ML20043A905

Rev 1 to Implementation of Reg Guide 1.99,Rev 2 for River Bend Station,Unit 1.

ML20043A905
Person / Time
Site:	River Bend
Issue date:	03/31/1990
From:	Caine T, Papandrea C, Ranganath S GENERAL ELECTRIC CO.
To:
Shared Package
ML20043A901	List: ML20043A900 ML20043A902 ML20043A905 ML20043A906 ML20043A908
References
RTR-REGGD-01.099, RTR-REGGD-1.099 SASR-89-20, SASR-89-20-R01, SASR-89-20-R1, NUDOCS 9005230265
Download: ML20043A905 (56)

Text

.. - . _ . _ _ _ . _______ __ _ - -__ _

i t

4 SASR 89 20

- DRF 137 0010.  ;

March.1990 ,

Revision 1

~;

b IMPLEMENTATION'0F REGUMTORY GUIDE 1.99-REVISION-2'FOR RIVER' BEND STATION UNIT 1 i

Prepared by: f T. A Caine, Senior: Engineer

. ' Materials Monitoring &

Structural Analysis Services Verified by: afd/v (L

• I j C. J. Papandrea, Engineer- .- j l Materials Monitoring &-

Structural Analysis Services 1

i. .

f l- 1 1

Reviewed by:

S. Ranganath,' Manager l l

Materials Monitoring & '

Structural Analysis Services ,

! I 1

i i- 9005230265 900514 l f,DR ADoCmovoongs GENuclearEnergy l

v -n.

• y

.c 1

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric. solely for the use 'f of Gulf States' Utilities. The information contained in.this report-is believed by General Electric to be an accurate and-true representation -!

of thefacts-known, obtained or provided to General Electric at the time this report,was prepared.

The only undertakings of the General' Electric Company respecting-information in this document are: contained in the contract governing 1 Gulf' States Utilities Purchase Order RS No. GE880003, . Item 5.01, and- ,

nothing contained in this document'ahall be construed.as changing'said1 contract, The use of this information - except as defined ' b'y - said contract, or for any purpose other than that for which it is intended,.  ;

is not authorized; and with respect to any such unauthorized use, neither ' General Electric Company nor any 'of cthe contributors to this i document makes any; representation or warranty (express or implied) as to the completeness, accuracy or usefulness 'of the: information .}

contained-in this' document or that such - use 'of such information may not infringe privately owned rights; nor do they' -assume any  ;

i responsibility for liability or; damage of any kind wnich may result j from such use of such information.

?

r h

s J

i I

f

(

11

}

6 TABLE OF CONTENTS I.811

1.0 BACKGROUND

11 l 2.0 CENERIC LETTER 88 11 EVALUATION 21 2.1 Chemistry 21 2.1 Initial RT NDT 2*I 2.3 Fluence 22 2.4 Surveillance Test Correction Factor 23 2$ SHIFT and Adjusted Reference Temperature (ART) 23 2.6 Results of Impact Evaluation 24 3.0 PRESSURE TEMPERATURE CURVES 31 3.1 Background 31 3.2 Non Beltline Regier.s 31 3.3 Core Beltline Region '32 3.4 Closure Flange R6gion 32 3.5 Core Critical Operation Requirements of 33 10CTR50, Appendix C 4.0 SUGGESTED USAR REVISIONS 41 5.0 SUGGESTED TECH SPEC REVISIONS 51

6.0 REFERENCES

APPENDICES A COMPARISON OF IRRADIATION EMBRITTLEMENT PREDICTIONS OF A1 RECULATORY CUIDE 1,99, REVISIONS 1 AND 2 FOR RIVER BEND i

B BELTLINE PRESSURE TEMPERATURE CURVE CALCULATION METHOD B1 l

iii L

L

1.0 AACKCROUND i i

The pressure. temperature (P.T) curves in the Technical Specifications are estabitshed to the requirements of 10CTR50, Appendix G [1] to assure that brittle fra.cture of the reactor vessel is prevented. Part of the analysis involved in developing the PT l curves is to account for irradiation embrittlement effects in the core region, or beltline. In the past, Regulatory Guide 1. 9 9,-

Revision 1 [2] bas been used to predict the shift in nil ductility reference temperature (RTNDT) as a function of fluence in the beltline region. Regulatory Guide 1.99, Revision 1 (Rev 1) was developed.  ;

assuming that copper (Cu) and phosphorus (P) were the key chemical elements influencing embrittlement, t

Regulatory Guidr. 1.99, Revision 2 [3] (Rev 2) was issued in May 1988. Rev 2 represents the results of statistical . evaluation of commercial reactor surveillance test data accumulated through about 1984 The basic elements of the regulatory guide, a chemistry factor and a fluence factor remained the same from Rev 1 to Rev 2. However, each factor is significantly.different. The chemistry factor (CF) has been changed from an equation based on Cu and P in Rev 1 to tables of Cr values based on Cu and nickel (Ni), with separate tables for plates and for welds. The fluence factor has been modified in Rev 2 to a somewhat more complex form. The overall effect of the changes from Rev 1 to Rev 2 has generally been to increase RTNDT shift predictions for relatively low fluences (below 10 19 n/cm2) and to decrease RT NDT shift predictions for higher fluences.

l '

1 A GE prepared response [4] to Generic Letter 8811 [5), which presented an evaluation of the impact of Rev 2 on existing P.T curves-for River Bend Unit 1 was submitted to Gulf States Utilities (GSU) in t

October 1988, e 1

l l

Revision 0 of this report documented the impact results from [4),

P T curves based on Rev 2 shifts, and recommended USAR and Tech Spec '

revisions. In March 1990, it was discovered that the Iower shell thickness of 5.81 inches, rather than the lower intermediate shell thickness of 5.41 inches, had been used in analyses. Revision.1 corrects this non conservatism. I 11 1

2.0 CENERIC 1.ETTER 88 11 EVAISATIDN Methods used to determine impact of Rev 2, and results of the determination are presented in this section. In addition to the j information provided previously in [4), discussions are provided justifying the use of og - O'F in the Rev 2 Margin term for the River  :

Bend beltline materials. Furthermore, a thickness of 5.41 inches was used, resulting in slightly higher 1/4 T fluence values.  !

The beltline region in the vessel consists of three lower intermediate shell plates and the connecting longitudinal welds.

Appendix A shows the details of the impact evaluation for the beltline materials. Since weld metal heats are not traceable to specific '

beltline welds, all weld heats were evaluated. The process followed for analyzing each beltline material is described below.

2.1 CHEMISTRY The chemistry data shown in Appendix A were taken from Table 5.3 1 of the River Bend USAR [6).

2.2 INITIAL RT NDT '

The values of initial RTNDT shown in the appendix were taken from the USAR [6). These values were based on 50 ft lb impact energy verification testing, with transverse Charpy specimens used for plate, as required by the ASME Code, Paragraph NB.2300. -

For beltline materials, the methods of calculating adjusted RTNDT in Rev 2 include a Margin term to be added to the calculated value ART The Margin term includes a component for uncertainty in NDT.

initial RTNDT' 'I. Rev 2 discusses determination of ag for two categories of initial RTNDT, measured values and generic mean values.

For generic mean valuer, a7 is simply the standard deviation calculated for the data set used to compute the mean. For measured values, requirements for determination of og are somewhat vague.

21 ,

i Rev 2 states, "If a measured value of initial RT NDT for the material in question is available, ag is to be estimated from the precision of the test method.** CE's position for RTNDI values derived from measured data, as is the case for the River Bend beltline materials, is that og is zero, as explained below.

The Charpy curves fit to surveillance data, which ultimately provided the ARTNDT data for development of Rev 2, were best estimate fits. An idealized example is provided as curve #1 in Figure 2 1.

However, the ASME Code approach to determining RTNDT is based on the lowest value of three specimens exceeding the required limits of impact energy and lateral expansion. A visualization of a Charpy.

curve drawn on the basis of the Code RTNDT approach is shown as curve

1. 2 in Figure 2 1. In comparing curves #1 and #2, it is clear that curve #2, which is based on the lowest value rather than the mean value, provides a conservative estimate of initial RT Therefore, NDT.

the ASME Code method of determining RTNDT from measured data is conservative, and ay - O'F is appropriate.

2.3 FLUENCE I

The values of fluence for 32 effective full power years (ETPY) shown in [4] were based on dosimetry results [7). The - dosimetry capsule, tested after the first fuel cycle, provides measured dosimeter flux data. 14ad factors, computed with two dimensional neutron transport methods, relate the dosimeter flux to the flux at the peak location in the beltline, During this effort,_ it was determined that the lead factors used in [7] were incorrect, so that j the inside surface lead factor of 0.50 should actually be 0.67. This changes the predicted 32 EFPY peak inside surface fluence to  !

6.6x10 18 n/cm2, rather than 8.9x1018 n/cm2 reported in (7) and [4). '

• In the Rev 2 draft which was circulated altar editing to incorporate public comments, the text stated, "og, the standard deviation for the initial RT NDT, may be taken as zero if a measured value of initial NDT for the material in question is available."

