ML19263C059
| ML19263C059 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 01/24/1979 |
| From: | Naughton W COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7902020234 | |
| Download: ML19263C059 (27) | |
Text
Commonwealth Edison One First National Plaza. Chcago, litenois Address Reply to: Post Oihce Box 767 Chicago Ilhnois 60690 January 24, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Zion Station Units 1 and 2 Additional Information on Proposed Expansion of Spent Fuel Storage Capacity NRC Docket Nos. 50-295 and 50-304
Dear Mr. Denton:
The NRC Staf f requested Commonwealth Edison Company to provide additional information in support of its request to expand the storage capacity of the Zion Units 1 end 2 spent fuel pool.
The request consisted of two sets of questions telecopied from the Staff on November 14 and 28, 1978.
Attachments 1 and 2 to this letter contain Commonwealth Edison's responses to these questions.
Please address any additional questions that you might have to this office.
One (1) signed original and thirty.1ine (39) copies of this letter are provided for your use.
Very truly yours, William F.
Naughton Nuclear Licensing Administrator Pressurized Water Reactors attachments 7902020 A37
ATTACHMENT 1 SPENT FUEL POOL CAPACITY EXPANSION ZION NUCLEAR POWER PLANTp, UNITS 1 AND 2 DOCKET NOS.-50-295 AND 50-304 ROUND 2 OUESTIONS 1.
QUESTION s
In regard to your response nwnber 20, a limit on the fuel assembly loading is more inclusive than a limit on the enrichment.
Also, this maximum fuel loading can be obtained from just an arithmetical calculation of quality assurance data.
Por these reasons, we find that a technical specification on these racks wh.ch limits the fuel loading to 39.4 grams of Uranium-235, or equivalent, per axial centimeter of fuel assembly is an acceptable method of limiting the uncertainty in keff whereas a limit on the enrichment is not.
ANSWER The appropriate technical specification change iu being drafted and will be submitted af ter approval by our On-Site and Off-Site Review functions.
1.1
4
,NRC Docket Nos. 50-295/304 ATTACFDEIPT 1 2.
QUESTION In regard to your response number 21, state the bases for the dimensions of the cylindrical supercell (Figure 3) for the first benchma'k calculation.
ANSWER Radii Ri and R2 of the cylindrical supercell (Figure 3) is, obtained by conserving the corresponding areas of the 9 x 9 is obtained by adding the basic fuel pin assembly. Radius R3 aluminum wall thickness to the fuel pin assembly and then conserving the area. Ra'dius Ru is obtained by adding to R3 half the thickness of the boral core.
Constants used for cylindrical supercel1 dimensions:
9x9 Basic fuel pin assembly array
=
0.75 inches Fuel rod pitch
=
.041 inch Thickness of AL wall
=
0.168 inch Thickness of boral plate
=
h(0.75-2x0.041-0.168)
Thickness of water layer
=
0.25 inch
=
Area of fuel region:
(fuel rod pitch x 9)2
~
=
(0.75 x 9)2 sq. in.
=
(6.75)2sq.in.
=
(6.75 x 2.54)2 sq. cm.
=
293.95102 sq. cm.
=
nR 2
=
1 (293.95102)h R1
=
cm.
9.67303 cm.
=
2.1
~
NRC Docket Nos. 50-295/304 ATTACHMENT 1 2.
ANSWER'(continued)
Area of (H O + fuel) region 2
HR 2
=
2 (6.75 + thickness of water layer)2 sq. in.
=
(6.75 + 0.25)2 sq. in.
=
(7 x 2.54)2 sq. cm.
=
316.12f;4 sq. cm.
=
(
4)
R 2
10.03129 cm.
=
Area of (H O + At + fuel) region 2
(7 + thickness of AL layer)2 sq. in.
=
(7.041 x 2.54)2 sq. cm.
=
319.84246 sq. cm.
=
nR 2
=
3 319.84246 g3 cm.
