ML20054G049
| ML20054G049 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/09/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20054G048 | List: |
| References | |
| NUDOCS 8206210021 | |
| Download: ML20054G049 (18) | |
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W ASHINGTo N. D. C. 2C555 g,......f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
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RELATING TO THE MODIFICATION OF THE SPENT RJEL ST0R' AGE POOL-O FACILITY OPERATING LICENSE NO. DPR-29 AND FACILITY OPERATING LICENSE NO. DPR-30_
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COMMONWEALTH EDISON COMPANY
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OUAD CITIES NUCLEAR POWER STATION UNIT-NOS.1 AND 2_
DOCKET NOS. 50-254 AND 50-265 Au thors':
R. Bevan; S. Block; J. Boegli; W. Bropks; F. Clemenson; 0. Rothberg; B. Turovlin; and P. Wu 1.0 INTRODUCTIOff By letter dated Marph 26, 1981, and supplemented by letters dated June 24, July 24, August 10, August 26, October 19, November 2, and
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December 8,1981, January 27 and March 12, 1982, Commonwealth Edison
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~ Company'(CECO, the licensee) requested amendments to Facility Operating Licenses DPR-29. and DPR-30 for Quad Cities Station, Units 1 and 2, respectively. The* request is to authorize increased storage capability in the spent fuel pools (SPF1 for the two nuclear units. The proposed modi-fications would increase the SFP storage spac.es from the currently licpsed 2920 spaces to 7584 spaces combined total for the twn pools.
This expanded storage capacity will allow the continued opera' tion of the two nuclear units with onsite storage of spent fuel to past. the H
year 2000. The licensees basic supporting document for this action is a,. ;
report, Spent Fuel Pool Modification for Increased Storace Capacity, Quad l
Cities Nuclear Unit 1, Docket No. 50-254, and Quad Cities Nuclear Unit
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No. 2, Docket No.' 50-265, Rev.1, date,d June,1981.
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~ DISCUSSION The licensee's proposal would increase the SFP storage capacity by replacing the existing spent fuel storage racks with new high density storace racks. The new racks will contain neutron absorber materia'i in the rack walls so that spacing between stored assemblies can be reduced while maintaining adequate criticality margin.
t The hich density racks are made up of $odules, each module being comp ~osed of six-inch square cells, each cell accommocating a single BWR fuel a s s embly.
The cell ' walls contain a neutron absorber material sandwiched between sheets of stainless steel. The cells making up the module ~ have 5.22-inch center-to-center spacing. The general arrangement of the r;odul es-in. the pool s is shown in Figures 2.1 and 2.2 of the licensee's i
application and basic supporting document.
The general. details.of 8206210021 820609 PDR ADOCK 05000254 t
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design and construction of the racks are contained in. Figures 3.1-through 3.S and are described in Section 3 of the licensees basic
. supporting document.
The racks are free standing in that they are neither anchored to the floor of the pool or walls, nor are the modules 6
interconnected.
The applicable codes, standards, and practices for this modification are set forth in Section 3.2.of the licensee's basic supporting document. A detailed structural analysis is described in Section 6 of the document to show the adequacy of the r.tcks to resist the postu-lated stress combinations for normal and postulated accident conditions.
Section 9 of the licensee's basic supporting document describes-the detailed analysis to show that the pool floor meets all struc'tural acceptance requirements when conservatively analyzed.
The safety considerations associated w'ith this proposed action are addressed below.
A separate environmental impact appraisal has been prepared for this action.
O.h 3.0
-- EV AL'UATION
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3.1 Structural and' Mechanical Desien Considerations Description Quae cities Units 1.and 2 each have fuel stdrage pools 33 feet wide x 41 feet long.
The Unit 1 pool will contain 19'high density fuel racks in'seven different module' sizes with a total of 3714 storage locations, while the
. Unit 2. pool will contain 3E70 storage cells arranged in 20 racks with six:
different module sizes in :nis pool.
All modules are free standing, i..e., they are not anchored to the pool walls.
The. minimum gap between adjacent racks is three inches at all locations
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and nine in'ches between the racks and the fuel pool walls.
Because of these gaps, the possibility of inter-rack impact, or rack, collision with pool wall hardware during the postulated ground seismic motion, is precluded.
