.1.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415 (3)

FSAR, Section 4.3.11.5 (4)

FTI Document Number 77-1258569-01 l

[

(5)

BAW-1543, latest revision j

(6)

BAW-2121P l

I (7)

FTI Calculation Numbers 32-1245917-00 and 32-1257716-00 l

l i

i i

l i

l 1

Amendment No. 3, GG,GB,43,M,M4, 20 1

l

FIGURE 3.1.2-1 RCS INSERVICE HYDROSTATIC TEST II/U & C/D LIMITS TO 32 EFPY 2400 l

2200 f

2000

/

e

@1800 r

w

%1600 O

W 4 1400 l

H O

\\

f1200 ACCEPTABLE REGION 1

Q1000 W

DC

@ 800 13 E 600 h

400 200 I

0 0

50 100 150 200 250 300 350 400 450 500 550 600 RCS COLD LEG TEMPERATURE (*F)

Notes:

1.

This curve is not adjusted for instrument error and shall not be used 1

for operation.

2.

All Notes on Figure 3.1.2-2 are applicable for heatups.

This curve is based on a heatup rate of < 90*F/HR.

l 3.

All Notes on Figure 3.1.2-3 are applicable for cooldowns.

i l

i Amendment No. GB,H,M4, 20a t

l l.

FIGURE 3.1.2-2 l

RCS HEATUP LIMITATIONS TO 32 EFPY l

2400 l

l 2200 1-I I

l l

i 2000 g

< 50*F/HR f

^

I Non-Critical

-#/

=

g1800 Curve f

< 90*F/HR f

S 4

f Critical f

Curve l

W d1400 J

)f O

1 x

< 1200 l

r l

8 i

b

< 1000 g

)f l

m 800 ACCEPTABLE

< 70*F/HR

)

REGION Non-Critical x j

/

6 g

/

Curve T

GC 400

/

h l

'd/

200

< 90*F/HR l

Non-Critical Curve l

0 i

i 0

50 100 150 200 250 300 350 400 450 500 550 600 RCS COLD LEG TEMPERATURE (*F) l Notes:

1.

These curves are not adjusted for instrument error and shall not be used for operation.

2.

When DHR is in operation with no RCPs ope ting, the DHR system return temperature shall be used.

3.

RCP Operating Restrictions:

RCS TEMP RCP RESTRICTIOh._

T > 300*F None 300*F 2 T 2 225'F 53 225'F > T 2 84*F 52 T < 84*F No RCPs operating 4.

Allowable Heatup Rates:

RCS TEMP H/U RATE 60*F < T s 84*F s 15'F/HR T > 84*F As allowed by applicable curve i

Amendment No. BB,84,M4, 20b l

i m

FIGURE 3.1'.2-3 RCS COOLDOMN LIMITS TO 32 EFPY 2400 2200 1

2000 2

E-1800 a.

$<1600 l

l W

l d

l H 1400 0

t 2l l

ACCEPTABLE

< 1200 REGION l

)

1000 w

QC l:D

$ 800 l

l l

' 600 i

400

/

1 200 j

0 0

50 100 150 200 250 300 350 400 450 500 550 600 RCS COLD LEG TEMPERATURE ("F) l Notes:

l

1. This curve is not adjusted for instrument error and shall not be used for operation.

l

2. A maximum step temperature change of 25*F is allowable when securing all l

RCPs with the DHR system in operation. This change is defined as the RCS temperature prior to securing all the RCPs minus the DHR return l

te:perature after the RCPs are secured. When DHR is in operation with no RCPs operating, the DHR system return temperature shall be used.

3. RCP Operating Restrictions:

RCS TEMP RCP RESTRICTIONS T > 255'F None 150*F s T s 255'F s2 (See Note 5)

T < 150*F No RCPs operating

4. Allowable Cooldown Rates:

RCS TEMP C/D RATE STEP CHANGE T 2 280*F 100*F/HR s 50*F in any 1/2 HR i

200*F > T 2 150*F 50*F/HR (See Note 5) s 25'F in any 1/2 HR 1

T < 150*F 25'F/HR s 25'F in any 1 HR

5. If RCPs are operated < 200*F, then the RCS cooldown rate from 150*F 5 T s 180*F is reduced to 30*F in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />.

Amendment No. M,4-3,M4, 20c

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MARKUP OF CURRENT ANO-l TECHNICAL SPECIFICATIONS (FOR INFO ONLY) i J

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3.1.2.7 Prior to rasching fift: n thirty two effectiva full powar yacts of l

l oparation, Figures 3.1.2-1, 3.1.2-2 and 3.1.2-3 shall be updated for the next service period in accordance with 10CFR50, l

Appendix C, S :ti:n V.S.

The service period shall be of l

l sufficient duration to permit the scheduled evaluation of a l

portion of the surveillance data scheduled in accordance with the latest revision of Topical Report BAW-1543(5). The highest predicted adjusted reference temperature of all the beltline region materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.8.

The provisions of Specification 3.0.3 are not applicable.

3.1.2.8 The updated proposed technical specifications referred

)

to in 3.1.2.7 shall be submitted for NRC review at least 90 days J

prior to the end of the service period. 7.pp: print: 2dditi:::1 l

Pnc ::vicu ti=

h:11 b; 211:x:d f : p::p:::d t :hnic:1 epecific:ti:n: cub =itted in :::::d:ne: uith 10 CFn 02:t 50, l

Appendia C.

S :ti:n V.C.

3.1.2.9 With the exception of ASME Section XI testing and when the core i

flood tank is depressurized, during a plant cooldown the core flood tank discharge valves shall be closed and the circuit j

breakers for the motor operators opened before depressurizing J

the reactor coolant system below 600 psig.

3.1.2.10 With the exception of ASME Section XI testing, fill and vent of the reactor coolant system, emergency RCS nakeup and to allow maintenance of the valves, when the reactor coolant temperature is less than 400162'F, the High Pressure Injection motor operated l

41ves shall be closed with their opening control circuits for the motor operators disabled.

3.1.2.11 The plant shall not be operated in a water solid condition when the RCS pressure boundary is intact except as allowed by i

Emergency Operating Procedures and during System Hydrotest.