10.09005 cm.
=
R3+
thickness of boral plate Ru
=
10.09005 + 4 x 0.163 x 2.54
=
10.30341 cm.
=
4 2.2
NRC Docket Nos. 50-295/304 ATTACHMEtrf 1 4
Al
.O BORAL 4
b
/y N
Fm k
Rg = 9.67303 CM R.g = 10.03129 "CM R = 10.09005 CM 3
R = 10.3 m 1 m 4
Figure 3.
1-0 Supercell Configuration for First Banc!= ark CalculaMans s.
2.3
NRC Docket Nos. 50-295/304 ATTACEIMENT 1 3.
QUESTION In your calculations of the two critical assemblies with Boral, describe how you accounted for the self-shielding of the boron carbide particles in the aluminum matrix.
ANSWER In the benchmark calculations, the Boral core was homogenized and then the cross sections were obtained from XSDRN, which is a one dimensional discrete ordinates spectral averaging code.
There was no account for particle self-shielding since the range of particle sire is60-200 mesh with a mean sire of 175 mesh and, therefore, self-shielding effects are neglig ible,
3.1
NRC Docket Nos. 50-235/304 ATTACHMENT 1 4.
QUESTION In regard to your response Number 23, the NRC requires an on-site neutron attenuation test to verify' the presence of the boron.
This is in addition to the Quality Assurance Program you described.
Provide a description of the neutron attenuation test that you will perform at the Zion plant to statistically show with 95 percent confidence that the boron is not missing from more than one out of every sixteen plates.
ANSWER A neutron posion verification test will be conducted at the Zion plant after the racks are installed in the pool.
This will be a qualitative test to statistically show with 95 percent confidence that the boron is not missing from more than one out of every sixteen plates.
This procedure is similar to the poison verification tests conducted at Montecello and TVA by National Nuclear Corporation utilizing their proprietary equipment.
4.1
' NRC Docket Nos. 50-295/304 ATTACHMENT 1 5.
QUESTION For the proposed type of racks, a surveillance program is required to show the continued presence of boron throughout the life of the racks.
Provide a description of the boron surveillance program that you will perform.
ANSWER See attachment "A",
Neutron Absorber Sampling Plan - In Pool.
i t
5.1 S
NRC Docket Nos. 50-295/304 ATTACHMENT 1 ATTACHMENT "A" NEUTRON ABSORBER SAMPLING PLAN - IN P0OL A sampling plan to verify the ability of a neutron absorber material employed in the high density fuel racks to withstand the long-term environment is described.
The test conditions represent a restricted flow of water over the neutron absorber material. The samples will be supported adjacent to and suspended from the f".i racks. Eighteen (18) test samples are to be fabricated ir.
accordan.e with Figure 1 and installed in the pool when the racks are in-stalled.
The procedure for fabrication t.nd testing of samples shall be as follows:
1.
Samples shall be cut to size and dried in an oven for five hours at 175'F, followed by a cycle at 600 F for three hours.
2.
Samples shall be weighed imediately following removal from the oven and weight in milligrams recorded for each sample.
3.
Samples shall be fabricated in accordance with Figure 1 and installed in pool.
4.
Two samples shell be removed per schedule shown in Table 1.
5.
Carefully cut samples apart at the weld without damaging the neutron absorber. Wash with a soft brush in a mild abrasive and detergent solution, imerse in nitric acid to remove surface products, followed by a rinse of clean water and alcohol. Dry in a 175'F oven for five hours, followed by a cycle at 600'F for three hours.
6.
Weigh the samples and evaluate the weight change in the neutron absorber material in milligrams per square centimeter per year.
5.A-1
NRC Docket Nos. 50-295/304 7.
If pitting is present, the depth of tha four ma or pits are to be a
recorded and the average pit penetration in mils of an inch per year detennined.
8.
Retain two (2) samples.
9.
Prepare report of sample test results and observations.