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The racks will be constructed from *STM 240 - 304, austenitic steel sheet material, ASTM 204-304 auscenitic steel plate material, and. ASTM le2 - F304 austenitic steel forging material. A typical module contains' storage cells which have 6 inch minimum internal cross-sectional opening'.
Skip welding at the top ensures proper venting of the sandwiched space c
in the sub-elements which make up the fuel racks.
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The rack assembly is ' typically supported on'four plate-type s leyel, thus creating the water plenum for coolant flow.
Further details of the spent fuel racks are i,11ustrated in the licensee's basic supporting document.
Evaluation and Conclusions _
In our evaluation of the lic'ansee's proposed action, established. codes, st'a
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and criteria were applied, consistent with the HRC's guid
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Accotdingly, the t'esign of.
dated April,1978 and revised January,1979.
the racks, fabrication, and installation criteria; the structural design and analys.f s procedures for all loadings, including seismic and impact.
loadings; the load combinations; the structural acceptance criteria; the quality assurance requirements for design, a,nd applicable industry codes were all reviewed in accordance with the appl-icaide-port-ions-of that NRC guidance.
i-For the design of the spent fuel modules, two sets of criteria were to be The first
- establishes requirements to ensure that, adjacent satisfied.
racks _will not' impact dusing the Safe Shutdown Earthquake (.SSE), assuming-It is thelower bound value of the poci surface friction coefficient.
by this criterion that the factors. of safety against tilting be 1.5 for t OBE and 1.1 for the 'SSE.
to ensure that loading combinations and stress allowables are in accordance The basic
'with Section III. Subsection NF of the ASME 1980 Edition.
materig allowables, fabrications, installations and quality contro The loading considered,in the modules also conform with the same code.
analysis involves dead loads, live loads, thermal loading, and seis loadings (03E or SSE). effects of a postulated accident involving the drop
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on the racks and on the fuel pool liner, and the fuel handling crane uplift i
accident.
A dynamic analytical model, consisting of beams, gaps, springs, da inertia coupling representing fluid coupling between rack and assemblies, an between rack and adjacent racks, was used to predict the maximum sli.dingThe distance and seismic forces resulting from the SSE.
used to predict the seismic stresses and displacements. The coefficient of friction between the stainless steel liner and the leveling legs of the N
racks used in the analysis was chosen based on the information contained in 'a report by E. Rabinowicz of Massachusetts Institute of Technology entitled " Friction Coefficients of Water Lubrication Stainless Ste The result of for a Spent Fuel Rack Facility" dated November 5,1976.
this analysis indicates that, although the proposed racks which are free-standing may slide toward each other during the SSE, sufficient gaps are provided between the modules a_nd the modules and the pool walls suc ldd the inter-rack impact, or the rack collision with the pool walls, is prec u e.
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The analysis, design, fabrication, and criteria for estab'lishing installation procedores of the proposed new spent fuel racks are in conformance with accepted codes, standards and criteria identified in.
the NRC guidance.
The ' structural design and analysis procedures for all loadings, including seismic, thermal, and impact loading; the acceptance
. criteria for the appropriate loading conditions and combinations; and the s
applicable industry codes are in accordance with appropriate sections of the NRC staff "0T position for Review and Acceptance of Spent Fuel Storage and Handling Applications."
Allowable stress limits for the combined loading conditions are'in
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accordance with the ASME Code, App. XVII.
Yield stress values at the-appropriate temperature were obtained from Section III of the-ASME.: ;--
Code.
The quality assurance and criteria for the materials, fabrication and installation of the new racks are in accordance with accepted requirements ~ of the ASME Code.
The effects of the additional loads on the existing pool structure due to the new fuel racks, existing fuel racks, and equipment have been examined.
The pool structural integrity is assured by conformance with the Standard Review Plajigection 3.8.4.-
l Resulis of the seismid ahd structural analyses indicate that the racks are capable of withstanding the loads associated with all design loading conditions. Also, i'mpact due to fuel assenbly/ cell interaction has been cons'idered, and will result in no damage to the racks or fuel assemblies.
Two typ,es of postulated fuel assembly drops onto the racks were analyzed by the ligensee and evaluated by the staff. The 'first drop is a straight drop of a fuel assembly from a maximum of 36 inches above the storage location and impacting the base.
The-second drop involves a fuel assembly drooning from a maximum of 36 inches above the rack and hitting the top
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of th'e rack.
In both cases, the impact energy is dissipated by local yielding; j
however, the sub-criticality of the fuel arrays is not violated.