1 1

l l-Amendment No M,64,96,HB,444, 18a

%4, %k,

(

,.. ~ - _ _ _ -

BASES All reactor coolant system components are designed to withstand the effects of f

cyclic loads due to system temperature and pressure changes. (N)

These cyclic loads are introduced by unit load transients, reactor trips, and unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table 4-8 of the FSAR.

The maximum unit heatup and I

cooldown rates satisfy stress limits for cyclic operation. (2)

The 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100*F satisfies stress levels for temperatures below the DTT. (3)

(

The major components of the reactor coolant pressure boundary have been analyzed l

in accordance with Appendix G to 10CFR50. Results of this analysis, including the actual pressura-temperature limitations of the reactor coolant pressure I

boundary, are given in FTI Document 77-12 58 5 69-01 S.*.F 2105 ( 4 ). The limiting weld l material is being irradiated as part of the B&W Owners Group Integrated Reactor Vessel Material Surveillance Program and the identification and locations of the l

capsules containing the limiting weld material is discussed in the latest revision to B&W report, BAW-1543. (5)

The chemical composition of the limiting weld material is reported in the B&W Report, BAW-1&l42121Pv,(6)

The effect of neutron irradiation on the RTNDT of the limiting weld material is reported in the l

S ch' n e p c r t, S."? 2075FTI Calculations 32-1245917-00 and 32-1257716 (7).

I Figures 3.1.2-1, 3.1.2-2, and 3.1.2-3 present the pressure-temperature limit curves for hydrostatic test, normal heatup, and normal cooldown respectively.

The limit curves are applicable through the fifteenth thirty second effective l

l full power year of operation. The pressure limit is also adjusted for the pressure differential between the point of system pressure measurement and the l

limiting component for all allowed operating reactor coolant pump combinations.

l l

The pressure-temperature limit lines shown on Figure 3.1.2-2 for reactor criticality and on Figure 3.1.2-1 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10CFR50 for reactor criticality and for inservice hydrostatic testing.

l The actual shift in RT;DT of the beltline region material will be established periodically during operation by removing and evaluating, in accordance with j

Appendix H to 10CFR50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region.

l The spray temperature difference restriction based on a stress analysis of the i

l spray line nozzle is imposed to maintain the thermal -tresses at the pressurizer spray line nozzle below the design limit. Temperature requirements for the steam generator correspond with the measured NDTT for the shell.

l i

1 i

Amendment No. M, M, M, H, M4, 19 l

l I

Tha haetup end cooldown rates statad in this spacification are intandid no tha msximum ch:ngas in temperature in one direction in a one hour period. The actual temperature linear ramp rate may exceed the stated limits for a time l

period provided that'the maximum total temperature difference does not exceed i

the limit and that a temperature hold is observed to prevent the total i

temperature difference from exceeding the limit for the one hour period.

Specification 3.1.2.9 is to ensure that the core flood tanks are not the source for pressurizing the reactor coolant system when in cold shutdown.

j l

Specification 3.1.2.10 is to ensure that high pressure injection is not the j

source of pressurizing the reactor coolant system when in cold shutdown.__Ihe l

LIQE_snable temperature has been ca_lgulated in accordance with Code Case N-514.

l Instrument error is not included in the reactor coolant temperature of 262'F.

)

l l

Specification 3.1.2.11 is to ensure that the reactor coolant system is not operated in a manner which would allow overpressurization due to a temperature I

transient.

REFERENCES (1)

FSAR, Section 4.1.2.4 (2) ASME Boiler and Pressure Code,Section III, N-415 (3)

FSAR, Section 4.3.11.5 (4)

S.""-2105 FTI Document Number 77-1258569-01 l

l (5)

BAW-1543, latest revision l

(6)

BAW-gl,2,1_RM44p l

l

\\

(

(7)

E.*3 2075, n:tisicr. 1 FTI Calculation Numbers 32-1245917-00 and 32-1257716-00 l l

1 1

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l i

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Amendment No. G,22,38,6-3,96,M4, 20 1

NMES:

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E

- 1. THE ACCEPTABLE PRESSURE TEMPERATURE COMBINATIONS ARE BELOW AND TO THE D

[

Og9 RIGHT OF THE LIMIT CURVE (S). THE LIMIT CURVES INCLUDE THE LIMITING PRESSURE 2400 _

IFFERENTIAL BETWEEN THE POINT OF SYSTEM PRESSURE MEASUREMENT AND THE P SSURE ON THE REACTOR VESSEL REGION CONTROLLING THE LIMIT CURVE.

THEY

[

s ARE T ADJUSTED FOR POSSIBLE INSTRUMENT ERROR AND THESE CURVES SHALL K

yE-2200 _.

NOT,BE ED FOR OPERATION.

2. APPLICABLE HEATUP RATES OF 550 DEG F/HR.

k

.30 T3. APPLICABLE FOR C LDOWN RATES OF:

E T2280 DE 100 DEG F/HR (550 DEG F IN ANY 1/2 HOUR PE OD) d 1800 Z.

280 DEG F> T2150 DEG F 0 DEG F/HR (525 DEG F IN ANY 1/2 HO ERIOD) h-T<150 DEG F 2 EG F/HR (s25 DEG F IN ANY 1/2 R PERIOD)

." 1600 2 4. REACTOR COOLANT PUMP RESTRICTI

/

EAh NONE.

Ey L 1400

~

8z U;I;

% 1200

~

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ie

/

/

k 1000

~

PRESS TEMP

]

POINT (psig)

(deg F)

N 800 A

388 70 B

422 100 g

C 455 125

,g 461 150 600 1

D,p 527 170 F

541 205 400

~

/

G 22 220 255 H

1 '

200 I

1382 280 J

1763 305 K

2286 0

L 2500 339

... i i i i.

.. i i i i i i i i i i

,.., i 0

50 100 150 200 250 300 350 Indicated Reactor Coolant Inlet Temperature, "F REACTOR COOLANT SYSTEM INSERVICE HYDROSTATIC TEST HEATUP AND COOLDOWN LIMITATIONS APPLICABLE FOR FIRST 15.0 EFFECTIVE FULL POWER YEARS Figure 3.1.2-1 Amendment No. 24,43, M+,

20a

7 i

i 260h a

gt:,.

