ML20205R698
| ML20205R698 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/31/1987 |
| From: | Selman M CONSOLIDATED EDISON CO. OF NEW YORK, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| IEB-84-03, IEB-84-3, NUDOCS 8704060417 | |
| Download: ML20205R698 (5) | |
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Indian Point Unit ',N'o.
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Docket No. 50-247*
s U.S. Nuclear Regulat'ry Commission o
ATTN: Document ~ Control Desk l
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Gentlemen:
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This lettdr responds to a request for additional information concerning our{
response to IE Bulletin No. 84-03 " Refueling Cavity Water Seal." ( This matter is considered unresolved as contained in NRC Inspeccion Report 50-247/86-27.
The resubmittal requested is included in Attachrtent A to
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Please contact us if you have further questions.
i.s Very truly yours, s
M 20.190.3.17.1 cc Dr. Thomas E. Murley
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Pagionap Administrator - Region I.
U.S. Nuclear Regulatory Commission-f31 Park Avenue 1,
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King of Prussia, PA 19406 Senior Resident Inspector U.S. Nuclear Regulatory Commission s
P.O. Box 38 Buchanan, NY 10511
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-s Attachment A Supplementary Response to I.E. Bulletin No. 84-03 Refueling Cavity Water Seal Consolidated Edison Company of New York, Inc.
Indian Point Unit No. 2 Docket No. 50-247 March 1987
s N Supplementary Response to I.E.Bulletin 84-03 Page 1 Refueling Cavity Water Seal In the Indian Point Unit No. 2 (IP-2) response to I.E. Bulletin No. 84-03 dt.ted November 30, 1984, it was concluded that the IP-2 Presray Seal installation will not result in a gross leak.
The following, further supports that a failure of the IP-2 cavity seal is not credible.
As discussed in the initial response, the refueling cavity water seal used by IP-2 is different from the seal used at Haddam Neck.
In addition to the dimensional differences in the annulus at Haddam Neck and IP-2, the seals themselves differ in size. The seals used at IP-2 are 4 inches wide across the top wedge portion, as opposed to 3 in:hes at Haddam Neck.
Both of these seals are used to secure a 2 inch area.
Therefore, the seal size will greatly reduce the possibility of the IP-2 seal pulling through the annulus.
The material difference between the seals also increases the margin of safety at IP-2.
The IP-2 seal is 60 durometer while the seal used at Haddam Neck was 40 durometer.
The elastomer-hardness is specified on Presray Seal purchaso orders at IP-2 to preclude the inadvertent receipt of a softer elastomer. Nuclear Power Quality Assurance at IP-2 also confirms the elastomer hardness to be 60 durometer. While the softer rubber used at Haddam Neck allowed the seal to extrude through the opening, the hardness of the IP-2 seal impedes it from bowing, bending, and being pulled through the 2 inch annulus opening.
To demonstrate this a bench test' of a 1k" piece of the 60 durometer seal was conducted at IP-2 on April 1, 1985., A 365 lb load was applied concentrated on a line at midspan of the seal piece. Negligable deflection was observed. The normal hydrostatic load on the surface area of the lh" piece was calculated to be 29.3 lb.
Thus it was shown that hydrostatic pressure greater than expected will not extrude the seal into the annulus.
A second test was conducted during our initial review and acceptance of the seal design.
The seal was deflated with a full head of water in the refueling cavity, and no leakage was observed. This is because the wedging effect of the seal created a secondary seal.
The results of the evalua-tions conducted to date demonstrate the acceptability of the current seal design acting as a passive device.
The analyses have also shown the seal to be a passive device capable of maintaining its integrity in a deflated condition. An inflated seal is expected to afford additional resistance to the hydrostatic forces present during refueling.
It was concluded that in view of these tests a seal failure resulting in a catastrophic draining of the refueling cavity is not a credible sequence.
Nonetheless, other possible refueling cavity drain paths were also evaluated.
A two-step process was used for this evaluation.
The first step determined the time it would take the operators to place the ' maximum number of elevated fuel assemblies into safe storage.