RT 22

a For purposes of comparing reference temperatures currently f applied under Rev 1 to those that are predicted with Rev 2, the 1/4 T fluenco used to develop the P.T curves currently in the I

Technical Specification was used in Appendix A for the Rev 1 shif t calculations. The inside surface fluence of 6.6x10 18 n/cm8'was used ,

in the Rev 2 shift calculations, as described below.

The Rev 2 method of calculating shift requires that the fluence at the vessel inside surface, fourf, be calculated and then attenuated to the depth x according to the relationship:

fx -f surf (* 0.24x), {

This method results in a fluence at the 1/4 T location of 4.8x10 18 n/ce ,

2.4 SURVEILIANCE TEST CORRECTION FACTOR Rev 1 allows for consideration of credible surveillance data when ,

it becomes available. Rev 2 requires that two sets of credible data be developed before considering their use. However, no surveillance testing has been performed yet, so surveillance test correction factors do not apply for either Rev 1 or Rev 2 calculations, and are set to 1.0 in the appendix, i

2.5 SHIFT AND ADJUSTED REFERENCE TEMPERATURE (ART)

The RTNDT shif t calculations in the appendix are based on the , ,

procedures in Rev 1 and Rev 2. For Rev 1, the equation for SHIFT is:

SHIFT - (STF)*[40 + 1000(%Cu .08) + 5000(%P .008))*(f)0.5 where STF - surveillance test correction factor f - fluence for the given EFPY / 1019 Tor Rev 2, the ?iHIFT equation consists of two terms:

1 23 e

i

SHITT - ARTNDT + Margin where ARTNDT - (CF)*f p 0. W log O Margin - 2(o3 2 + a32 ).5 Chemistry factors (CF) are tabulated for welds and plates in. Tables 1 and 2, respectively, of Rev 2. The margin term og has set values in  !

Rev 2 of 17'T for plate and 28'T for veld. However, og need not be greater than 0.5*ARTNDT' 9

The values of ART in the appendix are computed by adding the SHIFT terms to the values of initial RTNDT. ART versus EFPY is plotted for the most limiting beltline materials in Figure 2 2. The ART for the most limiting plate is applicable through 7 EFPY. The most limiting veld is the governing material beyond 7 ETPY.

2.6 RESULTS OF IMPACT EVALUATION >

The impact of implementing Rev 2 can best be determined by comparing the ART values based on Rev 1 and Rev 2. Table 2 1 shows l the ART values at 32 EFPY for each beltline material. The following conclusions are drawn from the results in the table:

1. The Rev 2 ART values at 32 EFPY are below 200'F, which- is the allowable limit in 10CFR50, Appendix G. Therefore, implementation of Rev 2 vill not result in any additional -

requirements for analysis, testing or provisions for thermal annealing.

2. Rev 2 increased the maximum ART value by 46'F. This, combined with the fact that the curves in the Tech Spec were based on a -

• thickness of 5.81 inches, results in the P T curves A',B',C'  ;

currently in the Tech Spec being valid for only 6 EFPY. not 32 EFPY as they were for Rev 1. Likewise, the P T curves A,B,C l

currently in the . Tech Spec are valid for only 2 EFPY, not 8.8 EFPY as they were for Rev 1.

Based on these conclusions GSU requested that the R1ver Bend P T curves be updated to reflect requirements of Rev 2.

24

(:

u I

Table 2 l' COMPARISON OF REV 1 AND REV 2 ART VA1,UES

- FOR RIVER BEND c

32.EFPY.

! Rev 1 Rev 2

. )eltline Connonant ART'(*F) ' ART ('F)

Plates:

C3138'2 57.0 83.5 C3054 1 20.0 60.0 C3054 2 42.0 82.0 l

l Welds:

i-l 492L4871/A421827AE -12.0- 25.7 492L4871/A421B27AF 30.0 15.1 SP6756/0342 (tanden) 2.0. 102;8

5P6756/0342 (single) -12.0. 92.8 I

s" r

I 25  :

S

,e , ,

150 140 -

130 - -

120 - O

@ 1 110 - #1

,5 100 - ,

(*

u 90 -

1' O #2 co d 80 -

a ti 70 _

R

_E. 60 - o N 50 -

0 40 -

/f l 6 8 O I l 30 - I

-I ,

20 - l 8 A -W RT 10 - I Determination A!rcady Conservative NDT I ,

O , , ,

i , , , , , , , ,

-80 -40 0 40 80 120 160- 200

. Charpy Test Temperature (F)

Firire 2-1. O m. .. . _ :..en o f Sea r ve i l l a nce Da t a l'I t anel RT r!DT APPf"#CI' 9 h - gesu y w - rv q +m' sw y -W w

• w a y qse <w-- m ey , , e

120 WELD SP6756 __

100 tot 0342 Tondem 0.09% Cu, 0.927. fii # - -

s 90 - -

p 80 /

g

/ "

PLATE 22-1-3 70 - r y e_f o.08: cu. o.63% tai C 60 #

50 3 7 -

U

< 30 ' [ ,

7 o

w 20 / /

N 10 ,

8 /

& G j

-10

-20 /

-30 __ --

I

/ -

-60 i 0 4 8 12 16 20 24 28 32 EFFECTIVE FULL POWER YEARS Figure 2-2. Predicted Adjusted Reference Ternperature as o Function of Effective Full Power Years of Operation

3.0 PRESSUREoTEMPERATURE CURVES ,

3.1 BACKGROUND

Operating limits for pressure and temperature are required for >

three categories of operation: (a) hydrostatic pressure tests and ,

leak tests, referred to as Curve A; (b) non nuclear heatup/cooldown '

and low level physics tests, referred i:o as Curve B; and (c) core critical operation, referred to as curve C. There are three vessel ,

regions that affect the operating limits: the closure flant,e region, j the core beltline region, and the remainder of the vessel, or non beltline regions. The closure flange region limits are controlling at lower pressures primarily because of 10CFR50' Appendix G  ;

[1) requirements. The non beltline and beltline ' region operating i limits. are evaluated according to procedures in 10CFR50 Appendix G, Appendix G of the ASME Code [8) and . Welding Research Council (VRC)

Bulletin 175.[9), with the beltline region minimum temperature limits adjusted to account for vessel irradiation.

Figure 31 has curves applicable per Rev 2 for 32 EFFY of operation, for use in the USAR. Figure 3 2 has curves applicable per Rev 2 for 8 ETPY of operation, for use in the Tech Spec. The requirements for each vessel region influencing the P T curves are discussed below.

3.2 NON BELTLINE REGIONS Non beltline regions are those locations that receive too little fluence to cause any RTNDT increase. Non beltline components include the nozzles, the closure flanges, some shell plates, top and bottom head plates and the control rod drive (CRD) penetrations. Detailed stress analyses, specifically for the purpose of fracture toughness analysis, of the non beltline components were performed for the BWR/6.

The analyses took into account all mechanical loadings and thermal transients anticipated. Detailed stresses were used according to [9]

to develop plots of allowable pressure (p) versus temperature relative to - the reference temperature (T - RTNDT). Plots were developed for the two most limiting regions; the feedwater nozzle and the CRD-I 31 I l

i l

o l

l penetration regions. All other non beltline regions are categorized under one of these two regions.

I L

The generic BWR/6 non beltline region results were applied to

) River Bend by adding the highest RT NDT for the non beltline discontinuities to the appropriate P versus (T RTNDT) curves for the WR/6 CRD penetration or feedwater nozzle. The limiting RTNDT values are 10*F for the CRD penetration 1imits and -20'T for the feedwater nozzle limits.

l 3.3 CORE BELTLINE REGION l

The pressure temperature (P.T) limits for the beltline region are determined according to the methods in ASME Code Appendix G [8). As l the beltline fluence increases during operation, these curves shift by an amount discussed in Section 2. Typically, the beltline curves shift to become more limiting than the non beltiino curves at some point during operating life. Using Rev 2 and the appropriate vessel thickncss of 5.41 inches for River Bend, this occurs after only 2 EFPY of operation. The curves resulting from shifting the beltline limits -

l are shown in Figures 3 1 and 3 2 as A', B' and C'.

The stress intensity factors (K y ), calculated for the beltline region according to - ASME Appendix 0 procerkves, were based on a combination of pressure and thermal stresses for c 1/4 T flaw in a flat plate. The pressure stresses were calculated using thin walled cylinder equations. Thermal stresses were calculated assuming the through wall temperature distribution of a flat plate subjected to a-100*F/hr thermal gradient. A 32 EFPY ART of 103'T and an 8 EFPY ART of 61'F were used to adjust the values from (T - RTNDT)

Figure G 2210 1 of [8).

3.4 CLOSURE FLANGE REGION 10CFR50 Appendix C sets several minimum requirements for pressure and temperature, in addition to those outlined in the ASME Code, based on the closure flange region RTNDT. In some cases, the results of analysis for other regions exceed these requirements and they do not 32

i

, i affect the shape of the PT curves. However, some closure flange ,

requirements do impact the curves. In addition, General Electric recommends 60'F margin on the required bolt preload temperature, i

As stated in Paragraph G 2222(c) of Reference 2, for application  ;

of full bolt preload and reactor pressure up to 20% of hydrostatic test pressure, the RPV metal te*m rature must be at R*NDT or greater.