2400 :_ NOTES.

{L, p

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N I

A d

- 1. THE A PTABLE PRESSURE TEMPERATURE COMBINATIONS ARE BELOW AND TO THE g.

RIGHT O E LIMIT CURVE (S). THE LIMIT CURVES INCLUDE THE LIMITING PRESSURE

}

2200 :_

DIFFERENTI BETWEEN THE POINT OF SYSTEM PRESSURE MEASUREMENT AND THE f

d PRESSURE ON T REACTOR VESSEL REGION CONTROLLING THE LIMIT CURVE. THEY I

p,SgN', C f ARE NOT ADJUSTED R POSSIBLE INSTRUMENT ERROR AND THESE CURVES SHALL

/

2000 --

NOT BE USED FOR OPE ION.

P

.,y

[

2. APPLICABLE FOR HEATUP RAT OF s 50 DEG F/HR.

p h,&1600:_

PUMPS OPERATING, THE INDICATED DHR TEM RETURN TEMPERA

3. WHEN THE DECAY HEAT REMOVAL SYST IS OPERATING WITH NO RC

['

EAb TO THE REACTOR VESSEL SHALL BE USED.

j

~ te i

$314003 4 REACTOR COOLANT PUMP RESTRICTIONS:

Oe NONE.

oo UZ K

PRESS TEMP g rf) 1200 - 5. CURVE N-O-P-Q IS THE CRITICALITY J

q POINT (psig)

(des r)

UU A

331 70

@d B

331 125 I

g *g 1000 -

C 369 150 D

415 170 c

3 2

\\

E 498 195 g

800 r

541 204 i

5

~

/

/

G 541 230

/

[

H 695 234

[

j 809 255 D

J 1011 280 K

I234 300

,yr-L 8

330 M

225 355 34 200 O

1234 5

P 1708 37 2250 395

- t 0~

t,

[

0 50 100 150 200 250 300 350 400 Indicated Reactor Coolant Inlet Tw.ture, 'F REACIOR COOLANT SYSTEM NORMAL OPERATION - HEATUP LIMITATIONS APPLICABLE FOR FIRST 15.0 EFFECTIVE FULL POWER YEARS i

Figure 3.1.2-2 Amendment No. 24,43,M4, 20b

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i.

.b m-

- +. -

-.w~,.n w

s,.,

e mn=-

i 2600 -

i i

i i

i 1.

ACCEPTABLE PRESSURE TEMPERATURE COMBINATIONS ARE BELOW AND TO THE 2400 RI OF THE LIMIT CURVE (S).

THE LIMIT CURVES INCLUDE THE LIMITING PRESSURE DIFFE IAL BETNEEN THE POINT OF SYSTEM PRESSURE MEASUREMENT AND THE K

PRESSURE THE REACTOR VESSEL REGION CONTROLLING THE LIMIT CURVE.

TF' p% Ap 2200 ARE Ny AD ED FOR POSSIBLE INSTRUMENT ERROR AND THESE CURVES SHALL f

d

-NOT BE USED FO OPERATION.

J

t..

s

[ [

.y 2000

2. APPLICABLE FOR COOL RATES OF:

Q.

T2280 DEG F 0 DEG F/HR (550 DEG F IN ANY 1/2 HOUR PERIOD) y ti 280 DEG F> T2150 DEG F 5 EG F/HR (525 DEG F IN ANY 1/2 HOUR PERIOD) 1800 T<150 DEG F 25 D F/HR (525 DEG F IN ANY 1/2 HOUR PERIOD)

A $"

E

3. WHEN THE DECAY HEAT REMOVAL SYST S OPERATING WITH NO RC 1600 PUMPS OPERATING, THE INDICATED DHR S TEM RETURN TEMPERATURE Eg TO THE REACTOR VESSEL SHALL BE USED.

/

La j'

1400

_4. A MAXIMUM STEP TEMPERATURE CHANGE OF 35 DEG IS ALLOWABLE

[

83 gg WHEN REMOVING ALL RC PUMPS FROM OPERATION WITH E DHR u

SYSTEM OPERATING. THE STEP TEMPERATURE CHANGE IS Sh Eag DEFINED AS THE RC TEMPERATURE (PRIOR TO STOPPING 1200 RC PUMPS) MINUS THE DHR RETURN LINE TEMPERA O

y*g THE REACTOR VESSEL (AFTER STOPPING ALL RC S).

y 1000 THE 50 DEG F/HR COOLDOWN RATE IS ALLOh o

BOTH BEFORE AND AFTER THE STEP TEMP TURE PRESS TEMP "A"U

  • j POINT (psig)

(deg F) 7 800

-j A

260 70

,5

5. REACTOR COOLANT PUMP RE ICTIONS:

B 287 100 C

312 125 gnn F

E 7

E

/

M F

30 195 i

400 A

B, C,

D,f G

5 220 H

907 255 p

20 y I

1192 280 i

J 2078 K

2250 339 0

=

0 50 100 150 200 250 300 350 400 Indicated Reactor Coolant Inlet T+4ure, 'F REACTOR COOLANT SYS~II!M NORMAL OPERATION - COOLDOWN LIMITATIONS APPLICABLE FOR FIRST 15.0 EFFECITVE FULL POWER YEARS Figure 3.1.2-3 Amendment No. M,M,M4, 20c

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m W

ca 4

.s.L m.

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6 9

ATTACHMENT 1 ANO-1 FLUENCE DATA I

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I l

1 l

l

l l

TABLE OF CONTENTS A. Objectives and Background Discussion............................................

......3 B. Results.....

.............4 C. Meth odol ogy....................................................................................

1 0 C-1 Generation of the neutron source................

........................................10 l

C-2 Development of the geometrical models...................

......................11 C-3 Calculation of macroscopic material cross sections..................................... 11 C-4 D O RT Analyses.......................................................................... 1 1 C-5 C/M Ratio s................................................................................ 1 1 C-6 Determine synthesized three-dimensional results..........

.................15 C-7 Estimation of the best-estimate fluence....................................... 15 D. U ncerta inty..................................................................................... 1 8 E. References.......................................................