Safe storage exists
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when elevated fuel assemblies in transit are placed in the spent fuel pool
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s Supplementary Response to I.E.Bulletin 84-03 Page 2 Refueling Cavity Water Seal allowing only one fuel assembly in the lowered containment upender.
Although the estimated time for relocating the elevated fuel assemblies is 1.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, a conservative assumption of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> was used for bounding possible drain paths to eliminate those which result in slow enough drainage to allow considerable time for the safe relocation of transient fuel assemblies.
The worst condition described below, resulting from the loss of shield water height due to a leak, can be administratively controlled by existing procedures as described on the last page, if the rate of the loss of water is less than 1000 gpm.
Then, the elapsed time for the water height to drop 10' is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The second step is to identi-fy all leakage paths not eliminated by the first step.
Examination of other postulated leak paths at IP-2 reveals the following results.
4" & 6" Pipe Breaks The reactor vessel refueling cavity contains 300,000 gallons at a water height of approximately 25'.
Therefore, a 4" pipe break posi-tioned at or near the base of the refueling cavity would initially drain the cavity at a rate of 300 gpm or an elapsed time of 40 min-utes/ foot drop in water level.
Similarly, a like 6" pipe break would drain the cavity at a rate of 900 gpm or an elapsed time of 13+
minutes / foot drop in water level.
This analysis of the 6" pipe break bounds all credible piping associ-ated leak paths as describcd above.
These postulated pipe breaks can be administratively controlled by existing procedures.
Primary Coolant Pipe Nozzle Access Covers Analysis of the primary coolant pipe nozzle access covers spaced around the Presray Seal was performed.
In this two part study, the analysis made of the 8 cover plates showed that the bending stresses due to the hydrostatic head of water were a nominal value about 7885 psi.
Next, these plates were analyzed for the stresses that would result from the impact of a dropped fuel element directly on the center of these large plates. With the drag force of the water, the deformation of the plate and the fuel assembly accounted for, the calculations showed that the plate will not fail and is not a credible leak path.
Fuel Transfer Canal The fuel transfer canal structure, located between the fuel transfer canal in the reactor vessel refueling pool and the fuel storage pool
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Suppplementary Response to I.E.Bulletin 84-03 Page 3 Refueling Cavity Water Seal in. - the fuel storage building,.is equipped with a metal expansion I~
joint.
In ~ the event the IP-2 fuel transfer tube expansion joint experienced gross failure, the limiting flow protection restriction is the outer sleeve. slip, joints. ' Calculated maximum leakage thru this~
slip joint, with gross failure of the expansion. joint, is 400'gpm. A loss of shield water. of this magnitude can be administratively.
controlled by existing procedures.
Furthermore,. closure of the spent fuel pool isolation gate will prevent drainage of the spent fuel-pool.
Further isolation is accomplished by ' closing the ' transfer tube gate-valve. --
As a result of the aforementioned evaluations of. potential leak paths,-it was determined that all are administrative 1y. controllable with leak ' retes-under 1000 gpm..The administrative; controls for IP-2 are outlined _in the 4
existing ' procedure, Abnormal Operating Instruction 17.0.3,
. entitled
" Undesirable Decrease in Refueling Cavity Water Level".
This -_ procedure considers a number of leakage paths including the annular vessel _. seal.
Actions which are available pursuant to this procedure-include' the following j
Containment Evacuation'.
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Fuel assembly placement within containment and/or spent fuel pit.
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Isolation of the spent fuel pit.
1 Attempt to re-establish the annular seal.
j Supplying makeup water to the refueling. cavity via RHR or-Recirculation pumps.
Each of these pumps has a nominal. capacity of j
3,000 gpm.
Containment isolation -(Refueling Integrity) and fuel. storage building ventilation requirements need_not ba established as they are required to be I
in effect'and thus would already be;in effect for all refueling operations-(refer to IP-2 Technical-Specification 3.8).
Based on the;-above foregoing analyses and the previous. submittal, we have concluded that additional actions are unnecessary.
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