The GE practice is to require (RTNDT + 60'F) for bolt preload, for two [

reasons: ,

s. The original ASME Code of construction required j (RTNDT + 60*F), and

b. The highest stressed region during boltup is the closure i flange region, and the flaw size assumed in that region i (0.24 inches) is less than 1/4 T. This flaw size is j detectable using ultrasonic testing (UT) techniques. In fact, CB&1 studies report that a flaw in the closure flange region of 0.09 inch can be reliably detected using UT.

l

[

For River Bend, the closure flange and attached shell values of (RTNDT + 60'F)

- 70'F, are consistent with the allowable lowest I service temperature for the bolting material.

10CFR50 Appendix G, paragraph IV. A.2, sets minimum temperature requirements for pressure above - 201 hydrotest pressure based on the l RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 90*F) and Curve B temperature no less than (RTNDT + 120'F) . .

These requirements - cause the steps in the curves at 201 hydrotest pressure (312 psig) shown in Figures 3 1 and 3 2, 3.5 CORE CRITICAL OPERATION REQUIREMENTS OF 10CFR50 APPEND 1X G Curves C and C', the core critical operation curves shown in Figures 31 and 3 2, are generated from the requirements of 10CFR50 Appendix G. paragraph IV.A.3. Essentially paragraph IV. A.3 requires that core critical P T limits be 40'r above any Curve A or B limits.

Curve B is more limiting than Curve A, so Curve C is Curve B plus 40*F.  ;

33

, ..m~

Another requirement of IV. A.3, or actually an allowance for the .

LVR, concerns minimum temperature for initial criticality in a i l startup. The BVR, given that water level is normal, is ' allowed 3 l

initial criticality at the closure -flange region (RTNDT + 60'F) at pressures below 312 psig. Above 312 psig, the core critical curve

, temperature must be at least that required for the pressure test l (Curve A' at 1100 psig). As a result of this requirement, Curve C' on ,

Figure 3 1 has a step at 312 psig to 198'F vbich is the temperature I required by Curve A' at 1100 psig.

l t

P i

P e

l l

l c

k i

3-4

1600  ;

i t

A B A'C B' C' 1400- ,

7 9 i ,

i i

, i J , t

$ 1200 ,'

l a '

$ J i [ l l z .I t , , ,

I , s 1

0 J J 'l * '

~ 1000- I '

f l{

d I - l* l

$ [ '

A',B',C' - CORE BELTUNE

, [ i AFTER ASSUMED 153'F 800 '

• SHIFT FROM AN INITIAL e

O [ / P 8 WELD RT Nor OF -50'F o '

b s I: / "

[ [ A - SYSTEM HYDROTEST LIMIT WITH FUEL IN VESSEL 600 ,

L-2 h

i

/ B - NON-NUCLEAR HEATING UMIT C - NUCLEAR (CORE CRITICAL) y / l U MIT

@ 400 ,' '

W ' VESSEL DISCONTINUITY ,

UMITS E 312 rsio ,, , , ,

1 ---

CORE BELTLINE WITH 153'F SHIFT 200 ,

soLTuP [ CURVES A',B',C' ARE VALID 70'r

) FOR 32 EFPY OF OPERATION CURVES A,B,C ARE VALID l FOR 2 EFPY OF OPERATION -

0 , ,

l

! O 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (oF)

Figure 3-1. Pressure-Temperoture Operating Limit's Volid to 32 EFPY 3-5

1600 AA' B B'C C' 1400- p,, ,

t ,

i 8 n o

' a

_t '

g -[ $ l

, u m '

3 1200 l ,

o 8 s 8 r g g I 'r U- f I .'

s

'J 8

0 1000

- I a I,lf tII l,;

$ 2 s e A',B',C' - CORE BELTLINE y / 8 8 AFTER ASSUMED 111'F SHIFT FROM AN INITIAL e 800 '

l [ ,

o ['8 , ,

WELD RT ror OF -50 F y ,

I  !

6 l 'l A - SYSTEM HYOROTEST LIMIT E 600 t' s' WITH FUEL IN VESSEL t- f f B - NON-NUCLEAR HEATING

s UMIT C - NUCLEAR (CORE CRITICAL) y UMIT

@ 400 m -- VESSEL DISCONTINUITY y mm UMITS l

CL

--- CORE BELTLINE WITH 11 l'F SHIFT 200 r

BOLTVP -

f CURVES A',B',C' ARE VALID 70'r

/ FOR 8 EFPY OF OPERATION CURVES A,B,C ARE VAllD' FOR 2 EFPY OF OPERATION l 0 , i 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (oF)

Figure 3-2. Pressure-Temperature Operating Limits Votid to B EFPY 3-6

FMT 88 212 0019 i

i s

Toble 2. Flur Density and Fluence Detsteinattena

• River Bond Statien Irradiatient Otteter 30. 199 5

• Bert ember 14 1967 3 l  !

l I r

4 e FP (t Wiro dye /g 21osent Reettien Rete Flus Density Fluence  ;

Wire Weltht tot end of Idpe/nusieve (n/cm 2,g3 gaf,,23  ;

(Element) a frradiation) tant'd11 gj_ggy *0.1 MeV gj_ggy *0.1 MeV p 3ren A 0.1469 1.00:10 0 6.22:10-18  ;

1 Iron B 0.1538 1.68:10 6 6.21:10-18 ,

Iron C 0.1261 1.95 10 6 6.44:10*I8 6.20:10-16 ( AV) 4.es10 8 7.1:108 1.4:10II 2.2 101T L

• At Full Power (2694 MW ).

i

. f f

i l

1 i

3 I

h a

4.0 SUGCESTED VSAR REVISIONS This section contains excerpts from the River Send USAR. '

includir.g suggested revisions to account for changes associated with  ;

implementation of Rev 2 provided in this report. Revisions are noted with margin sidebars, and added or revised wording is underlined. A sidebar with no underlined words indicates a deletion.

The specific sections revised are as follows:

Section 4.1.4.5 Section 4.3.2.8 Section 5.3.1.6 Section 5.3.2.1 Section 5.3.2.2 Related revisions were made in the following:

Table 4.3 5 Table 5.3 1 Figure 4.3 23 ,

Figure 4.3 24 (deleted)

Figure 5.3 4 Figure 5.3 4a (added) l Figure 5.3 5 [

l l

l

~

l i

b l 41

USAR Pane 4.1 17 i

4.1.4.$ Neutron T1uence Calculations Neutron vessel fluence calculations were carried out using a two dimensienal, discrete ordinates, S n transport code with general anisotropic scattering. This code is a widely used discrete ordinates code which will solve a wide variety of radiation transport problems. The program will solve both fixed source and multiplication problems. Slab, cylinder, and j spherical neometries are allowed with various boundary l conditions. The fluence calculations incorporate, as an initial starting point, neutron fission distributions prepared from core physics data as a distributed source. Anisotropic scattering was considered for all regions. The cross sections were prepared

with 1/E flux weighted, Pt matrices for anisotropic scattering j but did not include resonance self shielding factors. I Fast neutron fluxes at locations other than the core mid plane were calculated using a second two dimensional, discrete ordinate code. This second two dimensional code is used to solve smaller sired eroblems. and is similar to the two dimensional code used for the vessel neutron fluence calculations.

I 1

The fast neutron flux calculations are used to establish the  !

ratio of flux between the surveillance cannule locations and the i location of ceak vessel inside surface flux. known as the lead feetor. Use 6f the lead factor is discussed in Section 4.3.2.8.

i USAR Page 4.3 18 i

4.3.2.8 Vessel Irradiations l 1

The lead factor was calculated using the two dimensional discrete l ordinates tranport codes described in Section 4.1.4.5. The- l discrete ordinates code was used in a distributed source mode _

i 42 '

I

l 4

with cylindrical geometry. The geometry described seven recient with the core modeled as two homonenized reeiens. The coolant water region between the sng and the shroud was described containing saturated water at 550'F and 1050 psi, subcooled .

gegen at $30'T and 1040 mala was used for the coolant between the abroud and the vessel. The material compositions for the stainless steel in the shroud and the carbon steel in the vessel f contain the mixtures by weight as specified in the ASME material {

specifications for ASME SA240, 3041., and ASME SA533 Crade B. In the region between the shroud and the vessel, the presence of the jet pumps was ignored. A diagram showing the regions and dimensions modeled is shown in Figure 4.3 23.