5.A-2
~
NRC Docket Nos. 50-295/304 ATTACHMENT 1 TABLE 1 Date Installed INITIAL FINAL WEIGHT PIT SAMPLE WEIGHT WEIGHT CHANGE PENETRATION NO.
SCHEDULE (mg/cm2-Yr)
(mg/cm -Yr)
(mg/cm -Yr) mil /Yr 2
2 1
2 90 day v
3 4
180 day V
5 6
1 year V 7
8 5 year V 9
10 10 year v 11 12 15 year v 13 14 20 year V 15 16 30 year V 17 18 40 year y 5.A-3
NRC Docket Nos. 50-295/304 ATTACINENT 1
~
9 o
S G
!=
o R
E C
(TYP-SEAL WELD
$'7, 0.075X.125-7 FILLER 4 SIDES g
304 SST o
E 0.062" DIA.
k H0LE-TYPICAL TOP & BOTTOM ORILL BEFORE ASSEMBLY
'His!"a x
kh, g
x
.030" 304 SST 18 SAMPLES Figure 1 5.A-4
NRC Docket Nos. 50-295/304 ATTACHME17f 2 SPENT FUEL P0OL CAPACITY EXPANSION ZION NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-295 AND 50-304 ROUND 3 QUESTIONS QUESTION NUMBER 1:
Provide a more detailed description of the inter-tube welded connection; include drawings if possible. Specifically discuss if the tubes are welded continuously to each other the full length of the tube or only at discrete intervals. Also discuss the structural members or plates used in this connection.
RESPONSE
The tubes typically have bars (flat plates) attached to the specified corners as shown on Drawing No. 1000483. These bars are welded the full length of the tube.
The tubes with the bars attached are welded into cleter subassemblies per Drawing No. 1000484 Again, they are welded together the full length.
These clusters are then welded to other clusters and the base assembly as shown on Drawing No. 1000490, which is typical of the other rack size assemblies.
These cluster attachment welds are again the full length of the tube, 1.1
NRC Docket Nos. 50-295/304 ATTACHMENT 2 QUESTION NUMBER 2:
Provide a detailed description of the analysis or con-siderations used to establish that t he tube inner com-r'rtment containing the Boral remains sealed against 1 Jkage. What are the potential consequences of pool water leaking into the area containing the Boral?
RESPONSE
Consideration was given to maintaining the inner compartment containing the Boral sealed against leakage, and on the basis of the information available, it wan decided to vent the Boral containing compartment and allow pool water to enter *and exit without restriction.
The consequences of pool water in the area containing the Boral are discussed in the Brooks and Perkins' report, "The Suitability of Brooks & Perkins' Spent Fuel Storage Module for Use in PWR Storage Pool," Report No. 578 dated July 7, 1978, and did confirm the Boral panels are capable of meeting a forty year service life.
I 2.1
NRC Docket Nos. 50-295/304 LTTACHMENT 2 QUESTION NUMBER 3:
What considerations have been taken to prevent off-gas from the Boral and swelling of the tube?
RESPONSE
The off-gas from the Boral will not be a problem in a vented tube, thereby eliminating any swelling of the tube.
3.1
i
}
N.. Docket Nos. 50-295/304 i
ATTACHMENT 2 QUESTION NUMBER 4:
Provide the basis for concluding that an empty rack will slide further than a full rack under seismic loadings.
Provide a drawing of the equivalent stick model used in the sliding analysis and indi. ate the points where these loads were applied.
RESPONSE
Since the spent fuel racks are stored under water, their seismic movements are caused by the horizontal inertia of " virtual mass" which is the sum of the body mass and the " hydrodynamic nass".
The magnitude of the hydrodynamic mass depends on the shape of the rack body and the density of water, and so is independent of whether the rack is loaded or ampty. The friction force resisting the seismic movement is proportional to the buoyant weight of the rack and its contents, but because of larger horizontal " virtual mass" per unit weitnt, the ratio of inertia force to friction force is more for empty racks. For this reason; it was concluded that empty racks will slide further than a loaded rack under seismic loadings.
