ML19329B189
| ML19329B189 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/19/1977 |
| From: | Roe L TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML19329B188 | List: |
| References | |
| NUDOCS 8001310543 | |
| Download: ML19329B189 (35) | |
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Application for Amendment to License No. NPF-3 for Davis-Besse Nuclear Power Station Unit No. 1 Enclosed are forty-three (43) copies of the requested changes to the Davis-Besse Nuclear Power Station Unit i Technical Specifications, Appendix A to License No. NPF-3 together with the report " Safety i
Evaluation of the Spent Fuel Storage Capacity Modification for Davis-Besse Nuclear Power Station Unit 1" which states the reasons for the 4
requested changes and contains details of the design, design analysis, and safety evaluation of the proposed modification to increase spent fuel storage capacity.
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/s Vice President, Facilities Development Sworn to and subscribed before me this /9Y day of nacember, 1977, l
I f
MA Notary Public FRED W. GERMAIN Notary Public-State of Ohio t.ly Commission Expires Oct. 30,1002 l
soo131o5'r3
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A DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximun internal pressure of 40 psig and a temperature of 264*F.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 177 fuel assemblies with each fuel assembly containing 208 fuel rods clad with Zircaloy -4.
Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 2500 grams uranium. The initial core loading shall have a maximum enrichment of 3.0 weight percent U-235.
Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.3 weight percent U-235.
l The first cycle fuel loading shall contain 68 burnable poison rod assemblies with each assembly containing up to 16 burnable poison rods of sintered AL 0 -B C clad with Zircaloy-4.
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CONTROL ROOS 5.3.2 The reactor core shall contain 53 safety and regulating and 8 axial power shaping (APSR) control rods. The safety and regulating control rods shall contain a nominal 134 inches of abscrber manterial.-
The APSR's shall contain a nominal 36 inches of absorber material at their lower ends. The ncminal values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmiun. All control rods shall be clad with stainless steel. tubing.
DAVIS-BESSE, UNIT 1 5-4 er
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DESIGN FEATURES I.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
In accordance with tha code requirements specified in Section a.
5.2 of the FSAR, with allowance for normal degradation pursuant to applicable Surveillance Requirements.
b.
For a pressure of 2500 psig, and For a temperature of 650*F, except for the pressurizer and c.
pressurizer surge line which is 670*F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 12,110 + 200 cubic feet at a nominal T,yg of 525*F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tcwer shall be located as shown on Figure 5.1 -1.-
5.6 FUEL STORAGE CRITICALITY 5.6.1 The new fuel storage racks are designed and shall bc maintained with a nominal 21 inch center-to-center distance cetween fuel assemblies placed in the storage racks to ensure a k,ke,r.quivalent to e
< 0.95 with the storage pool filled with unborated wa The K of 7 0.95 includes a conservative allcwance of 1% ak/k for uncertai85,ies as described in Section 9.1 of the FSAR.
The spent fuel storage racks are designed and shall be maintained with a rectangular array of stainless steel cells spaced 12 31/32 inches on centers in one direction and 13 3/16 inches on centers in the other direction.
Fuel assemblies stored in the spent fuel pool shall be placed in a stainless steel cell of 0.125 inch nominal thickness or in a failed fuel container such that a K gg equivalent to 10.95 is maintained with the storage po-1 filled with e
unborated water.
The K,gg of 10.95 includes a conservative allowance of 17.ak/k for calculational uncertainty.__
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool belcw 9 feet above the top of the fuel storage racks.
DAVIS-BESSE, UNIT 1 5-5
DESIGN FEATURES 4
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CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 735 fuel assemblies.
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5.7 COMPONENT' CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limit of Table 5.7-1.
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i DAVIS-BESSE, UNIT 1 5-6 i
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m The Toledo Edison Company and The Cleveland Electric Illuminating Ccmpany Safety Evaluation of the Spent Fuel Storage Capacity Modification for Davis-Besse Nuclear Power Station Unit 1 Docket No. 50-346 License No. NPF-3
)
December 5, 1977
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Table of Contents 1.0 Introduction 2.0 Design of the Spent Fuel Storage Racks 2.1 Design Bases 2.2 Storage Rack Description 2.3 Storage Rack Safety Evaluation 2.3.1 Nuclear Criticality Analysis 2.3.2 Structural and Seismic Analysis 2.3.3 Storage Rac'< Thermal-Hydraulic Analysis 3.0 Spent Fuel Fool Structural and Seismic Analysis 4.0 Spent Fuel Pool Cooling System Evaluation 5.0 Radiological Consequences Figure 1 -
Spent Fuel Storage Rack General Arrangement Figure 2 -
Spent Fuel Storage Rack Module Appendix A - Davis-Besse Nuclear Power Station Unit 1 Modified FSAR Material 1
1.0 Introduction This report is prepared and submitted in support of Toledo Edison's request, on behalf of itself and Cleveland Electric Illuminating, to amend the Davis-Besse Nuclear Power Station Unit 1 Facility Operating License No. NPF-3, to reflect an increase in spent fuel storage capacity.
Due to the present and expected near-term shortage of reprocessing facilities and other uncertainties associated with the tail end of the i
LWR fuel cycle, it is necessary to replace the existing Davis-Besse Unit 1 spent fuel acorage racks (260 assembly capacity) with high capacity spent fuel storage racks (735 assembly capacity). This modification will be necessary before the first refueling, in order to replace the racks with no spent fuel in the pool and to maintain the ability to discharge spent fuel while maintaining full core (177 fuel assemblies) off-load capability.
l The Toledo Edison Company is responsible for the overall design, construc-tion, and operation of Davis-Besse Unit 1.
Toledo Edison has contracted with Nuclear Energy Services, Incorporated, to design, manufacture, deliver and install the proposed high capacity spent fuel storage racks.
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2.0 Dmsign of thn Sp, Fu91 Storags Racks 2.1 Design Bases The high capacity spent fuel storage racks are designed to provide storage locations for up to 735 fuel assemblies and are designed to maintain the stored fuel, having an equivalent uranium enrichment of up to 3.3 weight percent U-235 in UO, in a safe, coolable, and suberitical configuration during normal'and abnormal conditions.
2.2 Storage Rack Description The spent fuel storage rack consists of a 16 x 45 array of 720 square stainless steel cells spaced on a pitch of 13-3/16 inches in the 16 cell direction and 12-31/32 inches in the 45 cell direc-tion, plus 15 locations for failed fuel containers located adjac~ent to one side of the 16 x 45 cell arrangement.
