NLS8800063, Forwards Responses to NRC 880113 Questions on Proposed Change 48 & Revised Tech Spec Pages 155 & 156 Showing Increase of 2 F in Nonbeltine Portions of Curves of Figures 3.6.1.a & 3.6.1.b.GE Rept NEDO-24793 Also Encl

Forwards Responses to NRC 880113 Questions on Proposed Change 48 & Revised Tech Spec Pages 155 & 156 Showing Increase of 2 F in Nonbeltine Portions of Curves of Figures 3.6.1.a & 3.6.1.b.GE Rept NEDO-24793 Also Encl

ML20147D623
Person / Time
Site:	Browns Ferry, Cooper
Issue date:	02/22/1988
From:	Trevors G NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20147D629	List: ML20147D634 ML20147D670 NLS8800063, Forwards Responses to NRC 880113 Questions on Proposed Change 48 & Revised Tech Spec Pages 155 & 156 Showing Increase of 2 F in Nonbeltine Portions of Curves of Figures 3.6.1.a & 3.6.1.b.GE Rept NEDO-24793 Also Encl
References
NLS8800063, NUDOCS 8803040073
Download: ML20147D623 (13)

Text

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NLS8800063 February 22,-1988

U.S. Nuclear Regulatory Commission

. Document Control Desk Washington, D.C.-20555 Gentlemen:-

Subject:

Supplemental Submittal; Proposed Change No. 48 to the Cooper Nuclear Station Technical Specifications NRC-Docket No.-50-298, DPR-46

References:

1)

Telecopy from W. O. Long to M. T. Boyce, dated January.13, 1988, Transmittal of NRC Questions on Proposed Change No. 48.

2)

Telephone Conversation, B. Elliot (NRC),

K. Walden, J. Meacham, M. Boyce, G. Smith, and P. Ballinger (NPPD) and T. Caine (GE), January 19,11988,. Discussion of NRC

-Questions on Change No. 48.

3)

-Letter NLS8700560, from L. G. Kuncl to U.S.NRC, dated October 28,1987, "Proposed Change No. 48 to the Cooper Nuclear Station Technical Specifications."

4)

Letter NLS8700310, from G. A. Trevors to U.S. NRC, dated July 6,1987, Submittal of GE Report MDC-103-0986, dated May.1987, "Cooper Nuclear Station Reactor Pressure Vessel Surveillance Materials Testing and Fracture Toughness Analysis." provides Nebraska Public Power District's responses to the NRC questions transmitted in Reference 1.

The District's responses also reflect the' clarifications to the questions made during the January 19 conference call (Reference 2).

The NRC questions refer to Proposed Technical Specification Change No. 48, Reference 3, and GE Report MDC-103-0986, Reference 4.

Enclosed as Attachment 2 are revised Technical Specification pages 155 and 156, showing an increase of 2'F in the non-beltline portions of the curves of Figures 3.6.1.a anci 3.6.1.b.

The 2*F increase in maximum RTunT for the ASTM A 508 forgings resulted from a reanalysis of the data in Ecordance with Branch Technical Position MTEB 5-2, as discussed in Reference 2.

The no significant hazards consideration evaluation previously submitted in Pro)osed Change No. 48 remains applicable, since this added 2*F change moves tie non-beltline portion of the curves in the conservative (higher required minimum temperature) direction.

p 8803040073 860222 I

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Page 2' FCbruary 22. 1988 Also enclosed, in response to NRC Question 1 on the GE fluence evaluation, is a copy of General Electric (GE) Report NED0-24793, "Browns Ferry Unit 3 In-Vessel Neutron-Spectral Analysis.".This report-describes the origin of.

neutron spectrum and cross-section data -used in the GE! fracture toughness analysis.

Should you have any questions regarding the attached. responses, or require further information, please contact this office.