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The dropping of a heavy load onto the' protective pool liner of the pool floor l
l was also analyzed'.
Although local damage and plastic deformation may occur, the overall structural integrity of the liner is maintained.
l The effect of postulated stuck fuel assembly due to the attempted withdrawal was considered,' and the damage, if any, was required to be limited to the region above the active fuel elements.
Results of the stuck fuel assembly l
analysis show that the stress is below that allowed for the applicable ~ loading l
combinations.
'ie find that with respect to structural and mechanical design the subject l
modification proposed by the licensee satisfies the applicable requirements of Cene'ral Design Criteria 2, 4, 61', and 62 of 10 CFR, part 50, Appsndix A and is a c c epta bl e.
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5-3.2 Materials Considerations Discussion and Evaluation We have reviewed the co=patibility and chemical stability of the materials (except the fuel assemblies) wetted by the pool water.
In addition, our review has included an evaluation of the Borafier neutron absorber material used in the high density storage 1,ocations for environmental stability.
There wi11 be both the old and the new types of spent fuel stora'gc cells in the Quad Cities Station spent fuel pools during the transition time while new storage modules are being installed. The transition period.
is expected to last slightly over one year.
The spent fuel pool.is '
filled with demineralized high-purity, high resistivity water.
The new high-density spent fuel storage ra ks are of welded stlinless c
st' eel constr'uction with a "Boraflex" neutron absorber sandwiched between the stainless steel sheets. The neutron absorber is comp.osed of boron carbide powder in. a,. rubber-like silicone polymeric mat ix.
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'~ihe old low density fuel storage tubes provide for ths interim storage of fuel assemblies and are constructed of aluminum without neutron
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absorber material. The anticipated corrosion of the aluminum alloys, type 1100 or 6061, is negligible in water of spent fuel ' pool quality at temperatures up to the boiling point of water.; at 125 C (257 F) a cerrosion rate of 1.5 x 10-4 mils / day has been measured for alloy 6061 aluminum, in water of pH 7, which corresponds to a total' corr'osion of 1.1 mils in twenty years.
Since the oxidation rate will continue to decrease slightly over this period, this ' estimate i's considered to be conservative.,.
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The inherent high corrosion resistance of aluminum and stainless steel
-makes them well suited for use in demiheralized water. Aluminum and stainless steel fuel storage racks submerged in water have been in use for 10 years with no deterioration evident.
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n 13 e-Aluminum and 200-series stainless steel are very similar insofar asBecause their coupled potential ts concerned.
The use of low-conductivity, galvante corrosion should not occur. sta d successfully for corrosion is a recommended practice and has been use many years By the aluminum industry.
The pool liner, rack lattice structure and the high density fuel storage l
tubes are stainless steel which is compatible with the storage the corrosive deterioration of the type 304 stainless steel should not environment.
x 10-5 nches in 100 years, which is negligible t
exceed a depth of 6.0 Dissimilar metal conta'ct corrosion relative to the initial thickness.
(galvanic attackl 5etween the stainless steel of the pool liner, ra lattice structure, fuel storage tubes, and the Inconel and the Zircaloy in the spent fuel assemSlies will not.be significant because 'all of d are these materials are protected by highty passivating oxide films anThe Bo therefore ad.similar galvanic potentials >. composed of Boraflex has calvanic potential in contact with the metal components.
i undergone extensi'vditesting to study the effects of gamma irradiat o in vario'us enviro _q ents, and to verify its structural. integrity an
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"' suitability as a neutron absorbing material.
Venting The space which contains the Sor.aflex is vented to the pool.
licone will a11ew gas generated by the chemical degradation of the si ill polymer binder during heating and irradiation to escape, and w q[ event bulging or swelling of the stainless steel tube.
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To provide added assurance that no unexpected corrosion or h licensee of the materials will compromise 'the integrity of the racks, t e
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has committed to conduct a long term fuel storage cell surveillance Surveillance samples are in the form of removable stainless steel clad Boraflex sheets, which are proto-typical of the fuel sto program.
These specimens..will be removed and examined periodica cell walls.
t Concl usions_
From our evaluation as discussed above we conc h ld be that will occur in the spent fuel storage ' pool environment s ou Components of little significance during the remaining. life of the plant.
lloys which have a
.in the spent fuel storage pool are constructed of a i
low differential galvanic potential between them and have a h l
i ~ corrosion.
tance to general corrosion, locali:ed corrosion, and ga van c~t i
Tests under irradiation and at elevated temperatures in water degradation that the 5craflex material will not undergo significant during the expected service life of 10 years.