To achieve the 720 cell configuration, six 7 x 8 array and six 8 x 8 array modules are employed. The storage rack is shown in the general arrangement drawing, Figure 1.
Each storage rack module has two basic components:
the support structure and the fuel storage cell as shown in Figure 2.
The support structure consists primarily of the four corner storage cells which are secured to two levels of grid members which maintain the horizontal position and vertical alignment of the remaining inner storage cells. The support structures are supported from the sp:nt fuel pool floor pads through the four corner cells.
Diagonal beacing is provided on the structure principally to accommodate the loads imposed by rack installation, by fuel handling and by seismic events.
Each storage cell is basically a stainless steel box 9.25 in.
square (0D) by lo6 in. Long with 0.125 in, walls.
The cells are flared at the top to si=plify insertion of the fuel as9e=bly into the cell. Attached to the bottom of each cell are four stainless steel posts which support the weight of the cell and its contents.
The posts attached to the inner cells rest directly on the spent fuel pool floor and space the cells off the pool floor a sufficient distance to assure adequate area for cooling flow. To accccmodate any pool floor liner irregularities, the rack is designed to permit the inner cells to move vertically within the rack structure (i b in.
motion is provid:d). The cells, however, are positively locked into rna support structure so that they cannot be inadvertently lifted out of the rack.
Horizontal seismic loads ara transmitted from the rack structure to the walls surrounding tha rpent fuel pocl through seismic bracing extending from the perimeter of the 16 x 45 array to the pool wall.
3 racing is provided at each of the two grid levels (approximately 25 in. and 145 in. above the spent 'uel pool floor). Vertical dead weight and seismic loads are essentially transmitted directly to the pool floor by each storage call.
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2.3 Storage Rack Safety Evaluation 2.3.1 Nuclear Criticality Analysis A detailed nuclear analysis has been performed to demonstrate that for all anticipated normal and abnormal configurations of fuel assemblies within the fuel storage racks, the k,, of the system is less than 0.95 as confirmed by transport t$eary calculations.
The following conservative assumptions have been used in the criticality calculations:
1.
The pool water has no soluble poison.
2.
The fuel assemblies have no burnable poison.
3.
The fuel is fresh and of a specified enrichment (3.3 w/o) which is higher than that of any fuel currently scheduled for use in Davis-Besse Unit 1.
4.
The.ack configuration is infinite in all three dimensions.
5.
No credit is taken for structural material poison other than the scainless steel fuel rack cell.
6.
All fuel cells are assumed to be 0.125 in, thick, the minimum allowable thickness.
The normal configurations considered in the nuclear analysis included the reference configuration (fuel assembly centrally positioned within cell having ncminal rack dimensions), the eccentric positioning of fuel within the fuel storage cells, and the variations permitted in fabrication of the principal storage rack dimensions (fuel storage cell pitch and cell wall thickness). A transport theory analysis of the reference configuration was done to establish a diffusion to transport theory bias.
The abnormal configurations analyzed were:
a variation from the maximum water density at near 40 F to 260 F including a partly voided situation at 212 F, and a configuration where a complete cell and fuel assembly were displaced due to failure of the retaining clips.
Inadvertant placement of a fuel assembly adjacent to a fuel rack was not analyzed since structure is provided on the peripheral racks where required to =aintain a center-to-center spacing in excess of 17 inches, a clearly acceptable value (k, g less than 0.90).
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The principal calculational method used for the criticality analysis was diffusion theary using HA!9fER (a multigroup integral transport theory code) and EXTERMINATOR (a 2-D multigroup diffusion theory code). Verification calculatiens were doc by transport theory using GGC-3 (a consistent Bn i
or P1 coca for the calculation of fast neutron spectra and i
associated multigroup constants) and DOT (a 2-D multigroup discrete ordinates transport theory code).
The following diffusion theory results were obtained for the normal configurations:
Description eff 13.078 in, average pitch, 68 F 0.8874 1
1 13.016 in average pitch, 68 F 0.8929 13.078 in, average pitch, 68 F,
j fuel displacement 0.8944 i
ak due to reduced pitch
+0.0055 ak due to eccentric fuel placement
+0.0070 The worst case normal configuration k is obtained by staticticallycombining(squareroot$hsumofthesquares) the effects of the normal variations.
The result is k
=
- ff 0.S874 + 0.0089.
l The abnor=al configurations analy::ed gave the following results:
1 ak due to increased pool temperature
+0.0104 Ak due to can displacement
+0.0009 The worst case abnormal configuration combines the worst case normal configuration with the cost adverse abnormal condition (pool temperature rise).
Worst case abnormal configuration k,,, = 0.8874 +.0193
.0089 The reference case was also calculated using transport theory to establish a diffusion to transport bias of 0.0342.
This results in:
Worst case normal configuration k,f', = 0.9216 t.0089 (with bias)
Worst case abnormal configuration s f', = 0.9216 +.0193 (with bias)
.0089 Using a criticality calculational uncertainty factor of.01 combined statistically with the uncertainties due to normal variations produces a worst case k g, = 0.9454. This vel.ue meets the criticality criterion of*a'k
, of less than cout 0.95 (reference USNRC Standard Review $1an 9.1.2).
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,e 2.3.2 Structural and Seismic Analysis The Davis-Besse Nuclear Power Station Unit 1, high density spent fuel storage racks have been designed to meet the FSAR requirements for Seismic Category I structures.
Detailed structural and seismic analyses of the high density storage racks have been performed to verify the adequacy of the design to withstand the loadings encountered during installation, normal operation, the severe and extreme conditions of the i
operating basis and safe shutdown earthquakes and the abnormal loading condition of an accidental fuel assembly drop event.
2.3.2.1 Applicable Codes, Standards and Specifications The following design codes and regulatory guides have been used in the design / analysis of spent fuel storage 1
racks.
1.
A.I.S.C. Manual of Steel Construction, Seventh Edition, 1970.
2.
USNRC Regulatory Guide 1.61, " Damping Values for Seismic Design of Nuclear Power Plants", October, 1973.
3.
USNRC Regulatory Guide 1.92, " Combination of Modes and Spatial Components in Seismic Response Analysis, Rev. 1, February, 1976.
4.
USNRC Standard Review Plan, Section 3.3.4.
2.3.2.2 Loads and Load Combinations The following load cases and load combinations have been considered in the analysis in accordance with the require-ments of USNRC Standard Review Plan, Section 3.8.4.
