ML19341A199

Safety Evaluation Supporting Proposed Changes to Tech Specs 3.4.9 & 4.4.9 Re Heatup & Cooldown Limit Curves

ML19341A199
Person / Time
Site:	Farley
Issue date:	01/09/1981
From:	ALABAMA POWER CO.
To:
Shared Package
ML19341A198	List: ML19341A197 ML19341A199 ML19341A200
References
NUDOCS 8101220363
Download: ML19341A199 (2)

Text

.

.- SAFETY EVALUATION FOR CAPSULE Y SURVEILLANCE

' TECHNICAL SPECIFICATION CHANGES

Background:

O Six surycillance capsules for monitoring the effects of neutron exposures on the Plant Farley-Unit I reactor pressure vessel core region raterial were inserted in the reactor pressure vessel prior to initial p? ant start- ~

f up. The six capsules were positioned in the reacNr vessel betieen the C . peutron shielding pads .and the vessel wall at ic* ations shown in Figure f'

j 4-1 of WCAP-9717 " Analysis of Capsule Y from the Alabama Power Company Farley Unit No.1 Reactor Vessel Radiation Surveillance Program."

In accordance with Table 4.4-5 in the Joseph M. Farley Unit 1 Technical Speci-f fications Capsule U was scheduled for irradiation surveillance testing after -

f the replacement of the 1st Region. (The capsules have been relettered, see l Capsule Y was removed l Figure 4-1 in WCAP-9717, Capsule U is now Capsule Y).

f' I

after 1.13 effective full power years of plant operation. WCAP-9717 summarizes testing and post irradiation data obtained from the first Thematerial surveillance post irradiation l

I

. capsule (Capsule Y) removed from the reactor vessel. mechanical .

at the Westinghouse Research and Development Center Hot Cell Laboratory.

l

(

l

References:

(1) WCAP-9717, ~" Analysis of Capsule Y from the Alabama Power Company

" Farley Unit No.1 Reactor Vessel Radiation Surveillance Program."

WCAP-8810, " Southern ~ Alabama Power Company Joseph M. Farley Nuclear (2) Plant Unit No. 'l Reactor Vessel Radiation Surveillance Program."

(3) Joseph M. Farley Nuclear Plant Unit No.1, Technical Specifications, Section 3/4.4.9 and Bases 3/4 4.9.

E .

Bases:

The reactor vessel material irradiation surveillance specimens were removed, examined, and tested to determine changes in material properties. The results of these examinations were used to update Figures 3.4-2 and 3.4-3 concerning heatup and'cooldown limitations, which are applicable for the first 7.7 effective The results of the testing is presented in Section i.

full power years 5 of WCAP-9717.

(EFPY).A discussion of heatup and cooldown limit curves is pres in Appendix A of WCAP-9717.

Based on the new capsule to vessel inner wall . lead factors and the new capsule withdrawal schedule in ASTM E185-79, a modification to Table 4.4-5 " Reactor The Vessel Material Surveillance Program Withdrawal Schedule" is also proposed.

new lead factors represent updated values resulting from improved analytical techniques developed since the issuance of WCAP-8810.

Additionally, Figure B 3/4.4-2 "Effect of Fluence and Copper Content on Shift for Reactor Vessels Exposed to 5500F Temperature" has been deleted.

of RTNDT An adjusted reference temperature, based upon the fluence and copper content of the material can be predicted using Figure B 3/4.4-1 and the recomendations of 8101230%3

tw page 2

- Regulatory Guide [.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The proposed heatup and cool-down limit curves include the predicted adjustments for the shift in RTNDT at'

, the end of 7.7 EFPY. (Note, WCAP-9717 shows 10 EFPY).

Also Table 3/4.4-1, " Reactor Vessel Toughness" has been modified to include estimates per the NRC Regulatory Standard Review Plan, Section 5.3.2. In addition, the proposed changes have been tailored after the Farley Unit 2 Tech-nical Specifications which reflect the latest NRC approved version of the -

,- Standard Technical Specifications.  !

. . i

Conclusion:

The proposed changes to the Technical Specifications and Bases do not involve an unreviewed' safety question as defined hy 10CFR50.59.

5 SL-5 mw l

e e

0 e

4