Figure 4.1 shows the equivalent stick rodel used in the sliding analysis. Time history of SSE seismic ovement was applied at Node 8 which represents the pool floor.
4.1
NRC Docket Nos. 50-295/304 A_TTACHMENT 2 EL. 169.8 O1 113.0
() 2 66.0 g3 36.0
( )4 Sliding Element 0.2
=
p 16.0
) 5 6.0 6
0.0 7
0 F4GURE 4.1 LUMPED MASS STICK MODEL FOR SLIDING ANALYSIS 4.2
NRC Docket N08 50-295/304 ATTACHMENT 2 QUESTION NUMBER 5:
Provide the value of the " rattling factor" used in the seismic analysis.
RESPONSE
I Rattling factors account for the nonlinear effects of the fuel bundles moving within the spent fuel rack cells. The magnitude of these factors depend on the structural and damping properties of the rack and fuel bundles as well as on the level of excitation.
The rattling factors used for the Zion rack evaluation ranged from 1.10 (for SSE loading of 10 x 11 size rack in the direction parallel to the loncer side) to 2.57 (for OBE loading of 5 x 10 size rack in the direction parallel to the shorter side). These are upper bound factors computed using a conservative assumption that all the fuel bundles inside a rack " rattle" in phase.
5.1
NRC Docket Nos. 50-295/304 ATTACHMENT ?_
QUES._JN NUMBER 6:
Provide a description of the thermal gradient cnalysis considera-tions, include the thermal gradient considered and a discussion on why this was considered a conservative estimate of the worst case, i.e., the gradient between a full and empty cell.
RESPONSE
The thermal gradient due to the placement of a hot fuel bundle in an empty rack is as shown in Figure 6.1 (0 F at the rack bottom and 32.38of at the top). Stresses caused by this thermal gradient were computed using a finite element model which is also shown in Figure 6.1.
To minimize the computation cost, only the central part of the rack body was modeled with the sides restrained from lateral translation, thus representing the worst case and predicting conservative stresses.
It is important to note here that, near the top of the rack the thermal gradient, and hence, the resulting thermal stresses are maximun, but the dead load and seismic stresses are minimum. Therral gradient near the bottom is very small where the seismic and dead load stresses are maximum.
6.1
ATTACIIMENT 2_
NRC Docket Nos. 50-295/304 NUCLEAR SERVICES CORPORATION Hot Bundle E1. 169.B in
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154.0 -
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32.38*F s
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NRC Docket Nos. 50-295/304 ATTACIIMENT 2 QUESTION NUMBER 7:
The fuel bundle drop analysis considered a drop at the most " critical" location on the rack, provide a descrip-tion of this location and drawings to illustrate the postulated configuration of the fuel bundle at impact.
Discuss the procedure for limiting the height of the fuel bundles above the racks to 24 inches.
Discuss the conse-nuences of a fuel bundle dropping straight through the tube and impacting the bottom of the rack.
RESPONSE
The top corners of the racks were found to be the most critical locations for evaluating the consenuence of dropping a fuel bundle.
When tha fuel bundle drops on the rack, the cross-sectional area of the cell walls absorbing the impact energy increases as the load is transmitted downward.
Since this gradually-increasing cross-sectional area is minimum when the fuel bundle drops on a corner, the latter ccnstituted the most crit ical location.
For evaluating the consecuences of fuel bundle drop, the bundle configuration was assumed to be vertical at impact 7.1
NRC Docket Nos. 50-295/304 ATTACHMUNT 2 (Figure 7.1).
An inclined drop was judged to be less critical f rom the following considerations:
(a) The in pact area will be larger, (b) The br. pact will be " softer" because of the flexibility of the fuel bundle itself.
The length of the fuel handling tools and interlocks on the fuel pool bridge hoist limits the distance bitween the top of the rack and fuel assembly to less than 24."