Sincerely,

/

s G. 4. Trevors Division Manager Nuclear Support GAT /mb:mh25/3 (CHNG48)

Attachment cc: NRC Regional Office Region IV Arlington, TX NRC Resident Inspector Office Cooper Nuclear Station B. Elliot (NRC)

J t

m Att:chment 1 t3 NLS8800063 P gi1 ef 11 RESPONSES TO NRC QUESTIONS Estimate of Initial RTNDT

-Question 1 How many heats of ASME SA 533 Grade B Class 1 low alloy steel were evaluated to develop the relationships:

a) 2'F/ft.-lb.,

.b) the difference in longitudinal and transverse charpy energies is 30*F7

Response

The Charpy energy curve slope of 2*F/ft-lb is an upper bound estimate, based on test results of 24 A533 plates reported in Welding Research Council (WRC)

Bulletin 217 [1] and evaluation of 22 plates of SA 533 reported in the LaSalle FSAR [2). The 30*F conversion factor is based on the results in WRC Bulletin 217.

Question 2 How many heats of ASTM A 508 Class 2 low alloy steel were evaluated to determine the difference in longitudinal and transverse charpy properties is 30'F7

Response

There was little data available to establish the longitudinal-to transverse conversion for A508, so the 30'F factor for A533 was used, assuring that the resulting RT values were consistent with A508 NDT values (3).

For the sake NDT of prompt response, additional A508 data has not been collected and evaluated at this time.

Instead, the A508 forging data for the Cooper vessel have been reanalyzed using Branch Technical Position MTEB 5-2 [4].

The A508 forging (closure flange and nozzle) materials were subjected to Charpy tests at one temperature, 10'F.

For all forgings except one steam nozzle forging, NDT values were documented as well. Branch Technical Position MTEB 5 2, Section B.1.1 (4) states, "If limited Charpy V-notch tests were performed at a single temperature to confirm that at least 30 ft-lbs was obtained, that temperature may be used as an estimate of the RT provided that at least 45 ft lbs was obtainedifthespecimenswerNongitudinallyoriented.

If the minimum value 'obtained was less than 45 ft-lbs, the RT may e est mated as NDT 20*F above the test temperature."

This procedure was applied to the A508 forging data, as summarized in Table 1 on page 6 of 11.

The maximum RT is 30'F, compared to the non beltline NDT RT limit f 28'T used in the pressure temperature curves.

NDT

-1 d

Att:chment 1 to NLS8800063 P;go 2 cf 11 The 2'F increase in RT impacts the non beltline portions of the curves in Figures 3.6.1.a and 3N1.b of the Technical Specifications.

Revised Figures 3.6.1.a and b are included as Attachment 2, updated to reflect the 2'F RT in rease.

These revised pages replace the previous curves submitted wi$TProposed Change No. 48.

These two figures also appear as Figures 2-1, 2-2, 2-3, 7-5, 7-6, and 7-7 of GE Report MDE 103 0986. Also included in this Attachment as page 8 of 11 is a revised Table 7 3 from MDE-103 0936, showing the revised temperature values in the non beltline region of Figures 3.6.1.a and 3.6.1.b.

Question 3 How many heats of Combustion Engineering weld metal fabricated with Linde 1092 flux were evaluated to determine that the lower limit of its NDT temperature is -50'F?

Response

The 50'F limit set for weld metal was an administrative lower limit for RTND, based on the determination that the RT determined from Charpy energy wasbigherthantheweldNDT.

This appears t b he NRC to be an unacceptable basis for the -50'F RT values assumed for the Cuper welds.

The NRC NDT recommendation is to use the RT upper bound value used in pressurized thermalshock(PTS) evaluations,"w$ichis:

Upper Bound RT

~ "*""

NDT +

NDT where Mean RT 56*F, and Ng,17.p,,,

Upper Bound RT

~~

NDT in initial RT w

for which weld unavai1Ne,ould cause weld Heat 12420, This increase to become the limiting beltline material chemistry data were prior to 32 EFPY, However, a review of PTS evaluation data showed that a Linde 1092 weld was made with Heat 12420 in the beltline at Salem Unit 1 [5).

This is identical to the CNS weld.

The as-welded chemistry for Heat 12420, 1092 Flux Lot 3708 at Salem Unit 1 is 0.22% Cu and 1.02% Ni.

Phosphorus information was not provided.

For the purposes of using Regulatory Guide 1.99, Revision 1, a phosphorus content of 0.020% is assumed.

This chemistry data was used, along with the -22'F initial RT L"

1. • • • "*1"***

NDT' the weld metal fabricated with Linde 1092 flux.