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7-We further conclude that the environmental compatibility and stability of the materials used in the spent fuel storage pool are adequate, based on test data and actual service experience in operating reactors.
We have reviewed the surveillance program and we conclude that the' monitoring of the materials in the spent fuel storage pool, as proposed by the ifcensee, will provide reasonable assurance that the Boraflex material will continue to perform its function for the design, life o'f We therefore find that the implementation of a monitoring the pool.
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program and the selection of appropriate materials of construction by,
the licensee meet the requirements of 10 CFR -Part 50, Appendix A,.
Criterion 61, by having a capa,bility to permirappropriate periodic inspection and testing of components, and Criterion 62, by preventing criticality By maintaining structural integrity of components and of *
.the boron poison.
Installation and Heavy Load Handling Cbnsiderations 3.3 The results of the staff's generic review of handling heavy loads at nuclear power plants, i.e., NUREG-0612, " Control of Heayy Loads at Nuclear Power Flants;" is ongoing and will not be completed before the Therefore, we have
--Spent fuel pool meWfications are to commence.
limited this review and evaluation to the heavy load handling operations
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associated with'.the Quad Cities Unit 1 and 2 proposed spent fuel modifications.
The heaviest identified load with this modification is a 16 x 16
's* rage rack weighing 161/2 tons, whereas.the main hoist on the reactor bu iding crane is rated at 125 tons.
The overhead crane was previously modified and as documented in a NRC review dated January 27,1977,. we found it to be acceptable.
From this we^ conclude that the overhead load.,.;
handling system is acceptable.
'The licensee ha's stated that the travel paths of the storage racks will be established before moving the racks, and the travel paths will be based on the studies associated with NUREG-0612.
The handling procedures will be such that none of the storage racks containing stored fuel will be immediately adfacent to the empty rack being moved.
Consequently, a load handling mishap will not impact on stored fuel. Based on these considerations, we conclude the procedures are acceptable.
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8-The June 22, 1981 Commonwealth Edison response to our December 22, 1980 ceneric letter on control of heavy loads states that operator training i
qualifications and conduct for Quad Cities Units 1 and 2 comply with ANSI B30. 2-1976.
From this we conclude the qualifications and conduct of operators handling heavy loads are acceptable. The above submittal also states that the inspectio.n, testing and maintenance related to Quad Cities cranes ccmply with ANSI B30.2-1976.
From this we conclude that adequate measures will be taken to assure the operab.ility of the cranes used in handling the spent fuel pool modifications loads, and i*
'are therefore, in this ' respect acceptable.
A lifting yoke has been designed to handle the new storage racks.
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f will consist of a four-leg bridle hitch with turnbuckles, attached to a rectangular frame that supports four lifting rods that will be threaded into the four legs of the racks. The holes in the rectangular frame germit the lifting rod spacing to' be adjusted so as to permit them to remain vertical and yet accommodate Ahe seven different sized racks.
Figure.3-8 of 'the licensee's submittal indicates the lifting yoke is Based rated for 22.7 tons while the heaviest storage rack is 161/2 tons.
on the above, we. conclude that the lifting yoke is ade,quate for handling the,new storage racks, and therefore, acceptable.
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The existing aluminum open lattice r,torage racks will be removed using the overhead crane and a wire rope sling. The sling design complies with the requirements of ANSI B30.9-1971.
It's load rating is slightly more than twice the weight of the heaviest rack to be removed. The ends of the sling terminate with locking safety hooks which.ar'e' attached to lifting lugs on the storage rack. Based on the above we conclude
.' g that rigging interposed between the crane hook and.the load is acceptable for handling the old stor. age racks,.and that the crane meets the objectives of APCS 3 BTP 9-1 and, has sufficient. capacity for the described operatio.nss
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The travel paths, procedures, operator training and crane maintenance l
'are adequate to acccmplish the heavy load handling operations associated ~
with spent' fuel pool modifications and are therefore acceptable.
In re' gard to the handling of light loads over stored spent fuel, an analysis has been made assuming the channel measuring device, weighing 1000 pounds, was dropped 30 feet above the racks. The 'results indicate that deformation will occur but the kef'f remains equal to or less.than O.95, in conformance with SRp, Section 9.1.2.