................. 26 2

l

l A. Obiectives and Backaround Discussion

)

In late 1995, ENTERGY contracted with FTl to determine new operating limit curves for I

1 ANO-1, for operation to 21 and 32 EFPY. In order to define the new operating limits, knowledge of the 21 and 32 EFPY neutron fluence at all points of interest was required.

l This document describes the analysis that was used to determine the fluence and transmits l

l l

the results of the analysis to the interested parties.

l 6

{

The fast neutron fluences' at all points of interest were calculated in accordance with the requirements of the U. S. NRC Draft Regulatory Guide DG-1025 [1]. A detailed description j

of the methodology that was used is given below. Explicit values of the fluence were computed for the following locations:

f l

pressure vessel inside surface maximum location (IS),

j e

WF 18 longitudinal welds, upper shell and lower shell, l

e WF 182-1 circumferential weld (max location),

l l

e WF 112 circumferential weld (max location),

Maximum location on the lower shell, and e

. Maximum location on the upper shell (same as I.S. max.)

)

The cycle-average flux and the corresponding accumulated fluence at each of these locations was calculated for end-of-cycle 10'(EOC 10), EOC 11, and EOC 12.

The average flux over the three cycles was computed and used to extrapolate to 21 EFPY and 32 EFPY, l

l

' energy greater than 1 MeV 3

l

__.._.. _ _ _ _ __.._. _ ___ _ =._ _._. _ _. _ _ _. _. _ _ _

l j

i B. Results l

The results are presented as follows-i Flux and fluence results at all points of interest in Table B-1

)

I l

(data from Reference [2])

l l

C/M Results in Tables B-2, B-3, and B-4 e

i (data from Reference [2])

See Section D for uncertainty discussion.

Discussion Over the last eleven years, FTl has developed a calculational-based fluence l

analysis irmincdology [4,12] that can be used to accurately predict the fast neutron fluence in the reactor vessel using cavity dosimetry to benchmark the fluence predictions. The methodology was developed through a full-scale benchmark experiment that was performed at the Davis-Besse Unit 1 1

reactor [4,12]. The results of the benchmark experiment demonstrated that l

the accuracy of a fluence analysis that employs the FTl metiwk, logy would j

4 be well within the NRC-suggested limit of i 20% (1o)[4,13].

Based on the excellent agreement between bias results of the ANO-1 analysis and the bias data from the Cavity Dosimetry Benchmark Experiment, the accuracy of the ANO-1 analysis is comparable to that of the benchmark experiment, and is less than the 20% (1a) criterion. This is discussed in i

more detail in Sections C-5 and C-7.

4

TABLE B-1 FLUX (E > 1 MeV)(nkm'-see)

NCREMENTAL FLUENCE frWcm)

CUMULATWE FLUENCE trWcm')

FLUX LOCATION CY10 CY 11 CV 12 TIME AVG EOC it EOC 11 EOC 12 EOC S EOC 10 EOC 11 EOC 12 EXTRAPtX.ATION 383J 478M 488.12 SA2 EFPY 1e X 11R 13.22 21 EFPY 32 EFPY EFPD EFPD EFPD EFPY EFPY EFPY INSEE SURFACE (MAX) 9.334E+08 7.237E+08 8.503E+0e 7.500E+0e 3.00EE+17 2 RTE +17 2.657E+17 3.350E+18 3.880E+18 3.987E+18 4.213E+18 S M SE+18 8.710E+18 CAPSULE LOC 2.000E+08 1J50E+08 1 M 1E+08 MA MA MA MA MA MA MA MA MA MA PEAK ON WELD WF-1821 8.867E+00 8.721E+0e 5.963E+0e 7A20E+0e 2 R GE+17 2.784E+17 2.341E+17 2.900E+18 3.236E+18 3.511E+18 3.74sE+18 5.487E+18 7.982E+18 PEAK ON WELD WF-112 8.084E+0e 7.000E+49 8.204E+0e 7.200E+0e 2.935E+17 2.883E+17 2.440E+17 3.218E+18 3.510E+18 3.79eE+18 4A42E+3 522EE+18 8.340E+18 PEAK ON WELD WF-18 (UPPER) 5.901E+0e 5.19eE+0e 4.588E+0e 5.181E+0e 1A80E+17 9,138E+17 1.791E+17 2A12E+18 2.000E+18 2221E+18 3AstE+18 4E1E+18 8.000E+18 5

m mm --

--- 2

_a-,a-_---m_

m

PEAK ON WELD WF.18 m

5.000E+00 5.26SE+49 4177E+49 5.210E<40 1EF9E+17 2.162E+17 1J73E+17 2.300E+18 2.SesE+18 2.734E+18 2J72E+18 4.260E+18 8.067E+18 PEAK ON LOWER SMLL 9.121E+00 7.191E+00 S.379E+0S 7A78E+4e 3ASE+17 2.980E+17 2.508E+17 3.31E+18 3.812E+18 3JeEE+18 416eE+18 5.901E+13 SJe2E+18 6

m m

.m m m m

m

TABLE B2 C/M RESULTS -CYCLE 10 i

l l

l C/M AVG l

Dos Type DOSIM C

M C/M M/C BY TYPE Fe S

2.368 2.249 1.052912406 0.94974662 f

Fe R

2.368 2.265 1.045474614 0.95650338 l

Fe T

2.368 2.215 1.069074492 0.93538851 1.0558205 i

l Ni BC 4.972 4.932 1.0081103 0.99195495 Ni BD 4.972 4.945 1.005460061 0.99456959 1.00678518 Cu S

0.006035 0.006104 0.988695937 1.01143331 i

1 Cu T

0.006035 0.006344 0.95129256 1.05120133 0.96999425 i

Nb 9

0.4793 0.4782 1.002300293 0.99770499 Nb 18 0.4793 0.4442 1.07901846 0.9267682 1.04065938 U-238 NONE 0.006196 0.006582 0.941355211 1.06229826 0.94135521 1.0029229 I

