ML20091A273
| ML20091A273 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/17/1977 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20091A270 | List: |
| References | |
| NUDOCS 9105140471 | |
| Download: ML20091A273 (46) | |
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RETURN TO REGUUIORY CENTRAL ROOM 016 t MONTICELLO NUCLEAR GENERATING PLANT Docket No. 50-263 License No. DPR-22 l August 1977 1 l DESIGN REPORT AND SAFETY EVALUATION l l FOR REPIACEMENT OF SPENT FUEL POOL STORAGE RACKS i 4 Dctkd # gp ' A U B i 9105140471 770817 PDR ADOCK 05000263 P PDR pff gf j'/
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! TABLE OF CONTENTS i .
l 1.0 Introduction 1 2.0 Design Basis 2 3.0 System Description 2 3.1 Cene ral 2 3.2 Fuel Storage System Construction 7 4.0 Discussion and Evaluation 7 1 4.1 Criticality Analysis 7 4.2 Spent Fuel Pool Cooling and Heat Transfer Analysis 15 4.3 Seismic and Stress Analysis 21 ! 4.4 Mrterial Considerations 26 4.5 Radiological Cor.siderations 27 l 4.6 Fuel Storage Rack Replacement Safety 27 5.0 Environmentr.1 Considerations 27 5.1 Need for Increased Storage Capacity 27 i 5.2 Environmental Impact of Increased Storage Capacity 28 5.3 Environmental Impact of Postulated Accidents 37 5.4 Evaluation of Alternatives 38 5.'5 Evaluation of Proposed Action 43 & 6.0 Summary and Conclusions I G
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1.0 INTRODUCTION
l This design report and safety evaluation considers the installation of higher density, poisoned fuel storage racks in the existing I
. Monticello spent fuel pool. The control rod storage racks will be replaced in the process with racks of a different design for tempor-
, ary storage of rods. l . I The spent fuel pool currently contains racks sized to accommodate 1 740 fuel assemblies or about 1.5 cores. There are 484 fuel assemblies in the reactor core. It was originally assumed that about one
, quarter of the core would be discharged annually and that spent fuel 1
would be removed from the plant for reprocessing within approximately a year after discharge from the reactor. Because the reprocessing option is not available at this time, other alternatives have been , evaluated. It has been decided that the racks in the existing spent fuel pool should be replaced by new racks designed for higher density storage. An operating reactor should have the ability to temporarily unload the entire reactor core at any time. At the present time, this is not possible. It is therefore necessary to increase the storage capacity as soon as possible to avoid an extended outage should the
- need to offload the core arise. Installation of the first high density storage modules is scheduled to connence in April of 1978 to restore full core discharge capability and to allow the remaining existing racks to be emptied and replaced.
The high density fuel storage racks will provide 2,197 spent fuel storage locations. One of the existing fuel storage racks and two of the existing control rod / defective fuel storage racks will not be replaced, allowing storage of an additional 40 fuel assemblies for a total of 2,237 assemblies. This will provide full core discharge capability along with sufficient spent fue!. discharge capacity until the fall of 1987, assuming annual quartet cesre reloads. Recent trends in BWR fuel cycles have been toward smaller reloads (one fif th core) with greater burnups. However, for conservatimn, this evaluation cor.tinues to use the FSAR bcsis of annual, quarter core reloads, j Puture trends in BWR fuel cycles might incorporate greater than quarter core reloads with longer fuel cycles. Therefore, the analysis of spent fuel pool cooling assumes a discharge batch of 141 assemblies which would be characteristic of an 18 month fuel cycle. The following sections of this report describe the design of the high density fuct storage racks to be installed and contain a discus-sion of the environmental and radiological considerations of the installation.
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i I l
i i 2.0 DESIGN BASIS ~~ 4 . f The new spent fuelstorage racks were designed to conform to the j following crite-ia*: l 1) General Design Criterion 2 as related to components j important to safety being capable of withstanding the l
. effects of natural phenomena.
I 2) General Design Criterion 3 as related to protection against fire hazards.
- 3) General Design Criterion 4 as related to components being able to accommodate the effects of and to be compatible i' with the environmental conditions associated with normal operation and postulated accidents, i
i 4) General Design Criterion 62 as related to the prevention j of criticality by physical systems. l 5) Regulatory Guide 1.13 as it relates to the fuel storage { facility design to prevent damage resulting from the SSE
- and to protect the fuel from mechanical damage.
1 l 6) Regulatory Guide 1.29 as related to the seismic design classi-fication of facility components.
- l
! 3.0 SYSTEM DESCRIPTION l 3.1 General j The location of the spent fuel storage facility within the plant is
! shown in Figures 3.1-1 and 3.1-2. Drawings.with greater detail can be
- found in Section XII of the Monticello FSAR. The arrangements of the new
, fuel storage system in fuel storage pool is shown in Figures 3.1-3 and 3.1-4. l l Storage is provided for canned def ective fuel, ' used control rods and l spent fuel. Two existing racks are used for defective fuel cans or j control rods, each rack holding up to ten cans or control rods, as i needed. Up to 121 temporary spaces are provided for control rod
- storage on the periphery of the spent fuel pool. The spent fuel storage system is a modular design that provides storage spaces for fuel bundles or assemblies ** on approximately 6.5 inch center-to-center spacing. Each storage module has a capacity of 169 bundles.
I l *
- General Design Criteria per 10 CFR 50, Appendix A (General Design Criteria for Nuclear Power Plants) and USNRC Regulatory Guides a's noted.
l ** BWR fuel bundle consists of fuel rods held in a rectangular array by spacers and l tie plates. A BWR fuel assembly includes a fuel channel or shroud enclosing the ! fuel bundle. 2 , i
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3.2 Tuel Storage System Construction The fuct storage module is a fabricated stainless steel structure composed of f ael storage tubes, made by forming an outer tube ang an inner tube of 304 stainless steel with an inner core of Boral into a single f abricated tube. The outer and inner tubes are welded together af ter being sized to the required dimensional tolerances by a patented process. The completed 6torage tubes are f astened together to form a 13x13 storage module. Each 13x13 module is approximately 7 feet square and 14 feet high and provides storage space for 169 BWR fuel assemblics. The base plate of each storage module is raised above the floor of the fuci pool suf ficiently to permit natural circulation of cooling water to flow to the modules. 4.0 SAFETY EVALUATION 4.1 Criticality Analysis 4.1.1 The Principal Analytical Model The criticality analysis calculations were performed with the HERIT computer program. The MERIT program is a Monte Carlo program which solves the ncotron transport equation as an eigenvalue or a fixed source problem including the neutron shielding problem. This program is especially written for the analysis of fuel lattices in thermal nuclear reactors. Geometries with up to three space dimensions and neutron energies between 0 and 10 MeV can be handled. The HERIT program uses cross sections processed f rom the ENDF/B-IV library tapes. 4.1.2 The Model for Verification 4 The qualification of the MERIT program rests ulon extensive quali-fication studies including Cross Section Evaluation Work Group (CSEWG) thermal reactor benchtnarks (TRX-1, -2, -3, -4) the B&W UO 2 and Pu02 criticals, Jersey Central experiments, CSEWG fast reactor benchmarks (GODIVA, JEZEBEL), the KRITZ experiments, and in addition, comparison with alternate calculational methods. Boron was used as solute in the moderator in the B&W UO2 and B&W Puo2 criticals, and as a solid control curtain in the Jersey Central experiments. The MERIT qualification program has established a bias
, of .005 + .002 (lo) ak with respect to the above critical experiments.
Therefore, KERIT underpredicts k,gg by 0.5 percent ak.
- Product of Brooks & Perkins, Inc. Consisting of a layer of B C-Al 4 matrix bonded between two layers of aluminum.
7 ;
i 1 4 4 i 4 l 4.1.3 Assumptions i 4 The calculations were performed using the following apeamptions: 1 l 1. Maximum BWR fuel bundle neutron multiplication f actor * (k.) of 1.35 in standard core geometry at 20 0C. l . j 2. 8x8 BWR fuel.
- 3. Boron ( OB) equivalent go a gomogeneous eal concentration of d 0.013 grams (minimum) g B/cm
- 4. Cell pitch of 6.563 in.
4.1.4 Criticality Calculations 4.1.4.1 Cell Calculations i j The storage cell infinite multiplication f actor va9 calculated for 1 the high density fuel storage system as defined by the assumptions
; above and the geometrical layout shown in Figure 4.1-1. Table 4.1-1 summ:11res the results of the calculations. The maximum k. of a storage cell occurs at 20 00 with the fuel bundles centered and no ? flow channels present. Any variation such as increasing the pitch, j eccentric bundle positioning, reducing moderator density, and in- ! creasing the temperature to 65DC decreases the k. of the cell.
i a 4.1.4.2 Module Interaction Calculations The design of the High Density Spent Fuel Storage System has certain spaces on the periphery of the module fabricated with unpoisoned closure plates. At such locations it would be possible for several i fuel bundles to be stored opposite others with no intervening poison. 2 Criticality evaluations were made of this condition as defined by Figure 4.1-2 using various gap spacings. The results are shown in
, Table 4.1-2.