In this respect we find that a postulated light load drop will not cause a criticality accident.
The prcpesed modifications meet the guidelines of the applicable portions of the fo11cwing:
Regulatory Guides 1.13,1.29 and.1.71,1.55,1.92 and 1.124; and 10 CFR Part 50, Appendix A, General Design Cri.teria 1, 2, 61, 52 and 63; Standard Review Plan Sections 3.8.3 and 3.S.4 and' industry stand rds ANSI N210-1975, ACI 318-77, AISC, ASTM, ASME Section III t
.Divisioiil' Subsection NF 1980 and ASME Section IX-1980.
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Crit'icalitv Considerations Discussion and Evaluation The boron content in the neutron absorber material in the rack wall's
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is equivalent to.a 3-10 areal density of 0.01728 grams per square centimeter.
The muittplication factor of the racks is calculated for-an 8 x 8 assem51y having a uniform enrichment of 3.2 weight percent U-235. The infinite multiplication factor for.this assembly.in the stardard reactor configuration at cold clean conditions is 1.362l, For comparison the maximum value of the infinite multiplication factor for reload Bundles is 1.241 at the most reactive point in'the bundle life (NEDO-240ll-P-A," General Electric Generic Reload Fuel Application" Amendment 9, dated November 17, 1980).
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The rack design i's.~4hus conservative for assemblies which are anticipated,
d in ths-racks.
Other conservatisms present in the analysis
' to be store include the use oTthe minir.um (worst case) center-to-ten'ter spacing and a Boraflex. poison plate width less than the design value.
The criticality analyses of the racks were performed with the AMPX-XENO computer code package using the 123 group XSDRN cress-section set with.,the 11ITAW1. subroutine for U-238 resenance shielding efG: cts. This code n'as\\been benchmarked against experiments by Southern Science. Applications, Inc. and the results are reported in SSA-127 (Rev.1), " Benchmark Calculations for Spent Fuel Storage Racks" dated September 1980. The
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results of the comparison show that the Code set underpredicts the multiplication factor by 0.36 percent reactivity change with a deviation of 1.23 percent reactivity change at the 95 percent probability, 95 percent
. confidence level. Trend analyses were performed to obtain an estimate
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of the effect of varying amounts of boron between assemblies.
This analysis showed that AMPX-KENO should overpredict the reactivity of the Quad Cities racks By 3.1 + 1.2 percent reactivity change. No credit is taken forethis overprediction in the analysis.
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Sensitivity analyses were performed to cbtain the reactivity effect of the variation of stainless steel wall thickness, boron loading variations, and channel deformation (bulge).
The results of these studies indicate -
a total uncertainty of 0.E7 percent reactivity change due to these effects.
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The calculated value of the nominal case multiplication factor was 0.9155 +.0067 where the uncertainty is the statistical uncertainty in the Fonte-Carlo (KEN 01 calculation only.
To this value must be
~ added the calculational bias of Q.0036 and the statistical combination of the bias uncertainty (0.01231, the calculational uncertainty (0.0067). and the mechanical uncertainty (0.0097). The resulting v'alue for the maximum multiplication factor is 0.9361. This value meets the acceptance criterion that requires the keff be less than or equal to 0.95.
The criticality effects of various abnormal a'trd postulated accident conditions have been investigated.
This includes improper positionina of an assembly in its storage, rack, Sowing of the channel, variations in pool temperature, a droppcd fuel assembly, and a missing absor5er plate tin the racks.
These analyses show that.the criticality acceptance criterion iq not v'iolated when not more' t!)an one Boraflex plate out of fifteen is missing. Appropriate measures will be taken during manufacture of the racks and prior to installation in the pool to assure the presence of t,hg,boren absorber material as designe,d.,
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n the course of our revieU*, we have found that:
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State-of-the-art calculation methods which have been benchmarked against critical experiments h' ave been used.
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Credible abnormal configurations have been investigated,.
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Uncertainties.and biases have been treated, and.
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4 The result, including all uncertainties, rJeets our acceptance criteria for the nominal case and for abnormal and postulated accident conditions.