The distributed source which can be sacarated in sence and enerry. was obtained from the cera mover shame and the neutron  ;

sometra. The integral over position and energy is normalized to ,

t the total number of neutrons in the core region. The core region -

is defined as a 1 centimeter thick disk with no transverse L leakage. The power in this core region is set equal to the  ;

maximum power in the axial direction. The optimal axial power l l

distribution is shown in Firure 4.3 13. l f

i Dosimetry located on the inside surface of the vessel was removed i i

after the first fuel evele and tested to determine the flux at i that location. The lead factor relatina the desinecer location l

to the ceak location was used to calculate the neak vessel inside  ;

surface flux. Assumine an BOY _e m ity factor. or 32 effective full cover years (EFPY) in 40 years of oeeration. the fluence for this operatine ceriod was estimated. The measured dosimeter flux. and calculated ceak flux and fluence are shown in i Table 4.3 5. The calculated neutron flux leaving the cylindrical I core is listed in Table 4.3 6.

l 1 .

i 43 W

i Table 4.3 5 DOSIMETER AND VESSEL PEAX TLUXES AND FLUENCES i

e l

l t

Time at Power: e l EOC1 1.0 EFPY - 3.14x10 seconds 32 EFPY 32 ETPY = 1.01x10' seconds l

Lead Factors: ,

-r Inside Surface (I.D.) 0.67 l

2 Dosimeter Flux (n/cm s) 4.4x10' (nominal) 5.5x109 (upper bound)

FLUENCE (n/cm ):  ;

NOMINAL UPPER BOUND 17 17 EOC1 Peak 1.D. 2.1x10 2.6x10 6.6x10 IO 10 32 EFPY Peak I.D. 8.3x10 i

l k

e i

44

-.A-.., .. - . '. .- m ,mh*.aa -m2..e_a- .ews *4%.--w - - - - 4a - e m4 --,. -------m , - - ' - - - - --------.-,-e-- .-w -

+-Fu-n -----e. w a, -..--

k 1n.1 g g a.6 i I

gos.0 1DRTWIO i 3

~ 1 w

E 6WATI" 9 ',

b 93 % 6 VESSI' i

e 46HmoVD

{ ' is.5 gwATER g $4.0 I t

b.

w ,

{ 2CDRII*TI*i 0, "

5 60.0 " .

- k

\ $1.3 yigure - ggtt yoR W O~D g ALYSIS oy VISSEL TLUESCE h

45 1

- ,,,-,.. n... ~ ,,- , , - - - . . , n , . . - , - . ,...., , . . , , . ., , , ,.. .,,. , ,, , .. , .- -,.<

i UEAR Pare 5.3 6!

Last cararrach; in accordance with the requirements of the edition of 10CTR$0, -

Appendix H, that was current on the issue date of the ASME Code to which the reactor vessel was purchased, three capsules are  !

provided since the predicted end.of life adjusted reference  !

temperature of the reactor vessel steel, as credicted at the time  :

of desinn. was less than 100'r. The proposed withdrawal schedule -

is in accordance with 10CTR50, Appendix H, requirements specified in the 1983 revision.

USAR Paee 5.3-7: i Section 5.3.1.6.2, second paragraph:

A The peak fluence at the inside surface of the vessel beltline

! aht11 is 1.n(10)18 n/cm2 af ter 40 yr of service. All predictions of radiation damage to the reactor vessel beltline material were t l made using peak fluence values, l i Section 5.3.1.6.3, first paragraph: ,

I Estimated maximum changes in vessel beltline RT yp7 (initial ,

reference temperature) values as a function of the 32 effective full cower year (EFPY) fluence are listed in Table 5.31. The precicted peak 32 EFPY fluence at the inside surface of the '

vessel beltline is 1 i(10)18' n/cm2 Transition temperature changes and changes in upper shelf energy were calculated in accordance with the rules of Regulatory Guide 1.99, Revision 2. l Reference temperatures were established in accordance with  ;

10CTR50, Appendix C, and NB 2330 of the ASME Code, 46

.i i

USAR Pane 5.3 9t  !

l '

Section 5.3.1.6.5: -

CE RI.pzida,d a separate neutron dosimeter so that fluence  ;

, measurements 2.9.214 be made at the vessel ID af ter the first fuel cycle to verify the predicted fluence at an early date in plant  ;

operation. This measuremeD.e was made in 1908. with a measured flux that was about 25I lower than desian vredictions. which in consistent with dosimeter test results at other B1JRs . Dosimeter flux vires are also included in the surveillance caosules. so that fluence may be further calibrated af ter several eveles of normal operation. Thus. there is no need for additional seoarate i dosimetry. It is possible, however, to install a new dosimeter, l if required, during succeeding fuel cycles.  !

USAR Page 5.3 12t Section 5.3.2.1.2: -

1 i

5.3.2.1.2 Temperature Limits for ISI Hydrostatic or Leak ,

Pressure Tests The fracture toughness analysis- for system pressure tests resulted in the curves labeled A shown on Fig. 5.3 4 The curves labeled " core beltline" are based on an initial RT NDT Of E for the ytid material for River Bend Station.

The predicted adiusted RTNDT from Figure 5.3 5 (based on the y vessel inside surface neutron fluence attenuated to the 1/4 T deoth according to Rerulatory Guide 1.99. Revision ~ 2) has been und to adiunt the beltline curves to account for the effect of fast neutrons. Fig. 5.3 4 shows the beltline curve with an .;

assumed addition of 1D.'I for 32 EFPY of ooeration at River Bend [

Station. Firure 5.3 4a shows the beltline curve with an assumed addition of 111'F for B EFPY of ooeration. Figure $.3 4a also aceears in the Technical Soecification as Firure 3.4.6.1 1.

47 t

6

l '

USAR Pane 5.3 13-Section 5.3.2.2:  ;

.,, Of the design transients, the upset condition producing the most adverse temperature and pressure condition anywhere in the vessel head

-i and/or shall areas occurs in the bottom head. vieldina a minimum fluid temperature of 250'F and a maximum pressure peak of 1,180 psig. Scram s automatically occurs as a result of this event, prior to the reduction in bottom head fluid temperature, so' the applicable ope'tating limits are given by the non nuclear heating limits for vessel discontinuities such as the bottom head (curve B on rinures 5.3 4 or 5.3 4a),,,, f i

I f

T l

l l

l 48

RBS USAR Table 5.3-1 RIVER BEND STATION UNIT 1 REACTOR VESSEL CHARPY TEST RESULTS VESSEL BELTLINE EMBRITTLEMENT EFFECTS l

1) Vessel Plate (Beltline)

Percent Heat Number C Mn Si P S Ni Cu Mo._ V C3138-2 0.19 1.37 0.25 0.012 0.015 0.63 0.08 0.58 -

C3054-1 0.19 1.30 0.26 0.007 0.020 0.70 0.09 0.57 -

• C3054-2 0.19 1.30 0.26 0.007 0.012 0.70 0.09 0.57 -

2) Vessel Welds (Beltline)

Percent Heat / Lot No. C Mn St P S Ni Cu No V 492L4871/A421B27AE 0.07 1.06 0.37 0.018 0.025 0.95 0.04- 0.50 0.02

• 492L4871/A421B27AF 0.07 1.17 0.32 0.020 0.020 0.98 0.03 0.51 0.02 SP6756/0342 (1) 0.078 1.24 0.53 0.010 0.012 0.92 0.09 0.46 0.006 5P6756/0342 (2) 0.063 1.27 0.57 0.010 0.011 0.93 0.09 0.45 0.006 y 1) Vessel Plate (Beltline)

Starting R.G. 1.99, Rev. 2 32 EFPY E '

Heat Number RTNDT (*F) Extrapolated RTyg (*F) RTg (*F) U f C3138-2 +9 11 84 86,74,78 C3054-1 -20 10 6Q 94,93,93

• C3054-2 +2 10 12 92,102,92

2) Vessel Welds (Beltline)

R.C. 1.99,'Rev. 2 Starting 32 EFPY Heat / lot No. NDT ( F) . Extrapolated RTg (*F) RT (*F) U 1f 492L4871/A421B27AE -60 16 16 151,160,161

• 492L4871/A421527AF -50 65 15 126.129,136

-SP6756/0342 (1) -50 111 103 95,99,96 SP6756/0342 (2) -60 153 93 89,94,91

• Selected for reactor vessel test specimen.

(1) Tandem wire process (2) Single wire process

1600 9

9 L

A B A'C B' C' 1400 ,

y , ,

d i i I I f 1 ._

e I # #

.? t t

$ 1200 a

l jfl r,7 8

e r

$ i I

l x T o'

] j [*)' ,'

c1.

O 1000 r ,

a i si , ,

y  ! '

m e' A',B',C' - CORE BELTUNE

[ [ AFTER ASSUMED 153*F  ;

SHIFT FROM AN INfTtAL x

o 800

/ ,

/

• P WELD RT nor OF -50'F H r a U '

I i h

x o

6 600 ,'

, I s A - SYSTEM HYDROTEST LIMIT WITH FUEL IN VESSEL t-2

'f h'

i

/ 8 -'NON-NUCLEAR HEATING UMIT

' C - NUCLEAR (CORE CRITICAL)

$ i l U MIT

@ 400 ,' '

m ' VESSEL DISCDNTINUITY UMITS

$ si2 rso l 1 ---

CORE BELTLINE WITH 153'F SHIFT 200 ,

BOLTUP  ! CURVES A',B',C' ARE VAUD 70'r FOR 32 EFPY OF OPERATION CURVES A,B,C ARE VAUD FOR 2 EFPY OF OPERATION O , ,

100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (oF)

Figure 5.3-4 Minimum Temperatures Required versus Reactor Pressure 4-10

1600 AA'8 B'C C' 1400 7, , j ,

• s o

! e I g, 8 f s a ,

o 3 1200 l l 0

[

}j t

,lll a

1 CL Gl j # 8 l i 9 1000 l d

it f [;f ll 8

^~

A',B',C' - CORE BELTUNE y [ l AFTER ASSUMED 111 F e

o 800 ' , l8 ,

SHIFT FROM AN INITIAL WELD RT,e7 OF -50'F

['8 a G ,

s t 6

• l l .