1 4.1.4.3 Assembly Drop Accident i The effect on reactivity of an accidental fuel assembly drop onto or adjacent to the High Density Fuel Storage System (HDFSS) vas considered for a number of postulated cases. The conclusions are presented in Table 4.1-3. 9 ,
- The use of a maximum fuel k. as a criticality base assumption eliminates the
, , multiplicity of U-235 enrichment and burnabic poison combinations and clarifies the exact condition considered. i 4 I I F i 8
1 4.1.5 Discussion of Calculated Results The maximum values f rom both " single cell" and " module interaction"
. calculations have been extracted f rom their respective tables to identify the bias and uncertainty that has been included. T he k.
and bias values are summed and the independent uncertainties at two standard deviations are combined by the root mean square method to arrive at the total. Case 2 - Nominal dimensions, without flow channel, 20 0C
- k. .8624 Calculation convergence ok i .0038 MERIT Bias & Uncertainty Ak .005 i .002 Model Bias 6 Uncertainty Ak _None Total * .8674 1 0086 (20)**
Case 7 - Minimum Module Spacing
.8526 k.
Calculation Convergence Ak i .0036 MEkIT Bias & Uncertainty Ak .005 i .002 Hodel Bias & Uncertainty Ak .0017 i .0051 Total .8593 i .0131 (20)** l l l
- This maximum k. of .876 corresponds to approximately .89 for existing racks
** 20 corresponds to 95% confidence level 9
9
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- FIGtHE 4.1-1 HDFSS Cell Configuration 4
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( , ! 1 < l 4 i FIGURE 4.1-2 ) RDFSS Rack Module Array l 1
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6 TABLE 4.1 SINGLE CELL HIGR DENSITY FUEL STORACE CRITICALITY RESULTS [ i i I i CASE DESCRIPTION b(+ 20)* l 1 Nominal Rack Dimensions ** .8668 + .0075 - With Flow Channel 6 20'C 2 Nominal Rack Dimensions .8674 +- 0036 j Without Flow Channel e 20*C ! 3 Some as Case 2 except 9 65'c .8561 + .0084 j i 4 Increased Fitch Vithout alow .8364 + .0106 - l Channel @ 20 C 5 Same as Case 2 But with .8276 + .0123 l Eccentric Bundle Position t 4
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6 Same as Case 2 But Uith Reduced i , Moderator Density : ! 1 0.9 g/cc .8470 + .0100 ! O.5 g/cc .7066 - .0071 - 0.2 g/cc l
.6059 + .0072 :
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- A includes MERIT Program and Geometry Bias and Uncertainty at a 95% Confidence Level l ** 6.563" Fitch With Nominal Material Thicknessess 2
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i 1 i i l i TABLE 4.1-2 l HDFSS Criticality Analysis i . i Module Interaction j - Cane Description km (i 2o)* l j 7 Minimum gap between I modules (2A - 1.244 in.) .8593 i .0131 l 8 Intermediate gap between modules (2A = 2.100 in.) .8579 i .0130 9 Nominal gap between mod-t ules (2A = 2.967 in.) .8506 i .0134 i 1
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- k,, includes program and geometry biases and uncertainties
; at a 95% confidence level 1
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I J TAB 1.E 4.1-3 liigh Density Spent Fuel Storage Syrtem , Assembly Drop Accident j Case Summary
) ,
No. Came Description Effect on Reactivity l j 1 A fuel assembly drops 18' inches Analysis indicates that localized tube
- vertically and impacts the top damage or fuel support member damage of a fully loaded itDFSS module. will occur, but neutron absorber material i The dropped assembly comes to will not be removed irom its por don a rest horizontally on top of the between adjacent fuel assembliet 4 fuel llDFSS. assembly resting horizontally att. 'he ilDFSS does not increase the syste eactivity;
; because the reactivity aseumes an ..afinite j vertical length of fusi (no neutron leakage 3
in the vertical dimension). ... k,gg <0.90.
- 2. A fuel assembly drops from 18 inches above the HDFSS, enters j Structural analysis indicates that local-an empty storage position and [ ized tube damage vill occur and one neutron i falls to the bottom of the absorber plate may be damaged. A reactivity storage position. ( analysis of this case with the neutron f absorber plate between two fuel assemblies j 3. A fuel assembly drops from 18 totally missing, stows that k,gg remains l inches above the HDFSS and less than 0.90.
strikes a tube wall at an obliqu j angle.
- 4. A fuel assembly drops from 18 It is not possible for a fuel assembly inches above the top of a drop of 18 inches to drive four stored i fully loaded module and str*.kes assemblies through the bottom of the module.
the upper tie plates of 2, 3 or Even so, the reactivity ef fect of this i 4 fuel assemblies in storage, impossible event was calculated as a l limiting value. An 18 inch section of f uel l in four bundles in an unpoisoned square array. was found to have a k,gg approximately equal ' to that of the system. There would be no l increase in the overall t'eactivity. keff
<0.90.
- 5. A fuel assembly drops from 18 This case was analyzed for normal handling inches above the !!DFSS falls conditions. . k gg e <0.90.
outside of the loaded llDFSS
. and lodges adjacent and parallel to an unpoisoned, occupied fuel storage position.
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)
I l The thermal analysis of the HDFSS considered two cases. First l 1 the normal cooling case considered refueling discharges, storage ! between regular discharges and full core discharge ... assuming j 1 - that the Spent Fuel Pool Cooling and Cleanup system is operable ] and the RHR system is available as required. Second, the impact j of loss-of-cooling was examined. Little dif f erence in the ability ; i* J to control pool water temperature was found between the Hb/SS , j results and those of the existing fuel storage system. [ t l 4.2.1 Normal Cooling Case I f The greatest cooling requirement on the Spent Fuel Pool Cooling and l j Cleanup System occurs af ter a refueling discharge. The following r j assun.ptions were used as, a basis for heat load determination. Discharge frequency 18 months * , ! Fuel assemblies per discharge batch 141 assemblies,
- Average burn-up 25.0 GVD/MT
! Specific power 30 IN/kg ;
! Cooling before discharge 72 hours ; j Time required to discharge 24 hours From dischargethese assumptions, batch using was determined to bethe 9.7ORICEN x 10 B Codeg This*TU/ hour.* the he is the l l value of the first peak on the Heat Load vs Cool Time curve in Figure 4.2-1. Ascanbeseen,thehegtloadfallsoffrapidly,andeventuallyreaches a minimum of 0.55 x 10 BTU / hour at the time of the negt refueling l discharge. The second discharge adds another 9.7 x 10 BTU / hour to the ! accumulated heat load and so on through twelve discharges to the HDFSS. l The incremental heat load attributable to the increase in fuel stogage
- capacity is that shown on Figure 4.2-1 by Note 1, 0.3 to 0.45 x 10 BTU / hour. This incremental load is the dif ference between 5 batches (as defined above) and 12 batches. Twelve batches will fill the llDFSS except i
for space for a full core discharge. l-The horizontal dotted lines on Figure 4.2-1 show the cooling capacity l of the Spent Fueg Pool Cooling and Cicanup System. Thissysgenwill j remove 2.87 x 10 BTU / hour from each of two loops (5.74 x 10 BTU / hour total) with entering pool water at 125'F, returning water to the pool ) at 99.6 F. Cooling water frca the Reactor Building Closed Cooling l Water System is assumed to enter the coolers at 95 F and exit 103.5"F.
- The 484 assemblies discharged to date have been on an annual basis. An 18 month j cycle and large discharge is used here to provide a bounding value for the evaluation.
i ** ORNL-4628 ORIGEN Code - The ORNL Isotope Generation and Depletion, M.J. Bell. 15
- . . - _ _ _ _ _ _ ~ - __ - _- - - -_ _ _-
Af ter a refueling discharge as shown in Figure 4.2-1, additional cooling irom the Residual liest Removal (R}tR) system, is required , for threg to four weeks untti the pool heat load decreases to
; 5.7 x 10 BTU / hour (in practice it has not been necessary to use the ! RitR system af ter past discharges, indicating the many conservatisms between the uaximizing heat load assumptions, above, and actual conditions). The RilR system has a capacity of 57.5 x 10 6 BTU / hour in each of two independent loops. This requiressent to use the RRR system is not af fected by the installation of the llDFSS because the l
heat load contribution from a f reshly unloaded refueling discharge j is controlling and the amount of load contributed by other fuel in ) the pool is insignificant.
- A full core discharge is evaluated with nomal cooling available and the llDFSS to show that there le little difference between high density storage and the present fuel pool storage system. Additional l assumptions for full core discharge at is
- 1. Full core discharge required 30 days following last refueling discharge and fills last 484 spaces in llDFSS.