Fuel assemblies thus cannot be raised above the 24" limit.
Consequences of the fuel bundle dropping straight through the tube and impacting the bottom of the rack have been invesitgated.
The method of analysis and the results obtained are briefly described below:
The fuel bundle will drop approximately 164 inches from the top of the rack to the rack base plate.
If the fluid drag on the bundle is neglected (a conservative assump ion), the imoact e c ~,; will be approximately 254,000 in-lbs.
This energ: w
- absorbed by the following mechanisms:
~:
(a) Since the fuel bundle is " soft" as compared to the rack, a large part of energy will be absorbed by the collapsing of t h.
fuel bundle, thus limiting the maximum load transmitted to t; rack.
7.2
wrRC Docket Nos. 50-295/304 A_T_TACHMENT 2 (b) A part of the enrgy will be absorbed in bending the base plate inside the fuel cell.
If it is conservatively assumed that, in the extreme case, the bending of the base plate causes a localized plastic hinge to form at the intersection of the tube wall and the base plate, the upper bound stress due to the accidental fuel bundle drop can be evaluated by applying at the cell wall the load required to form such a localized plastic hinge.
This was done using a finite element model of a portion of the rack in the vicinity of the bundle drop.
Loads were computed and stresses were determined at the bottom of the tube wall.
The poison material is capsulated at a height of 4.26 inches from the base plate.
Maximum stress in the outer tube wall at that level was computed to be 18.9 ksi, well below the yield stress limit of the material.
Also, it has been observed that the loads dissipated rapidly in the structural panels, indicating that the overall structural integrity of the rack is not impaired.
7.3
NRC Docket Nos. 50-295/304 ATTACIIMENT 2 Dropping Fuel Bundle Q
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Dl Cm l '-
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,N,N o
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o All dimensions in inches FIGURE 7.1 SPENT FUEL RACK SHOWING C;:7ICAL LOCATION OF POSTULATED FUEL BUNDLE 0?::
7.4
NRC Docket Nos. 50-295/304 ATTACHMEffr 2 QUESTION NUMBER 8:
The results from the sliding analysis indicated that one rack could potentially slide 1.31 inches and that the minimum gap between any two adjacent racks is 2.4 inches. Discuss the basis for concluding that two adjacent racks could never slide out of phase, actually slide towards cach other, and impac..
They potentially could close a gap of 2.62 inches (180 degrees out of phase).
RESPONSE
Each rack can potentially slide a distance of 1.31 inches towards each other. However, providing a space between the two adjacent racks less than twice this distance was justified from the following considerations:
(a) 1.31 inches is the peak movement of the rack obtained from a time history analysis. Under identical conditions, the adjacent rack would be in phase and would also move 1.31 inches in the same direction, in which case the original gap between the two racks would remain unaltered. However, since the adjacent rack is not likely to have identical conditions, the gap status is likely to change. Only if the two racks have identical conditions and their movements are exactly 180o out of phase, the minimum required gap to preclude impact would be the absolute sum of the movements of two racks, i.e., 2.62 inches. However, the probability of satisfying both these conditions simultaneously is extremely small, which justifies the use of a lesser gap.
If SRSS method is applied to account for the low probability of the phenomenon, the required gap would be 1.414 times 1.31, i.e.,1.85 inches which is less than the 2.4 inches gap provided.
8.1
NRC Docket Nos. 50-295/304 ATTACED4ENT 2 (b) 1.31 inch is the predicted movement of the empty rack computed using the minimum coet ficient of friction.
It is judged that the loaded racks would slide significantly less than the empty racks. The reasons have been outlined in the response to Question No. 4, 8.2
NRC Docket Nos. 50-295/304 ATTACHMENT 2 Q'JESTION NUMBER 9:
Provide the type or grade of stainless steel used in the construction of these racks.
RESPONSE
All structural materials, with the exception of Boral, are stainless steel grade 304.
a 9.1