ML20085H375
| ML20085H375 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 07/08/1969 |
| From: | Neidig R SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
|
| References | |
| FOIA-91-17 NUDOCS 9110280256 | |
| Download: ML20085H375 (5) | |
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SAXIO:s INCLi'J.R EXPERIlif! ITAL COniOnt.TIO!!
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DOCRET !!O. 50 146 LICLl SL 11PR-4 Ainendnzent 1;o. I to Change Eeport No. 15 3.
On !!avetaber 13,196P, hpp3 f cont submitted Change Report lio. 15 deretJbine, chenges being tr.ade to the prescurizer cafety relici valve inctn31ation.
The ocction. " Safety Considerationc", has been revired and thf u revision is being subtnitted as Amendment 1;o. I to Change Report !?o. 15.
SAXTON INClEAR EXPERIMD;TAL CORPORATION l
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H. E. lieldJc R. E. !!cidig, President l
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t July 8, 1969 FDj g){g6910424 DEMOR91-t7 PDR
, Docket lio. 50-146 DPh-4 Channe Report No. 15 Ainendment No. 1 4
4 CHANGE REPORT TOR Tilt INSIALLATION OF NEW PRESSURIZER SATETY VALVES MiD WATl:R SEAL e
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,Dochet !?o. 50-146 DPg-4 Change geport No. IS i
Page 1 of 3 Pagen Amendaant No. 1
_Descru tion of Channe The pressurir.er safety valves PSV-372 and PSV-373 have been flanged and reinst'alled with a water seal.
Purther, at the first opportunity the valves will be replaced with the neu safety valves, a drawing of which is shown in I
Picure 1.
Some of the valve operating parameters are given in Tabic 1, and a list of valve material specifications in Tabic 2.
purpose of Chtmte The present safety valves __have leahed frequently resulting in severe crosion of the seating surfaces and considerable delays during plant startups.
To eliminate these delays, a vator seal has been installed between the top of the pressuriret and-the safety-relief valves.
The unter seal piping will-accumulate condensate to form, a water seal which will keep steam and gas out of direct contact with the valve seats.
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gfy_t,v_ ConrJ d e ra t i nr 5.
The design, fabricat. ion and crection of the water seal piping was made in accordancc with USAS b31.1 - 1967 Edition.
The piping f. rom the presourizei to the safety relief valves is designed for 2500 psia and 650 P.
The safety valves satisfy the requirements of Article 9. Section Ill, ASMC Boiler and Pressurc Vessel Code.
The integrity of the water seal piping was assured by conforming to the follouing Code requirements:
1.
Wclders and weld procedures were qualified in accordance with ASME Boiler and Pressure Vessel Codo,Section IX.
2.
Reot and final pass of all wc3ds were liquid penetrant inspected and all butt welds were 100% radiographically inspected. -The
_ procedures and acceptance standards were in accordance with ASME Boiler and Prescurc Vessel Code, Section 1.
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) Docket lio. 50-146-bPR-4 Change P.eport l'o.15 s
page ? of 3 Pages Amendment !;o. 1 Paragraph 137:1 of 1131.1 requires that all piping systerns designed, fabricated and erected under the Code denonstrate Icak tightneos, which raust be inet ',y a
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hydrostatic test prior to initial operation. L'here a hydrostatic test is not practicable an initial service 3cah test, a vacuum test or 100% radiography of all velded jointr. in an all uelded system may be subr.tituted.
Paragraph 137.4.1(b) further specifico that the hydrostatic test, if performed, r;hn11 be conduct ed at a test prest,ure of 1.5 x decign pressure unicss a lesser pressure is indJcnted by Paragraph 137.4.1(a).
Paragraph 137.4.1(a) specifies that the test pressure l
1 shall tml exceed the maxinum test pressure of any vecr.cl or couponents in the piping system.
Reference to Pigure 2 shows that in order to hydrostatienlly test the water scal piping, the reactor coolant system will he subjected to'the scme hydrost atic pressure.
Therefore, the. practicebility of conducting such a test is Jimited by the maximum pressure capability of the reactor coolant system.
In dctortdning the iaaximun pretsure capability of the reactor coolant syst ern the follouing conditions were considered:
1.
Pressure retaining components in the reactor coolant system include auntenitic stainlens steel piping and fittings, and low alloy stcel.
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NDT chif t and thermal strecr. considerations for the low alloy stec3 in the reactor veosc1 limit minitaum systera temperature to 5201 for 0
approximately system design pressure.
The allouabic stress for austenitic stainless steel in both the ASME Boiler and Pressure Vessel Code and USAS B31.1 recognizes the work-hardening characteristics of these materials by permitting the allowable strens at cicvated temperature to reach 90% of the minimum 0.2% offset yield strength at the specific dcoign temperature. A footnote to Tables A-1 and A-2 of USAS B31.1 c'autions that allouable stress is 90% of yic3d strength, and for some loading conditions undesirabic plastic deformation could occur.
Section III of the ASME Boiler and pri.ssure Vessel Code provides a more quantitative warning by limiting any test prer. cure to the lesser of 1.25 x design pressure or that which produces a strecs o
V Iliothet lio. $0-146 ITU-4 Change Report 1;o. 15 t
page 1 of 3 parcs Amendment No. I equal to 90% of the staterial ylold strength at the test t ettpe ra t ure.
Therefore, I
the limiting pressure for the reactor coolant system at a temperature of $20 T 0
i is set by the ausenitie etainicus steel piping and fittings in the systcu.
I The rcactor coolant system piping (centrifuga11y cast) and the reactor coolant system fittings (static cast) were designed, fabricated and crected in accordance with LIGAS E31.1 - 1955 Edition, and nucicar piping Cases 11-9 and 11-10 i
renpectively.
The material property data curves are the same for both components; j
therefore, the al}oable stresses as set forth in Cases li-9 and I;-10 are identical, j
The maxiicum allowable stresces for the system design temperature and the Icak test temperature arer i
for a design temperature of 650 r, Sa = 15,000 psi For a 1cah test temperature of $20 r, S, = 15,720 psi Thus, the onximun leak test pressure to which the reactor coolant system nay be j
subjected to btay withia the 90% yield stungth Ilmit int x '2485 2604 psig a
73 where 2485 psig is reactor coolant system design pressure.
'In vicu of the gun 11ty control exercised during fabrication and crection. and the use of radiographic and 11guld ponctrant examinations to verify the structural adequacy, the use of 2604 psig is an uncessarily high pressure to demonstrate Icak tJghtness of the water seal piping.
Therefore, the piping will be leah tested at a pressure of 2575 psia.
This test will be conducted in conjunction uith the pressure test to be conducted on the recirculation and safety injection system piping.
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