- 2. Full core discharge is complete 150 hours siter shutdown.
0 The total heat cc.mputed with ORICEN for these conditions is 27.2 x 10 BTU / hour. The RitP cool'ing system will be required to cool a full core dis-chargewhetherthellDFSSisgnplaceornot. The incremental heat
, load difference of 0.45 x 10 BTU / hour is not a significant factor in the heat removal requirement. A graph of Ileat Load vs Cool Time -
is shown in Figure 4.2-2. it shows that approximately 150 days cooling are required to lower the heat load to the point where the Spent Fuel Fool Cooling and Cleanup system alone can handle the heat removal re-quirements. l During all of the above analyzed nonnal conditions of fuel storage the cooling or heat removal is controlled and is consistently the ! same as previously noted: pool vater temperature entering coolers, 0 I
< 1251; pool water return temperature, 99.6 F. Under these conditions, the maximum water temperature reached for the hottest bundle will be less than 115 F. Maximum cladding tesperature is 120.3 F and the maximum Boral temperature in the storage tube vill be 104.3 F. These temperatures are small relative to structural integrity or corrosion limiting temperatures. In no case does the cooling water approach boiling.
4.2.2 Loss of cooling case l The redundant, safety grade RilR system is available for ultimate cooling of the Monticello spent fuel pool, assuring that loss of cooling will not occur. Three conditions were analyzed anyway, to compare t.he temperature inertia of the 11DFSS and the existing systesa in approaching boiling. Condition 1 - IIDFSS with full core discharge
-Frevious normal discharge made 30 days prior to cooling system failure.
16
-Pool is full of fuel discharged under the assumed schedule (4.2.1) except for 464 slots reserved for full core dis-charge. Full core is then loaded into the reserved slots 150 hours af ter reactor shutdown.
-Pool is at 150 F at the tbne of failure.
- -Core irradiation levels are those specified on Figure 4.2-2.
(The core is assumed to consist of three normal 141 assembly batches with a remaining batch of 61. To maximize core exposure, the maall batch is assumed to be the one most recently replaced. To maximize the heat load of the most recent discharge batch in the spent fuel pool, it was assumed to be a 141 assembly batch. These inconsistent conditions are chosen to be conservative.)
-Pool is assumed to contain 2.45 x 106 lbgofwater.
-Heat load from fuel in pool is 27.2 x 10 BTU / hour.
Condition 2 - Existing Capacity with Pull Core DischarEe
-All conditions the same as in Condition 1 except only one nornal discharge batch is in the pool prior to full core discharge
-iteat load in pool is 26.5 x 10 6 BTU / hour.
Condition 3 - Existing Capacity without rull Core Discharge
-Pool cooling loops fail immediately after the 6th normal disc.harge batch is transferred into the pool.
-Pool is at 1500F at the time of failure. 6 / hour.
-Heat load at dtis time in the pool is 10.8 x 10 BTU The calculational method
- allows for evaporation from the pool surface, but does not allow for conduction through the pool walls since con-duction losses are negligible. Results of the calculation are plotted in Figure 4.2-3. Condition 1, representing 11DFSS capacity with full core discharge, results in the pool reaching boiling in about 5.6 hours. Condition 2, which represents the pool in its present configura-tion but with full core discharge, would result in boiling in about 5.8 hours. The difference between the two is negligible. Condition 3 which represents the present pool filled by refueling discharges, results in the pool boiling af ter about 14 hours.
4.2.3 Conclusions The HDFSS does not add significant cooling requirements to tiose t ha t exist. The present heat removal systems have adequate capacity
. to maintain the pool temperature at 125 F. The RHR system can be relied on as a backup cooling system with no need for consideration of loss of cooling.
* "How to Calculate Heat, Water Losses" - B.L. Thomas Chem. Eng. August 8, 1960, p 129 17 l
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4.3 Seismic and Stress Analysis 4.3.1 Analytical Criteria and Assumptions Analysis was based upon the criteria and assumptions given as follows: 4.3.1.1 ASMI Boiler and Pressure Vessel Code Section III, Subsection NT. 4.3.1.2 USNRC, Reg. Guide 1.92, Combining Modal Responses and Spatial Components in Scismic Response Analysis. 4.3.1.3 Honticello Final Saf ety Analysis Report, Appendix A Seismic Design Criteria. OBE - Operating Basis Earthquake SSE - Safe Shutdown Earthqucke 4.3.1.4 Honticello Reactor Building Seismic Evaluation of Spent Fuel Structure Bechtel. Inc. , Tecimical Report, Job No.12085, Rev.1, dated January 1977. 4.3.1.5 " Light-Cage Cold-Formed Steel Design Manual," 1961 Edition, American Iron 6 Steel Institute. 4.3.1.6 Acceptance criteria were based ont Normal and upset (OBE) Appendix XVII, ASME, Section III. Faulted (SSE) Para. F-1370, ASME Section III, Appendix F. Local buckling stresses in the spent fuel storage tubes were calculated according to " Light-Cage Cold-Formed Steel Design Manual" of American Iron 6 Steel Institute la lieu of Appendix XVII, ASME, Section III, because of its appliability to these light-gage tubes. 4.3.1.7 Stresses due to neismic loading in the X, Y and Z directions are combined by the Square Root of the Sum of the Squares Method. Eo " ' eX2 + n y2 + og2 4.3.1.8 The analysis assumed that the total mass matrix for each storage module consists of the structural mass of the fuel and the module plus the virtual or added mass matrix (hydrodynamic mass) due to included and displaced water and its resistance to movement. Added damping due to fluid ef fects were conservatively neglected. Virtual mass values were derived using WATER-01, a CE Proprietary, qualified level 2 computer program. 21
f l l 4.3.1.9 The base plate of each storage module is raised above the floor i of the fuel pool suf ficiently to permit natural circulation of cooling , water flow to the modules. Analysis has confirmed that frictional ! forces between module support and the floor and the low seismic , I overturning moment of the racks make them stable under all conditions of storage.
- 4.3.2 Analytical Methods Appropriate modeling of the fuel storage module was developed for ,
- each structural component and mass values assigned over the height in eleven mass nodes. The modules were combined into an idealized 8-module array and the pool wall was included to determine hydro-dynamic mass ef f ects. The modules vera analyzed as a cantilever beam attached to a rigid base, using DYSEA, a GE-developed version (qualified level 2) of SAP-IV modified to derive loads in a water -
filled rectangular pool. These loads were derived for the horizontal and vertical accelerations specified in the General Electric BWR Systems Department seismic criteria document and were compared to the allowable stresses in the reference documents. The analysis indicates that the derived loads do not overstress the module; thus, it can be concluded that the modules are not overstressed for the Monticello application < since the Monticello accelerations at the fuel pool elevation are 0.2g (SSE) and the analysis was done for 3g (SSE). Monticello accelerations are those in Ref. 4.3.1.4, Bechtel, Inc. > for the Monticello Building fuel pool elevations. Generic accelerations are those in GE Document 384RA137, Rev.1, "BWR-6 Seismic Design Specification". Table 4.3-1 compares the loads for the GE specified seismic accelerations I at the module f requency of 12.17 hz with those for Monticello. , Stress analyses were made by classical methods for both OBE and SSE conditions, based upon the shears and moments developed in the finite-element dynamic analysis of the seismic response. These values are compared with allowable stresses referenced in ASME Section III, Paragraph NF in Table 4.3-2. 4.3.3 Discussions of Results l 4.3.3.1 The lowest natural f requency of the 13x13 modules was 12.17 hz l for the primary mode. The natural frequency for the module and its supporting base will be specified later. , 4.3.3.2 Maximum displacement at the top of the modules fur X direction * - or T direction (the modules are symmetrical) is 0.07 inch. 22 -
i ! 1 i l 4.3.3.2 (Cont'd) Nominal spacing between modules is 2-in. so ao interaction between modules as a consequence of SSE is considered. , 4.3.3.3 The only applied loads to the module are the seismic loads. 'Ihese
- were calculated to be at the top of the fuel support members (bottom
! of tube). The loads in the X, Y and Z direction occur simultaneously.
- 4 Since the OBE loads are 490% of SSE loads and the OBE stress ,
! allowables (with the exception of the buckling.allowables) are ! $0% of SSE allowables, OBE is limiting and only buckling loads are 3 derived for SSE condition. i I Table 4.3-2 compares derived loadings for the seventeen analyzed I stress locations with the allowable values. In every case, the
- calculated stresses are lower than the allowables.
j 4.3.3.4 Thermal stresses were calculated and found to be insignificant. 4.3.3.5 The building floor loading was re-analyzed by the plant architect- !) F engineer and found to be acceptable per 4.3.1.4. t i l t i i i l i 5 a I i I 4 i l l ! 23 l
TABLE 4.3-1 COMPARICON OF SEISMIC ACCELERATION LOADS: GE SPECIFIED VS HONTICELLO OBE (8 = 0.02) Analysis-Basis Monti:ello (GE 384 RA 137)
! Vertical 1.17g + 1 grav. 0.09g + 1 grav.