Table 2 on page 7 of 11 shows the adjusted reference temperature (ART) for the limiting plate and the beltline welds.

As shown, using the chemistry data from Salem 1, the welds are less limiting than the plate, even considering the higher initial RT Veld Heat 12420 could have a copper content as high as 0.28% witN.

In fact.

t becoming limiting.

Therefore, the pressure-temperature curves in the surveillance report, and submitted in Change 48, are considered bounding for the beltline.

2

7

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)

'. 'Attcchment l~ta N1.54800063.-

-Paga 3'ef 11

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vf

.-Question 4

'}

i

.Are_the relationships 1in 1) and 2) mean values or upper bound? -Was there a

! statistical evaluation of the,deta?

.e RResponse:

. Refer,to Questions 1-~nd 2.

.l a

(

Fluence Evaluation 9

Question 1.

How were the cross-sections, E>1MeV, for the copper and iron wires determined?

l

.What spectra were used-for the iron 54 and copper-63 reactions?

. Response-l Fe 54 (n.p) Mn 54 and Cu 63 (n.a) Co 60 cross section values for Cooper'were obtained from cross section functions and reaction rate ratios developed from i.

itest measurements. These measurements were taken from neutron energy spectral determinations for,BWRs and for the GE Test Reactor at Vallecitos.

GE Report j

NEDo 24792 [6), enclosed. describes the experiment in which neutron spectra -

were determined at eight locations between the core shroud and the vessel inside wall at the Browns Ferry 3 plant. From this experiment, data functions were' derived _ relating cross section to water gap (distance from fuel to vessel

~

wall) for BWRs.- The use of test measurements to determine cross sections is

.I expected to be more accurate than a method which uses a calculated spectral

'l hape.

s Question 2 i

Do the 2-signa deviations reported on pp 4-3 include uncertainties for I

counting, weighting, water gap and capsule location with respect to the core j[

and the pressure vessel? If it does not, what is the 2-signa deviation for

-the dosimetry estimates including all sources of uncertainty?

l

Response

The 2a uncertainty reported includes uncertainties in determining the flux l

from the dosimeter flux wires (e.g.,

counting, weighting, cross sections, etc.).-

The uncertainty associated with the lead factors is not reported, and was not calculated.

Instead, the lead factors were computed using conservative assumptions to assure that the vessel flux predicted from the l

dosimeter results is conservative. The computational model used to determine lead. factors assumed the minimum vessel diameter and allowed for no water gap between the dosimeter o g the jnside vessel wall. The computational value of j

dosimeter flux, 0.96x10 n/cm, 18 88 expected, slightly less than the i

9 2

measured flux of 1.05x10 n/cm. Therefore, combining the measured flux with

[

its uncertainty factor of 1.25 and the conservative lead factor, gives a i

conservative estimate of flux at the vessel 1/4 T.

i e

j 1

3

Att:chment 1 to NLS8800063 Pega 4 cf 11 Question 3 Did your boundary condition input into the computer codes (DOT and SN1D) account for albedo neutrons at kT? Do the course mesh used in the computer programs provide adequate solutions for the neutron flux?

Response

The boundary conditions used in the computer codes did not account for albedo neutrons at the 1/4T.

Comparisons of flux decrease through the vessel wall were made between the DOT two dimensional model, which stopped at the 1/4 T, and the SN1D one-dimensional model, which extended several centimeters further into the vessel vall. The rates of flux decrease at the 1/4 T in each model match closely, indicating insignificant flux build up due to reflected neutrons.

Past flux analyses which modelled the entire vessel wall thickness, the biological shield wall and the annulus between have shown the reflected neutron contribution at the 1/4 T to be insignificant for neutron ener5 es i

>1MeV.

There are lower energy neutron groups where reflected neutrons may be significant, but the >1MeV flux is not significantly affected.

The flux computation is a combination of two dimensional and one-dimensional models.

The two dimensional model aesh is described in the surveillance report. The one-dimensional model had about 110 radial intervals (versus 43),

with about 15 intervals for the vessel vall (versus 4).

This combination of a somewhat coarse two dimensional model and a finer one-dimensional model provides adequate flux solutions, as shown by the agreement within 9% with the Cooper flux wire test results.