' A - SYSTEM HYDROTEST LIMIT E 600 l' WITH FUEL IN VESSEL t- f f B - NON-NUCLEAR HEATING g U MIT C - NUCLEAR (CORE CRITICAL) y UMIT l

l @n 400 l VESSEL DISCONTINUITY UMITS E 312 rso .

Q- ---

CORE BELTLINE WITH 111'F SHIFT 200 r i BOLTUP  ! CURVES A',B',C' ARE VAUD l 70'T

/ FOR S EFPY OF OPERATION l CURVES A,B,C ARE VALID FOR 2 EFPY OF OPERATION 0 i i 0

l 100 200 300 400 500 600-MINIMUM REACTOR VESSEL METAL TEMPERATURE (oF) i Q

Figure 5.3-4o Minimum Temperotures Required versus Reactor Pressure i 4-11

120 110 WELD SP6756 "

100 tot 0342 Tandem - -

o.097. Cu. 0.92% Nb y /

90 e

/

80 g 70 -

m / '

PLNE 22-1-3 7 p O. oar. co o.63?. Ni

^

h- 60 #

7 i* A '

w*

< 30 ~ / ,/

O [ /

r \1 10 / / ,

'(

C 5 -

E

' O

-10 /

-20 1

-40

-50

-60 0 4 8 12 16 20 .24 28 32 EFFECTIVE FULL POWER YEARS Figure 5.3-5 Predicted Adjusted Reference Ternperature or o Function of Effective Fun Power Years of Operation

i s

5.0 SUCCESTED TECH SPEC REVISIONS This section contains excerpts from the' River Bend Tech Spec related to P.T limits, with suggested revisions related' to implementation of Rev 2 and the updates of this report.. Revisions are [

noted with margin sidebars, and added or revised wording is L underlined. .A~sidebar with no underlined words indicates a deletion.

I The following sections of the Tech Specs have been revised:

• t Surveillance Requirements Section 3/4.4.6 Bases Section 3/4.4.6 Related revisions have been made to the following:  ;

'I.

Figure 3.4.6.1 1' t Table 4.4.6.1.3 1-Table B 3/4,4.6 1 i

Figure B 3/4.4.6 1 i

L r

(

i 1

I 51 u

'i JtEACTOR C001 ANT SYSTEM 3/4.4. 6 PRESSURE / TEMPERATURE LIMITS REACTOR COOIANT SYSTEM LIMITING CONDITION FOR OPERATION 1

3.4. 6.1 The reactor . coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4. 6.1_1 - (1) curves A and A' for hydrostatic or leak testing; (2) curves -B and B' for l heatup by non nuclear means, cooldown following a nuclear shutdown and low i power PHYSICS TESTS; and (3) curves _C and C' for operations with a critical ,

core _other than low' power ~ PHYSICS TESTS, with. -

a.- -A maximum heatup of 100*F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, 'i b.- A maximum cooldown of'100'F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period,

c. A maximum temperature . change of 10'T in any one hour period during inservice hydrostatic and leak testing operations above. .i the heatup and cooldown limit curves, and

d. The reactor vessel flange and head flange temperature greater (

than or equal to 70'F when reactor vessel. head bolting studs are-under tension.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or ,

! pressure to within ' the limits within 30 minutes; perform an engineering '.

l- evaluation to determine the effects of the out of limit condition . on.- the structural integrity of the reactor coolant system; andi determine' that the' I reactor coolant system remains. acceptable for continued operations. ,

Otherwise, b6.in at least HOT SHUTDOVN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(

SURVEILIANCE REOUIREMENTS 4.4,6.1.1 During- system heatup, cooldown and inservice leak Land hydrostatic testing errations, the' reactor coolant system temperature.-and pressure shall be daermined, at least once per 30 minutes, t_o be within.

the above required heatup and cooldown limits and to the right of the limit i lines of Figure 3.4.6'.1 1 curves . A and A', B and B ' ,- or C and ' C ' , ~ a s  !

applicable.

Li RIVER BEND - UNIT 1 3/4 4-21

REACTOR C001 ANT SYSTEM i SURVEILIANCE REOUIREMENTS (Continued) i 4.4.6.1.2 The reactor coolant system temperature - and pressure shall. be ,

determined to be to the right of the criticality limit line of

' Figure 3.4.6.1 1 curves C.and c' within 15. minutes prior to the withdrawal, of control rods to bring the reactor to criticality'and at least once per'

• 30 minutes during system heacup. '

4.4.6.1.3 The reactor- vessel material surveillance specimens shall be removed and examined to determine changes in reactorL pressure vessel '

material properties as required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3 1. The results of these examinations shall be used to update the curves of Figure 3.4.6.1-1.

4.4.6.1.4 The reactor vessel flange and head flange ' temperature shall be verified to be greater than or equal to 70'F:

a. In. OPERATIONAL CONDITION 4- when reactor coolant- system =

temperature is: -

1. s 100'F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2. -s 80*F, at~least once per 30 minutes,

b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vesse1~ head bolting, Ltuds, i

f RIVER BEND UNIT 1 3/4 4-22 h

, + .

l l

1 l- 1600 l

~

AA' 8 B'C C' 1400 fe , ri e s e

i .* * *  ;

I'e ' ' '

a ,

.? 4'

$ 1200 l

o it ,

' l ' ',,

6 1 I

e l .x I a- [ ,hl,ljl,l_

[ R 1000 ,

y ,( '

d j, o :s

$ 8

[ t A',B',C' - CORE BELTUNE -

y [

• 8 AFTER ASSUMED .111 F. -

e 800- '

l ,

. SHa'T FROM - AN ' INITIAL -

-l l o ,- i WELD RT er OF -50'F . .l 6 [l8 < '

6

" l 'l 8

A - SYSTEM HYDROTEST.UMIT ' '!

E 600 d d - WITH FUEL IN VESSEL 1 t- f f 8 - NON-NUCLEAR HEATING'-

3 UMIT

- C ---NUCLEAR (CORE CRITICAL) y UMIT

@ 400 -j-v) VESSEL ' DISCONTINUITY -a UMITS

$ 312 PSIG m 1

~

CORE BELTLINE WITH ':

111'F SHIFT l 200 r j BOLTup [. CURVES A',8',C' ARE VALID I 70'F

] FOR 8 EFPY OF OPERATION - .l CURVES A,8,C ARE VALID $

FOR 2. EFPY OF OPERATION  !

' 0 i i 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (oF) l 1

FIGURE 3.4.6.1 - 1 MINIMUM TEMPERATURE REQUIRED VS REACTOR PRESSURE  ;

RIVER BEND - UNIT 1 3/4 4-23 i

i

TAALI 4.4.6.1.3 1 REACTOR VESSEL MATERIAL SURVEILIANCE PROGRAM VITHDRAWAL SCHEDULE CAPSULE VESSEL LEAD FACTOR AT  : WITHDRAWAL TIME NUMBER LOCATION 1. D . /1/4 T (EFPY)

.,f 1 3' O.67/0'.89 6 2 177' O.67/0.89 15  ;

1 3 183' O.67/0.89 Spare- i f

(

l l

l o

RIVER BEND - UNIT 1 3/4 4-24 -

L l

l  :

l l REACTOR COOIANT SYSTEM

{

BASES l 3/4.4.6 PRESSURE / TEMPERATURE LIMITS' All components in the reactor coolant system are designed to withstand '

i the effects of cyclic loads due'to system temperature and pressure changes.

l' These cyclic loads are introduced by normal load transients,. reactor trips, ,

l and startup and shutdown operations. The various categories of load' cycles used for design purposes are provided in Section 3.9 of the FSAR. During '

startup and shutdown.: the rates of temperature and pressure changes _ are limited so that the maximum specified heatup and cooldown rates are con-sistent with the design assumptions'and satisfy the stress limits for >

i l

cyclic operation.

l During heatup, the thermal gradients in the reactor vessel-wall produce thermal stresses which vary'from compressive.at the inner wall to tensile at the outer wall. .These thermally induced compressive stresses- ,

tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure temperature curve based on steady state conditions, A.e., no thermal stresses, represents a lower bound of all similar. curves:

for finite heatup rates when.the inner wall of the vessel is treated-as the governing location.

The heatup analysis also covers-the determination'of pressure. .

temperature limitations:for the case in which the outer. wall of the' vessel ,

becomes the controlling location. The' thermal gradients. established during heatup produce tensile stresses which are already present.- .The~ thermally .;

induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the~ time along the heatup_ ramp;.

therefore, a lower bound curve similar to that. described for the heatup of '

the inner wall cannot be defined. Consequently, for the cases in which the outer wall of the vessel becomes-the stress-controlling location,-each. <

heatup rate of interest must be analyzed on an individual basis.

The reactor vessel materials have been tested to determine their initial RT . The results of these tests,are shown in Table B 3/4.4.6 1.

Reacteropbtionandresultantfastneutron(Egreaterthan.1MeV)-

irradiation will cause-an increase'in the RT ere re, an adjusted '

ND reference temperature, based upon the fluence,T. nickel content and copper l  ;

content of the material in question, can be predicted using Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel ,

Materials." The pressure / temperature limit curve, Figure 3.4.6.1 1, curves  !

A', B' and C', includes predicted adjustments for this shift in RT _ #

NDT ,

the conditions at 8 EFPY.  !