1 Horizontal 1.5g 0.12g 5 P lb. (down) 3.5x10 1.71 x 10 5 Mx , Z in-lb . 2.1 x 10 1.68 x 10 0 l Vx, Z lb 1.81 x 10 1.45 x 100 i l _ SSE (B = 0.04) Analysis-Basis Monticello (GE 384 HA 137) Vertical 1.75 g + 1 grav. 0.12g + 1 grav. 1 Horizontal 3.0g 0.22g P lb (down) 4.4 x 10 1.8 x 10 Mx, Z in-lb. 2.25 x 10 2.57 x 10 6 ! Vx, Z lb. 2 gC 1 x 10 2.99 x 10 0 i ; I I J e f 24 4
TABLE 4.3-2 Comparison of Derived Loadings with A11ovables Stress for Allowable for Location / Type OBE OBE
, Tube Side Lateral 5,368 psi 18,150 psi Close Plate Lateral Bend 1,884 18.150 Close Plate Weld Bend 7,671 18,150 Tube, Local Shear at Bottom 3,165 11,000 Support Plate Bending 4,827 18.150 Support Plate Weld Bending 14,846 18.150 Support Plate Weld Shear 545 11,000 Bending of Fitting Lip 1,848 18.150 Bearing on Support Plate 854 24,750 Combined Compression on Tube Corner at Bottom 15.100 16,500 Combined Fitting Stress 4,886 16,500 Fitting to Base Weld 4,106 8,250(3)
Close Plate to Base Weld 4,185 10,000(1) Horit./ Vert Shear in Tube 7,720 11,000 Assy. Due to Bending SSE Tube local instability at 13,228 13,822(2) Bottom, Corner Tube for SSE for SSE NOTES
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(1) Allowable = 0.5 S g per Fig. XVII-2211 (C) of Ref. 4.3.1.1 (2) Allowable = 0.67T crit. Per Para. F-1370, Ref. 4.3.1.1 (3) Allowable = 0.3 S yper Fig. XVII-2211 (C) of Ref. 4.3.1.1 25 _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ )
1 i 4.4 Material Considerations 4 I All structural material used in fabrication of the new HDTSS is i type 304 stainless steel. This material was chosen due to its corrosion resistance and its ability to be formed and welded with consist ent quality. Boral*$1ates, used as a neutron absorber, are i an integral non-structural part of the basic fuel storage tube. I i ' These plates are sandwiched between the inner and outer wall of the storage tube and are not subject to dislocation, deterioration
; or removal, deliberate or inadvertent. The inner and outer valls of the storage tube are wrided together at each end, thereby isolating the Boral plates from direct contact with Spent Fuel Pool (STP) water. At Operating pool water temperatures there is no significant deterioration or corrosion of stainless steel or Boral.
Presence of the neutron absorber material in the f abricated fuel storage module will be verified by visual examination and dimensional ins pe ct ion. The tube design allows visual verification during fabrication that material exists in the space provided for place-ment of the neutron absorber material. The thickness of the Boral plates is nonstandard, providing a statistical significance betvcen , the thickness of the Boral and commercially produced aluminum or steel sheets. In addition, use of non Boral plates in fabrication of fuel storage tubes will cause tube deformation. Deformed tubes will not be accepted by inspection or by fabrication fixtures used for assembly of a fuel storage module. Dimensional inspection or the neutron absorber plate locations will be performed at the pool site. These data will be compared with dimensional results obtained during f abrication of the individual tubes and of the module assembly. A visual reinspection of each Boral plate location will also be pe rf o rmed. Acceptance of the above inspections will ensure that Boral plates are contained in the fuel storage module in sufficient , quantity and proper location to maintain the neutron multiplication f actor at or less than 0.90 with a 95 percent confidence level. Boral has corrosion resistant properties similar to standard aluminum sheet. Corrosion data and industrial experience confirm that aluminum and Boral have acceptable corrosion resistant properties
- for the proposed application. Although experience indicates that it is unnecessary, an inservice test program will be conducted, consisting of periodically removing and examining samples of Boral plate which have been suspended in the storage pool.
- USNRC Safety Evaluation for Yankee Rowe, dated 12/29/76 page 4. Structural and Material Considerations.
** Product of Brooks & Perkins, Inc. Consisting of a layer of B C-A1.
4 l matrix bonded between two layers of aluminum. 6 26
4.5 Radiological Considerations Section 5.2.3.4 includes a discussion of the environmental radio-logical c onsiderations. There are no significant radiological consequences identified which affect the health and safety of the general public. Occupational personnel exposure will be maintained as low as reasonably achievable and within the 10 CPR Part 20 limit for an occupational exposure of 3 rem per quarter. The re-sults of the conservative projections discussed in Section 5.2.3.4 are summarized below; actual do.es experienced at other facilities performing similar rack replacement projects have been substantially less than these projections. Annual Recurring Hon-recurring Occupational Dose Occupational Dose Task Individual Total Individual Total Rack Replacement 0 0 <3 rem <22 man-ram ruel Handling not distinguishable 0 0 Spent Resin Ibndling 0 0 0 0 4.6 Puel Storage Rack Replacement Safety Installation of the firt t high density storage modules is scheduled to ccenend in April of 1978, ruel will be emptied from the existing racks intclfl.e new modules so those sacks can be removed and the re-maining higa density modules installed. The rack replacement procedures vill be written in such a manner that loads veighing in excess of a fuel assembly will not be moved over new or spent fuel in storage. Administrative controls will also prohibit the the mov2 ment of such loads over fuel 5.0 ENVIRONMDUAL CONSIDERATIONS 5.1 Need for Increased Storage capacity Monticello cocnenced power operation in early 1971. Since that time there have been four refueling outages during which a total of 484 spent fuel assemblies have been discharged from the reactor, none of
, which has been shipped from the site. There are two reasons for the current need to increase the spent fuel storage capacity.
Full core discharge capability does not presently exist, creating an immediate and current need for the proposed increased storage capacity. While the inability to unload the core does not present a safety problem, it could result in an extensive economic penalty in terms of contingene.y plans and replacement energy costs during a plant outage awaiting core unload. At the present time, 484 of the 740 storage spaces in the spent fuel pool are filled, leaving only 256 27
i 1 .
< l open spaces. To unload the 484 fuel assemblies presently in the *
, core would require replacing 12 of the original 13 control rod storage racks with the interchangeable, standard BWR 20-element fuel storage racks. Af ter the fall 1977 refueling outage there l will be 616 spent fuel assemblies tu the pool and full core dis-charge will no longer be possible even if all control rod racks
, are replaced. It is therefore neceenary to expand the storage capacity as soon as possible to regain full core discharge capability with minimal exposure to a potential economic penalty.
Irrespective of full core discharge capability, increased spent 1 fuel storage capacity is required for continued operation of the Monticello plant at its Itcensed operating power level. After 1 the fall 1977 refueting outage there will be 616 spent fuel assemblies in the spent fut i pool. If another 121 assemblies are discharged in a fall 1978 quarter-core reload, the spent fuel pool will contain 737 assemblies. Operation could continug until late 1979 at which time the reactor would have insufficient reactivity to maintain full power operation, ne plant could conceivably be operated two additional years (to late 1981), assin without full core discharge c pability, if the 13 control rod storage racks are replaced with the 20-element fuel storage racks as are presently in use at Monticello. In either case the need for increased spent fuel storage capacity is necessary to allow continued generation of electrical power by the plant. 5.2 Environmental Impact of Increased Storage Capacity 5.2.1 Land Use The spent fuel storage pool for the Monticello facility is an in-
, tegral part of the existing reactor building structure. We pro- ; posed modification will not alter the external physical geometry
! of the spent fuel pool. We existing fuel storage racks will simply be replaced with racks allowing storage of fuel in a configuration of higher density, ne pool was intended and licensed for the storage of spent fuel. h is proposed modification will not alter
- the use of land but will make more efficient use of land already designated for spent fuel storage.
l 5.2.2 Water Use
. There will be essentially no change in plant water usage as a result
. of the proposed modification. We increased storage will add a small but relatively insignificant amount of heat to the spent fuel pool vster. As discussed in Section 4.2, under normal conditions the maximum pool water temperature will not exceed 125 F. We increase in water makeup atte!.buted to the modification because of
! increased spent fuel pool evaporation will be undetectable in the total plant water makeup requirement. l l 28
l 5.2.3 Radiological l 5.2.3.1 Radioactivity Released to the Spent Puel Fool Water The Spent Puel Pool Cooling and Cleanup System is designed to pro-vide cooling for decay heat removal and to maintain pool water clarity and purity. The cooling capacity is discussed in Section 4.2 of this report. Cleanup of the pool water will be discussed in this section. The Spent Fuel Pool Cooling and Cleanup System consists of circulu-ting pumps, heat exchangers, filter demineralizers, piping, valves and instrumentation. Water is skimmed from the top of the spent
' fuel pool by fixed orifice plates on the east side of the pool, flowing by gravity to the skimmer surge tanks. The pumps take suction from the surge tanks and pump water through the heat exchangers and filter demineralizers before discharging it through diffusers at the bottom of the spent fuel pool.