From the above considerations, se find that fuel assemblies of the 8 x 8 two-water r6d design, having. average enrichments less than or equal to 3.2 weight percent U-235, other fuel designs containing less than 15.49 crams of.U-235 per axial centimeter, or BWR assemblies having cold clean infinite multiplication factors in the Quad Cities reactor geometry of less than 1.36 may be safely stored in the Quad Cities 1 and 2 storage pool.
Conclusion We conclude that any number of spent fuel assemblies of a design likely to be used in the Quad Cities reactors can be safely stored in the spent Gel racks with adecuate criticality margin.
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3.5 Scent Fuel po'o1 Coolino Considerations Descriotion and Evaluation
- I Quad Cities Units 1 and 2 each has a stainless steel lined.reinforc'ed~
concrete spent fuel storage pool.
The two pools are. joined by a transfer canal.
Fuel can be transferred between the two pools via the transfw canal after opening the two gates, located at the sides of the re-ctive pools. A normal fuel discharge, i.e.,
a5out 200 assembiws, occurs at 18 month intervals.
To the extent possible the discharge cycles of the two units are phased such that the refueling-operations on the two units wi,11 not occur simultaneously.
Separate spent fuel pool cooling systems are provided for each of the '
two pools.
The FSAR states that each of the two separate cooling systems
'was designed to be capable of maintaining the pool water temperature of their respective pools below 125 degrees F during maximum normal discharges, 'when the reactor building closed cooling wate'r system-is, at its maximum temperature of 105 degrees F.
This assures that a comfortable workip,g environment can be maintained during normal conditions.
Further,.on those infrequent off normal conditions.where;
-for example, a fuTUcore di.scharge occurs, the pool wat.er. temperature will not exceed 150 degrees F.
Analyses of the pool water tem'peratures following this proposed spent fuel expansion shows the maximum pool water temperatures does not exceed 134.6 degrees F when the pool is completely filled with normal discharges.
This is nearly a 10 degree increase over that stated in the FSAR.
This.is less than the 140 degrses F t given in the Standard Review plan Section 9,.1.3 - Spent Fuel Pool Cooling limjhieanup System and is acceptable. Further, the analysis of the and maximum pool water temperature following a full core discharge, at any point until the pool is filled with spent fuel, will not exceed 145.4 This is less than the 150 degrees F stated in the FSAR, and',. ;
degrees F.
is acceptable.
The spent fuel pool cooling system (SFPCS) for each unit consists of one cooling loop having two parallel, 50 percent capacity, pumps placed in series with two, 50 percent capacity, parallel heat exchangers. Each pump is rated at 700 gpm, i.e., 350,000 pounds per hour, and assuming the pool water temperature is at 125 degrees F each heat exchanger is rated at s
3.55 x 10o BTU /hr.
Therefor.e each unit's, spent fuel pool cooling system has a total design flow of 700 000 pounds per hour and a total heat b
removal capability of 7.3 x 10 BTU /hr at a pool water temperature 'of 125 degrees F.
.By allowing the pool water temperature to rise to 134.6 degrees F the total heat removal capability of each spent fuel pool 6 STU/hr.
ecoling system increases to approximately 10.9 x 10
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'!B In addition to the above spent fuel pool-cooling s[ stem, provis. ions have been made to cros.e Me the spent fuel pool cooling system 'to the residual heat remova:
...nR) system. This is, accomplished by installing two 6 inch pipe size spool pieces in the two legs of the _ spent fuel p'ool cooling loop.
The stx inch. RRR tie-in line will prov~ide an additional spent fuel pool cooling water flow of 1,000 gpm f.e., 500,000 pounds-4 per hour. While it has not been stated by the licensee, we note that it appears feasible to use the cooling system in one unit to assist cooling the pool water in' the adjacent unit pool. This~ could be accomplished by opening the two gates in the. transfer canal and allowing,an
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interchange of water Setween the two pools.
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,-s "i'3 Decay Heat ihe licensee has analyzed five different, cases of spent, fue1I'po'ol: decay heat loads and the resultant pool water temperatures with and'without7 the additional cooling provided by the residual heat removal system (RHR).
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The cases investigated are as follows:
(.1) The pool is YiL)ed wits normal d'scharges of 240 fuel asse' blies m
M-only provided b'y the. SFPCS (decay heat equals
.and cooligg YtT7Erl..