Pressure Temperature Curves Question 1 Where are thermocouples located that will be used to measure the minimum metal temperature of the beltline and closure flange (vessel and head) regions i

during the leak and pressure tests (Figure 7 5)? During these tests what is the difference in temperature between the thermocouplca and the beltline and closure flange locations? Do administrative controls iden*ify to the operator which thermocouples must exceed the PT curves in Figure 7-6'.

Response

Figure 1 to this Attachment provides g'.neral locations of the thermocouples attached to the pressure vessel, fl.nge, and head regions.

As shown, i

thermocouples are located at various locations on the vessel shell, flange, and heads, although none are placed on the vessel beltline region. Table 3 is a tabulation of all thermocouples which are read out on recorders in the Control Room.

Metal temperature on the beltline regions will be correlated with an existing vessel shell thermocouple (TE-69 J1) reading during the next cutage prior to performance of the hydrostatic pressure test.

This vessel shell thermocouple reading, with the determined correction factor, will be used by the Control Room Operators to monitor beltline temperature during subsequent hydrostatic pressure testing.

Separate thermocouples will be used to measure the metal temperature of the bottom head region.

The vessel bottom head region is at a lower temperature than the remainder of the pressure vessel due to the influx 4-l 1

Attcchment 1 e,o NLS8800063 P:go 5 cf 11 of cooling water for the Control Rod Drive hydraulic units and the poor mixing characteristics brought about by the internal structure configuration of the region.

The District recognizes that different pressure-temperature limitations apply to certain pressure vessel regions and that appropriate thermocouples must be monitored for each identified vessel region.

The appropriate thermocouple outputs will be identified to plant Operators for monitoring vessel temperatures and procedurally controlled during hydrostatic testing and other plant evolutions where thermal and pressurization limitations apply.

5-L

Attcchment' 1 to NLS8800063 Pcg3 6 cf -11' 2,..g 3

4 f.

Table 1 ESTINATES FOR COOPER-VESSEL A508 CLASS-2 FORGINGS INITIAL RTST t.'u Minimum.

Charpy Test

-Charpy L-Temperature.

Energy NDT' RTNDT.

c - onant

(*F)

-(ft-lbg1 M

Q

?

-Head Flange

,10 74 20-

'20 a Vessel Flange 10 62 10 10 Mozzles t

Recire. Outlet 10 58 0-

.10 Recirc. Inlet' 10 35 10 30 Steen Outlet 10 --

31 N/A 30 Feedwater 10 48 0

10 Core Spray 10 45 0

10 Head Spray 10 69 0

10 Vent 10 74 0

10 q

Jet Pump Instrument 10 134

-10 10 CRD Ret. urn in 87 19 to l

l 1

1 l

l i

i l

• The RTNDT was taken as 20'F because of the NDT. RT NDT Per MTER 5-2 is 10'F.

i

__.___.__._..__._____d

- = - - -

A t ta:hme2L 1 te CLS8800063 g;

Paga 7 cf 11.

e i

/

Table 2 ADJUSTED REFERENCE TEMPERATURES FOR BELTLINE MATERIALS AT 32 EFPY Survalliance Initial HDr

. Shift ART 1092 Test RT l

'i ligat Flux Lot 1Gu 12.,

Factor

(

• F).

L*f.L M

Plate:

C2274 1 0.20 0.010 2.39 14 157 171 I

r l

Welds:

l I

l 27204 &

l 12008 3724 0.19 0.013 1.62 22 110 88 b

12420 3724 0.22" 0.020 1.62

-22 151

' 129 I

i l

i b

12420 3708 0.22" 0.020 1,62 22 151 129 l

21935 3869 0.20 0.016 1.62 22 125 103 Hypothetical Veld:

0.28 0.020 1.62

-22 188 166

• Based on data from Salem 1 PTS evaluation, b Assumed upper limit for weld process.

