ML20024G210
| ML20024G210 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 09/27/1977 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Stello V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9102080333 | |
| Download: ML20024G210 (3) | |
Text
.
o.
r h? "Q~U M')
01
~~~p4 e
[j j g r 9,.[,, T[
~,'
ct A v.
(j
- +- :
-w N -3 R T H E R N STATES POWER COMPANY MlN N E A PO U S. M LbsOTA 5 5401
' phy '$
D.8e 737 September 27, 1977 IO Tsir7 i
/s Mr Victor Stello, Director Q'
D Division of Operating Reactors JMp I
c/o Distribution Services Branch, DDC, M
U S Nuclear Regulatory Comission Washington, DC 20555
Dear Mr Stello:
MONTICELLO NUCLEAR GENERATING PIANT Docket No. 50-263 License No. DPR-22 Supplement No. I to License Amendment Request Dated September 19, 1977 4
On September 19, 1977 we submitted a License Amendment Request to revise Table 4.6.1 of the Appendix A Technical Specifications, "In-Service Inspection Requirements for Monticello." One of the regnested changes would substitute an examination during I
system pressure tests for the volumetric examination now specified for control rod housing pressure boundary welds. This change is consistent with Section XI of the ASE Code (1974 Edition through Sumer 1975 Addenda).
5
't I
We have been requested to supply additional information to justify our de::ermina-e
]
tion that these welds are exempt frce volumetric and surface examination by Article IWB-1220(b)(1) of Section XI of the Code. Ihe purpose of this Supplement is to provide that information.
Attached you will find a drawing of the weld in qttestion and a sumary of the
,l analysis showing that under postulated conditions of loss of coolant from the j
ccuponent during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner assuming makeup is provided using normal systems supplied by onsite power.
Yours very truly,
\\ g '4M s= o A i'
L 0 Mayer, PE N
Manager of Nuclear Support Services IDM/DMM/ak L
cc: J G Keppler G marnoff MPCA - Attn: J W Ferman l
MECCA - Attn: R J Hatling S J Cadler 910200 33 770927 l
PDR ADOCK 0S000263 l' h P
rn.\\
J Attachment to letter dated September 27, 1977 Page 1 of 2 L 0 Mayer, NSP, to Victor Stello, USNRC Control Rod Housing Pressure Boundary Weld - IWB-1220(b)(1) Analysis Summary Refer to Figure (1).
Failure of control rod housing to internal nozzle veld results in maximum break area of:
D1 = Maximum internal nozzle diameter-6.000 inch
=
D2 = Minimum CRD housing outside diameter 5 985 inch
=
d Am =](D1 + D )(D1 - D ) = 0.14 in2 2
2 4
The CRD support structure (Reference 1) prevents the housing from dropping far enough to clear the vessel penetration.
The maximum two-phase vessel blovdown through this break area (Reference 2, Figure 2-h) assumir.g saturated liquid at 1025 psia is 7.8 lb /800*
m f
g The capacity of the RCIC system is 400 gallons / minute of water (Reference 3).
Assuming a water temperature of 1h00F, this.;orresponds to a makeup rate of 54 lb /sec. The coolant loss through a failure of the control m
i rod housing veld is therefore well within the capacity of the system.
I I
I
\\
l t
Re fe rences:
- 1. Van Zylstra, E. H., " Analysis of Potential Control Rod Drive System Failures Resulting in Cor. trol Rod Withdrawal," General Electric Topical Report GECR-5089, Furch,1%6.
h
- 2. Moody, F.
J., "Fhximum Two-Phase Vessel Blowdown f rom Pipes," General Electric Topical Report APED-4827, April 20,1965 l
- 3. Monticello Final Safety Analysis Report, Tab 3e 10-2-3, page 10-2.13
6
^T r
V
,,,,b,
Page 2 of 2 t-Category
- E-2 Weld
. s_ b k_7 Category *
>f k
j E-1 Weld-4 l
y x
x Thermal Sleeve s
k' N
\\
f
?
\\
Internal Nozzle s
b' '
.,. s N/
/
'; H s/
s.
8/
t
/
^
i _
'l
/
s Q /
j.'
/
/
/
/
s N
/
j
/
l d Drive
,./
,, - y s.
x l
N '/
s
/
/
}
/ TQ
) /
e'
/
N
\\
e l l/
,/ ';
r
/
,e s
N X'
/ N s f
/
// x N /
yl -
/
y N
l
,/
N j
v s
s g
-)
x
{
s Contr:1 Rod s
s Drite Housing Reactor Vessel s
s
,f,,
j Bot'.om Head s
i s
e s
t 1
}
N s
N N
N l
\\
l N
3 J.
[
- Befer to Table 4.6.1, "In-Service Inspection Requirements for Monticello,"
of the Appendix A Technical Specifications for the Monticello Naclear Generating Plant.
f
' EC PPAM 195
( T U 3 NucLEAA REGULATONY ccup ON oopuiNeosa (2-78) -
d
@ 8 "" d M
)
^
- NRC DISTRIBUTWN pon PART 50 DOCKET MATERI AL y
_TJ 'Mk, fictor Stello FROM:NSP OATE 0" DOCUMENT Minneapolis, Minnesota 55401 f
,,'EsTTan ONoTORIZED P R oP INPUT PORM NUMBER oF COPIES RECEIVED blOINAL KVNCLASSIFIE D acon go c e,myg ossC'CTION ENCLOSU RE Consists of additional information to 09/19/77 submittal of License Amendment Request to revise Table 4.6.1 of Appendix Tech Specs, "In-Service Inspection Requirements for Monticelle concerning control rod housing pressure boundary welds...
1p 2p i
PLANT NAMEDNTICELLO jem 09/30/77 Yb CY3
$UCL 22L SAFETY FOR ACTION /INFORMATION BRANCH CHIEF: (7)
.) Dn3 'tS N_
I AN INTERNAL DISTR!BUTION l
l M FIBE T l
l I
I i
I C E (2) l l
I I
t OELD HANAUER l
l CHECK l
l l
l STELLO i
EISENHUT l
l l
SHA0 l
l BAER l
l BUTLER l
l GRIMES
{
l J. COLLINS
{
l l
1 I
I I
I I
I I
I I
l l
EXTERNAL DISTRIB' TION l
QOf9 TROL NUMBER l
J l __ l LPDR:/EkWcA sa L ali ! /M s NN.
I I I
l l16 CYS ACRS SENT CATEbOIlY M l l l
. M l
C I i l
m 4l l l l l l 1
n s
o a