ML19319D959
| ML19319D959 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/19/1975 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19319D956 | List: |
| References | |
| NUDOCS 8003270642 | |
| Download: ML19319D959 (17) | |
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ATTACHMENT III RANCHO SECO UNIT I FUEL STORAGE RACK REPLACEMENT SAFETY ANALYSIS f
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r TABLE OF CONTENTS P. age A.
Introduction 1
1.
History and Need for Replacement 1
2.
General Description 1
B.
Environmental Aspects 2
C.
Safety Analysis 3
1.
Criticality Considerations 3
2.
Fuel Handling Considerations 3
3.
Cask Drop Consequences 6
4.
Mechanical Considerations 6
5.
Material Considerations 8
6 Thermal' Considerations 9
7.
Installation Considerations 12 Table 1.1 - Assumptions Used in Criticality Analysis 13 Table 1.2 - Theory - Experiment Correlations 14 Table 6.1 - Thermal Hydraulic Parameters for 100 KW Bundle Heat Load in Proposed Rancho Seco Spent Fuel Rack 15 l
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A.
Introduction i
1.
History and Need for Replacement The Rancho Seco Unit I spent fuel storage pool was designed on the assumption that a fuel cycle would be in existence that would require the storage of a single batch of spent fuel for less than one year in the spent fuel pool.
Therefore, a pool storage capacity of 1-l/3 cores was considered adequate.
This would allow the complete unloading of the reactor for main-tenance or inspection, even if one batch (1/3 core) were in the pool.
Currently, spent fuel is not being reprocessed on a commercial basis in the United States.
In addition, the Sacramento Municipal Utility District has been unable to contract for off-site storage of spent fuel at the present time.
It is therefore desirable to modify the existing spent fuel storage facility to allow storage of additional spent fuel on the Rancho Seco Unit 1 site.
The Rancho Seco Unit 1 spent fuel pool has a present capacity for 244 fuel assemblies including four failed-fuel assemblies. A proposed modification would expand the storage capability to 579 fuel assemblies, including three failed-fuel assemblies.
It is highly desirable that this modification be com-pleted prior to the first refueling of Rancho Seco Unit 1 so the storage pool can be drained and the installation made under essentially clean conditions without fuel present.
The District has entered into a contract with Exxon Nuclear Company for the desion, analysis, and fabrication of replacement spent fuel storage racks. These replacement spent fuel storage racks will provide storage for slightly over three full cores of fuel. Therefore, nine annual discharges may be accommodated, or six annual discharges may be accommodated while still maintaining the capability for a full-core discharge.
2.
General Description The replacement spent fuel storage racks will be fabricated from 304 stainless steel, and do not use a poison material such as boron-impregnated stainless steel or aluminum.
The individual fuel assemblies will be stored in square fuel i
guides fabricated from 14-gauge stainless steel.
The fuel guides are mounted in the rack structure on a square 15-inch center-to-center lattice.
The design of the racks-has been 4
analyzed to show that the racks are adequate for their intended use and that the design meets all applicable codes, standards, i
and regulatory guides.
'I B.
Environmen_tal Aspects On September 16, 1975, the Nuclear Regulatory Commission announced its intent to prepare a generic environmental impact J
statement on handling and storage of spent fuel storage from light water power reactors. The Sacramento Municipal Utility District is requesting a license amendment to allow modification of the l
Rancho Seco Unit 1 spent fuel storage pool in advance of this en-vironmental impact sta'ement. This licensing action would enable the District to store five additional annual discharges of fuel.
This would give the District greater operating flexibility which j
would be desirable even if adequate off-site storage facilities i
should later become available.
Also, it is not likely that this j
licensing action would constitute a committment of resources that would affect the alternatives available to other nuclear power plants or future actions taken by the industry in the future to alleviate fuel storage problems.
The proposed modifications will require custom-made racks made of stainless steel. This material is readily available in abundant supply. This material requirement is insignificant and does not present an irreversible commitment of natural resources.
The District has assessed the environmental impacts of this modification.
There are no potential effects on the environs outside of the fuel storage builing that will result from the pro-posed construction work. Within this building, the impacts are expected to be limited to those normally associated with metal-working activities.
In addition, there are no adverse effects that will occur on-site or in the surrounding environs that can be associated with an increase in the number of fuel assemblies stored in the pool.