Load Cases Load Case 1 - Dead Weight of Rack Plus Corner Fuel Assemblies, D + L (Normal Load)
Under normal operating conditions the rack is subjected to the dead weight loading of the rack structure itself plus the loads resulting from four fuel assemblies stored in the four structural corner cells. The loads resulting from the individual storage cells and contained fuel assemblies transmit their load directly to the pool floor and not through the structure.
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Load Case 2 - Dead Weight of Rack Plus 1 g. Vertical Installation. Load, D + I.L. (Normal Load)
Daring installation the rack is subjected to the loading resulting from its own structural weight plus a 1 g.
vertical load resulting from a suddanly applied crane load.
Load Case 3 - Dead Weight of Rack Plus Uplifting Load, (D + U.L.) (Abnormal Load)
The possibility of the fuel handling bridge fuel hoist grapple getting hooked on a fuel storage cell was con-sidered. The uplif t force considered for this load case was 500 pounds which conservatively exceeds the maximum load of 2750 lbs. allowed by the hoist load limit cell and the fuel assembly, storage cell and hoist grapple combined weight of %2500 lbs.
Load Case 4 - Operating Basis Earthauake, E (Severe Load)
The rack, fuel assemblies, and virtual water = ass react to the si=ultaneous loading of the horizontal and vertical components of the seismic response acceleration spectra specified for the Operating Basis Earthquake. The seismic loading is applied to the fully loaded rack.
Load Case 5 - Safe Shutdown Earthauake, E' (Extreme Load)
Same as Load Case 4 except that the seismic response acceleration spectra corresponding to the Safe Shutdown Earthquake was used in the analysis.
Load Case 6 - Assembly Drop Imoact Load, (Abnormal Load)
The possibility of dropping a fuel assembly on the rack from the highest possible elevation during spent fuel handling was considered. A 1685 pound weight was postu-laced to drop on the rack from a height of 24 inches.
This height was determined based on an assu=ed minimum water cover of 8 feet to be maintained during fuel asse=bly handling. Two cases were considered:
- 1) a direct drop on top of a single storage cell and 2) a subsequent tipping of the assembly onto surrounding storage cells.
Thermal Loading, T (Normal Loadi The stresses and reaction 1 afb to thermal loadings are insignificant since cit..ral.c-t are provided between racks to allow unrestrains: grcwth di tra racks for the maximumexpectedtemperaturedifferentia# kased on a maximum pool temperature of 185 F.
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Load Combinations (a) For service load conditions, the following load combinations are considered using elastic working stress design methods of AISC:
(1)
D+L (la) D + L + T (2) D + I.L.
(3)
D+L+E (3a) D + L + T + E (b) For factored load conditions, the following load combinations are considered using elastic working stress design methods of AISC:
(4) D + L + T + E' (5) D + T + U.L.
2.3.2.3 Design and Analysis Methods Static Analysis The response of the rack structure to specified static loading conditions has been evaluated by means of linear-elastic analysis using the finite element method. The rack was mathe=atically modeled as a three-dimensional finite-element structure consisting of discrete three-dimensional elastic beams.
Six degrees of freedom (three translations and three rotations) were permitted at each nodal point. Appropriate boundary conditions were assumed for each load case.
Dynamic Analyses The response of the rack structure to specified seismic loading conditions has been evaluated by =achematically 4
modeling the storage rack as a lumped = ass, multi-degree-of-freedom system. Masses are lumped so as to represent the dynamic characteristics of the storage racks.
Tae eigenvalues and eigenvectors (frequency and made shapes of vibration) of the lumped mass model have been calcu-laced using the Householder-QR technique.
The Seismic Response Analyses are then performed using response spectrum modal superposition methods of dynamic analysis, using the Davis-Besse Unit 1 Amplified Response Spectra and appropriate damping for welded steel structures in accordance with Regulatory Guide 1.61.
Individual modal responses of the systeu are combined in accordance with Section 1.2.1 of Regulatory Guide 1.92.
The maximum response of the systes for each of the three orthogonal spatial components (two horizontal and one vertical) of an earthquake has been combined on a square root of the sums of square (SRSS) basis (Regulatory Guide 1.92).
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The sloshing effects of water on the fuel racks have been evaluated using the analytical methods given in ASCE's
" Structural Analysis and Design of Nuclear Plant Facilities".
The " rattling" effects of the fuel inside the cell have been accounted for by using suitable impact factors.
The static, seismic and stress analyses for the fuel storage racks were performed utilizing the STARDYNE cs=puter code.
The assembly drop load case (Lead Case 6) was performed with linear and non-linear analysis techniques using energy-balance methods.
2.3.2.4 Structural Acceptance Criteria The following allowable limits constitute the structural acceptance criteria used for each of the loading comoina-tions presented in Section 2.3.2.2.
Load Combinations Limit 1, 2, 3 S
la, 3a 1.5S 4, 5 1.6S S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC " Specification for the Design, Fabrica-tion and Erection of Structural Steel for Buildings",
February 12, 1969. The yield stress value for stainless steel is taken as 30.0 ksi.
The acceptance criteria for Load Case 6, the accidental fuel assembly drop onto the rack, is that the resulting impact will not adversely affect the leak tightness integrity of the fuel pool floor and liner plate and that the deformation of the impacted storage cells will not adversely affect the value of k,ff.
2.3.2.5 Results of the Analysis The results of the static, seismic and stress analysis of the rack structure show that the stresses and deflections for each load combination are nominal and within the applicable acceptance criteria.
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The evaluation of the water sloshing effects in the pool resulting from a seismic event indicates that water sloshing will have insignificant effects on the fuel storage rack due to the depth of the pool relative to the rack height.
The results of the analysis for Load Case 6 indicates that the drop of a fuel assembly onto a fuel storage cell will not collapse or buckle the cell, thereby precluding a significant change in rack geometry. The analysis shows that the external kinetic energy of the dropped fuel assembly is absorbed in the local deformation of the flare at the top of the fuel storage cell, in the partial shearing of the cell leg weld, in the local crumbling of the pool floor concrete and in the minor deformation of the pool floor liner plate under each leg. The leak tightness of the fuel pool, however, will be maintained since the deformation is insufficient to result in a tear or puncture of the liner.
It has, therefore, been concluded from the results of the seismic and structural analysis that the deflections and/or stresses in the rack structure resulting from the various loadings meet the deflection and stress acceptance criteria for Seismic Category I structure.
2.0.3 Storace Rack Thermal-Hydraulic Analysis The adequacy of natural circulation flow to cool the spent fuel assemblies in the rack matrix was verified by establish-ing for the East /%'est rack row with the maximum number of assemblies, a thermal-hydraulic balance between the driving head produced by decay heat generation and the pressure losses existing in the natural circulation flow path.
