ML20024G797
| ML20024G797 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/27/1973 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Oleary J US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9104300424 | |
| Download: ML20024G797 (3) | |
Text
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- yyy, NORTHERN STATES POWER COMPANY M I N N E A PO 4.l S. M I N N E S OT A 95409 h)
July 27, 1973 S$"
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' l\\ g, Mr. J F O' Leary, Director ttG M ORT
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p' Directorate of Licensing WA g
United States Atomic Energy Consnission
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Dear Mr. O' Leary:
HONTICELLO NUCLFAR GENERATING PLANT Reyktory 30 Cy.
Docket No. 50-263 License No. DPR-22 Supplementary information on Instrument Line Break Analysis A summary of an additional investigation of the instrument line break acci-dent performed for NSP is reported herein for your information. This analysis vad conducted by an independent architect /engineet snd included a determination of the engineering practicability, as well as an independent evaluation of the dose consequences, of installing flow restricting orifices.
This independent analysis adds support to the previous analysis submitted in the Monticello Full Term Operating License application.
Both analyses assumed a complete break of an instrument line between the drywell i onetration and the isolation valve.
Radiological assumptions of the two anal?. 2.8 were compatible and generally followed the reconsnendations of Safety Guides 3 and 11.
Both analyses concluded that secondary containment integrity was maintained, and that doses were substantially below the guidelines of 10CFR100.
Dose calcu-lations of the two studies are stmanarired as follows:
Whole Body
'Ihyroid w/ori fice w/o orifice w/ orifice w/o orifice FTOL Analysis 7.5 mR 13.6 R Ind. Analysis 0.03 mR 2 mR 0.034 R 23.8 R Ind. Analysis
- 0.09 mR 6 mR 0.102 R 71.4 R
- Using Technical Specification iodine concentration in the primary coolant.
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.ERN CTATEO POWER CO 8ANY 2-These values were calculated on the basis of 300,000 uCi/sec off-gas rate and FSAR design primary coolant iodine concentration.
The independent analysis also took into consideration dose consequences using maxLmum (per Technical Specification) iodine concentration in the primary coolant.
These results are indicated by an asterisk in the above table.
The independent analysis showed that it was technically feasible to install flow restricing orifices, however with a quertionabic improvement in overall safety.
The present system haa been tested and operated successfully. The installation of orifices would require 130 new welds in the 65 affected lines, wouad entail additional rad:.ation exposure to plant personnel and some risk of spres' of contumination in the plant.
An examination of the plant physical arrangement has shown tha*. there is virtually no poteatial for physical damage to the exposed sections of 1-inch instrument lines oatside of primary containment.
The sections are screened from rotating machinery by shield walls and thus, are not subject to potential missile damage. No foreseen maintenance operations in the areas could detrimentally affect the line sections.
The Type 304 Stainless steel, used in the 1.nch lines, is not subject to brittle fractures and the stress icvels are generally small as a consequence of the line sizes.
The exposed sections of instrument lines outside of primary containment are nominally 24 inches long and extend from the containment penetration through the excess flow check valve and gate valve up to the 3/8 inch tubing leading to the instrument racks. A break in the 3/8 inch stainless steel tubing with concurrent failure to close of the excess flow check valve would result in the flow characteristics similar to that through the proposed orifices.
It is concluded on the basis of the two cnalyses performed that no backfit is necessary ia regard to potential instnnnent line breaks because 1) there is minimal risk of breakage; 2) dose consequences are moderate even under conservative assur,tions (substantially below the guidelines of 10CFR100);
and 3) the integrt y and functional performance of secondary containment can be maintained.
Yours very truly, h.
L 0 Mayer, PE Director of Nuclear Support Services LOM/DWJ/br cc:
B 11 arier G Charnoff Minnesota Pollution Control Agency Attn. K Dzugan I
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AEC DST IEUT70N FOR PART 50 DOCKET MAT (AL (TEMPORARY FORM)
CONTROL NO: 5797 FILE:
s FROM:
DATE OF DOC DATE REC'D LTR MEMO RPT OTHER Northern States Power Company Minns apolis, Minnesota 55401 7-27-73 7-30-73 x
L. O..lia v e r TO:
.l. F. O' Leary SENT LOCAL PDR E 1 signed 39 CLASS UNCLASS PROP INFO INPUT NO CYS REC'D DOCKET NO:
XXX 40 50-263 DESCRIPTION:
ENCLOSURES:
Ltr furnishing addl info regarding analysis j
of postulated pipe failure outside containment..
j A, CKNOWLEDGED PLANT NAME:
Monticello FOR ACTION /INFOR".ATICN 7-30 73 fod BUTLER (L)
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