The existing spent-fuel pool cooling system has been de-termined to be adequate for the increased number of fuel assemblies.
Therefore, the original heat rejection estimates are still applicable.
Continuous water purification is used to remove liquid wastes from the spent fuel pool water..This same filtration and demineralization will be used after pool modification to maintain the quality of the water at the same high level as originally planned. Therefore, there will be no increase in radiation levels inside the spent fuel pool building.
The District feels that this report answers any technical issues relative to the fuel storage pool modification.
If this 4
modification were to be deferred, the rack replacement would have to be made after spent fuel is already in the Rancho Seco spent fuel pool.
The District feels this is not in the best public interest. Operation of the Rancho Seco plant would also be jeopardized because the capa-bility to discharge an entire core for maintenance or inspection would be lost following the second annual discharge of fuel in the spring of 1978. Accordingly, deferral of this modification would result in a more expensive modification and would~ result in substantial harm to the public interest.
i.
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C.
Safety Analysis 1.
Criticality Considerations An analysis was performed of the potential maximum reactivity of the fuel stored in the proposed fuel assembly storage faility.
This analysis considered the minimum possible spacing under normal and earthquake conditions, the maximum fuel enrichment level, the most reactive conditions of fuel density, and the most reactive water temperature.
The parasitic neutron characteristics of the structure were also considered in the analysis. The limiting conditions assumed for these various parameters are given in Table 1.1.
(a) Calculational Methods The KENO-II Monte Carlo codel/ was utilized to calculate the reactivity of the Rancho Seco fuel storage array.
Multigroupcrosssectiondata(18energygroups)utgized 3
in these calcylations were averaged using the CCELL__, BRT-1 /,
and GAMTEC-II't/ codes.
Specifically, the cross section data for various regions within the storage array were obtained as follows:
CCELL - Utilized to obtain cell averaged multigroup cross section data for fuel rod-water lattices.
Such calculations included both the bundle averaged cell parameters and the actual lattice cell parameters.
In addition, CCELL was used to (1) examine the effects of UO2 pellet density, moderator temperature, and fuel temperature on the infinite media multiplication factor of the fuel assembly, and (2) calculate epither-mal multigroup cross section data for stainless steel (E4.0.683 ev) averaged in a neutron energy spectrum characteristic of the water regio.ns within a fuel assembly.
jf G. E. Whitesides and N. F. Cross, " Keno - A Multigroup Monte Carlo Criticality Program," CTC-5, Union Carbide Corporation Nuclear Division, September 1969.
y W. W. Porath, "CCELL Users Guide," BNW/Jil-86, Pacific liorthwest Laboratories, February 1972.
y C. L. Bennett and W. L. Purcell, "BRT-1:
Battelle Revised THERMOS,"
BNWL-1434, Pacific fiorthwest Laboratories, June 1970.
4/
L. L. Carter, C. R. Richey, and L. E. Hushey, "GAMTEC-II: A Code.
for Generating Consistent Multigroup Constants Utilized in Diffusion i
and Transport Theory Calculations," BNWL-35, Pacific Northwest Laboratories, March 1965. !
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l l
l BRT Thermal group (6 0.683 ev) cross section data for the stainless steel fuel guides were averaged using the Batelle Revised THERMOS code.
Such data were averaged assuming a 0.078 inch thick region of stainless steel.
GAMTEC-II - Multigroup cross section data for water were averaged over a neutron energy spectra characteristic of an infinite media.
In addition to the codes identified above, the XMC Monte Carlo code was utilized to verify the accuracy of CCELL i
calculated values of k. for rod-water lattices.
The XMC code.is a energy groups) pseudo-point energy Monte Carlo code (* 90 which permits the discrete representation of the entire fuei assembly, j
(b) Results The reactivity of both the nominally spaced (15 inches center-to-center) and minimum spaced (14.5 inches center-to-center) storage arrays were corputed.
The infinite multiplication factors for the arrays were calculated to be 0.916 +.003 and 0.932 +.003 respectively.
(The uncertainty Tactors refer to the Monte Carlo calcula-tional statistics at the 95% confidence level.)
In addition to the above, a calculation was performed for a condition of pool water and fuel assembly tem-perature to bound the limiting effects of pool tempera-ture variations.
Specifically, the reactivity of the array was computed assuming the fuel assembly and associated moderator were at 20 C while the water between adjacent fuel assemblies was assumed to be at 100 C.