1

{

l 7

l TABLE B3 1

C/M RESULTS -CYCLE 11 l

C/M AVG Dos Type DOSIM C

M C/M M/C BY TYPE l

l Fe D

2.268 2.324 0.975903614 1.02469136 Fe AU 2.268 2.409 0.941469489 1.06216931 l

Fe E

2.268 2.36 0.961016949 1.04056437 0.95946335 N1 BI 4.237 4.928 0.859780844 1.16308709 1

Ni BJ 4.237 5.036 0.841342335 1.18857682 0.85056159 Cu E

0.006305 0.007087 0.889657119 1.12402855 l

Cu F

0406305 0.007077 0.890914229 1.12244251 0.89028567 1

Nb NBA 0.4943 0.5109 0.967508319 1.03358284 Nb NBB 0.4943 0.513 0.963547758 1.03783128 0.96552804 l

U-238 U-1 0.006447 0.007453 0.865020797 1.15604157 0.8650208 l

0.90617189 I

i l

i 8

TABLE B4 l

CIM RESULTS - CYCLE 12 l

l l

C/M AVG Dos Type DOSIM C

M C/M M/C BY TYPE i

l Fe Al 1.943 1.905 1.019947507 0.98044261 Fe AA 1.943 1.95 0.996410256 1.00360268

)

i Fe AJ 1.943 1.908 1.018343816 0.98198662 1.01156719 Ni AT 3.826 3.785 1.010832232 0.98928385 j

Ni AU 3.826 3.826 1

1 1.00541612 Cu AE 0.005407 0.005883 0.9190889 1.08803403 l

Cu AF 0.005407 0.006039 0.895346912 1.11688552 0.90721791 1

Nb NBA 0.4439 0.4133 1.074038229 0.93106556 l'

Nb NBB 0.4439 0.4145 1.07092883 0.93376887 1.07248353 U-238 U-1 0.00553 0.00708 0.781073446 1.28028933 U-238 U-2 0.00553 0.007011 0.788760519 1.26781193 0.78491698 l

0.95632035 1

l l

i l

l l

i

)

9

.. -.... - -. - - -... - -. - -. - -. -.-_ - _. - -... - -. ~. -. - -.. -. -

C. Methodoloav l

The two-dimensional discrete ordinates transport code, DORT [3], was used to calculate the fluence at the points defined in the background discussion, above. The fluence i

analysis methodology is defined in detail in BAW 2205 [4], and is conveniently summarized i

in the following discussion.

f f

Figure C-1 is a global view of the analytical procedure that was used to determine the i

incremental fluence accumulated over each cycle. A separate " Figure C-1 analysis" was required for each cycle (10,11, and 12) because there were separate sets of dosimeters for l

each cycle.

Now, referring to the flow chart (Figure C-1), the analysis is divided into seven main tasks:

(1) generation of the neutron source, (2) development of the DORT cavity models, (3) calculation of the macroscopic material cross sections, (4) synthesis of the results, and (5-

7) estimation of the calculational bias, the calculational uncertainty, and the fina! fluence.

l Each of these tasks is discussed below.

l C-1 Generation of the neutron source: The pin-by-pin relative power density l

distribution for the cycle of interest was calculated using the' SORREL [5]

t l

code The effects of bumup on the spatial distribution of the neutron source i

were accounted for by calculating the cycle average fission spectrum on an assembly-by-assembly basis, and by determining the cycle-average specific neutron emission rate. These data were used with the normalized time-weighted-average pin-by-pin relative power density (RPD) clistribution to determine the space-and energy-dependent neutron source. This neutron source was input to DORT as indicated in Figure C-1.

f l

l 10 1

i

. ~ > - - -,, -, - -

- ~, -

- -. -. =. -

C-2 Development of the Falcal models:

The ANO cavity geometry models for the mid-plane (RO) and the vertical plane (RZ) were developed and input to the DORT code as indicated in Figure C-1. These models were used in all three analyses and will be used in all subsequent ANO-1 i

Appendix H and PT curve analyses.

]

C-3 Calculation of macroscopic material cross sections: In accordance with DG-1025, the BUGLE-93 cross section library was used. The GlP [6] code j

was used to calculate the macroscopic cross sections for all materials used in

]

the analysis, from the core out through the cavity and into the concrete.

l l

l C-4 DORT Analyses The cross sections, geometry, and appropriate source were combined to create each model for each cycle. This resulted in a total of twelve DORT runs being executed. Each DORT run utilized a cross section Legendre expansion of three (P3), forty-eight directions (S.), and the appropriate boundary conditions. All outer boundaries were voids. A theta-weighted flux extrapolation model was used for all runs. The bootstrapping of the R-Theta and R-Z models was accomplished using the FACT [8] code with standard input.

C-5 C/M Ratios Often, the terms " Calculations" and " Measurements" are used in the fluence analysis area, but an adequate definition of the terms is seldom provided.

l l

The following discussion will clarify the meanings of the terms,

" Measurements" ('M') and " calculations" ('C') as used by FTl/BWOG.

I i

11 i

I i

1 i

l i

Measurements When FTl or BWOG refers to " Measurements" (or "M"), we refer to the measurement of the physical quantity of the dosimeter that responded to the neutron fluence, not to " measured fluence". For example, for an iron dosimeter, reference to "the Measurements" means the specific activity of"Mn,'which is the product isotope of the dosimeter reaction:

"Fe(n,p)"Mn For example, M = 1100 Ci("Mn)/g("Fe)

Calculations The calculational methodology produces two primary results: the calculated dosimeter activities and the neutron flux at all points of interest.

When we refer to the " calculations", or to "C", we are referring to the calculated dosimeter activity. Calculated advities are determined in such a way that it is directly comparable to M, without recourse to the measurements (e.g. we calculate C in Ci("Mn)/g("Fe) which is then directly comparable to M). ENDF/B6-based dosimeter reaction cross sections (7], (i.e. response functions) were used in determining C for each individual dosimeter.

It should be noted that in the FTl approach, the calculated activity is totally independent of the measurements.