9

Attachrent 1 to NLS8800063 PsTs 8 of 11 Revissd Tabis 7-3 GE Report MDE-103-0986 Table 7-3 PRESSURE-TEMPERATURE,VA1.UES FOR FICURE 7-6 (CURVE B) AND 7-7 (CURVE C)

. Pressure Curve B Curve C (esi)

Temo. (*F)

Temo. (*F)

Remarks 0

80' 80 Boltup Temperature 48 8 0 80 Non-beltline limits Curve C i

50 80 82 60 80 94 70 80 102 80 80 110 90 80 117 95 80 120 Non-beltline' limits Curve.B 100 84 124 110 89 129 120 94 134 130 99 139 140 104 144 150 108 148 160 112 152 170 115.5 155.5-180

~119 159 190 122 162 200 125 165 210 127.5 167.5 220 130 170 230 132.5 172.5 i

240 135 175 250' 137 177 260 139 179 270 141 181 280 143 183 290 145 185

-300-147 187 Increment pressure 20 psi 320 150 190 340 153 193 360 157 197 380 160 200 400 162.5 202.5 420 165 205 440 168 208 460 170 210 480 172 212 500 174.5 214.5 520 176.5 216.5 540 178.5 218.5 550 179.5 219.5 Beltline becomes limiting 560 181 221 580 184.5 224.5 7-11

m:.

+

Attschment 1 to'NLS8800063-

Pcga 9 of-11~

L1 -

TABLE 3-

Recorder TR-90

. Thermocouple

-Recorder Input-

' Thermocouple Location

' Identification Black'

Vessel Flange

TE-69-A2 Red.

Vessel Wall Adjacent.to Flange TE-69-82

. Recorder TR-89 Thermocouple:

-Recorder Input Thermocouple Location Identification ^

00

'Feedwater Nozzle N4B End-

.TE-69-D1 01-Feedwater' Nozzle'N4B Inboard

<TE-69-D2-L 02 Feedwater Nozzle N4D End.

'TE-69-El L'

'03 Feedwater: Nozzle N4D Inboard TE-69-E2 04 Vessel Top Head' Adjacent to Flange TE-66-Al-05 Vessel; Top Head. Flange.

TE-66-B1-06

-Vessel Bottom Drain Line

'TE-106 07,

Vessel Wall Below Feedwater Nozzle TE-69-J1 08 Vessel Wall Above Botton Head TE-69-H1 09 Vessel Above Skirt Junction 1TE-69-Fl' 10

-Vessel Botton Head TE 69-L1 Ell-Vesnel Skirt Near Junction TE 69 K1 9-

Attachm:nt 1 to NLS8800063 Page 10 of 11 2 MAGNETIC ALLY CLAMPED TO HE AD 2 TO HEAD FLANGE iVESSEL HEAD (2)

M HEAD FLANGE ]

A w

o o

TE-69-A2 O ll

[

TE-60-82 O VESSEL FLANGE C

a.

a.

2 2

e e

U U

1 E

E

_g @

VESSEL SHELL N

MS) 34 LOCA flONS C

(

THERMOCOUPLE TR 2 INPUTS 89 F.W.

HOZZLE TR 12 INPUTS 90 9-21 P ANEL VESSEL SKIRT

(

/

\\

t

\\

/

l

/N

,/

- DR AiN tine s

DOTTOM HE AD X

[

s i t t Y

TO CLE ANUP SYSTEM l

t l

l THERMOCOUPLE PAD LOCATIONS l

FIGURE 1

'Attichm2nt 1 to'NLS8800063:

Pags 11 of 11,

.-REFERENCES

.[1] Hodge, J. M. ', "Properties of Heavy Section Nuclear Reactor Steels,"

. Welding Research Council Bulletin 217,- July.1976.

~

-[2].LaSalle ' County ~ Station Final Safety Analysis Report, Response to'NRC

~

Question'121.11,-Amendment 46, August.1979.

1[3].LaSalle County Station Final Safety Analysis Report, Response to NRC Question'121.17, Amendment 54, January 1981.

[4] Nuclear: Regulatory ' Commission Branch - Technical Position MTEB 5-2, "Fracture Toughness requirements," Revisions 1, July 1981.

[5]

Pressurized Thermal Shock Evaluation for the Salem Units, Public Service Electric & Gas Company Report NRJ-060, Revision 0, January 1986.

[6]' Martin, G.

C., "Browns Ferry Unit 3 In-Vessel Neutron Spectral-Analysis',"

3 General Electric Company Report NEDO-24793, August 1980.

l-h E

r 9 -. -,. -. =. -, _.. - - -.

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