Such as-sumptions maximize both the reactivity of the fuel assembly and the interaction between adjacent assemblies, i
For this boundary case, the reactivity was calculated to be 0.935 +.004.
Additional calculations were made to show that the calculations are conservative with respect to systematic uncertainties.
No experimental data are known to exist
)
that provide a direct comparison for calculations of fuel assemblies widely spaced in water (i.e., grossly i
overmoderated arrays of fuel assemblies).
Theory j
experiment correlations are limited to calculations of small water-moderated critical arrays of fuel. rods. ?
Such critical experiments have been evaluated using the KEf10 Monte Carlo code with cross section data averaged as for this criticality safety evaluation. The results of those calculations are shown in Table 1.2.
Although a specific bias cannot be established on the basis of these calculations, it appears that the calculational method yields conservative results relative to the experimental data.
In addition, the KEtt0 and DTF-IVE/
transport theory calculations agree within the statis-tical uncertainty of the Monte Carlo calculations.
Finally, to quantize potential systematic errors associated with the stainless steel, a calculation was performed with the stainless steel guide replaced by a void region.
The resulting decrease in reactivity associated with the 0.078 inch stainless steel fuel guides is + 2.7%A P.
If one assumes a 10% error in the calculated effect of stainless steel in the storage array, the array reacti-vity would increase by ~.27% A P which is not statis-tically significant.
(c)
Independent Audit of Calculations An independent review of the criticality safety calcu-lation performed by Battelle florthwest confirmed the adequacy of the calculations.
2.
Fuel Handling Considerations An analysis was performed by the consequences of a fuel-handling accihnt in the Final Safety Analysis Report for Rancho Seco Unit 1.
That analysis identified the maximum fission product 5/
K. D. Lathrop, "DTF-IV - A FORTRAfi-IV Program for Solving the Multigroup Transport Equation with Anisotropic Scattering," LA-3373, Los Alamos Scientific Laboratory, July 1965. -
release as that associated with the total bundle rupture during handling near the surface of the pool. The Nuclear Regulatory Comission's Safety Evaluation Report for Rancho Seco Unit 1 stated that the FSAR assumption was acceptable.
The modifica-tion proposed for the fuel storage pool would not affect the consequences or probability of that accident or introduce a different or more severe accident.
An examination was made of the possible positions which a fuel assembly could assume if it were inadvertently placed hori-zontally on the top of the filled racks. This examination showed that the reactivity of the array could not exceed the reactivity of the normal bundle array as determined in the criticality analysis.
3.
Cask Drop Consequences The Nuclear Regulatory Commission concluded in its Safety Evaluation Repcrt for Rancho Seco Unit 1 that the spent fuel shipping cask storage area had been designed to minimize the loss of water due to an accidental drop of a spent fuel shipping cask and was acceptable.
The proposed fuel storage rack modifica-tion does not involve the spent fuel shipping cask area.
There-fore, the proposed modification does not affect the original cask drop evaluation.
4 Mechanical Considerations A comprehensive dynamic seismic analysis and stress evaluation of the Rancho Seco modified spent fuel storage racks has been performed. This evaluation includes calculation of seismic loads, stress analysis for all applicable loading com-binations, and determination of structural adequacy of all load carrying numbers, including local and distributed stresses of the spent fuel pool liner, floor, walls, and cask catcher.
The seismic loads for the fuel racks were obtained from the response spectrum analysis performed with a model of the most limiting rack. The dead loads were obtained from a self-weight analysis using the seismic model.
The crane uplift live load was considered. Thermal loads have been included in the evaluation.
Stress evaluation for seismic load combinations was performed for each type of structural member of the most limiting rack, for the pool walls and floor, and for the cask catcher.
Based on the calculated stress margins, it has been determined that the Rancho Seco modified spent fuel storage racks, spent fuel pool lining, walls, and floor, and the cask catcher are structurally adequate.
Structural design criteria were developed to assure conformance with recognized codes and ap-plicable U.S. NRC regulatory guides.
The following regulatory guides apply: '
(a) Regulatory Guide 1.13 - Design is in conformance with the stated provisions for spent fuel equipment.
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(b) Regulatory Guide 1.29 - The spent fuel pool storage racks' are designed as Category I structures.
(c) Regulatory Guide 1.61 - Damping values for welded steel structures are used for the storage racks.
