ML20133C574

Proposed Tech Spec Change Request 104,Rev 3,replacing Current Reactor Vessel Matl Surveillance Requirements w/BAW-1543, Integrated Reactor Vessel Matl Surveillance Program

ML20133C574
Person / Time
Site:	Three Mile Island
Issue date:	10/02/1985
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20133C560	List: ML20133C556 ML20133C574
References
NUDOCS 8510070432
Download: ML20133C574 (4)

Text

_

' ~. 0,,

I. Technical Specification Change Request No.'104, Rev. 3 It is requested that the attached revised pages replace the following pages of the existing Technical Specifications:

pages 4-11, 4-12 Also, pages 4-13 and 4-27a are to be eliminated.

This request supersedes the previously submitted TSCR 104, Rev. 2 in its entirety.

II. Reason for Change By letter dated May 8,1985, the NRC informed the affected B&W licensees that they may request a license amendment to remove the current reactor vessel material surveillance requirements from the Technical Specifications and request consideration of the NRC accepted BAW-1543,

" Integrated Reactor Vessel Mat 2 rial Surveillance Program" for their plant. GPUN requests consideration of the program for TMI-1 in accordance with Section II.C of Appendix H, 10 CFR 50.

III. Safety Evaluation Justifying Change The Babcock and Wilcox Owners Group Materials Committee Report BAW 1543, Revision 2 and 2A, " Integrated Reactor Vessel Material Surveillance Program" has been accepted by the NRC as nreting the criteria of Appendix H of 10 CFR 50. The Surveillance Program described in the TMI-1 Technical Specifications is no longer valid and is being deleted.

TMI-1 will maintain the program described in BAW 1543 therefore continuing to fulfill the requirements of Appendix H.

IV. No Significant Hazards Consideration The proposed change is administrative. The Reactor Vessel Material Surveillance Program will now be set forth in an NRC approved B&W topical report instead of the TMI-1 Technical Specifications. Any changes to the topical report would be required to be approved by the NRC. GPUN would then request approval for TMI-1. This proposal will not involve significant hazards issues for the following reasons:

1. it does not affect plant design or operation, and therefore would not involve a significant increase in the probability or consequences of an accident previously evaluated;

2. it does not involve modification to existing plant equipment, and therefore would not create the possibility of a new or different kind of accident from any accident previously evaluated;

3. it does not involve changes which would affect the safety analysis of the plant, and therefore would not involve a significant reduction in a margin of safety.

P

r L.._.-

The proposed amendment is in the same category as Example (1) of amendments that are not likely to involve significant hazards consideration (48 FR 14870) in that this is an acministrative change to how the Reactor Vessel Material Surveillance Program is maintained.

V. Implementatton 'I It'is requested that this amendment become effective upon issuance.

VI. Amendment Fee (10 CFR 170.21) /

. The fee for this amendment was provided with TSCR No. 104, Rev. O, submitted June 8, 1981.

~

e i

L, ' ) '

..I k., l (

i i

e i

s t 1 .

e

?

1 i

(  %

s rq.

4

, . , - , .--- - - , , - , , _ , . ,. - ,. _r ..-, , . - , , - - - , - , -

,,---,y,---, -,

4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION Applicability This technical specification appl %s to the inservice inspection of the reactor coolant system pressure boundary and portions of other safety oriented system pressure boundaries.

Objective lhe objective of this inservice inspection program is to provide assurance of the continuing integrity of the reactor coolant system whil( at the same time minimizing radiation exposure to personnel in the performance of inservice ,

inspections.

Specification 4.2.1 Inservice Inspection of ASME Code Class 1 Class 2, and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC.

4.2.2 Inservice Testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC.

4.2.3 (Deleted) l 4.2.4 The accessible portions of one reactor coolant pump motor flywheel assembly will be ultrasonically inspected within 3-1/3 years, two within 6-2/3 years, and all four by the end of the 10 year inspection interval. However, the U.T. procedure is developmental and will be used only to the extent that it is shown to be meaningful. The extent of coverage will be limited to those areas of the flywheel which are accessible without motor disassembly, i.e., can be reached through the access ports. Also, if radiation levels at the lower access ports are prohibitive, only the upper access ports will be used.

Amendment No. 15, 29, 54, 60, 71 4-11

% . .o, , ,

.4.2.5 (Deleted) 4.2.6 (Deleted) 4.2.7 A surveillance program for the pressure isolation valves between the primary coolant system and the low pressure injection system shall be as follows:

1. Periodic leakage testing (a) at test differential pressure greater than 150 psid shall be accomplished for the valves listed in Table 3.1.6.1 for the following conditions:

(a) prior to achieving hot shutdown after returning the valve to service following maintenance repair or replacement work, and (b) prior to achieving hot shutdown following a cold shutdown of greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> duration unless testing has been performed within the previous 9 months.

2. Whenever integrity of a pressure isolation valve listed in Table 3.1.6.1 cannot be demonstrated, the integrity of the other remaining valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other valve located in the high pressure piping shall be .'

recorded daily.

I (a)

To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

Amendment No. 29, 54, 60, Order dtd. 4/20/81, 71, Corr. Ltr. dtd. 11/2/81.

4-12 (Page 4-13 Deleted)

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