Filter demineralizers remove collodial sized materials such as copper, iron, and non-reactive silica as well as exchanging ions from the resins for undesirable ions in the spent fuel pool water. Each backwash process generates approximately 5.5 cubic feet of 4 solidified waste. A backwash is required when the pressure drop acroso a filter demineralizer unit reaches 25 psi or when the effluent conductivity increases to 1.5 umho/cm. The spent fuel pool water volume is 40,000 cubic feet. The capacity , of each cleanup pump and filter dumineralizer is 450 gpm resulting in a processing time for the entire pool volume of six hours with boch trains in service. During refualing outages two trains are generally operated continuously except for weekly backwa.hing of the
- filters. After the refueling outage, only one train is generally operated with monthly backwashing. The main reason for operating the t cleanup system late in an operating cycle is to remove dust which settles on die pool surface and maintain low turbidity. An equili-brium concentration of radioactivity in the spent fuel pool water occurs within two to three months following completion of fuel handling.
- Radioactivity and turbidity in the spent fuel pool water are caused, in part, by the release of crud from fuel assemblic; and their channels.
Operating experience has shown that the greatest quantity of crud is released when loose deposits are shaken off during handling, particulsely when diey are first rasoved from the reactor. Once the handling of
. spent fuel assemblies is finished, the concentration of particulates rapidly decreases to a low background level.
The presence of defective fuel in the spent fuel pool also contributes to the level of radioactive contamination. The initial Monticello core experienced a greater number of fuel defects than expected, re-sulting in an orderly program to prematurely discharge the fuel from the reactor. All initial core fuel is presently in the spent 29
- i i
l fuel pool. No defective fuel ht; been observed to date in the four subsequent reload batches. Experience shows that the radio-i activity in the water attributable to defective fuel has masked the contamination resulting from crud. The reverse is expected in the future. i Most solid and gaseous fission products are retained in the ceramic UO2 pellets within the zircaloy encapsulated fuel rods. Some of the gaseous itssion products are generated by fissions occurring on the pellet surface and are not restrained within the 1attice structure of the fuel pellet. These gases generally migrate ' to an upper plenum area in the fuel rods. During operation the thermal expansion of the fuel pellets and the heating of plenum gas pres-l surizes the fuel rods. The fission gases, along with a small quantity 1 of solid fission products may be released to the coolant when a l cladding perforation develops. As the initial pressure of a dis-l charged defective fuel rod decays, and as the decay heat decreases, , contamination no longer vents to the spent fuel pool. There are no water currents sufficient to cause teaching of the fuel pellets. Data presented below support the conclusion that spent fuel pool i contamination from defective fuel is significant only for a short period of time following removal from the reactor. The majority of
; the radioactivity attributed to fuel defects is 1-131 which is very soluble in water. Because of its relatively short eight day half-life the iodine activity decreases very rapidly.
Experience at Monticello over the last four years supports the conclusion that increasing the period of storage will not increase the radio-activity in the spent fuel pool water. Shortly after each refueling outage the radiation field over the pool surface has returned to an apparent equilibrium of one ar/hr; there has been no detectable trend showing an increase in radioactivity in the water resulting from the increased inventory of fuel in storage. Typical radioactive isotope . concentrations in the spent fuel pool water are presented below to compare recent data (5/18/77) to that taken after 3 weeks of fuel l handling during the last refueling outage (10/6/75). Fuel Pool Activity (uc/ml) 4
- Isotope 10/6/75 5/18/77 cc-141 5.7x10-4 3,3xto-6 5
ce-144 --- cs-134 1.8x10~3 6.4x10 1.9x10-4 3 4.2x10-4 co-137 Mn-54 2.1x10 1.tx10- 4 2.3x10-5 Zr-65 2.0x10-4 5.1x10-5 Co-60 --- 3.4x10-4 Nb 95 1.3x10'3 --- 30
I The half-lives of the above isotoses are relatively long with respect to the period of fuel handling; the 10/6/75 data are therefore represen-tative of the maximum concentrations expected during a refueling outage. The relatively short half-life 1-131 (eight day) is the major contributor of activity from defective fuel. It measured 7.Ox10-4 ue/ml on 10/6/75, only a quarter of that measured on 9/20/75. The residual 1-131 is currently near the detection thres-hold of about 10-6 ue/ml. The proposed modification will not affect the amount of radioactivity released to the spent fuel pool water. The majority of the radio-activity released to the pool water occurs at the time fuel is discharged from the reactors the proposed modification does not affect the size of frequency of fuel discharges from the reactor. Release of crud is only dependent on the amount of fuel handling and not the length or time in storage. The radioactivity of crud decreases with time, resulting in a decrease in radioactivity released to the pool water if handling is deferred. Release of radioactivity from defective fuel occurs prior to and shortly after discharge and is not affected by long term storage or handling. All fuel presently in the fuel pool will experience an additional movement because of the rack replacement orogram. If the fuel were loaded into a shipping cask rather than a storage rack, as was expected would be the case when the plant was licensed, the same potential for crud release would exist. Crud dislodged during ultimate rem > val of fuel presently in storage will be an incremental increase in the amount of radiation released to the water. Crud dislodged from fuel discharged after the modification will have a net incrementat. decrease in radioactivity released to the water since it will have up to 12 additional years for radioactive decay before handling. The number of tbmes a fuel assembly is moved at the plant site is not affected by the length of time it is in storage. The overall effect is that no net change is expected in the amount of radioactivity released to the spent fuel pool water as a result of the proposed modificati3n. 5.2.3.2 ' Adioactive Material Released to the Atmosphere The spent fuel pool water surface is at approximately the same level as the refueling floor in the reactor building. Ventilation of the spent fuel pool area is accomplished by the Reactor Building Ventila-tion System and the Stand-by Gas Treatment System. The Reactor Building Ventilation System provides filtered, tanpered outside air to all areas of the reactor building. Air flow is
, directed from areas of least potential for radioactive e ontamination to a monitored plenum from which it is released through a vent stack.
, Ventilation flow from the spent fuel pool goes directly into the plenum.
The reactor building is maintained at a slight negative pressure such that any building Icakage is drawn into dhe building and exhausted through the vent stacks such that all releases can be monitored. The Reactor Building Ventilation System is designed to provide one air volume change per hour. If high radiation is sensed by the reactor building ventilation monitors, the ventilation system isolates and 31
l the Standby Gas Treaunent System is initiated automatically. This system is designed to provide one air change of the reactor building each day. 'Ihe exhaust passes through charcoal filters to the main plant stack, significantly reducing the potential offsite dose. Contamination from the spent fuel pool would have to enter the ventilation system in the form of man 11 airborne crud particles swept away from the water surface, amall fission product particles from defective fuel swept from the water surface or gaseous fission products from the defective fuel. Experience has shoun that the fuel pool demineralization systs effectively removes particulates over the first few weeks following discharge of spent fuel from the reactor. As discussed in Section 5.2.3.1, telease of radioactive crud to the water is not affected by the proposed modification. Therefore the ebility to store spent fuel for a longer period of time as afforded by the new rack design will not result in an increase in die maount of radioactive material retrased to the environment due to crud particles. Noble gases generated in the fission process renain sealed in the fuel rods. If fuel rods are defective, the gases are released to E the environment by way of the ventilation system. With the exception ( of krypton-85, these noble gases have short half-lives and decay to negligible amounts in a short period of time. The short half-live gases originally present will have decayed to insignificant amounts during the two years allowed by the existing storage capacity; increasing the 1e:.gth of time during which spent fuel resides at the site will not significantly affect the inventory of shert-lived noble gases in spent fuel. With respect to the 11 year half-life Kr-85, it only can be released to the atmosphere if it escapes from a fuel rod through a cladding perforatica. Despite the presence of defective fuel at Monticello, Kr-85 has never been observed in detectable quantities in the ventilation system. Section 5.2.3.1 concludes that the nochanism of leakage from defective fuel is such that it only persists for a short period of time following discharge from the reactor. Increasing die quantity and duration of fuel storage will therefore not have a noticeable effect on the release of radioactive material to the atmosphere in the form of Kr-85 or short-lived noble gases released from defective fuel. 5.2.3.3 Radioactive Solid Waste Generation
~
. As discussed in Section 5.2.3.1, the amount of fuel handling requiref.
each year is not affected by the proposed sodification. Therefore, there will be no increase in the annual solid waste generation in the form of spent fuel pool demineralizer resins. The rack replacanent program will require movement of spent fuel from dhe existing storago racks to the racks proposed to be installed. The incranental increase in crud, considering daat fuel would otherwise be loaded into a shf eping cask as discussed in Section 5.2.3.1 if the proposed modification were not pursued, is equivalent to that released 32
A l l during the final site handling of the 616 assemblies expected to be in the old racks at the time new storage modules are installed. It is conservative to assume that the resin consumption resulting from the transfer of fuel to the new racks is no greater than that during a refueling outage. This one-time resin consumption is conservatively estimated to result in 80 cubic feet of solid waste or about 0.02 )
- percent of the projected solid waste generation for the life of the
. plant. Because of the radioactive decay of crud prior to its release from the fuel, as discusaed in Section 5.2.3.1, this incremental increase is restricted to the volume of spent resin only; there will be no measurable change in the radioactivity in spent resins consumed as a result of the proposed modification despite the increase in volune.
