JPN-89-044, Responds to Generic Ltr 88-11, Radiation Embrittlement of Reactor Vessel Matls. Mod of pressure-temp Limits Contained in Tech Spec Necessary in Order to Continue to Satisfy Requirements of 10CFR50,App G

Responds to Generic Ltr 88-11, Radiation Embrittlement of Reactor Vessel Matls. Mod of pressure-temp Limits Contained in Tech Spec Necessary in Order to Continue to Satisfy Requirements of 10CFR50,App G

ML20246A944
Person / Time
Site:	FitzPatrick
Issue date:	06/30/1989
From:	Brons J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-88-11, JPN-89-044, JPN-89-44, NUDOCS 8907070143
Download: ML20246A944 (6)

Text

,.

123 Main strom

,a.

White Plains, NewYork10601 j

6, M4 681.6240 l

l NmW&%er aosa c.s,e -

1 4# Authority RJ%';E2lar?"'

June 30, 1989 JPN-89144 U.

S. Nuclear Regulatory Commission Mail Station'P1-137 i

Washington, D.C.

20555 l

Attn:

Document' Control Desk

Subject:

James A.

FitzPatrick Nuclear Power Plant I

Docket No. 50-333 l

Response to Generic Letter 88-11 Radiation Embrittlement of Reactor Vessel Materials l

References:

1.

NRC Regulatory Guide 1.99 Revision 2, dated May 1988, " Radiation Embrittlement of Reactor Vessel Materials."

2.

NRC Generic Letter 88-11, dated July 12, 1988, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on 1

Plant Operations."

3.

NYPA letter, J.

C.

Brons to NRC, dated i

January 12, 1989, JPN-89-004, concerning the same subject.

Dear Sir:

Regulatory Guide 1.99 (Reference 1) describes methods used to evaluate neutron embrittlement of reactor vessel beltline materials.

Generic Letter 88-11 (Reference 2) requests that licensees use these methods to evaluate predicted i

embrittlement and to submit results of this evaluation and a schedule for any proposed actions.

Reference 3 requested a change in the response date to June 30, 1989.

The Authority has completed the required analysis, and concluded that a modification of the pressure-temperature limits contained in the FitzPatrick Plant Technical Specification is necessary in order to continue to satisfy the requirements of 10 CFR 50, Appendix G.

g0 l' \\

8907070143 890630 1

PDR ADOCK 05000333 p

PDC L

m-l l

i

)

l The vessel plate, G-3415-3, and weld, 2-233, are the most J

limiting components of the reactor vessel. This is based on the chemistry factor for coppe5 ^"d "i k 1 and initial reference 8

temperatures (RTNDT) of 24 F and -22 F, respectively.

Both materials are in the lower shell course, near the boundary of the core beltline. (See attached Figure 1). As part of this analysis, 4

FitzPatrick plant specific lead factors have been determined for l

the inside surface of the vessel at the peak lower-intermediate shell location (0.61) and at the peak lower shell location (0.72).

Pressure-temperature curves were developed using the most conservative chemistry data available for operation to 12, 14, and 16 Effective Full Power Years (EFPY).

The ngn-beltline region curves are based on the RTNDT value of 30 F calculated for the recirculation inlet nozzles and head flanges and are not affected by the changes to RTNDT shift associated with Regulatory Guide 1.99 Revision 2.

The beltline curves apply to the plates and welds and are limiting above 500 psig at 16 EFPY. For the specific example of ghe 1000 psig leak test, the required test temperature is 192 F through 16 EFPY (Figure 2) based on Regulatory Guide 1.99 g

Revision 2 methodology, compared with 157 F (Figure 3) based on Regulatory Guide 1.99 Revision 1 methodology.

The Fit-Patrick Technical Specifications will be revised, as required, in accordance with the results of the analysis.

All changes will be implemented before start-up from the next refueling outage (currently planned for May 1990) in compliance with Reference 2.

Should you or your staff have any questions regarding this matter, please contact Mr. J.

A.

Gray of my staff.

Very truly yours, John C.

Brons Executive Vice President Nuclear Generation Encl.

cc:

See next page 4

i

i L

J L

b 4

l cc:

U.

S.