Pressure losses in the downcomers, in the rack inlet plenum, and along the fuel assemblies were explicitly considered in the analysis.
Cross-ficws in the inlet plenum area were conservatively neglected.
The results of the thermal-hydraulic analyses indicate that even with the most conservative assumptions, the natural circulation in the spent fuel pocl is adequate to preclude local boiling by a substantial margin.
The maximum tempera-ture increase in the assembly with the minimum flow is less than 21 F.
This results in a max 1=um outlet temperature of less than 206 F for the maxi =um heat load (assumed maximum bulk temperature of 135 F), which is still below the pool water saturation temperature of 239 F at the highest assembly elevation.
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3.0 Spent Fuel Pool Structural and Seismic Analysis The spent fuel pool, located inside the fuel handling area in the auxiliary building, is a reinforced concrete pool lined with k inch thick stainless steel.
The auxiliary building, as well as the storage pool, is a seismic Class I structure which is designed to withstand seismic, tornado, and thermal loads as discussed in Sections 3.7 and 1.8 of the Davis-3 esse Unit 1 FSAR.
The spent fuel pool structure has been analyzed to determine the effects due to the proposed design for the high capacity spent fuel storage racks.
The spent fuel racks interact with the spent fuel pool structure through thermal stress loads, dead weight loads, and seismically induced loads.
As discussed in Section 2.3.2.3, thermal stress loads are insignificant.
As stated in Section 2.2, vertical seismic and dead weight loads are essentially transmitted directly to the pool floor by each storage cells.
The four corner cells of each module transmit the load of the 4
grid structures and bracing, and t.:us transmit somewhat higher loads than interior cells. All vertical.eeismic and dead weight loads are 1
within design limits for the pool structure.
Horizontal seismic loads are transmitted to the pool walls from seismic bracing attached to each of the two levels of the rack grid structure.
l Figure 1 shows a plan view of the locations for the bracing.
The hori-zontal seismic loads for the Operating Basis Earthquake (OBE) and Safe 1
Shutdown Earthquake (SSE) were analyzed and found to be well within design limits for the Sh foot walls on the north, east and south sides of the pool.
The loads on the west wall, the 3 foot wall between the spent fuel pool and the fuel transfer tube pit, were found to be well within design limits for the 03E but were found to exceed design limits (0.9 of yield strength) for the SSE. To maintain a conservative =argin j
of safety for this wall, the Applicant will modify the pool structure by j
providing adequate struts spanning the fuel transfer tube pit to transmit a portion of the loads to the 5b foot wall on the west side of the fuel transfer tube pit (see Figure 1).
The struts will be in place whenever there is; (1) spent fuel stored in the pool and (2) the fuel transfer tube pit is not filled with water. With the fuel transfer tube pit filled with water, the hydrostatic forces acting on the 3 foot wall I
between the spent fuel pool and fuel transfer tube pit are equalized.
Under these conditions all affected walls are well within design limits
-for SSE loadings without the additional bracing. The fuel transfer tube pit is filled with borated water during refuelings and normally drained otherwise.
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4.0 Spent Fusi Pool Co.iing System Evaluation The spent fuel pool cooling system is designed to maintain the borated spent fuel pool water quality and clarity and to remove the decay heat from the stored fuel in the spent fuel pool. A detailed discussion of the system is presented in Section 9.1.3 of the Davis-Besse Unit 1 FSAR -
(see Appendix A of this document for applicable changes to this material).
The capability of the existing spent fuel pool cooling system to handle heat loads resulting from the expanded spent fuel starage has been calculated for both the normal and the emergency heat load cases.
Decay heat generation is calculated according to NUREG-75/087, Section 9.2.5, Branch Technical Position APCSB 9-2, " Residual Decay Energy for Light Water Reactors for Long-Term Cooling", 11-24-75.
The values used in this analysis include recommended uncertainty f actors and contributions of actinides (U-239 and Np-239).
For the decay beat calculations it was conservatively assumed that the average discharge batch was 60 fuel assemblies and the batch underwent an equivalent continuous operating period of 3 years at a core thermal power of 2772 Mwt with an average cooling time before discharge to the spent fuel pcol of 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />.
Refuelings were assumed to occur en an annual basis.
The maximum normal heat load results with the pool filled with one freshly discharged batch in addition to 11 batches from previous re-fueling outages. Under these conditions, the decay heat generated is calculated to be 12.4 x 1.G Btu /hr. With this heat load the existing spent fuel pool cooling system, with both pumps and both heat exchangers operating, is capable of maintaining the spent fuel pool at 125 F or less. With one pg=p and two heat exchangers operable the pool can be maintained at 140 F or less, and with one pump and one heat exchanger operable the pool can be maintained at 155 F or less under the maximum normal heat load conditions.
The maxi =um abnormal heat load would result when one entire core (177 assemblies) is discharged 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown, 65 days after the last of 9 batches from previous refueling outages.
Under these condi-tions the decay heat generated is calculated to be 29.5 x 10 Btu /hr.
The station is designed such that the decay heat removal system (see Section 6.3 of the Davis-Besse Unit 1 FSAR) is used to remove the decay heat from the spent fuel pool under full core discharge conditions and serves as back-up system to the spent fuel pool cooling system under normal conditions. Each of the two decay heat removal trains is designed for a heat removal capacity of 30 x 10 Btu /hr. at an inlet temperature of 140 F.
The seismic Class I decay heat system is permanently connected to the Class I boundary of the spent fuel pool cooling system. The decay heat system thus will serve to make up the spent fuel pool water by supplying the borated water from the BWST to the fuel pool to prevent uncovering of the fuel should the need ever arise.
Based on the discussion above, it is concluded that the existing spent fuel pool cooling system provides adequate cooling capability to safely handle the additional heat loads caused by the expanded storage capacity.
The designed availability of the decay heat removal system as a back-up heat removal and =akeup system for the spent fuel pool provides assurance of maintaining the pool in a safe ther=al condition.
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5.0 Radiological Consecuences The radiological consequences of increased fuel pool storage capacity were evaluated based on a conservative refueling scenario.
Spent fuel equal to 1/3 core was assumed to be deposited in the fuel pool at one year intervals for a period of 9 years to simulate normal refueling capacity. This was assumed to be followed by a final full core unload.
Fuel was assumed to be stored in the high density spent fuel racks.
The following doses of interest were evaluated.
1.
The dose obtained directly from the stored fuel assemblies, assuming the technical specification requirement of 23 ft. of water above the stored fuel assemblies.