The existing fuel storage racks a re planned to be dispored of as low level activity solid waste. This is a one-time activity in the life of the plant which involves approximately 10,000 cubic feet of solid vaste or 2 to 3 percent of the projected solid waste generation for the life of the plant. 5.2.3.4 Occupational Exposure - There are three tasks which contribute to personnel exposure which are to be considered replacement of the existing racks with high density racks, normal refueling operations over a larger inventory of spent fuel and handling of spent fuct pool cleanup resins.
- 5. 2. 3. 4.1 Radiation Exposure During Rack Replacement Rack replacement activities. will be done using techniques designed to maintain the occupational dose as low as reasonably achievable.
Plans are generally to work remotely from above the fuel pool; however, it may later be determined that the objective of minimizing exposure may be better accomplished using divers. The existing racks will be unbolted from the fuel pool floor using long-handled tools. Workmen will stand over the pool, utilizing the pool water as shielding from any contaminants on the racks as well as from - irradiated fuel in the unaffected racks. The racks vill be decon-taminated during and following removal by hosing, hydrolasing or scrubbing prior to preparations for off-site shipment. The high density racks will be installed using tools remotely controlled by workmen standing over the pool, again using the pool water as shielding. The various tasks associated with the relocation of fuel in storage, the removal and disposal of the existing racks, and the installation of the new racks were evaluated. The time required for each task and the radiation Icvel in which each task is to be performed was conserva-tively estimated based on past experience of similar tasks. The total non-recurring exposure is conservatively estbaated to be 22 man-rem. The projected vot"k force includes eight construction workers and three supervisors. The majority of the exposure will be taken by construction workers. The maximum individual dose will be under three ram using conservative projections. Realistically, the maximum individual dose is not expected to exceed 2 ram. 33
he total exposure of 22 man-rem for the proposed action must be compared to other non-routine plant activities to put it in the proper perspective. One such comparison is the 1975 feedwater sparger replace-ment project wherein 474 man-rem was accumulated. Even the conservative projection for exposure from the proposed action is not excessive relative to other activities in the industry. 5.2.3.4.2 Radiation Exposure During Fuel 11andling Operations Personnel exposure during fuel handling operations over tt.e spent fuel is due to radioactivity in the spent fuel pool water and shine from the radioactive fission products contained in spent fuel in s torage. Section 5.2.3.1 concludes that there will be no increase in the amount of radioactivity released to the spent futt pool water. Therefore, there will be no increase in personnel exposure during fuel handling operations from this source. The perronnel exposure attributed to shine from irradiated fuel assemblies in the spent fuel pool is a function of the time spent by personnel in the vicinity of the fuel, the source strength, and the attenuation of radiation between the source and the receptor.
%e time spent in the vicinity of spent fuel will not be affected by the presence of the high density fuel storage racks. No time consuming activities take place in the areas beside and below the spent fuel. This fact, along with the fact that radiation levels are low in those areas due to the water shielding and the effective shielding of the thick concrete pool wallo, results in a negligible exposure relative to that incurred dutin', operations over the fuel pool.
The radiation level over the spent fuel pool where personnel are located during fuel handling is the combined shine from radioactivity j in the fuel pool water and the radioactivity in the irradiated fuel assemblies in storage. Despite the high level of radioactivity in the irradiated fuel, self-shielding by the fuel and the 21 feet of water between the top of the stored fuel and personnel working over the pool provide sufficient shielding such that shine from the fuel contributes a smaller dose than the radioactivity in the fuel pool water. he radiation shine over the spent fuel pool is greatly reduced by the self-shielding inherent in the geometry of the fuel assemblies and the storage racks. Since the uranium fuel serves as a very good
. radiation shield, a substantir.1 fraction of the radistion originating
.within an irrediated fuel assembly is absorbed in the assembly. he radiation level over the fuel pool is therefore approximately related to the visible surface area of irradiat.ed fuel in storage. If fuel is stored in a very sparse array the addition of fuel will cause an increase in the radiation level over the fuel pool because of the increased source term with relatively little assembly-to-assembly shielding. As the density of fuel in storage is further increased, the additional source term is counteracted and becomes of little significance as assembly-to-assembly shielding tends to reduce the 34
4 radiation level existing over the pool. The proposed storage racks are designed having the maximum practical density. Calculations performed for odher similar situations have verified that due to self-shielding, the radiation level over a spent fuel pool vill decrease as the density of fuel assemblies in storage increases. . 1 Realizing that calculations of personnel exposure attributed to shine from irradiated fuel in the proposed high density storage racks could be expected to show a decrease, an effort was made to define an empirical, conservative upper bound of the exposure with i no credit for self-shielding. The radioactivity of fission products in discharged spent fuel assemblics decays at the half-life of the respective isotopes. Approximately ten days are generally required to remove the spent fuel from the reactor. The radiat ion from the freshly discharged batch is a factor of five greate r than the batch discharged the previous year and a factor of 25 greater than that discharged seven to eight years earlier. The FSAR considered two annual quarter core discharge batches in the spent fuel pool. The reracked pooi would allow storage of 12 such additional batches. Because of the radioactive decay, the source term of the 12 Long term batches is only about 50% of that of the two recent batches. (It is conservative to assume the actual time period of 10 days for the removal of spent fuel from the reactor, rather than the minimum tLac assumed for analyzing the adequacy of the Spent Fuel Pool Cooling System, j Radioactive decay of the discharge batch is substantial during these ten days whereas it is quite insignificant for the older batches. Using the actual ten-day coo'down period therefore results in a 4 realistic interpretation of che data.) In the past the radiation level over the spent fuel pool has generally reached 10 mr/hr early in the refueling outage but decreased to one ' mr/hr by the end of the outage and for the duration of the cycle. The decade decrease is not attributed to radioactive decay of the spent fuel in storage since 30 days after shutdown the deca) power is still 70% of that at ten days. Rather, the prLae contrilutors of the spike and the decade decrease are considered to be crud acti-vation products and fission products from defective fuel which are removed by dhe cleanup system or decay rapidly as is the case for I-131. The residual one mr/hr that remains after cleanup of the fuct pool water and I-131 decay is used as a conservative upper bound for direct shine from the spent fuel. Increasing dbe source teca by 50% will rtise the upper ibnit of the shine dose to 1.5 mr/hr over the spent fuel pool. (This is not to suggest that a 50% increase is expected or that it is even a realistic upper bound. It is used to 4
' show with existing data that even an ultra-conservative, upper bound does not result in a significant incremental increase in occupational exposure.)
Activity over the spent fuel pool requires approximately 3,000 man hours annually. An incremental increase of up to 0.5 mr/hr will i result in a total of up to 1.5 man rem / year averated over the 30 men involved, or an average of 0.05 rea/ year each. These numbers, although very small already, should be recognized as conservative, 1 upper bounds that completely disregard self-shielding due to higher 35
i
)
density storage. It should be noted, for comparison, that the average exposure to plant operator in 1976, a year in which there was no refueling outage, was one rem. Since the unrealistically conservative outer bound itself is a small fraction of this exposure, the effect of the shine from additional fuel in storage is not expected to be disting-uishabic. The effect of the proposed action due to shine from spent fuel in storage is therefore insignificant. 5.2.3.4.3 Spent Resin Handling Radiation Exposura As discussed in Section 5.2.3.3. the fuel storage rack replacement will not result in an increase in the radioactivity in resins consumed. Therefore there will be no diange in personnel exposure from spent resin handling attributed to this modification. 5.2.4 Nonradiological Effluents There will be no change in the chemical or biocidal effluents from the plant as a result of the proposed modification. The only potential offsite nonradiological environmental impact that could arise from this proposed action would be an additional dis-charge of heat to the local atmoophere and to the Mississippi River. An evaluatb n of the augmented spent fuel storage facility i was made to determine the effects of the increased decay heat genera-tion on the environment in the vicinity of the plant. (This decay heat will uitbnately enter the environment in any situation; the only question is whether this occurs at the Monticello Plant, a reprocessing plant, an alternate storage location or a vaste re-pository.) The increased heat load due to the ability to store spent fuel for a longer period of tbne after dischage from the reactor is calculated in Section 4.2 to be 0.45 x 100 BTU /hr. The design heat goad of the Monticello condenser at maximum turbine load is 3.7 x 10 BTU /hr. This single load makes up the najority of all heat rejected from the plant to the environment. The incre-mental increase of 0.45 x 106 BTU /hr resulting from the increased spent fuel storage capacity is an 0.012% increase. This change will be indistinguishable in the environment. 5.2.5 Impacts on the Consunity The new storage racks will be fabricated offsite and shipped to the plant. No environmental impacts on the environs outside the reactor o building are expected during the rack replacement. The impacts within the reactor building are expected to be ILmited to those normally associated with metal working activities. No significant environmental
- impact on the cmamunity is expected to result from the rack replace-ment or from subsaquent operation with the increased storage of spent fuel in the spent fuel pool. The connunity will experience indirect l economic benefits !.n that the construction force will be drawn from l the labor market of the area.