C/M Benchmarks The fluence values for the reactor vessel beltline region are l

determined by best estimate calculations, which are, by definition, the 12 l

l

1 DORT results.

The FTl cavity dosimetry database, which was developed in the Cavity Dosimetry Benchmark Experiment, demonstrated a slight bias in the calculations. The energy-dependent bias removal function, h,, was developed to remove biases from the DORT results to provide "best-estimate" calculational results:

i 4BEST = 3 g ( MT), * ( h )~'

l where:

$sEst best-estimate fast neutron flux (n/cm,

j 2

=

sec)

($"'),

DORT-calculated fast neutron flux in

=

energy group "g"(n/cm'-sec) h, bias removal constant for group "g"

=

1 The bias is determined by statistical analyses of the CM ratios. CM ratios were calculated for each ANO-1 dosimeter and are reported in Section B.

The role of the ANO-1 CM data in the calculational process is, (1) to determine whether the ANO-1 calculational results are within the FTl database bias limits that were defined in the Cavity Dosimetry. Benchmark Experiment and (2) to ensure that the best estimate fluence is consistent with the requirements in DG-1025 (1).

l l

j 13 I

i j

l l

Discussion of the C/M results Neptunium-237 A large bias in CM was found in the Neptunium radiometric dosimeters in the BWOG Cavity Dosimetry Program [4], and a similar bias was found in this analysis. It is suspected that the Neptunium bias is related to the calculations (as opposed to the measurements).

Although other possibilities exist, the evidence suggests that the problem is either with the ENDF/B6 dosimeter reaction cross sections or results from the fact that the energy-dependence of the Np fission yield was not explicitly accounted for in BUGLE-93 [7]. Tentative plans call for investigation of this question in 1997. Np dosimeters were not considered in the statistical evaluation of the CM data.

i Uranium-238 All CM ratios are within the NRC-defined limit [1] of i20% (1o), that is consistent with the embrittlement criteria.

However, the U-238 ratios in cycles 11 and 12 indicated a suspicious departure from normal behavior, although they are statistically satisfactory.

It is interesting to note that the departure from normal behavior occured in Cycle 11. This was the first cycle in which FTl started using a new type of U-238 dosimeter. Tentative plans call for an investigation of -

this question by the BWOG in 1997.

Even though a suspicious l

departure from the normal behavior appeared evident, the deviations were less than the NRC suggested 30% criterion for cavity dosimetry and within 95%/95% of the FTl database values. Consequently, all U-238 dosimeters were used in the statistical evaluations of the CM data.

i s.

l 14 e

L

. - -~-..- -

?

i l

Solid State Track Recorders (SSTR)

At this time, SSTRs are being tested as experimental dosimeters in numerous cavity dosimetry irradiations. The evaluation of the SSTR results for ANO-1 will be carried out as part of an ongoing BWOG progam. SSTRs are not considered te be " production" dosimeters at this time.

C-6 Determine synthesized three-dimensional results:

i The DORT analyses produced two sets of two-dimensional flux distributions, l

i one in the vertical plane and one in the honzontal plane. The yertical plane, l

which will be referred to as the "RZ analysis" is defined as the plane bounded i

axially by the upper and lower grid plates and radially by the center of the core and a vertical line located two feet into the concrete biological shield.

The horizontal plane, referred to as the "R,0 analysis" is defined as the plane f

bounded radially by the center of the core and a point located two feet into the concrete and azimuthally by the major axis and the adjacent 45* radius.

The vessel flux, however, varies significantly in all three cylincrical-coordinate directions (R,0,Z). This means that if a point of interest is outside 5

the planes of both the RZ DORT and the RO DORT, the true flux cannot be determined from either DORT run. Under the assumption that the true 3-dimensional flux is separable, the two two-dimensional data-sets can be mathematically combined to estimate the flux at any three-dimensional point (R,0,Z). The synthesis procedure outlined in DG-1025 was used for this task.

C-7 Estimation of the best-estimate fluence The calculational bias was determined using a statistical combination of the o

calculated dosimeter activities (C) and the corresponding measured 15 l

dosimeter activities (M) using the procedures discussed in detail in BAW-2205 [4). The resulting bias in the ANO-1 results was compared to the benchmark bias (given in BAW 2205) and found to be within the acceptance criteria..This means that there is no significant bias associated with this

)

analysis beyond that identified in the Cavity Dosimetry Program, and accordingly, the energy-dependent benchmark bias was used with the DORT-calculated flux to determine the best estimate flux at each point of interest in the reactor vessel.

i i

)

1 l

i i

1

+

4 s

i l

I I

16 1

l l

l

FIGURE C-1 O

FLUENCE ANALYSIS (GLOBAL VIEW)

RPD Distribution Reactor Geometry Materials of Construction V

1f l

SORREL Code Construct Bugle-93 DORT Cross Section U

Models Library Neutron Source l I

k 1[

Geometry GIP Code Models i

Dosimetry g

T Counting Cross Sections DORT Analysis Analysis R0 and RZ c

(NESl) l if Y

Synthesized 3 0 Measured Results Dosimeter Y'

Activities

-U I

Calculated Vessel Fluence i

Dosianoter I

Activtti T a

Statistical l

Analysis

'l Calculational 4

Blas Estimate Fluence Uncertainty P

l D. Uncertainty l

The uncertainty in the ANO-1 best-estimate fluence was evaluated with respect to (a) the specific measurements and calculations performed for ANO-1 in this analysis and (b) with l

respect to the FTl generic uncertainties in the measurements and calculations. The reason for using both the ANO-1 specific uncertainties and the FTl generic set is to ensure consistency with Regulatory Guide 1.99 Rev. 2 [11]. Regulatory Guide 1.99 applies a

" margin" to embrittiement analyses that includes a confidence factor of 2.0. This implies a very high level of confidence in the fluence uncertainty.

The thirty-three dosimeter i

measurements from ANO-1 cycles 10,11, and 12 irradiations would not directly support t

this high level of confidence. However, if the ANO-1 dosimeter measurement uncertainties are consistent with the FTl data base, then the ANO-1 measurement uncertainties are j

supported by more than 590 additional dosimeter measurements from thirty-three capsules l

and 140 dosimeter measurements taken from the beltline region in the BWOG Cavity l

Dosimetry Benchmark Experiment. The 730 measurements from the thirty-three capsules l

and the experiment are sufficient to support a very high confidence level with a confidence factor of 2.0.

i The FTl generic uncertainty in the capsule and cavity dosimetry measurements has been l

estimated to be within an upper bound of 14%. This value was recently re-validated by independent estimates of the measurement uncertainties and defined by a standard deviation of 7.0%. (The 7.0% standard deviation evaluated recently and the 14% bounding uncertainty evaluated in the 1970s support a confidence factor of 2.0.)

l The ANO-1 cycles 10,11, and 12 dosimetry measurement uncertainties were evaluated to i

determine if any biases or inconsistent uncertainties were evident in comparison to the FTl i

database; none were found. The mean measurement uncertainty associated with the ANO-1 cavity dosimetry is 7.44%.