2.
The dose obtained from those radionuclides which become suspended or dissolved in the fuel pool water.
3.
The dose obtained from those radionuclides which escape the fuel pool into the fuel building atmosphere.
Direct Gamma Dose From Stored Fuel The buildup of fission products within the spent fuel assemblies was based upon ane year of full power operation.
These products were assu=ed to decay for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before being transferred to the spent fuel pool to account for the technical specification required 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> decay period. The dose rate after a final fit 11 core un1.oad was calculated at the fuel pool water surface. Previously stored fuel from normal refuelings was assumed to have decayed for a =aximum of one year.
In accordance with the above assumptions, the maximum direct ga=ma dose rats at the pool water surface from stored spent fuel is on the order of 10 ' mrem per hour.
The gamma dose rate through the side walls of the spent fuel pool will be greatly dependent upon the loading arrangement of the spent fuel in the racks. Use of proper radiation control measures in the affected areas will minimize the potential for any additional occupational exposure due to the increased number of spent fuel assemblies.
The dose rate through the bottem of the pool will approximately double for a period of one =onth under the one third of core just removed from the reactor. However, much of the area under the pool is already an "E" zone, the rest is a "C" zone which can be temporarily posted and barricaded if necessary for such a short period.
(Reference Chapter 12 of the DavisBesse Unit 1 FSAR.)
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Doss From Dissolvrd Radionuclides Increasing the number of assemblies in the pool has not led to increased concentrations of radionuclides in the spent fuel pool water at other operating plants. Measurements taken of the activity in the pool water before and after refueling have indicated essentially no change in concentrations. As a result, it is not expected that there will be more frequent changing of deminerali:er resin or filter cartridges. Therefore, little increase in annual-man rem is expected from either the radionuclides in the pool water or accumulated on the resin or filters.
Dose From Airborne Radionuclides Almost all of the leakage of fission products from the fuel will occur for each batch of fuel within a few =onths of when it is removed from the core. Therefore, most of the inhalation and submersion doses will be due to the latest batch of fuel placed in the pool, the increase in dose due to the additional fuel stored in the pool with several years de; y will be negligible, (I-131, Xe-133, and other shorter half life iodines and noble gates will have decayed, so that Kr-85 will be the only volatile fission product left in the fuel).
As discussed previously, there will be a slightly inc; eased heat load en the pool due to the storage of additional fuel witit several years decag.
The incremental increase in the te=perature of the pool water from 120 F to 125 F due to this increased heat loading will lead to a slightly higher tritium concentration in the air due to increased evaporation.
This increase in temperature above the seiginal design of 120 F will only be for a period of approximately 10-15 days af ter a refueling.
After that time the newly discharged spent fuel will have decayed suffi-ciently to reduce the heat load such that the spent fuel pool cooling system can maintain the pool less than 120 F.
Calculations show that the resulting small increase in tritium concentration in the air will be well within the MFC limit in 10 CFR 20 Appendix B, Table I, Colu=n I and will allow normal occupancy in the spent fuel pool area.
Licuid and Gaseous Releases The storage o# additional spent fuel assemblies in the spent fuel pool will not result in any additional liquid release from the plant.
To determine the potential for gaseous celeases due to the spent fuel stored in the pool, the activity of important iodine and noble gas isotopes were calculated and compared to the Davis-Besse Unit 1 FSAR analyses. Almost all of the releases occur frcm each one third of a core within a short period after it is removed from the reactor. Very little of the releases will be from the older fuel which has been stored for several years. The increase in the dose to an individual at the site boundary will be insignificant.
13 i
Radiation Exposure During Modifications This modification will require the disposal or storage of the existing spent fuel racks which are not expected to be contaminated, since they will be replaced before the first refueling.
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Appendix A Dcvis-3 esse Nuclaar Power Statien Unit 1 Modified FSAR Material
MODIFIED D-3 1.2.8.2.2 Chemical Addition System Che=ical addition operatiens are required to alter the cencentration of varicus chemicals in the reacter ecolant and auxiliary syste=s.
The system is designed to add beric acid to the reactor ecclant system for reactivity control, lithium hydroxide for pH control, and hydrazine for oxygen centrol.
1.2.8.2.3 Cooling Water Syste=s The eccling water systems re=cve heat from the statien equip =ent to permit a sustained operation and safe shutdown of the statien.
Condenser Circulating Water System The condenser circulating vater syste= is sized to handle the maxi =us ccndenser heat loads and consists of a closed syste utilizing a hyperbolic neutral draft cooling tower and the associated circulating vater pu=ps, piping and valves. Fill and makeup water is taken frcs Lake Erie through the intake water syste= and intake structure.
Four circulating vater pu=ps with suction frem the cooling tever discharge channel pump through the condenser and back to the cooling tover.
Service Water Syste=
The service water systes takes lake Erie water frc= the intaka
-~~~a punp suction pit after the traveling screens. This syste= supplies eccling vater to the ec=penent cooling vater system, ICCS pu=p recs ecolers, contain-
=ent air coolers, the turbine plant cooling vater syste=s, and is a source of water to the auxiliary feed pu=ps.
It also provides a scurce of nakeup water to the cocling tower.
Centenent Ccolin.- Water System This syste= is a closed loop syste= vhich provides ecoling vater to the nuclear and engineered safety features syste=s and also acts as an inter-
=ediate barrier between the radicactive systen and the service water syste=.
The syste= censists of three circulating pumps, three heat exchangers, a surge tank, associated valves, piping, instru=entation, and centrols.
Turbine ?lsnt Ccoling Water System The recirculated closed icop system furnishes purified and treated eccling vater to =ain turbine and turbine plant pump oil coolers, varicus pu=p seals, generator hydrogen equipment auxiliaries including generator hydrogen ecclers and stator liquid cool)r, isolated phase bus, air cc= pressor jackets and coolers, and turbine plant sample coolers.
The engineered safety features equip =ent is not dependent en the turbine plant cooling water syste=.
i i
1.2.S.2.i Spent Fuel ? col Coolir. System i
The spent fuel pool ecoling system is designed to maintain the berated spent fuel pool water at 125 ? or less with heat lead based en rencvin6 decay heat l
frcn 1-lh
MODIFIED D-3
~'
1/3 core, which is assu=ed to have undersene irradiation fo[ an equivalent centinuous operating peried of 3 years at full core ther=al pcver and to have been cooled for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />, and the decay heat frc= 11 batches of spent fuel from-previous refueling outages.
