ML20207D040
| ML20207D040 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/03/1988 |
| From: | Bailey J GEORGIA POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| GN-1477, NUDOCS 8808150226 | |
| Download: ML20207D040 (9) | |
Text
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Georgia Pbwer Company Fbst Offico Box 2G2 Waynesboro, Georgia 30830 Idephono 404 554 9961 404 724 8114 Southern Company Services. Inc.
Post Office Box 2625 Birmingham, Alabama 35202 Tei pnone 20s am "
Vogtie Project August 3, 1988 U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk File: X7BC35 Washington, l' C. 20555 Log: GN-1477 PLANT V0GTLE - UNIT 2 NRC DOCKET NUMBER 50-425 CONSTRUCTION PERMIT NUMBER CPPR-109 RESPONSE TO SPENT FUEL RACK QUESTIONS Gentlemen:
Your letter of June 24, transmitted two additional requests for information related to the Vogtle Electric Generation Plant - Unit 2 spent fuel racks and fuel handling accidents.
The responses to these questions are attached.
We are also including additional information in response to questions that were asked by members of your staff during a telephone call on July 19, 1988.
If you have any questions concerning these responses, please do not hesitate to contact me.
, Sincerely,,
4 J. A. Bailey Project Licensing Manager HM/sem Enclosure xc: NRC Regional Administrator NRC Resident Inspector P. D. Rice J. P. Kane R. A. Thomas B. W. Churchill, Esquire J. B. Hopkins (2)
G. Bockhold, Jr.
J. E. Joiner, Esquire R. J. Goddard Esquire R. W. McManus i
Vogtle Project File OI 8808150226 880803 1 (j DR ADOCK 050 S
d Question 470 #6 Perfonn an analysis of the radiological consequences of postulated fuel handling accident (NUREG-0800 "Standa rd Review Plan," Section 15.7.4) and spent fuel cask drop accidents (NUREG-0800, "Standard Review Plan,"
Section 15.7.5) inside containment.
Response
Fuel handling accidents, within the containment and within the Fuel Handling Building, have been analyzed in accordance with the Standard Review Plan and the results are presented in FSAR section 15.7.4.
Regulatory Guide 1.13 requires the use of the assumption that the cladding of all of the fuel rods in one fuel bundle might be breached.
The analysis presented in section 15.7.4 assumed that 1.2 fuel assemblies would be affected by the accident.
The parameters used in evaluating the radiological consequences of a fuel handling accident are presented in FSAR table 15.7.4-1, a copy of which is attached for your convenience.
The design of the spent fuel racks does not alter the assumptions or conclusions of the analysis presented in section 15.7.4 of the FSAR.
The spent fuel cask will not be handled inside the containment.
Cask handling in the Fuel Handling Building is limited by interlocks such that it cannot be moved over the spent fuel pool.
A type 1 single-failure-proof crane designed according to NUREG-0554 and branch technical position APCSB 9-1 is used in handling the spent fuel cask.
Therefore no cask drop will occur, and thus no radioactivity will be released.
The design of the spent fuel racks for the Unit 2 spent fuel pool does not affect these conclusions.
The allowed spent fuel cask movement, relative to the Unit I and Unit 2 spent fuel pools, is illustrated in the FSAR Figure 9.1.5-1, a copy of which is attached for your convenience, i
l
i Question 470 #7 If changed from previous assessment provide the number of spent fuel assemblies that 'could be damaged by dropping each typical load carried over the pool.
Response
The design of the spent fuel racks to be provided for the Unit 2 spent fuel pool -is _ similar to the design of the spent fuel racks that were licensed for the Unit I spent fuel pool, in terms of dimensions and materials.
The typical loads that are carried over the Unit 2 spent fuel pool are the same as for the Unit 1 spent fuel pool.
In order to demonstrate that the effects of dropping objects onto the Unit 2 spent fuel racks are consistent with the previous assessment, the Unit 2 spent fuel racks are being analyzed to demonstrate that the impact of the combined weight of a fuel assembly, control rod assembly, and handling tool will be within acceptable limits.