RIVER BEND - UNIT 1 B 3/4 4-5

9 ItEACTOR COOLANT SYSTEM BASES l f

PRESSURE /TEMPERATi%E LIMITS (Continued)

The actual shift in RT of the vessel material will be determined.

periodicallyduringoperati$kbyremovingandevaluating,inaccordance ,

with ASTM E111 and 10 CFR 50, Appendix H, irradiated reactor vessel l material specimens installed near the inside wall of the reactor vessel in the core area. The irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift.- The

. operating limit curves of Figure 3.4.6.1 1 shall be' adjusted, as required.=

on the basis of the specimen data and recommendations of Regulatory Cuide 1.99, Revision 2 l :.

The pressure temperature l limit lines shown in. Figure 3.4.6.1 1, curves C, and 'C', and A and A',. for reactor criticality' and for inservice leak and hydrostatic testing have been provided-to assure compliance with the minimum temperature requirements of Appendix C to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing,'

l RIVER BEND - UNIT 1 B 3/4 4 6

BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUCHNESS LIMITING HEAT /SIAB AVG. UPPER 3

$ BELTLINE WELD SEAM OR OR RT ART

• RT

o COMPONENT MAT'L TYPE ILEAT/IET Cu(Z) Ni(I) NDT(* F) NDT(*F) (FT-LBS) NDT(*F)

$ Plate SA-533 GR B CL.1 C3138-2 0.08 0.63 +9 75 79 84 5

i Weld SHELL COURSE No. 2 SP6756/ 0.09 M -50 153 17 103

c. ' Vertical Seam 3 lot 0342 5'

s NOTE: These values are given only for the benefit of calculating the 32 EFPY RT NDT' -

HEAT /SIAB as NON-BELTLINE OR

^

w COMPONENT MT*L TYPE HEAT /IDT NDT N

3 Shell Ring .SA 533 GrB C1.1 ALL HEATS' +10 5, Bottom Head Dome SA 533 GrB C1.1' ALL HEATS +10 Bottom Head Torus , SA 533 GrB C1.1 . ALL HEATS +10 Top Head Dome 'SA 533 GrB C1.1 ALL HEATS +10-Top Head Torus SA 533 GrB C1.1 'ALL HEATS +10 Top Head Flange SA 508 CL.2 ALL HEATS +10 Vessel Flange 'SA 508 CL.2- ALL HEATS +10 Feedwater Nozzle SA 508 C1. 2 ALL HEATS -20 Weld. IDE ALIDY STEEL 'ALL HEATS

-20 Closure Studs SA 540 GRADE B23 ALL HEATS or B24- Meets requirement of 45 ft-lbs and 25 mils lateral expansion at +10*F

- -- - _u_-- a . -_ _._ _ .

_=_____ . . - .

k t

7 -f I 6-5 i 9

S 5-X d

4-2 LJ 3-

)

d f2-5 1-0 , _, , , , , , , I

O 10 20 30 .40. l l

I 1' SERVICE LIFE (YEARS *)-

a o

l i

ERSES FIGURE B 3/4.4.6-1 FAST NEUTRON FLUENCE (E>1 MEV) AT VESSEL l.D. ,

AS A FUNCTION OF SERVICE LIFE'  !

l 4

AT 907. OF RATED THERMAL POWER AND 907. AVAILABILITY l

RIVER BEND - UNIT 1 3 3/4 4-9

6.0 ~ REFERENCES

[1) " Fracture-Toughness Requirements," Appendix G to Part 50 of Title.

10 of the Code of Federal Regulations,. July 1983.

[2]

• Effects of Residual Elements onf Predicted Radiation Damage to Reactor Vessel Materials," -USNRC Regulatory- Guide' 1.99, f

Revision 1, April 1977.  ;

[3] ' Radiation Embrittlement of Reactor: Vessel Materials," USNRC' Regulatory Guide'1.99, Revision 2, May 1988.

[4] Papandrea, C.J . , " Impact of Regulatory Guide l'.99, Revision 2 on . 1 River Bend Station, Unit 1," GE Report SASR 88-83, 0ctober 1988. ifi

[5] "NRC Position on' Radiation Embrittlement of Reactor' Vessel j Material ~and Its Impact - of - Plant Operations ," ; USNRC Generic Letter 88 11, July 1988.

[6] River Bend Station Updated Safety Analysis Report, August 1987.

i

[7] Caine, T. A. , " Flux Wire Dosimeter. Evaluation ~ for River Bend Station," GE Report SASR 88 49, June 1988'. ,

[8] " Protection Against .Non Ductile Failure," Appendix LG Lto Section III of the' ASME Boiler & Pressure Vessel Code, 1986 Edition with 1988 Addenda. >

[9] "PVRC Recommendations on Toughness Requirements. for 'Ferritic Materials," Welding Research Council Bulletin 175, August 1972.

l

-i 6-1

i

-r t

t APPENDIX A COMPARISON OF IRRADIATION EMBRITTLEMENT PREDICTIONS'-

OF RECULATORY GUIDE 1.99, REVISIONS 1 AND 2 FOR -<

RIVER BEND UNIT I'-

F u

6 s

t A1

COMPARISON OF REG. GUIDE 1.99 REVISIONS 1 AND 2 FOR RIVER BEND UNIT 1 BELTLINE MATERIALS Vessel Plate-(Beltline) : 22-1-3 Thickness 5.41 inches Material Heat: C3138-2 Chemistry: C Mn P S Si Cu Ni Mo 0.19 1.37 0.012 0.015 0.25 0.08 0.63 0.58 Initial RTndt: RTndt-I = 9 F, Sigma-I = 0F 32 EFPY Fluence - (f) : . Calculated Peak 1/4T f = 6.4E+18 n/ca^2 (used with Rev 1)

Calculated Peak I.D. f= 6.6E+18 n/ca^2

.Rev 2~ Attenuated 1/4T f = 4.8E+18 n/ca^2 (basis for Rev 2 delta RT)

Surveillance Testing Affecting Rev'1 shift Calc 61ation:

Surveillance testing not yet done.

Correction factor applied = - l' Chemistry Factor for Rev 2 Shift: CF= 51

~

Comparison of Rev 1 and Rev '2 SHIFT and ART (degrees F) versus EFPY:

Rev 2 Rev'2 . Rev 2 Fev 2 Rev 1 Rev l' EFPY Delta RT Margin SHIFT ' ART SHIFT ART 4 16.4 16.4 32.8 41.8 17.0 26.0 8 23.1 23.1 46.2 55.2 24.0 33.0-12 27.7 27.7 55.4 ' 64.4 29.4 38.4 16 31.2 31.2- 62.5 71'5. 33.9 42.9 20 134.1 34.0 68.1- 77.1 37.9 . 46.9 24- 36.5 34.0 70.5 79.5 41.6 50.6

'28' ' 3 18 . 6 34.0 :72.6 81.6 44.9 53.9 32 40.5 34.0- 74.5 83.5 48.0 57.0.

e..i%e -

pt .a g -

a.*g fy , yay

.%,, iyg .y,g ge gg. 9 y 9

COMPARISON OF REG. GUIDE 1.99 REVIdIONS 1 AND 2 FOR RIVER BEND UNIT 1 BELTLINE MATERIALS Vessel Plate (Beltline) : 22-1-1 Thickness 5.41 inches Material Heat: C3054-1 Chemistry: C Mn P S Si Cu Ni Mo O.19 1.3 0.007 0.02 0.26 0.09 0.7 0.57 Initial RTndt: RTndt-I = -20 F, Sigma-I = 0F 32 EFPY Fluence (f): Calculated Peak-1/4T f = 6.4E+18 n/ca^2 (used with Rev 1)

Calculated Peak I.D. f= 6.6E+38 n/ca^2' Rev 2 Attenuated 1/4T f =

4.8E+18-n/cm^2 (basis for Rev 2 delta RT).

Surveillance Testing Affecting Rev'1 Shift Calculation:.

Surveillance testing not yet'done.

Correction factor applied = .1-Chemistry Factor for Rev 2 Shift: 'CF= 58 Comparison of Rev 1 and Rev 2 SHIFT and ART (degrees ' F) versus EFPY:

Rev 2 Rev 2 Rev 2 Rev 2 Rev 1 Rev 1 EFPY Delta RT Margin SHIFT ART SHIFT ART

. 4 18.6' 18.6 37.3 17.3 14.1' -5.9

! 8 26.3 26.3 52.6' 32.6 20.0 -0.0 12 31.5 31.$ 63.0- 43.0. 24.5' 4.5.

16 .35.5 34.0- 69.5 49.5 28.3 8.3 20, 38.8 34.0 '72.8 52.8 31.6 11.6-24 41.5 -34.0 75.5 55.5 34.6 14.6 28 43.9 34.0 77.91 57.9. 37.4~ 17.4' 32 46.0 34.0 180.0 60.0' 40.0 20.0

, m. , - . , - -, -. _

w. , . - - - , _.- , ,- . ~ -

m_. __ ._ __ _ _ _ _ _ _ .