The unit is, hcvever, designed such that if it becc=es necessary at sc=e time to off-load an entire ccre into the spent fuel pool, the ecoling capa-city can be previded < the decay heat renoval syste=.
During ner=al cperation, bcth pu=ps and both heat exchangers are in centinu-ous operation; as the de: ay heat e=itted by the spent fuel decreases One penp and cne heat exchanger can be shut down. The spent fuel vater ovipera-ture is nor= ally =aintained at 125 F or less.
l During cold shutdown and refueling conditions, the reactor refueling canal is filled with water frc= the bcrated water stcrage tank.
1.2.8.2.5 Decay Heat Rc=cval Syste=
The nor=al function of this systen is to remove reactor decay heat during the latter stages of ecoldevn and =aintain reactor ecclant te=perature during refueling.
1.2.8.2.6 Sa=pling Systen The sanpling system provides samples for labora:Ory analyses which serve te guide the cperation of the reacter coolant systen, the nakeup and purifica-tien syste=, the chemical additien system and the pcVer ccnversion stea:
system. These sa=ples flew to central locatiens in the auxiliary * ' -"-bine buildings; access to the centain=ent vessel for this purpcse is nct required during pcVer operation. Typical cf the analyses perfor ed en such sa=ples 4
are reactor ecolant beric acid concentratien, pH, fissicn product activity levels, dissolved gas centent, corrosion prcduct concentratien and activity and =ain stea= gross activity.
Analytical results are uced fcr regulating bcron concentration adjust =ents, evaluating the integrity of fuel reds and the perfor ance of the decineralizers, and regulating chemical additien to the the reacter coolant.
1.2.3.2.7 Station Ventilation Syste=s The heating, ventilating, and air-conditioning syste=s for the station are designed to provide a suitable enviret=ent for equipment and persennel with equipment arranged in :enes so that pctentially conta=inated areas are separated frc= clean areas. The path of ventilating air in the auxiliary building is frc= areas of lov activity tcvard areas of progressively higher activity.
Conditioned air is recirculated in clean areas only.
The contain=ent air recirev' ' tion systen is used to circulate air within the contain=ent vessel.
T_s e=ergency ventilation systen ventilates the shield building and penetration rec =s.
1-15
MODIFIED D-B 9.1.2 SFINT FUEL STCRAGE 9 1.2.1 Design Bases The spent fuel storage is designed to stcre the irradiated fuel assenclies under water for decay prior to shipment offsite for reprocessing. The storage pool is sised to store 735 irradiated fuel assemblies which includes storage for 15 failed-fuel containers. The spent fuel storage cells are i
installed in parallel revs with center-to-center spacing of 12 31/32 inches in ene direction, and 13 3/16 inches in the other orthesenal direction.
This spacing and " flux trap" construction, whereby the fuel assemblies are inserted into neutron abscrbing stainless steel cans, is sufficient to main-tain a K,ff of 0.95 cr less. Shielding and seismic classification are discussed in subsection 9.1.2.2.
The design of the spent fuel storage area closely follevs the intent of j
Safety Guide 13.
9 1.2.2 Descriptien 2
After renova' "*** 'he reacter, the spent fuel is stcred under water within i
the spent-fuel stcrage pool. The storage pool is a reinforced concrete pcol lined with 1/k-inch-thick stainless steel.
It is located inside the fuel-handling area in the auxiliary building. The auxiliary building, as well as the storage pool, is a seissic class I structure which is designed to with-stand seismic, tornado, and ther al loads as discussed in sections 3.7 and 3.8.
The spent-fuel storage racks are also seismic class I structures which are designed to withstand seis=ic leadings. The' mass =cdel is shewn in figure 9-25a.
The fuel-handling area is also prctected against ternado-generated missiles and other potential missiles.
Adequate shielding is provided for staticn personnel by the 5-1/2-foc;-chick concrete valls and berated water in the pool.
The radiation :cnes arcund the spent fuel pool are shown in figures 12-2 and 12-3.
1 The spent-fuel racks (not including the failed fuel centainer 1ccatiens) are 4
arranged in a 16 X h5 array constructed of six 7 X S =cdules and six 3 X 3
=cdules. The arrangement is shown in figure 9-3A and 9-33.
The 1ccation of I
the storage pool within the statien complex is shown in figures 1-6 and 1-7.
A separate space is provided for leading the spent fuel shipping cask. The spent fuel cask pit is independent of and separated frcm the spent fuel pcol by a 3-foot-thick concrete vall.
The cnly ecmmunication between the spent fuel pcol and the cask pit is thrcush the 2h-inch-vide slot opening previded for the transfer of the spent fuel assemblies frca stcrage to the shipping cask. This opening is prcvided with a vatertight bulkhead which can isolate the spent fuel pocl when needed.
Following sufficient decay, the spent fael assenblies can be removed frcm storage and loaded into the spent fuel l
shipping cask under water for re= oval frcs the site. Casks up to lh0 tens in veight can be handled by the spent fuel cask crane.
l A cask-wash-and-decentaminatien area is also provided adjacent to the cask pit.
In this area, cutside surfaces of the cask can be decentaminated before l
shipment.
9-9
MODIFIED i
D-B 9 1.2.3 Safety Evaluation The spent fuel storage facility is designed for noncriticality by use of adequate spacing and " flux trap" construction whereby the fuel assemblies are inserted into neutron abscrbing stainless steel cans. The spent fuel storage racks are designed to prevent accidental insertien of a fuel assembl,'
in Other than the prescribed locations,thereby ensuring a safe gec=etric array.
All spent fuel asse=bly transfer operations are conducted under a =ini=um of 9-1/2 feet of borated water above the top of the active fuel asse=bly.
All piping penetrations into the spent fuel pool penetrate at least 9 feet above the top of the fuel asse=blies to avoid any possibility of co=pletely draining the pool in case of a pipe rupture. Isolation valves are provided en the pipes penetrating the pool. These valves are located as close to the concrete vall as practicable to minimize the possibility of pipe failure between the isolation valves and the pool.
The spent fuel pool vater is cooled by the spent fuel pool cooling syste= as discussed in subsection 9.1.3.
The spent fuel cask crane is electrically interlocked to prevent the crane fro = traveling over the spent fuel pool while any load is hanging on the =ain hook. This interlock can be bypassed only with a key. Even upon bypassing this interlock, the =ain hook stays inoperative; only the auxiliary hook can be used.
The cask pit is separated and isolable fro = the pool to preclude the possi-bility of draining the spent fuel pool in case of damage to the cask pit by an accidental drop of a cask in the pit.