This analysis conservatively assumes a weight of 2300 lbs. for the dropped load.
As stated in our letter of July 21, 1988, the results of this analysis will be submitted to the NRC by about August 12, 1988.
s NRC Staff question from telephone call of July 19, 1988 Describe the load path used to bring the spent fuel storage racks to the j
Unit 2 spent fuel pool.
j J
Response
Two cranes will be used to install the Unit 2 spent fuel storage racks.
The spent fuel cask handling crane moves the racks from the auxiliary building train bay into the pool area and places them on top of the cask loading pit.
This crane is single failure proof and the rack travel path is within the existing heavy loads path envelope as described in FSAR l
Section 9.1.5.
The racks are then moved west and lowered in the Unit 2 spent fuel pool with a temporary construction crane.
This crane is prevented from moving its hook with 15 feet of the Unit 1 spent fuel storage pool by rail clamps and is procedurally prohibited from moving over the new fuel storage pit (Figure 1).
It does not pass over any safe shutdown equipment in the fuel handling building.
The tunnels under the fuel handling building, which contain unit 2 safe shutdown components, are protected by the 6 foot thick building basemat.
I
..g NfiC Staff quescion from telephone call of July 19, 1988 Provide a brief summary description o.
the thermal hydraulic evaluations perfonned for heat transfer within the spent fuel racks.
_ Response The evaluation of local temperature effects using Holtec's QA validated three dimensional natural convection code THERP00L* determines the maximum pool local temperature and maximum cladding temperature.
The heat generat$on calculation is based upon NUREG 75/087 (NRM BTP APCSB 9-2).
The mass and momentum equations are solved simultaneously in an integral
- form to obtain radial (lower plenum) and axial (in-channel) velocity fields.
The computed velocity fields are then used to obtain the fuel assembly cladding temperature distribution.
The THERP00L analysis includes the effects of void generation.
Also accounted for in the calculations are the effects of voids (if formed) on the conservation equation, a crud layer on the cladding, and the gamma heating effect on the flux trap temperature.
l 1
- THERP00L has been used to qualify spent fuel racks for Fermi 2. Quad Cities 1 and 2, Oyster Creek, V.
C.
Summer, Rancho Seco, Grand Gulf 1, and Diablo Canyon 1 and 2.
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VEGP-FSAR-15 TABLE 15.7 4-1 (SHEET 1 OF 2)
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT In Fuel Building In Containment Source Data 3S(or #
Core power level ihE&&
3565 (MWt)
Radial peaking factor 1.65 1.65 Decay time (h) 100 100 Number of fuel assem-1.2 1.2 blies affected Fraction of fission RG 1.25 RG 1.25 product gases con-tained in the gap region of the fuel assenbly Atmospheric Dispersion Table 15A-2 Table 15A-2 Factors Activity Release Data Percent of affected 100 100 fuel assemblies gap activity released Pool decontamination factors Iodine 100 100 Noble gas 1
1 Filter efficiency (%)
No credit 0
l30 Building mixing vol-0 25 umes assumed (%
total volume)
HVAC exhaust rate 5000 15,000 (ft*/ min)
O T p o g r o.p {O C o.\\ e.b to r W kith dd\\
he.
Amend. 30 12/86 kw+ wtR AMtd MO\\i Cottecied ')A A
i VEGP-FSAR-15 TABLE 15.7.4-1 (SHEET 2 OF 2)
In Fuel Buildi_n_g V
In Containment,
j Building isolation No isolation 10+5 l30 time (s)
Activity release 2
period (h)
Release termina-tad 10 s after containment isolation signal with 5 s allowed for signal generation Activity released to the environment Isotope O to 2 h (Ci)
O to 2 h (Ci)
I-131 4.4E+1 4.4E+0 I-132 3.7E+1 3.7E+0 c-I-133 4.5E+0 4.5E-1 Xe-131m 6.5E+2 3.9E+0 Xe-133m 1.2E+4 7.2E+1 Xe-133 1.5E+5 9.0E+2 Xe-135 2.6E+2 1.5E+0 Kr-85 2.8E+3 1.7E+1 Amend. 30 12/86
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