(d) Regulatory Guide 1.92 - Seismic load combinations of vibrational modes and earthquake components conform with the accepted methods.
Combinations of structural loads conform with the U.S. NRC Structural Design Criteria for Category I Structures Other Than Containment (Revision 1, June 1975).
The guidelines of this guide are applied for the following industry codes:
(a) AISC (1969),Part 1, Elastic Working Stress Design, is applied to all stainless steel structures.
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(b) ACI (1971) is applied to the concrete pool wall.
The storage racks are designed for the applicable com-binations of dead load, live load, operating basis earthquake, safe shutdown earthquake, thermal and hydraulic loads.
The Rancho Seco FSAR ground response spectra, as found acceptable by the Nuclear Regulatory Commission in its Safety Evaluation Report, are used for seismic loads.
The replacement spent fuel storage racks consist of frames supporting storage receptacles (fuel guides) for spent fuel.
The basic structural function of these storage racks is to maintain safe geometric spacing between spent fuel assemblies during and after all applicable loading combinations and transients.
Within the spent fuel pool are fourteen rectangular spent fuel racks and one rectangular failed fuel storage rack. All adjacent racks are interconnected.
Thermal expansion of the racks is accommodated by a small lateral clearance of approximately 1/8" provided during the initial installation between the racks and the pool walls.
The structural frame consists of two horizontal grids (one at the top of the rack and one at the bottom of the rack),
four corner columns, and diagonal braces to stabilize the racks against side way motion.. The upper and lower grids are constructed from specially fabricated stainless steel tubing with a rectangular cross section.
Grid numbers are located to provide convenient attachment points for the vertical fuel storage receptacles which are rigidly secured to the grids.
The guide tubes which -
=
y hold the spent fuel assemblies are square tubes supported by the 4
rack frame at the top and bottom. The spent fuel assembly is supported by a plate at the bottom of the guide tubes. Lateral motion of the fuel assembly is limited by guide tube supports located near the top and bottom of the fuel assembly. The fuel assembly rests in the guide tube under its own dead weight.
The racks are each supported by four leveling screws.
Horizontal motion of the racks is limited by side-adjusting screws extending toward the pool walls at an elevation of ap-proximately 6-3/8" above the pool floor. All adjacent racks are interconnected at the top. The racks rest on the pool floor.
The stability of the racks to overturning seismic loads is in-sured through the dead weight of the racks and fuel and by rack interconnection plates which intertie the tops of all adjacent racks. The function of these plates is to provide a connection between racks which reacts to counteract any tipping moments generated by seismic activity. These interconnection plates i
operate in shear carrying negligible tension and compression forces.
Basic structural analyses were carried out by Exxon fluclear Company. A dynamic seismic analysis of the racks was conducted by EDS fiuclear Company.
This analysis was carried out using the seismic spectrum defined for the spent fuel pool floor for Rancho Seco Unit 1 for both the SSE and OBE. Seismic loads obtained from this analysis were combined with other loads as applicable to calculate stresses on all structural members.
The calculated stresses on the structural members was then compared with the applicable allowable stresses to determine structural adequacy.
The seismic and structural analyses show that the spent fuel racks are structurally adequate and meet the design criteria presented earlier. Analyses were also made to show that the spent fuel pool structure and the existing cask catcher are structurally adequate for the increased load caused by the higher density of fuel storage.
5.
Material Considerations All permanent structural material used in the fabrication of the new spent fuel storage racks is 300 series stainless steel, mostly 304.
This material was chosen for compatibility with the spent fuel pool water which contains boric acid at a nominal con-centration of 1,800 ppm boron.
At the normal operating temperature of 120*F there is no deterioration or corrosion of stainless steel in this environment.
There is also no corrosion problem at temperatures up to and including l
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pool boiling. All other structural components in the spent fuel pool system, such as the pool liner, cooling system pipe connections, etc.,
are made of stainless steel.
No poison material such as boron carbide has been used in the design of the fuel storage racks.
The 304' stainless steel utilized at fabrication of the racks shall comply with ASTM speci-fications A276-71 or A167-74. All weld electrodes shall be 308 or 4
308L stainless steel.
In summary, the material used in the new spent fuel storage racks is identical to present components and does not affect or alter previous evaluations.
6.
Thermal Considerations (a) Fuel Assembly Heat Removal The decay heat from a spent fuel assembly discharged from the Rancho Seco reactor has been determined using ANS Standard 5.1.