COMPARISON OF REG. GUIDE 1.99 REVISIONS 1 AND 2

FOR RIVER BEND UNIT 1 BELTLINE MATERIALS Vessel Plate (Beltline)

22-1-2 Thickness 5.41 inches Material Heat: C3054-2 Chemistry: C Mn P S Si Cu Ni Mo 0.19 1.3 0.007 0.012 -0.26 0.09 0.7 0.57 Initial RTndt: RTndt-I = 2 F, Sigma-I = 0F 32 EFPY Fluence (f): Calculated' Peak 1/4T f = 6.4E+18 n/ca^2 (used with Rev 1) i Calculated Peak I.D. f= 6.6E+18 n/ca^2 Rev 2 Attenuated 1/45' f = 4.8E+18 n/ca^2 (basis for Rev 2 delta RT).

Surveillance Testing Affecting Rev 1 Shift Calculation:

Surveillance testing not.yet done.

Cerrection factor applied = 1 Chemistry Factor for Rev.2 Shift: CF= 58 Comparison of Rev 1 and Rev 2 SHIFT and ART (degrees 1F) versus EFPY:

Rev 2 'Rev 2 Rev 2 Rev 2 -Rev 1 Rev 1 '

EFPY Delta RT Margin SHIFT- ART SHIFT. ART-

. 4 18.6~ 18.6 37.3. 39.3- 14.1 16.1 8 26.3 26.3 .52.6 54.6 20.0 22.0-12 -31.5 31.5 63.0 65.0 24.5 26.5 16 35.5 34.0 69.5 71.5 .28.3 30.3

20 38.8 34.0 72.8 '74.8 31.6 33.6  ;

24 .41.51 34.0 15.5 77.5 34.6 36.6 l j

28 43.9- 34.0- 77.9 J79.9 37.4 39.4 32 46.0 34.0 EJ.0 82.0 40.0 42.0 ,

i

. _ , __ _ ._ - . . , .- . . . , . . ~ . . . .- - . ~ . . . ~ . . - -. ..

COMPARISON OF REG. GUIDE 1.99 REVISIONS 1 AND 2 FOR RIVER BEND UNIT 1 BELTLINE MATERIALS Vessel Weld (Beltline) : BE , BF , BG Thickness 5.41 inches Material Heat: 492L4871 / Lot A421B27AE Chemistry: C Mn P S Si Cu Ni Mo O.07 1.06 0.018 0.025 0.37 0.04. 0.95 0.5 Initial RTndt: RTndt-I = -60 F, Sigma-I = 0F 32 EFPY Fluence (f): Calculated-Peak 1/4T f = 6.4E+18.n/ca^2 (used with-Rev 1)

Calculated Peak I.D.-f = 6.6E+18 n/ca^2-

, Rev 2 Attenuated 1/4T f = 4.8E+18 n/ca^2 (basis for Rev 2 delta RT).

Surveillance Testing Affecting Rev 1 Shift Calculation: .

Surveillance testing not yet done'.-

Correction factor. applied = 1 Chemistry Factor for Rev 2 Shift: CF= 54 Comparison of Rev 1 and Rev 2' SHIFT and ART (d_egrees-F) .versus EFPY: -

i Rev:2 RevL2 Rev 2 Rev-2 Rev 1 Rev 1 EFPY Delta RT . Margin ' SHIFT ART SHIFT ART.

4 17.4- 17.4 34.7. -25.3 25.5 -34.5-8 24.5- 24.5 48.9 --11.1 36.0 -24.0 12 29.3 29.3 .58.7 -1.3 44.1 -15.9 16 33.1 33.1 66.1 6.1 50.9 -9.1 20 36.1 36.1 72.2 12.2 56.9 '-3 .1 '

24 L38.7- 38.7 77.4- 17.4 62.4. -2.4 28 40.9 40.9: .81.8 21.8 67.3 7.3 32 42.9 42.9: -85.7 25.7 72.0 12.0

, .- .,. , .: . .,- -. - - . - . . .  ;. . , . ~ . . .

l 1.

l l

I COMPARISON OF REG. GUIDE 1.99 REVISIONS 1 AND 2

FOR RIVER BEND UNIT 1 BELTLINE MATERIALS-i Vessel Weld (Beltline)

BE ,-BF , BG Thickness 5.41 inches Material Heat: 492L4871 / Lot A421B27AF Chemistry: C Mn P S Si Cu Ni Mo 0.07 1.17 0.02 0.02 0.32 0.03 0.98 0.51 Initial RTndt: RTndt-I = -50 F, Sigma-I = 0F 32 EFPY Fluence (f): Calculated Peak 1/4T f = 6.4E+18 n/cu^2 (used with Rev 1)

Calculated Peak I.D..f = 6.6E+18 n/ca^2 Rev 2 Attenuated 1/4T f = 4.8E+18 n/ca^2 (basis for Rev 2 delta RT)

Surveillance Testing Affecting Rev-1 Shift Calculation:

Surveillance testing not yet done.-

Correctionsfactor applied = 1' ,

Chemistry Factor for Rev 2 Shift: CF= 41 Comparison of Rev 1 and Rev 2 SHIFT,and' ART (degrees F) versus EFPY:

Rev 2 Rev 2 'Rev'2 Rev 2 Rev 1 Rev 1.

EFPY Delta RT Margin SHIFT ART SHIFT. ART 4 13.2 13.2 26.4' -23.6 28.3 -21. 7 -'

8 18.6 18.6 37.1 - -12.9 40.0 -10.0 12 22.3 22.3- -44.5' -5.5 49.0 -1.0 1 46 25.1 2 5'.1 > 50.2 0.2 56.6 6.6' 20 27.4 -27.4 54.8 4.8 63.2 13.2 24- 29.4 29.4 58.7- 8.7 69.3 19.3 28 31.1 31.1 ~. 62.1- 12.1 74.8 24.8 32- '32.5 32.5' -65.1 15.1 80.0 30.0 m 4%-#., v e

1. v g- 9 .- - - -

r.#.t %.. ,ev.-

<w.- r- ..n.;_-_ _ j ,s _ _2c

L i.

t I

I l COMPARISON OF REG. GUIDE 1.99 REVISIONS 1 AND 2 FOR RIVER BEND UNIT 1 BELTLINE MATERIALS Vessel Weld (Beltline) : BE , BF , BG Thickness 5.41 inches Material Heat: SP6756 / Lot 0342 (Tanden Wire)

Chemistry: C Mn P S Si Cu Ni Mo 0.078 1.24 0.01 0.012 0.53 0.09 0.92 0.46 Initial RTndt: RTndt-I = -50 F, Sigma-I = 0F 32 EFPY Fluence (f): Calculated Peak 1/4T f = ~6.4E+18 n/ca^2 (used with Rev 1)

Calculated Peak I.D.:f = 6.6E+18 n/ca^2 Rev 2. Attenuated 1/4T f = 4.8E+18 n/cm^2 (basis for Rev 2 delta RT)

f. Surveillance Testing Af fecting Rev 1' Shift Calculation:

- Surveillance testing not yet done.

Correction ~ factor applied = 1 Chemistry Factor for Rev 2 Shift: CF= 122 Comparison of Rev 1 and Rev-2 SHIFT:and ART (degrees F) versus EFPY:

Rev 2 'Rev 2 ~ Rev 2 Rev 2 -Rev~1 Rev l' EFPY Delta RT Margin -SHIFT ART' SHIFT ' ART 4 39.2 '39.2 78.4 .28.4 '17.0 -33.0.

8 55.3 55.3 110.5 -60.5 24.0 -26.0

'12 66.3 56.0 122.3 72.3 29.4 -20.6 16 74.7 56.0 130.7 -80.7 33.9 -16.1-20 81.6 56.0 - 137.6: 87.6 37.9 -12.1-24 ~87.4 56.0 - 143.4 93.4 41.6 -8.4 ,

28 92.4 56.0 -148.4 D98.4 44.9 -5.1 32 96.8 5'6. 0 152.8 102.8 ^48.0 -2.0 4

4 m _ _ _ -3 w- s-wa- 4 en r- r -(e-c', m + 4,-e-_,-y. p e y , , . , , . _ _ _ _ _ _ _ _ _ ._ _ m

I~

f l COMPARISON OF REG. GUIDE 1.99 REVISIONS 1 AND 2 FOR RIVER BEND ' UNIT 1 BEUTLINE MATERIALS Vessel Weld (Beltline) : BE , BF , BG Thickness 5.41' inches Material Heat: SP6756 / Lot 0342 (Single Wire) l- Chemistry: -C LMn P S Si O2 Ni' Mo 0.063 1.27 0.01 0.011 0.57 0.09 0.93 0.45 l

Initial RTndt: RTndt-I = -60 F, Sigma-I = 0F

-32 EFPY Fluence (f): Calculated Peak 1/4T-f.= 6.4E+18 n/ca^2 (used with Rev 1)

Calculated Peak I.D..fg= 6.6E+18-n/ca^2 Rev 2 Attenuated 1/4T f = 4.8E+18 n/ca^2 (basis for Rev 2 delta RT) '

Surveillance Testing Affecting Rev 1 Shift Calculation:

Surveillance testing;not yet'done.