l 36
4 0 5.2.6 Transportation and Handling Delivery of material for the new high density storage racks and dis-posal of the existing racks for offsite burial will involve truck and/or rail transportation activity. The number of such shipments will be less than would be required to ship the spent fuel offsite at this time. By deferring offsite shipment of spent fuel a number
. of factors can be considered that will reduce the overall enviromnental impact: more fuel might be loaded per shipping cask, reducing the number of miles in transport; a lighter shipping cask may be used, reducing the tonnage in transport; the reduced radiation level of spent fuel vill furthe r reduce the already minimal environmental impact of spent fuel shirments which are covered by the Final Envinammental Statement.
As long as fuel is stored on site, transportation and associated fuel handling is climinated and no environmental impact from transportation or handling results. Although the environmental impact of trans-portation to a storage facility is slight, even diis Lapact is not justifiable if it can be avoided. Further, onsite storage eitminates the possibility of double shipments which would result from shipping the fuel to an interim point offsite and then to a terminal point such as a reprocessing or final disposal facility. Therefore, the proposed action avoids the slight environmental bmpact of transportation associated with offsite storage facilities. 5.3 Environmental Dnpact of Postulated Accidents The design basis accidents evaluated in the Monticello FSAR have been reviewed. Neither the consequences nor probability of occurrence of those postulated events is affected by the proposed modification. A fuel handling accident is postulated to occur over the reactor core where fuel is handled at a greater distance above other fuel than in the spent fuel pool, resulting in the more severe consequences. This con-dition will not be changed. The proposed modification does not affect the movement of fuel over the reactor core and therefore does not change the probability of this design basis accident. The Monticello Final Environmental Statement considers the consequences of three spent fuel handling accidents using the assumptions set forth in the Annex to proposed Appendix D of 10CFR Part 50. They are each discussed in the following sections. The proposed modification does not affect the assumptions on which each of these postulated events is based and does therefore not increase the consequences or probability
, of the analyzed events.
5.3.1 Fuct Assemb'.y Drop in Fuel Storage Pool It is assumed that one row of fuel pins in a freshly discharged fuel assembly is damaged in this event, releasing fission products to the pool water. The release would ultLmately enter the environment via the Reactor Building Ventilation System. The FES concludes that less th.n 0.1 percent of 10CFR20 lLnits will be experienced at the site boundary and about 0.18 man-ren whole body exposure will be experienced by the population within a 50 mile radius of the plant. The proposed 37
1 s
- l modification does not alter these numbers. The probability of the
; event will not be increased because the amount of spent fuel discharged from th* reactor remains unchanged.
5.3.2 Heavy Object Drop onto Puel Rack i For this event, the FES assumes that all fuel pins in a freshly dis- ! charged fuel assembly are damaged and release fission products to the . pool water. The release would again ultimately enter the environment I via the Reactor Building Ventilation Systou. The calculation concludes } that the dose would be less than 0.1 percent of 10CF12010mits at the j site boundary and about 0.34 man-rem within a 50 mile radius of the plant. l There are no loads greater than the weight of a fuel assembly that j need to be routinely moved over spent fuel. Installation of new fuel storage racks will be done in a way that does not requite them to be moved over spent fuel. It is an established practice that l heavy objects such as refueling shield plugs which are routinely handled on the refueling floor are not moved over spent fuel. The i consequences and the probability of this postulated event are therefore not altered by the propcsed modification. 5.3.3 Fuel Cask Drop . The postulated fuel cask drop event considered in the FES assumes all the fuel in a shipping cask ruptures and rele.ses fission products to the environment. The fuel is assumed to have been discharged from the reactor only 120 days prior to the accident. By allowing for longer storage of spent fuel, the proposed action will greatly reduce the consequences of this event. The rack replacement does I not in any other way affect the consequences or probability of a fuel cask drop. Since the FES was prepared, the reactor building crane has been
- modified to bnplement single-failure proof features which comply with l Regulatory Ouide 1.104 as applicable to operating plants. This change t was reviewed and found acceptable by the NRC Staff. Implementing these features has reduced the probability of the fuel cask drop i event discussed in the FES.
l 5.4 Evaluation of Alternatives Alternatives having the potential to alleviate the current need for additional spent fuel storage capacity were evaluated. The evaluation considered the availability, the benefits, the environmental impact, and the cost which ultimately affects the cost of electrical i s power to customers in the NSP service area. The alternatives are compared in the summary to this section. 5.4.1 Reprocessing
, There is preser.tly no licensed nuclear fuel reprocessing plant in the United States. In addition, on April 7,1977, President Carter issued a statement outlining his policy on continued development of nuclear power in the United States. He said, "We will defer indefinitely l
38 l
- e the commercial reprocessing and recycling of plute ntum produced in a the U S nuclear power programs. From our own experience, we have l concluded that a viable and economic nuclear power program can be sustained without such reprocessing and recycling."
Because of this indefinite deferral, reprocessing is not considered
- an available alternative for the current need for spent fuel storage at Monticello 5.4.2 Throw-away Fuel Cycle There is presently no national policy on the disposition et spent fuel and
, there is no repository for spent fuel from commercial nuclear reactors. The throw-away fuel cycle is therefore not an available alternative to the current need for increased storage capacity of Monticello spent fuel. The national policy and the repository design that ultimately evolve are expected to require a longer time period for radioactive decay than allowed in the present fuel pool. For this reason, the throw-away alternative will never fully replace the Monticello need for increased storage capacity. 5.4.3 Store Spent Fuel at Another Reactor Facility Northern States Power Company's other nuclear plant, Prairie Island, is a dual unit PWR which has experienced the same need for spent fuel storage as Monticello. Prairie Island is in the process of in-creasing their storage capacity based solely on their own needs. Since the PWR fuel assemblies are larger than BWR assemblies, storing Monticello fuel at Prairie Island would make inefficient use of those storage spaces designated for Prairie . aland. Furthermore, Prairie Island is not licensed to receive spent fuel. Approximstely half of all operating reactors in the United States have announced plans to increase their on-site fuel storage capacity. There is no known reactor facility that is licensed to receive Monticello fuel and/or has an interest in providing long term spent fuel storage. There is no reason to believe that this situation will change in the future. A May, 1977 ERDA Report, "1977-1986 LWR Spent Fuel Disposition Capabilities", shows that the spent fuel storage problem is industry-wide (rather than affecting only Monticello) when it concludes as follows: "With domestic nuclear fuel repcocessing indefinitely deferred, and the probable nuclear power growth past 1986, all of the planned storage capability will be needed. If the
' plans of GE for storage basin expansion and the plans of Exxon for a
, new basin, or equivalent plans, do not materialize, there will be a significant number of reactors that will not have storage capabilities for scheduled discharges." Storage of Monticello spent fuel at another reactor facility is therefore not considered an available alte rnative.
r 5.4.4 Existing Independent Spent Fuct Storage Installations (ISFSI) Three potential ISFSI sites in existence were considered. 39
, + Nuclear Fuel Services (NFS) has announced that it has withdrawn from the fuel reprocessing businces. Existing contracts for the reprocessing of spent fuel are under question, and new contracts are not being i offered. NFS is not even accepting fuel for storage from reprocessing
- customers. Northern States rower has no contract to store or reprocess Monticello fuel with NFS.
The Allied General Nuclear Services (AGNS) reprocessing plant presently under construction is not yet licensed to receive fuel for storage or reprocessing. Utilities with contracts of AGNS fuel reprocessing presently have spent fuel in storage in excess of the AGNS capacity.
; Spent fuel is being generated at a faster rate than the AGNS re-processing through-put, leaving a net requirement for additional storage. No contract exists, nor does the possibility exist for making such arran8cments for the storing of Monticello fuel at AGNS in the foreseeable future.