The base of the cask pit is solid concrete extending down to the foundation. Thus, a cask drop is not postu-laced to do any significant da= age to the structure.
The storage racks are designed to eitninate any pessibility of fuel 1ssambly sticking in the racks.
All projections and corners are pr:perly tapered and rounded ofr. The spent-fuel assemblies are placed into, and re=0ved frc=,
the racks by the spent-fuel handling bridge crane.
Since the fuel asse=blies
=ake free contact with the storage cells, there vctli be no uplift feree exerted on the racks. The spent-fuel-handling bridge crane is provided with an overload interlock en the hoist which shuts off the pcver to the hoist any ti=e the load en the hoist exceeds 2700 pcunds. The racks are designed o withstand this uplift force.
9-10
-=
MODIFIED s
D-3 9.1.3 SParT FUEL FCOL cc0Lno MID CLEXIUP SYSTDt 9.1.3.1 Design Bases i
The spent fuel pool ecoling systen is designed to maintain the borated spent fuel pool water quality and clarity and to remove the decay heat frca the stored fuel in the spent fuel pool.
It is designed to maintain the spent fuel pool water at approxi=ately 125 F, with a heat lead based en re=cving i
the decay heat generated frc= 1/3 of the core fuel asse=blies which are
- assumed to have undergene infinite irradiation and to have been eccled in the reactor for an. average of 150 hcurs prier to being stored in the pcci, plus the decay heat generated by the previous 11 batches frc= prict annual refuelings.
The decay heat re= oval syste described in section o.3 serves as a back-up system to the spent fuel pool cooling syste= under ner=al conditions and is used to re=cve the decay heat frc= the spent fuel pool shculd it be necessary to off-load the entire core into the spent fuel pool.
4 In additien to its prinary function, the spent fuel pool cooling system pre-vides for purification of the spent fuel pool water, the fuel transfer. canal water, and the centents of the berated water stcrage tank to re=cve fission and ccrresien prcducts and to maintain water clarity.
The radiation level and shielding are described in chapter 12.
9.1.3.2 syste= Descrittien The spent fuel pcci cooling syste= is shcvn in figure 9-1.
It consists of two half capacity recirculating pu=ps and two half capacity heat exchangers, asscciated valves, piping and instru=ents.
A bypass syste=
consists of a de=ineralizer and a filter, d
Syste perfor=ance data are shcun in table 9-2.
Majer ec=ponents of the syste= are briefly described belev.
4 i
9.13.2.1 spent Fuel Fool Heat Exchangers i
j The spent fuel peel heat exchangers are.~ designed te =aintain the temperature l
of the spent fuel pool vater noted in section 9,1,3,1, 4
i 9.1. 3. 2. 2 spent Fuel Pcol Pu=ps s
The spent fuel pool pu=ps take suction frc= the spent fuel pcol and recirculate the vater back to the pcol after it passes through the heat l
exchangers, demineri:er and/or filter in various ec=binations, depending on conditions.
i 9 1.3.2.3 spent Fuel Fool De=ineralizer The spent fuel pcol de=ineralizer can re=ove approxi=ately fifty l
percent of the fission products contained in the spent fuel pcol vater in 3h hours.
4 I
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~
MODIFIED D-B 9.1.3.2.4 Spent Fuel Pool Filter The spent fuel pool filter is designed to remove particulate =atter from the spent fuel pool water. The filter is sized for the same flow rate as the demineralizer (100 gpm).
9.1.3.2.5 Borated Water Storage Tank Recirculation Pump The borated water storage tank recirculation pump recirculates water from the borated water storage tank through the spent fuel pool cleanup system for demineralizing and filtering.
The pump may also be used for demineral-ization and filtering the water in the fuel transfer canal during a trans-fer of fuel.
The pump will be used for draining a portion of the refueling canal, fuel transfer pit and cask pit after completion of the fuel transfer operation.
During the winter, the pump will also serve to maintain the borated water storage tank temperature by circulating water through an external heater to prevent the tank water from freezing.
9.1.3.2.6 Spent Fuel Fool Ski =mers Surf ace skimmers are provided in the spent fuel pool to facilitate the re-moval of accumulated particulate matter from the surface of the spent fuel pool water.
9.1.3.3 Modes of operation 9.1.3.3.1 Normal Operation I
The spent fuel pool cooling system serves two =ain functions. The first is to r move the decay heat generated by spent fuel stored in the pool as a reau t of normal refueling conditions and, the second function is to provide purilication of the spent fuel pool water for clarity during fuel handling l
operations.
The first function is accomplished by recirculating spent fuel pool cooling water from the spent fuel pool through the pumps and heat exchangers and back to the pool.
The spent fuel pool pumps take a suction from the pcol and deliver pool water through the tubeside of two heat exchangers arranged in parallel back to the pool. The maximum normal heat load results with the pool filled with one freshly discharged batch in addition to 11 batches from previous refueling' odEages'.-~'41tE Ehis heat load'and both pumps and heat exchangers operating, the spent fuel pool cooling system is capable of maintaining the spent fuel pool at 125 F or less. With one pump and two heat exchangers operable the pool can be maintained at 140 F or less and with cne pump and one heat exchanger operable the pool can be maintained at 155 F or less under the maximum normal heat load conditions.
The second function is accomplished by providing a bypass purification system.
The bypass loop branches off from the spent fuel pool pu=p discharge cross-connect line, bypassing the heat exchangers. After decineralizing and filter-ing, the bypass flow is directed into the normal line downstream of the heat 1
exchanger and returns to the pool.
9-12
e MODIFIED p_3
%'pe y
The purificatien vill also be utili:ed to purify the water in the horated vater storage tank folicving refueling, and to maintain clarity in the fuel transfer canal during refueling. Water frc= the borated water storage tank or fuel transfer canal can be purified by using the borated water storage tank recirculation pu=p.
9 1.3.3.2 Abncrsal Conditiens The maxi =um abnormal heat lead would result chen ene entire cere (177 assen-blies) is discharged 150 hcurs after shutdevn, o5 days after the last of 9 batches frcm previocs annual refueling cutages. The unit is designed such that under these conditiens or other full core discharge conditiens, the decay heat removal system is used to remove the decay heat frem the spent fuel pool. One decay heat train alone is capable of maintaining the spen:
fuel pool at abcut ihG F cr less under the maximu= abnormal heat load des-
- cribed, s
i 9.1.3.k Reliability Considerations The spent fuel pool ecoling systen provides adequate capacity and ec=penent redundancy to ensure the reliable cooling cf spent fuel stcred in the spent I
fuel pool.