Correction factor-applied = 1 Chemistry-Factor for Rev.2 Shift:- CF= 122 Comparison of Rev 1 and Rev 2 SHIFT and ARTc(degrees F)-versus EFPY:

~ Rev 2 Rev 2 Rev'2 Rev 2 Rev 1- Rev 1 EFPY Delta RT Margin SHIFT -ART SHIFT ART-4 39.2 -39.2,  ; 78.4 -18.4- 17.0 -43.0 8' 55.3 '55.3 *

.110.5- 50.5 24.0 -36.0

.12 66.3 H56.0 122.3 62.3 29.4 -30.6- '

16- 74.7 .56.0 130.7 .70.7 33.9 -26.1 20 81.6 56.0- -137.6 77.6: '37.9. -22.1:

• -24 87.4 56.0 143.4 - 83.4 41.6' -18.4 28' 92.4 56.0 148.4- 88.4 .44.9 -15.1

] 32 96.8 56.0 -152.8L _ 92.8f 48.0 -12.0 h

a-- _ ___ _ __m _ _ _,__ -

v = , = ~ - , , ~ ~ . . y ..- ,.y , G., . , , , , y_,,, ,,,.Q,, , ,_ . . , _ ,.

l l

APPENDIX B~ j BELTLINE PRESSURE TEMPERATURE CURVE CALCULATION METHOD The beltline is the region of the vessel that will accumulate more than 10 17 n/cm2 fluence during . operation. ' The : vessel 'wal1~ . f rom the bottom of active fuel to the top of active fuel meets these conditions. The River Bend beltline consists of one shell of plates and the connecting walds. Therefore, there are no discontinuity-regions to consider in the beltline curve analyses. The methods used ,

for the pressure-test and heatup/cooldown curves are described below.

The core critical operation curve.is. simply the'heatup/cooldown curve plus 40*F, as required-in 10CFR50' Appendix G [1], so the methods; for . 4 the heatup/cooldown curves apply to the core critical curves as well.

-i B.1 PRESSURE TEST t

In general, the methods of ASME Code Section III,= Appendix G [8]

are used to calculate the pressure test. beltline-limits. The vessel shell, with an inside radius (R) to minimum thickness (tmin) ratio.of-about 20, is treated as a thin walled cylinder. 'The maximum stress is the hoop stress, given as a ,- PR/tmin' '

The stress intensity factor, Ky ,, is calculated .using Figure G-2214-1 of [5), accounting for the proper ratio of stress to yield strength. Figure G 2214-1 was taken from Velding Research Council (WRC) Bulletin 175 .[9), and is based on a:1/4 T radial flaw with a six to-one aspect ratio (length of 1. 5 _ T) . The- flaw is- I oriented normal to the maximum stress, in ' this case a vertically oriented flaw. This orientation is used even in the case where the:  !

circumferential weld is the limiting _ beltline material', aa mandated by the NRC in the past.

l t l

B1 s

i t

Pressure test K IR is . the calculated value Kg , multiplied) by a safety factor of 1,$, per [8), The relationship between. KIR ""d #

temperature relative to reference temperature ((T - RTNDT) is shovn in )

Figure G 2210 1 of [8), represented by the relationship K

IR 26,78 - 1,233 e [ 0.0145 ( T RTNDT + 160 )) (3,y) j This relationship is derived in (9) as the lower bound of all dynamic fracture toughness and crack arrest toughness data. This_ relationship provides values of pressure (from KIR) versus T (from'(T RTNDT))  ;

B.2 HEATUP/COOLDOWN i

The beltline curves for heatup/cooldown conditions are influenced by pressure stresses- and thermal stresses, ,according to the relationship in (8)

KIR - 2.0 Kg ,+ Kyt, .(B 2) where Kg ,is primary membrane K due to' pressure and K gg is radial thermal gradient K due to heatup/cooldown, The pressure stress intensity factor Kg , is calculated by the method described in section B.1, the only difference being the ' larger safety factor applied, The .the rmal gradient- stress intensity factor calculation is described below. '

The thermal stresses in _ the vessel wall' are caused by a radial thermal gradient which is created by changes in the adjacent reactor-coolant temperature in heatup or cooldown conditions. The stress  !

intensity factor is computed by multiplying the coefficient Mf from ,

Figure G-2214-2 of [8] by the through-wall temperature gradient ATw, given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214 3 of [8).

B-2

-~

l The relationship used to compute through wall AT, is based on one dimensional heat conduction through an insulated flat plate:  ;

62T(x,t)/6x2 - 1/$ (6T(x,t)/6t), where (B 3)

~

T(x,t) is temperature of the plate at depth'x and time t.

A is thermal diffusivity (ft 2 /hr).

.j Maximum stress will occur when the radial thermal gradient reaches- a :- .

quasi steady state distribution, so that 6T(x,t)/6t. - dT(t)/dt - C-where G is the heatup/cooldown : rate', -in . this . case 100'F/hr. -The differential equation is integrated over x-for the following boundary conditions, shown in Figure B 1:

1. Vessel inside surface , (x - 0) temperature ' is the same as the -

coolant temperature,_To.

i

2. Vessel outside surface (x - C) 'is ' perfectly insulated, so - the thermal gradient dT/dx 0, i

The integrated solution results in the:following relationship for wall temperature:

T - Gx2/2p . GCx/$ + T o, where .(Bd)

This equation is normalized to plot (T ' - T o)/ATy versus x/C in Figure B 2. The resulting through wall gradient compares very closely l

l l with Figure G-2214 3 of [8). Therefore, .AT, . calculated from l

Equation B-4-is used with the appropriate Mt of Figure G 2214-2 of [8]

to compute Kyt for heatup and cooldown.

l l The M t relationships were derived in [9] for' infinitely long cracks of 1/4 T and 1/8 T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.

l B-3

l The stress generated by the thermal gradient is a bending stress that changes sign from one side of ' the plate to the other. In d combining pressure and thermal stresses, it is . usually necessary to l evaluate stresses at the 1/4 T location (inside surface flaw) and the- l 3/4 T location - (outside - surface flaw). This is because the thermal I gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a  ;

conservative simplification, the thermal gradient stress at the 1/4 T t

is assumed to be tensile for both heatup and cooldown- This results in the maximum tensile stress being applied for heatup and cooldown at'

+

the 1/4 T' location where irradiation effects are more 'significant.

This conservatism causes no operation difficulties, since the ' BWR is at steam saturation conditions during normal heatup or cooldown, well; above the heatup/cooldown curve limits, i

Given the form of the equation by which AT, is determined, the heatup/cooldown rate of 100'F/hr for brittle fracture purposes refers ,

to an instantaneous rate. Instantaneous : rates in' excess of 100*F/hr are allowed for in the Tech Spec, as long:as a temperature . ' change' of' 100*F in a one, hour. period is not exceeded. This is based on the fact that the 1/4 T location of the assumed . flaw sees little if any effect t

of _ small perturbations in the 100*F/hr rate, due to - the thermal-inertia of the vessel' wall, It is understood in this ' Tech Spec allowance that operators will track vessel coolant heatup or cooldown.

to stay as close to a 100'F/hr rate as possible.

B.3 EXAMPLE CALCULATION - 8 EFPY PRESSURE TEST AT 1000'PSIG The following inputs were used in the beltline limit calculation:

RTNDT ......................... 61*F [

Vessel Height ................. 832 inch i Bottom of Active Fuel Height .. 209 inch Vessel Radius ................. 110.2 inch Vessel Thickness .............. 5.41 inch Beltline Material Sy .........., 62.2 ksi B4 y-

i Pressure was calculated to include- hydrostatic pressure for a full' l s

vessel:

P - 1000 psi +-(832 209) inch

• 0.0361 prifinch - 1022.5 osig 0

Pressure stress:

?

o - PR/t - 1022,5 psig

• 110,2 inch / 5.41 inch - 20828 osi-

-The factor M, depends on'(a/S y ) and /t:

o/Sy - 20828 / 62200 - 0,3

/t - (5,41)1/2 - 2,33 M - 2.11 1'

The stress intensity factor .Kym, is M,* a:

Ky , - 2.23

• 20828 - 46446 psi /in - 46 5 ksilin e

Equation (B 1) can be rearranged to solve for (T - RTNDT)*

(T - RTNDT) ~ 1"I(KIR - 26.78)/1,233)/0,0145 160 (T - RTNDT) - in[(1.5*46.5 ;26.78)/1.233)/0.0145 160 l (T RTNDT) ' E ,

Adding the adjusted RTNDT f r 8 EFPY of 61'F: .

{ -

T - 146'F l

l' l

B-5 i

l r K _

+

\

Reactor Coolant N

Cooldown Rate ,

G= 100 - F/hr - e77ex=o insulation ,

3 f-T=To P

Ou'tside Inside Sudace Sudace y )

x=0 x=C L

e Figure B-1. Boundary _ Conditions for Hectup/Cooldown Temperoture l

l B-6 ,

1 r~

i foo 90 . / i 2

/

m E /

d e 70 ,

E e

60 /

8 [.

f#0

[

tg 30 ./

a k ' 20 f t to -

f  :.5 i

o 0 20 - 40 60 80 100 l

l- waii Thickness, x ,

a

-t 1

Figure B-2. Assumed Through-Wall Temperature During Heatup/Cooldown 1

i B-7 i

s 9

• 1 ,

1

1 1

Attachment 2.'

4 General Electric Report SASR 88-49 " Flux Wirs Dosimeter Evaluation For River Bend Station"~

4 1 -

l. .

9 1' y J

S i

(. .

3 s

i l.

l l.

I