This value also shows consistency with the FTl i

database. Consequently, the ANO-1 cavity dosimetry measurement uncertainty can be 18

t j

t combined with the FTl database which is estimated to have a standard deviation of 7.0%

and a 95% confidence level that there is a 95% probability that the measurements are

'within 114% of the true value, j

i i

The FTl benchmark results for 730 dosimeters from thidy three capsules and the Cavity l

Dosimetry Benchmark Experiment are shown in Table D-1. These results indicate that the l

l generic uncertainty for FTl comparisons of calculations to capsule and cavity dosimetry l

. measurements is less than a standard deviation of 10%. This implies that the standard deviation between (a) the FTl calculations of the ANO-1 cavity dosimetry and (b) the

{

measurements, should be approximately 10% in general and bounded by i25% for a j

l 95%/95% confidence interval with thirty-four independent benchmarks. The weighted mean value of the ratio of calculated dosimeter activities to measurements (CM) in Tables

)

B-2, B-3, and B-4 have been statistically evaluated. The evaluation determined if a bias existed in the greater than 1.0 MeV fluence and estimated the uncertainty in the calculations of the ANO-1 fluence. The fluence with an energy greater than 1.0 MeV does not appear to be biased. Therefore the uncertainty in the ANO-1 cycles 10,11, and 12 I

benchmark of calculations to measurements for each respective cycle is.29%, 9.38%, and

)

4.37%. These values and the CM comparisons for each dosimeter indicate that the ANO-1 benchmark uncertainty can be combined with the FTl database.

With the ANO-1 specific measurement uncertainties consistent with the FTl database, and the benchmark uncertainties also consistent with the database, the fluence uncertainty for l

the calculations of the ANO-1 dosimetry is thereby also consistent with the FTl database and defined by a standard deviation of 7.0%. The calculated vessel fluence would have an uncertainty that is no greater than a standard deviation of 10% based on the generic FTl uncertainty evaluations. Extrapolating the calculations of the fluence to the end of life (EOL) for the vessel would result in an uncertainty that is less than a standard deviation of j

11% based on generic FTl uncertainty evaluations. These evaluations provide a 95% level I

19 l

. ~

.. ~..

of confidence that there is a 95% probability that the predicted EOL vessel fluence is within i22% of the calculated results, assuming each cycle is monitored by calculational l

modeling.

The uncertainty in the calculations of the ANO-1 fluence predictions are shown to be bounded by the FTl standard uncertainties in Table D-2.

The FTl standard set of uncertainty values was accepted by the NRC as referenced in the " integrated Reactor.

Vessel Material Surveillance Program"[9].

The NRC Safety Evaluation of the integrated surveillance program [9] states:

Uncertainties in neutron fluence estimates were discussed by the staff in its review of the B &W owners group request for exemptions to the requirements of l

Appendix H,10 CFR 50. The dosimetry methodology and vessel fluence analysis l

have been reviewed and accepted by the staff in a memorandum dated a

l December 5,1984 from L.S. Rubenstein to W.V. Johnston, " Review of Response to the Request for Additional Information on Capsule RSI-B for Rancho Seco, Reported in BAW-1702."

l In the staffs review of BAW-1702 it was reported that this methodology resulted in a maximum uncertainty in end-of-life vessel fluence of 34 percent. This uncertainty j

may be reduced for vessels not containing in-vessel dosimetry by inclusion of dosimetry devices in the reactor cavity. The B & W Owners Group has indicated l

that they have begun testing of these types of dosimeter devices. However, until these devices are installed, plants without dosimetry in the reactor vessel will have to rely on the methods of neutron fluence analysis documented in BAW 1702.

20 L

1 s

I r

l i

l l

i The NRC Evaluation of BAW-1702 provided the Table D-2 data [10].

{

l

. (NRC) CONCLUSION l

We have reviewed the Sacramento Municipal Utility District response dated i

t September 27,1984 regarding Rancho Seco surveillance capsule dosimetry. Due l

l to the capsule rotation the computational uncertainty of the flux as applied to the maximum location of the pressure weld should be increased by a small amount i.e.,

(

from 33.0% to 34.0%.

1

\\

\\

l I

i The standard uncertainties in Table D-2 are based on bounding values from evaluations l

performed in the 1970's. The uncertainties were estimated in concert with the NRC's " LWR j

Pressure Vessel Surveillance Dosimetry Improvement Program." When this program was initiated in 1977, the NRC needed to know the uncertainties in the capsule fluence i

i predictions in order to develop an industry embrittlement correlation suitable for safety l

analysis. With the very limited data available in the 1970's, FTl found that the only I

uncertainties that could be estimated with any validity were bounding values. Therefore, FTl provided the NRC and its contractors with capsule specimen echiliement data, fluence predictions, and the bounding capsule fluence uncertainty derived from measured dosimetry activities. The bounding uncertainty value is 14% as shown in the first row of Table D-2.

l ANO-1 vessel fluences that are appropriate for the 21 and 32 EFPY PT limits have a maximum value at the inside surface. Reinsertion and rotation of capsules is not applicable to the ANO-1 fluence analyses because cavity dosimetry was used. Therefore, the fluence j

uncertainties that are applicable to the ANO-1 fluence values are the first three values in the first column of Table D-2 (14, 20 and 22). These are (95% / 95%, 2o) maximum i

21 i

4

uncertainties. Table D-3 repeats the listing of these three ~ i 2a uncertainty values and lists the comparable standard deviations that relate to the NRC criteria of a 20% fluence 1

i uncertainty. The difference between the bounding column with i 2o and the a column, l

that provides consistency with the Reg. Guide and PTS embrittlement uncertainty, is the 2 confidence factor.