11.2 x iCS B (21 The pool is filled with rarmal discharges of 240 fuel assem511es and cooling is-provided 3y the SFPCS and the RHR system (decay 6 BTU /hr).
heat equals 11.2 x 10 (M The pool is fil. led with normal discharges of 200 fuel assemblies and cooling.is provided only By the SFPCS (decay heat equais 9.65 x 106 BTU /hrl.
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(.4). The pool is filled with normal discharges of 200 fuel assemblies and cooling is providgd By the SFPCS and the RHR system (dei:ay heat equals 9 65 x 10 BTU /hr).
(5) The po61 is filled, with normal discharges plus a recently discharced full core and cooling is provided by,the SFPCS and RHR 6 BTU /hrl.
system (decay heat equals 24.7 x 10 In the case of normal discharges and a full core discharge it. is assumed 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> will Be required to prepare the reactor for refueling. The transfer of a normal discharge of either 200 or 240 assem511es can be accomplished in two days.
In the case of a full core discharge,, six days will 5e required to transfer the fuel to the storage pool.
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According to the licensees analysis, the maximum bulk temperature of the pool. will not exceed 134.6 degrees when a normal fuel discharge of spent fuel is placed in the pool.
Although no safety problem is created by a s:mewhat higher pool temperature, the' higher temperature encroaches upcn marcin assumed in cur analysis of the licensee's ability to provide makeup water in the event that pool cooling capability is lost.
Similarly, in the event of a full core discharge to the pool, the -
iicensees analysis shows that the pool temperature will not. exceed 145.4 degrees.
Should the pool bulk temperature exceed this value during a full core discharge, further placement of spent fuel into the pool should be suspended until the temperature is brought to-below 145 degrees F.
The licensee has agreed to include this limit,in its operating procedures.
Makeuo Water. -
The spent fuel pool,, system is designed to minimize the ~1o'ss of water from the pool and to prev.ent the water level from falling bslow a safe level
_ a5cve the stored:f,el.
For example all penetrations into the pool, except for valved drains, are located at a height s'uch~that there will always be a safe level of water above the fuel.
Each pool has a high and low water l'evel monitor.
Both monitors actuate local annunciators and the low level monitor also actuates a control room low level annunciator.
In the event makeup water is needed, there are two sources ef. makeup water, the condensate storage tanks and the fire system.
Ap:dbximately 550 gpm of condensate water can Be delivered to1the pools via the condensate transfer pumps and skimmer surge tanks within a few minutes.
In addition as much as 1,000 gpm of condensate storage tank water can be supplied to the pools using the RHR pumps following the installation of a spool piece joining the RHR system to the spent fuel pool cooling-system. AEcut three hours would be required to install
-the spool piece.
In the event that the above identified sources of water become unavailable, the fire system hoses are capable of providing makeup water from the river within approximately 30 minutes.
The two pumps, each rated at 3,200 gpm, can provide water to the pool far in excess of any reasonable need.
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We conclude the makeup water system is adequate and acceptable because makeup water is availa51e from the condensate storage tanks and river via the fire system,and their respective makeup rates exceed the Soil off rate descriti.ed below.
Further, this _ makeup water can Se
.made availa51e before 5'otiing would occur.
Boil Off Rate The minimum time before boiling occurs and the maximum boil off rate were established assuming that:
O } the heatup follows a full core discharge in. Unit 2 stogage pool (_i.e., the pool with the least water ~
inventory of 44, 471 ft of waterl., (2) the pool water Bulk temperature is at its maximum temperature of 145.4 degrees F. (3)-there is e
no exchance of water Setween Pool 1 and Pool 2, (4) all. pool cooling is lost and (5) no credit is taken for heat lost to the pool walls and floor. Under the a5ove conditions,aSout 71/2 hours would elapse before bulk boiling would occur. The maxihum Sotloff rate would be 51 gpm.
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Based on the 'aSove, we conclude that the available sources of makeup water are adequate, the time required to activate the makeup system is sufficiently.less'than the time required to reach boiling,and the makeup rates;from Soth maXssp sources exceed the Soil off rate', and therefore Me provisions fof" makeup water are acceptable.
Local Boilina Using a conservative thermal hydraulic circulation 'model of rico1 water flowing down along the walls, laterally across the pool floor in the gwater plenum and up. hrough the stored fu'el assemblies, the maximum t
calculated water temperature at the oJtlet of the fuel assemblies was shown not to exceed 167 degrees Fahrenheit.
The sat'uration tempera'ture at this point is 240 degrees F.