Nuclear Regulatory Commission Region I i

475 Allendale Road King of Prussia, PA 19406 l

Office of the Resident Inspector l

U. S.

Nuclear Regulatory Commission P.

O.

Box 136 Lycoming, NY 13093 Mr. David E.

LaBarge Project Directorate I-1 Division of Reactor Projects - I/II U. S. Nuclear Regulatory Commission Mail Stop 14 B2 Rockville, MD 20852 i

i i

i M

TOP HE AD ENCLOSURE CL OSU A E

\\

FLANGE. i 71 A E GION jf u N

H%X ffwh a

l

)

UPPE R SHE LL

?

r' 6g j @

f'

(

l UPPE R INTE RME DI ATE l

$ HELL j

l l

l l

l LOWER

]

INTERMEDIATE f

G34142 G34117

$ HELL g

G 34141 h

CORE

/

s BE LTLIN E 3

LONGITUDINAL l

REGION

$E AM WELD l

{

t.233

{

{

l

/CIRCUMFE RENTI AL GIRTH WE LD l

d 1 240 lf 233 2

=

' ' ' ^

y

)

l G 34152 LOWER l

'b G

51R 1

11 1

xr 11 1

1 1a BOTTOM HE AD ENCLOSURE mr:

1 1

1 1

mr r r

r r

1y N

/

I I

Figure 1 Schematic of the RPV Showing Arrangement of Vessel Plates and k' elds l'

,o

.s.

'1600 VALID TO 16 EFPY

- A - SYSTEM HYDROTEST UMIT w!TH FUEL IN VESSEL B - NON-NUCLEAR HEATING A

B C

1400 -

uurT C - NUCLEAR (CORE CRITICAL)

!/

J, uurr 73; 1

/

/

/

1200 b

j l

o

.x l

l o_

r o

e

'H 1000 f

J BELTUNE 7

/

ART =11(fF Lo>

(r 800 w

/

/

O

/

b

/

tr V

EE 600 Ulf' r

H h

NE.t*F d2 to r

.a

\\

g N

NA NON-BELTUNE d

N3 c

j sea.s*F

" NDT 8

n 3,3 nig Q

DOLTUP

[

/

• *r

,i, -

8 o'f f

5) 0--

j j

i 0

50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Figure 2,-.

Pressure-Temperature Limits for FitzPatrick through 16 EFPY 6med on Re3ala4e y Cuide I99 ReAsien 2 i

o.;,.

t-1500 -

I VALID TO 16 EFPY I

I A

B C

-.---.).--~.--.g90 22g

'l goy s

1400_

1400 1400 1400

_l l

At). LUSTED BELTLINE

• WITH 1/47 FLAW LIMITING l

=

I I

l 1250 1250' i

e n

[----._

i 212.5 252.5 1200~

- - - - - - * - - ~ ~ ~ -

l I

I 1, -

i i

i j

g l-l I

l l

I 1-1000 l

1 s

fI I

I I

I l

I

._____..J.___...-

._4.___._p___._.

,00 I,

D

'755 l, f/a f FLA W LIMITI&fG FW N0ZZLE WITH lyg0 wQ l

l' n.

i l

l i

i w

l

__q__.-. _ m _.__. g.__. _ ___.

e

,0 0 _

l l

D n

FLANGE REGION l

l 9

WITH 0.24 INCH l

~

FLAW LIMITING

.w i

I j

e

__.-_[--.__._.I o.

400 _

l CURVE A -lNSERVICE HYDROSTATIC &

l LEAK TESTING l

O' l CURVE B - NON NUCLEAR HEATUP e r

0

'20 CooloowN l

CURVE C - CORE CRITICAL l

..4.. __ _ _. _;..

j 0 0._

I l.

SA TURA TION l

I l

l N

x 112 l

N l

90 l

s7 l

l 80 l

i 0

i i

i i

i I

6 I

I I

I I

50 75 100 125 ISO 175 200 225 250 275 300 325 350 TEMPERA TURE (* F)

FIGURE 3 RE ACTOR VESSEL PRESSURE - TEMPER ATURE LIMITS S A5E D OM R E c.u t A r o A.y c.ulO E 199 R EVt310M l

_ _ _ - -. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.