The Morris Operations owned by General Electric was planned as a re-processing plant. General Electric has since withdrawn from the reprocessing business and operates the facility as an ISFSI. The unused space at Morris Operations is less than half of the amount required to store the spent fuel which will be discharged from domestic reactors throughout 1977. If Horris space is used up indiscriminate 1y, there v111 be no contingency storage available to the nuclear industry. Originally, under a 1967 contract with General Electric for the supply of nuclear fuel for operation of the Monticello facility .it was con-templated that spent fuel would be shipped to General Electric, which in turn would undertake to make reprocessing arrangements. However, that contract did not encompass spent fuel now scheduled for discharge beyond 1983. Moreover, legal questions have becu raised as to the extent of General Electric's obligations with respect to spent fuel 4 under that contract in the absence of reprocessing capability in this country. As a result, for the purpose of minimizing the possibilities of plant shutdown in the future due to lack of spent fuel storage space, NSP entered into a supplemental contract with Gent- . Electric with respect to the interim torage of spent fuel. Under .he supplemental contract, General Electric will provide financial and technical assis-tance to NSP for the installation of the new poisoned fuel storage
- racks which will provide Monticello with full core discharge capability until the fall of 1987.
! 5.4.5 Construction of an Independent Spent Fuel Storage Facility (ISFSI) Construction of an ISFSI could take a number of fonas; an NSP project, a joint utility venture, the recent application for the expansion of
- , Morris Operations, the announced plans for an Exxon facility in l Tennessee, or any number of conceptual design studies that have been
! done. Storing fuel in a new offsite spent fuel storage facility is presently not a real alternative because there is insufficient tbne to design, license and construct such a facility and to ship sufficient fuel from Monticello. The earliest an ISPSI could become availcble for receipt l of spent fuel is 1980. Even the ERDA concept of a surface retrievable I fuel facility (SURFF) would not pennit acceptance of spent fuel before l 1985, and then only after 10 years of storage elsewhere. 40
I An offsite fuel storage facility requires a large financial investment. Industry estimates of costs for new offsite capacity range from $30-million for a 500 tonne facility (2500 BWR bundles) to over $100 million for a 3,000 tonne facility (15,000 bundles). (See " Spent Fuel Storage Study", AIF, April,1977. ) If these estimates were applied to a BWR storage facility, the investment cost alone would range from $6,657 to $12,000 per bundle.
- Offsite storage would require transporeation of fuel to the offsite 4
facility using spent fuel shipping casks and truck or rail trans-po rta tion. Current estimates of the shipping costs (including cask charges) for the annual quarter-core Monticello discharge range from j
$300,000 to $500,000. Reshipments may also be required before the fuel arrives at a yet to be established reprocessing or final disposal facility.
- Using published data and estbnates of operating costs, typical commercial prices can be calculated to range from about $3,700 to $5000 per assembly per year over the time frame of interest assuming an ISFSI of 3,000 tonne capacity. Based on these prices the levelized annual cost for an additional offsite capa-city of 1497 additional spaces as proposed would be about
$3,600,000 per year, including transportation to the ISFSI.
Constructing an ISFS1 would have a greater environmental impact than the proposed action. The new facility would require an acceptable site, access by truck and/or rail, a seismically quali-fied containment structure, a cask handling crane, fuel handling equipment, decay heat removal equipment, water purification equip-ment, a containment ventilation system with isolation capability, storage racks and numerous auxiliary equipment, systems and structures. By comparison, the only environmental impact of the proposed action is the material necessary for the storage racks and the modest personnel exposure during installation in the operating reactor facility. 5.4.6 Plant Shutdown , If high density racks are not installed in 1978 as proposed, the plant could have one additional quarter core discharge that fall, allowing operation at rated power until the fall of 1979. If an immediate need developed to unload the core to allow continued operation, the plant may have to be shut down earlier. In the fall of 1979, there would be 737 spent fuel assemblies in the 740 spent
, fuel pool spaces leaving insufficient room for another discharge batch. A plant shutdown would require that in 1979 the plant be shut down until such time that spent fuel is shipped offsite. Fuel j from the reactor could then be replaced with new fuel and power j operation could resume.
If Monticello were shut down, the replacement energy would be generated predominately by coal-burning electric generating plants. Loss of the base loaded Monticello unit would not only result in an increased 41
, F Store at Existing ISPSI -
Not an available alternative. construct ISPSI $3,600,000 Not availabit until 1980; j needed in 1978. Berefit of d < plant operation and pu sduction of electrical ererty. Greater environmental impact . nan the
] proposed action.
1 Plant Shutdown $55,700,000 No benefit of electrical energy generation. Results in consump-tion of less abundant energy j sources, higher cost of electrical energy to the consumer and de-creased system reliability Redesign Puel $960,000 Benefit of plant operation Storage Racks and production of electrical energy. No significant environmental impact From the above comparison, it is obvious that proposed action of redesigning the fuel storage racks is the most cost effective while being environmentally acceptable. A plant shutdown is definitely not cost effective. Constructing an ISFSI, while not a present alternative, should not be considered until after the spent fuel pool storage capacity at the reactor facility has increased to its maximum. 5.5 Evaluation of Proposed Action Unavoidable adverse environmental bnpacts have been evaluated and found to be insignificant. Replacing the existing fuel storage racks with racks of a higher density will allow spent fuel to be stored on site for a longer period of time. This does not result in any significant adverse environmental bmpacts on the land, water air, or biota of the area. The replacement of the fuel storage modules will result in a non-recurring occupational radiation exposure conservatively estimated to be 22 man-rem. The recurring occupational exposure resulting from the increased period of spent fuel storage is expected to be insignificant. No increase in personnel exposure beyond the confines of the plant site is expected. o
, p The proposed action, which will permit operation of the plant at its s licensed power level with full core discharge capability until late 1987 will not change the evaluation in the Monticello Final Environ-mental Statenent.
The proposed action will not result in any irreversible and irretrievable i commitments of water, land and air resources as identified in the Final Environmental Statement. No additional allocation of land would be made; the land acea now used for the spent fuel pool will be l used more efficiently by increasing the density of fuei storage spaces. ! The irreversible commitment of materials used to construct the pro-43
0 v posed storage racks is compared to the annual consumption of these materials in the United States as follows: < Amount Consumed Annual US daterial in Racks (1ha) Consumption (1bs) Stainless Steel 2,.25x105 2.82x10ll 3 go 9xto5 Boron carbide 5.55x103 ! Aluminum 2.54x100 8x10 9 l The material required is seen to be insignificant with respect to the annual U S consumption and does not represent a significant irreversible commitment of material resources. In any event, an equivalent amount of these or similar materials would be required wherever the fuel is stored. 6.0 SletiARY AND CONCLUSIONS , This safety evaluation has considered plans to replace the spent fuel storage racks in the Monticello spent fuel pool with racks , of a higher density. The re-racked pool will provide storage of up to 2,237 fuel assemblien, allowing continued operation of the plant at its licensed power level with full core discharge capability ! until 1987, at which tLae odher options for fuel disposal are expected to be available. A current need for such a modification is shown to exist. The integration of the proposed modification into the plant has been found acceptable. The k-infinity of the modified racks will have an increased margin to the current Technical Specification Itmit and will be within that limit under postulated accident conditions more severe than
- assumed in the Monticello FSAR. The spent fuel pool cooling and cicanup capability was found to be adequate. The seismic response of the fuel storage racks and the fuel pool meets governing criteria.
Materials used in the high density racks are similar to those used at other facilities. In particular, boron is to be used as a neutron i absorber in dhe proposed Monticello racks in the same form as is used elsewhere. Measures taken provide sufficient assurances that an adequate quantity will be installed and remain in place throughout the life of the plant. _
=
t < The radiological impact, along with other environmental impacts have
, been thoroughly evaluated. The non-recurring radiation exposure , during replacement of the fuel storage racks is conservatively esti-I mated to be 22 man-ran. This dose is reasonably low with respect to other plant activities. All environmental impacts are insignificant.
No additional land must be committed; the modifica tion will simply opthnize the utilization of land already desiFnated for this purpose l within the confines of the reactor building. 44
i No postulated accidents considered in the FSAR 6.1 FES are affected by the replacement 6f fuel storage racks. The safety evalvation has not identified any new postulated accidents resulting from this action. The potential for damage to stored fuel during the rack replacement was considered; it was concluded that adequate measures can and will
- be taken to avoid such an occurrence.
A cost-benefit evaluation supports the proposed modification. The only altentative presently available is to shut the plant down at an annual cost of over fif ty times that of the modified racks. Storing fuel offsite is presently not an available alternative. Should offsite storage become available after 1980, it will not be cost effective re,lative to replacement racks by a factor of three. The Technical Specifications and the Operating License have been reviewed; no revisions are found necessary because of the proposed action. No unreviewed safety questions were identified in this safety evaluation. L e w 45}}