The heat generation six days after shutdown is approximately 50 KW; however a heat generation rate of 100 KW has been used in analysis of heat removal from a single fuel bundle.
In the new fuel rack design, heat is removed from the stored fuel by the natural convective flow of water up through the fuel guide.
The rack utilizes a 14-gauge stainless steel shroud with a square cross-sectional shape as the vertical fuel guide.
In the design, flow is provided to each bundle through a 6-inch diameter hole located in the bottom of the shroud.
The shroud acts as a chimney.
Coolant is heated within the chimney, expands slightly, and raises from buoyancy upon leaving the chimney. The coolant mixes with water in the region above, is cooled, and then returns to the bottom of the vertical fuel guide through the open spaces between the fuel guides.
Analyses show that natural convection flow provides good heat removal from fuel bundles stored in the spent fuel guides.
The calculations were made using fuel bundle flow resistance data developed experimentally for similar fuel bundles.
Assuming a maximum heat generation in a bundle of 100 KW as a basis, calculatior;s show that the water flowing through a fuel guide will be heated about 18 F and that the maximum fuel pin to water temperature drop is 31*F.
The resulting thermal hydraulic parameters for a 100 KW bundle stored in the fuel racks are shown in Table 6.1 for the normal pool operating temperature of 120 F, and the maximum expected pool temperature of 180 F.
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ee
Table 6.1 shows that the fuel cladding will not exceed 169*F under normal conditions, or 229*F under maximum postulated loss-of-cooling conditions.
Since cladding integrity is not degraded at these temperatures, the rack design meets the heat removal design objectives. Ho coolant boiling occurs anywhere within the fuel storage racks even during loss-of-cooling conditions.
Even if the fuel pool were to increase to 212 F at the surface, there would be no bulk water boiling within the fuel guides because of the hydraulic head of water over the racks.
Fuel cladding temperatures under this condition could approach 260*F.
(b) Fuel Pool Cooling An evaluation was made to determine the adequacy of the existir.,
spent fuel pool cooling system for the increased capacity of 579 storage spaces.
For purpose of the evaluation, ANS Standard 5.1 was used for decay heat rate calculations and the following worst case conditions were evaluated:
Case A --Hormal refueling with eight fuel batches in the spent fuel pool and the ninth batch in the reactor core awaiting transfer to the pool.
Case B1--Complete core unload with six batches of fuel in the spent fuel pool and an anomaly requiring removal of the entire core with transfer to the spent fuel pool.
Case B2--Same as Case Bl except that neither decay heat removal system is operable. Cooling is supplied by just the spent fuel pool cooler.
Following is a summary of the results of the evaluation made for the three cases described above:
Case A - Normal Refueling:
Eight batches of fuel are cooling in the spent fuel pool (SFP)_with the ninth batch in the core await-
,ing transfer to the pool.
Assumptions (1) Three years irradiation time on all fuel.
(2)
SFP heat exchanger only (3) Pool temperature maintained at 120*F..
I 1
i Results (1) With 95*F cooling water, the ninth batch must cool in the core for 14 days.
(2) The pool will reach 212 F in 36.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> if all cooling to the pool were lost with the ninth batch in the pool.
The heat generation rate is that at 5 days.
Case B Core Unload:
Six batches of 0 Il are cooling in the SFP.
Due'to some anomaly, the entire cor, must be moved to the SFP.
1 Assumptions (1) Three years irradiation time on fuel already in the SFP; 2-1/4, 1-1/4, and 1/4 years irradiation on fuel in the reactor core.
(2) SFP heat exchanger and one decay heat removal (DHR) heat ex-changer available (the redundant DHR Hx is assumed inoperative).
(3) 95'F cooling water.
Results (1) For maintaining a pool temperature of 120'F, the core must cool for five days prior to transfer.
(2)
If all cooling was lost following transfer, the pool would reach 212*F in 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> using heat generation rate at five days.
Case B Core Unload Assumptions (1) Three years irradiation time on fuel already in the SFP; 2-1/4,1-1/4, and 1/4 years irradiation on fuel in the re-actor core.
(2) SFP heat exchanger available (both redundant DHR Hx's are assumed inoperative).
(3) Core is cooled 5 days prior to fuel transfer.
Results With 95*F cooling water, the pool equilibrium temperature is 180*F in 9.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />..