Explanations of (a) the dosimetry calculational uncertainty, (b) the Pressure Vessel calculational uncertainty supported by dosimetry data, and (c) the Pressure Vessel l

calculational uncertainty for extrapolations to the end of life, are presented to clearly define the vessel fluence uncertainty.

These explanations include the fact that the three l

calculational uncertainties in Table D-3 represent statistically independent variables.

Beginning with the cavity: (1) the fluence uncertainty is independently defined from the FTl

]

l standard measurement uncertainty (7%), (2) the calculational uncertainties resulting from 1

the additional variables affecting the vessel fluence are statistically added to the dosimetry l

measurement uncertainties to define the pressure vessel uncertainties (10%), and (3) the calculational uncertainties resulting from the time dependent variables are statistically added to the pressure vessel uncertainties to define the end of life vessel fluence uncertainties (11%).

l l

l l

1 t

22

l i

Table D Benchmark Comparison of C/M I

Plant / Capsule C/M Standard Dev.cu I

l Arkansas Nuclear 1/ B 1.0087 0.0087 Arkansas Nuclear 1/ C 0.9700 0.0300-Arkansas Nuclear 1/ E 1.0163 0.0163

{

Calvert Cliffs 2 / 97 degree 0.9602 0.0398 Crystal River 3 / F 1.0247 0.0247 Crystal River 3/ LG1.

1.0003 0.0003 Crystal River 3 / LG2 1.0281 0.0281 Davis Besse / Exp.

1.0034 0.0034 Davis Besse 1/ D 0.9892 0.0108 Davis Besse 1/ F 1.1341 0.1341 Davis Besse 1/ LG1 0.9696 0.0304 Millstone 2 /104 1.0317 0.0317 Millstone 2 / 97 1.0803 0.0803 North Anna 1/V 1.0694 0.0694 North Anna 2 /V 1.0632 0.0632 l

Oconee 1/ C 1.0231 0.0231 l

Oconee 1/ E 1,1417 0.1417 Oconee 2 / A 1.1845 0.1845 Oconee 2 / C 1.0688 0.0688 Oconee 2 / E 1.0231 0.0231 Oconee 3 / A 1.0553 0.0553 Oconee 3 / B 1.2696 0.2696 Oconee 3 / D 1.0277 0.0277 PCA /12/13 0.9275 0.0725 PCA/ 8/7 0.8900 0.1100 Rancho Seco 1/ B 1.0780 0.0780 Rancho Seco 1/ D 0.9575 0.0425 Rancho Seco 1/ F 0.9823 0.0177 Shearon Hanis 1/ U 1.1981 0.1981 St. Lucie 2 / 83 degree 0.9341 0.0659

)

Three Mile Island 1/ E 1.0394 0.0394 Three Mile Island 2 / LG1 0.9902 0.0098 Zion 1 0.9112 0.0888 l

Davis Besse 1.0044 0.0044 i

4 23

- _ = -._... -...

1 Table D-2 FLUENCE CALCULATION UNCERTAlfjTf Calculation Uncertainty '4 Without Capsule With Capsule Rotation Rdtation Capsule (derived from measured activity) i14 115 l

l l

Pressure vessel (maximum location i 20 i 21 for capsule irradiation time interval) l Pressure vessel (maximum location, 22 23 long term extrapolation)

Pressure vesselwelds i 33

- 34 24 i

l l

l i

Table I i:Tl Standard Fluence Uncertainties Calculation Uncertainty %

Standard 95 % /95 %

Deviation Confidence

~ i 2a Dosimetry (derived from benchmarked 7

14 calculations )

Pressure Vessel (maximum location 10 20 l

for capsule irradiation time interval)

Pressure Vessel (maximum location, 11 22 i

long term extrapolation) l l

l l

l l

i l

25 s

e

E E. References i

1.

U.S. NRC Draft Regulatory Guide DG-1025, " Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", September 1993.

2.

32-1244469-00, "ANO-1 PT Fluence Results", S. Q. King.

3.

BWNT-TM-107, "DORT, Two Dimensional Discrete Ordinates Transport Code", Ed.

M. A. Rutherford, N. M. Hassan et. al. May 1995.

4.

77-2205-00 "BAW-2205, B&WOG Cavity Dosimetry Benchmark Program, Summary Report", C. Garat, et. al. December 1994.

5.

NPGD-TM-427, Rev. 8, " SORREL, DOT Input Generation Code User's Manual",

L. A. Hassler & N. M. Hassan, July 1992.

6.

NPGD-TM456, Rev.11, " Gip Users Manual for B&W Version, Group Organized Cross Section input Program" L. A. Hassler & N. M. Hassan, August 1994.

7.

ORNL-DLC-175, " BUGLE-93, Production and Testing of the VITAMIN-86 Fine Group and the BUGLE-93 Broad Group Neutron / photon Cross-Section Libraries Derived from ENDF/B-VI Nuclear Data", D. T. Ingersoll et. al. April 1994.

8 NPGD-TM481, Rev. 5, " FACT User's Manual - A computer Code to Process the Neutron, Gamma angular flux and Fold Regular and Adjoint Results on a Coupling Surface for Response Analysis", October 1991.

i 9.

BAW-1543A Rev. 2, " Integrated Reactor Vessel Materials Surveillance Program",

A. L. Lows, et. al. May 1985.

10.

BAW-1702, " Analysis of Capsule RS1-B, Sacramento Municipal Utility District,

Rancho Seco Unit 1", A. L. Lowe, C. L. Whitmarsh, et. al., February 1982.

11.

USNRC Regulatory Guide 1.99 Rev. 2, " Radiation Embrittlement of Reactor Vessel Materials", May 1988.

12.

BAW-1875A, "The B&W Owners Group Cavity Dosimetry Program", S. Q. King, August 1985.

13.

L. Petrusha and C. Garat, " Evaluation of the Results of the B&W Owners Group Cavity Dosimetry Benchmark Experiment", Reactor Dosimetry. ASTM STP 1228.

ASTM, Philadelphia,1994.

26 i

, M A

s 3

4m as.

m 1

0 ATTACHMENT 2 ANO-1 PRESSURE / TEMPERATURE LIMITS-32 EFPY l

l i

I

!

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