Ample time is available to ensure that eccling can be restcred even in the unlikely event of multiple cc=penent failures or cc=plete ecol-ing loss. The system is so arranged that no uncontrolled, cceplete less cf water frca the pcol is possible by piping er ec=penent failures. The syste=
perfor=s no emergency functions and is not directly connected to the reacter coolant syste=.
The decay heat re=cval syste=, which has a higher heat re=cval capacity, serves as a back-up syste= to the spent fuel pocl eccling syste=,
9.1.3.5 codes and standards Iach ec=penent of this syste= is designed to the ecde or standard, as applicable, as noted in table 9-1.
9 1.3.6 Fuel leakage consideratiens If a leaking fuel asse=bly is transferred frca the refueling canal to the spent fuel pocl, a small quantity of fissicn product activity =ay enter the spent fuel pool ecoling vater, even though the asse=bly's clad-ding temperature is levered, and leakage shculd be =inimized. The nurifica-tion loop re= oves these fission products and other conta=ina ts frc$ the pcci Radiological evaluatica is presented in Chapter 11 and Chapter 12.
vater.
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9-13
MODIFIED D-3 renoved and a thorough visual inspection is made of the esse =bly.
The fuel asse=bly is then placed into the new fuel elevator. The new fuel elevator is provided to lower the new fuel assembly under water, thereby aT'
- -a
g the need to lower the crane hook into water. After the new fuel assembly has been lowered to the bottee of the fuel transfer pit, it is picked up by the fuel grapple on the spent fuel handling bridge crane and placed into the fuel transfer =echanism. The tilting =echanism on the transfer =echan-iss rotates the fuel assembly frca a vertical to a hori: ental position. The transfer carriage transfers the fuel assembly through the transfer tube to the inside of containment.
Inside the coreainment a second tilting =echan-ism rotates the fuel assembly back to the vertical position. The main fuel handling bridge inside the containment removes the fuel asse=bly frem the transfer carriage and places it in the reactor.
The procedure for the re=cval of the spent fuel fro = the reactor is si=11ar to the one above in reverse order. The spent fuel assembly is re=cved frca the transfer carriage in the auxiliary building by the spent fuel handling bridge crane and is placed into the spent fuel storage racks for decay prior to off-site shipcent.
Once refueling is cocpleted, the refueling canal water is drained and pumped to the borated water storage tank.
9 1.i.3 Shi;cing Spent Fuel _
The spent fuel assemblies will be stcred in the spen; fuel pool prior tc their shipment offsite.
The spent fuel shippiEg cash can be received at the site either by truck or railroad.
Upen arrival, the cask, en the railrcad car (cr truck), is inspected fer any evidence of physical damage. The cask is then unloaded frem the railroad car (or truck) with the spent fuel cask crane and placed in the cask wash area. The cask is washed, scrubbed, and stea= cleaned
-=- ve a
all road dirt and grime. After thercugh cleaning, the lid en the cask is unbolted, re=oved and sected. The cask is lifted frem the wash area (Figure 9-29) and lowered into the cask pit.
If the cask pit is e=pty to start with it is filled with the borated water from the borated water storage tank to elevation 601 feet 5 inches. The bulkhead between the spent fuel pool and the cask pit is re=oved to establish co==unication betweer the two.
The spent fuel is new picked up from the storage racks by the spent fuel bridge crana and placed into the cask. Depending on the size of the cask, as =any as 10 spent fuel assemblies may be shipped in one cask. When the cask is fully loaded, still in the cask pit, the lid is placed on top of the cask to pro-vide shielding when the cask is lif ted out of the water. When the cask is partially out of water, two or three bolts are icesely installed to keep the lid in place. The cask is now lifted out of the pit and placed in the cask wash area. The cask is connected to a ecoling system for the re=cval of decay heat frcs the fuel asse=blies. After all of the head bolts are in-stalled and properly torqued, the cask is washed and decentaninated, and the surface radiatica level is checked. When it is below the Department of Trans-portation limits specified in h9 CFR Part 171-178, it is ready for ship =ent.
The 9-21 i
MODIFTED D-3 cask is then placed on the railroad car (or truck) and connected to its cool-ing system and shipped offsite to the reprocessing plant.
9.1.4.4 Safety Provisions Safety provisions are designed into the fuel handling system to prevent the development of hacardous conditions in the event of component malfunctions, accidental damage, or operational and ad=inistrative failures during refuel-ing or transfer operations. A mechanical lock prevents disengagement of the fuel assembly grapple latches as long as a fuel assembly weight is suspended from the grapple mechanism.
Bridge and trolley controls are interlocked to prevent movement until the fuel assembly has been completely withdrawn into the protective mast tube.
The new and spent fuel assembly stcrage facilities are designed for ncncrit-icality by use of adequate spacing and, in the case of the spent fuel racks, by use of a stainless steel " flux trap" design. The new and spent fuel stcrage racks are designed to prevent insertion of a fuel assembly in cther than the prescribed locations, thereby ensuring a safe gecmetric array. A safe conditien is ensured even if new fuel is immerced in unbcrated water.
Under these cenditicns, a criticality accident during refueling er s:crage is not credible.
All spent fuel assembly transfer operations are conducted under water. The water level in the refueling canal provides a minimum of 9-1/2 ft of water over the top of the active fuel in the spent fuel assemblies during movement from the core into storage. The depth of the water over the fuel assemblies, as well as the thickness of the concrete walls of the refueling canal, is sufficient to limit the maximum continuous radiation levels in the working area to values consistent with the radiation coning described in Chapter 11.
The spent fuel storage pool water is cooled by the spent fuel cooling system as described in section 9.1.3 A pcwer failure during the refueling cycle vill create no im=ediate ha:ardous condition owing to the large water volume in both the refueling canal and spent fuel storage pool.
During the refueling period the water level in both the refueling canal and the spent fuel storage pool is the same, and the fuel transfer tube valve is continuously open. This eliminates the necessity for an interlock between the fuel transfer carriage and fuel transfer tube valve operations except to verify full open valve position.
The simplified movement of a transfer carriage threugh the hori: ental fuel transfer tube minimizes the danger of ja= ming or derailing.
All operating mechanisms of the system are located in the fuel handling crea for esce of main-tenance and accessibility for inspection before the start of refueling operations.
During reactor operation, a bolted closure plate and gasket on the con-tainment vessel flange of the fuel transfer tube and the fuel transf er tube valve on the fuel handling area end of the tube provide containment vessel isolation as described in section 6.2.4.
Both the spent fuel storage pool 9-22
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