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It is concluded that the existing spent fuel pool cooling system is adequate for the proposed increase in storage capacity since the original 120*F design temperature is not exceeded under normal conditions. Under Cases A and 81 at least one loop of decay heat removal system would be available for cooling during fuel transfer operations.
However, even without at least one decay heat removal system loop available for spent fuel pool cooling, it is unlikely that preparations could be completed for fuel transfer in less time than the required core cool-down time for both cases.
7.
Installation Considerations Rancho Seco Unit 1 is currently operating in its first fuel cycle and the reload fuel for the second cycle will not be received until the rack replacement has been completed.
Therefore, there is no fuel, spent or new, now in the fuel storage building.
The borated water now in the spent fuel pool will not be discharged from the site because the boron content would exceed discharge limits.
This water will be removed from the fuel pool and stored on site during the removal of the old racks and installa-tion of the new racks.
Possible temporary storage locations for this water are:
i (a) Waste receiver tanks (b) Concentrated boric acid storage tank (c) Fuel transfer pit area of the spent fuel pool (d) Top off the borated water storage tank (e) A new spent regenerant holdup tank now under construction before it is used'for its intended purpose (f) Temporary rented storage such as rubber bladders or clean railroad tank cars Once the water is removed from the storage area, normal construction activities may be started.
Ilo underwater work will be i
required, and since no spent fuel has ever been in the pool, radiation levels will be minimal.
The new racks will be provided in modules which will fit into the building through the cask shipment door.
Once inside the fuel storage building, normal construction procedures would.be followed for the location and alignment of the racks.
The fuel handling bridge will be reindexed to the new fuel storage locations; however, no modifications will be made to the fuel handling bridge l
itself.
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TABLE 1.1 ASSUMPTIONS USED IN CRITICALITY ANALYSIS
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Parameter Value Fuel Bundle Design 208 Fuel Rods (15 x 15 array -
standard design used for Rancho Seco Enrichment All fuel rods 3.5% U235 unirradiated UO2 Density 92.5% of Theoretical Maximum Temperature 20*F for fuel and moderator Field Bundle Spacing 14.5 inches
- center-to-center for 3
four bundle cluster Stainless Steel Thickness 0.78 inches surrounding each bundle
- Nominal spacing 15.0 inches, minimum spacing with adjacent bundles in corners of guide tubes 14.6 inches, maximum vibratory deflection of spacer during SSE 0.05 inches.
The reactivity calculation is for a four-bundle cluster with 14.5 inch center-to-center spacing.._
1 TABLE 1.2 THEORY - EXPERIMENT CORRELATIONS EXP'TL
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RESULTS MODERATOR-CYLIND.
CCELL-DTF-IV CCELL-KENO-II EXPERI-FUEL CLADDING SQUARE T0-FUEL CORE CALCULATED CALCULATED
's MENT DENSITY WT %
PELLET THICK.
LATTICE VOLUME RADIUS REA{eff)
TIVITY REA{eff)
TIVITY 3
NO.
(g/cm ) 235U DIA.(IN) MAT'L.
(Ill)
SPACING (IN)
RATIO CM
(
(
1 10.18 2.70 0.300 304 SS 0.0161 0.435 1.405 26.820 1.016 1.023 +.006 2
10.18 2.70 0.300 304 SS 0.0161 0.470 1.853 24.294 1.015 1.023 f;.005 3
10.18 2.70 0.300 304 SS 0.0161 0.573 3.357 23.600 1.011 1.017 +.007 4
10.18 2.70 0.300 304 SS 0.0161 0.615 4.078 24.771 1.009 1.012 f;.006 5
10.18 2.70 0.300 304 SS 0.0161 0.665 4.984 27.172 1.005 1.001 f;.007 1 '
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r TABLE 6.1 Thermal Hydraulic Parameters for 100 KW Bundle Heat Load in Proposed Rancho Seco Spent Fuel Rack Subcooled Pool Subcooled Pool (Outlet Temp.
(Outlet Temp.
Parameter 120*F) 180'F)
Coolant Mass Flow Rate (ib./hr) 18,000 18,000 Coolant Bulk Temperature Rise (*F) 18 18 Bundle Bulk Discharge Temp. ('F) 138 198 Fuel Pin Film Temperature Drop (*F) 31 31 Fuel Pin Peak Cladding Temperature ('F) 169 229 9..
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