Du'e to the margin between these two temperatures we conclude that-nucleate Boiling will not occur and in this respect the design is acceptabl e.
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i Conclusion Cool'ing capability for the spent fuel pools for the two nuclear units has been evaluated for the maximum expected loading conditions for the s
We conclude that the presently installed pool cooling capability new racks.
- is adequate to handle the heat load under any reasonably expected conditions of operation.
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. n-c 3.6 Soent Fuel Pool Cleanuo System Descriotion and Evaluation The spent fuel pool cleanup system consists of a filter demineralizer (precoat filter material and powdered anion and cation res-in), filters, and associated piping, valves, and fittings. The system is. designed to remove corrosion products, fission products, and impurities from the pool water.
Pool water purity is monttored by a continuous. conductivity meter installed on the inlet to the fuel pool deminerali2e'rs, and.by periodic gra5 samples for laSqratory analysis? Once a week a repre-sentative gra5 sample is o5tained from the fuel pool demineralt.zer inlet line 'for pH, for c5loride, stitca, and turbidity analysis. Weekly activity checks are made for gross Seta and gross alpha activity. Once -
-a month a sample from the same location is obtained for a gamma isotopic.
a nalysis. All. peaks are identifted. All identified isotopes are.
quantified, 'and an LLD is determined for Kr-85.
The criterion for,a7demineralizer backwash and precoat ts a consistent excursion from the~. chemistry limits, or high differential pressure
-.95 psidl across Tne deminpral tzer. We agree with the; licensee that a
the proposed high density fuel storage will not significantly alter the chemistry or radiochemistry of the spent fuel. pool water.
Past experience shows that the greatest increase in radioactivity and impurities in spent fuel pool water occurs during refueling and spent fuel handling.
The refueling frequency, the amount of core to be
% g replaced for each fuel cycle, and frequency of operat'ing the spent fuel pool cleanup system are not expected to increase as a result of high density fuel storage. The chemical and radionuclide composition of the.. :
spent fuel pool water is not expected to change as a result of the proposed high density fuel storage.
Past experience also shows that no significa'nt leakage of fission, products from spent fuel stored in pools
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occurs after the fuel has cooled for several months. To maintain water quality, the licensee has established the frequency of chemical and l
I radionuclide analysis that will be performed to monitor the water quality and,the need for spent fuel pool cleanup system demineralizer resin and filter replacement.
In addition, the licensee has also set the chemical and radiochemical limits to Be used in monitoring the-s l.
spent fuel pool water quality and initiating corrective action.
. We agree with the licensee that the increased quantity.
of spent fuel to 5e stored will not contribute significantly to the amount of radioactivity from fission products in the sper.t. fuel pool
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water.
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The proposed expansion of the spent fuel pool will not appreciably affect the capability and capacity of the existing spent More frequent replacements of filters or fuel pool cleanup system.
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. demineralizer resin, required when the differential pressure excee s 25 psid or decontamination effectiveness is reduced,as impurities.in the pool water as a result of the expansion of s spent fuel.
ides the
. system with the proposed ftigfi density fuel storage (1 products, and impurities from the pool and thus meets the requirements of General Design Criterion 61 in Appen. dix A of 10 CFR Part 50 as it relates to appropriate fuel s'torage systems, (2) is capable of reducing occupational exposures to radiation by removing radio of 10 CFR Part 20, as it relates to maintaining radiation' exposures as low as reasonably achievable; (3) confines radioactive materials in the pool water into the filters and demineralizers, and thus meets
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Regulatory Position C.2.f[c) of Regulatory Guide 8.8, as it relates to the source; and (4) removes reducing the spnead of contaminants from suspended impuritids from pool water by filters, and thus meets 1
" Regulatory PositTfh C.2f(,3,'ds through physical action.
to removing crud from flui Conclusion _
On the basis of the above evaluation, we conclude that:
(1) The existing spent fuel pool cleanup system meets General Design Criterion 61 of 10 CFR Part 50, Appendix A, Section 20.l(c) of 10 CFR Part 20 and the appropriate Sections of Regulatory Guide 8.8 and, therefore, is acceptable for the proposed high density fuel sto. rag (2) The existing spent fuel pool' cleanup system is adequate for the -
proposed modification.
(3) The conclusions of the, evaluation of the waste trea Report (August 25, 1971], are unchanged by the modification of the
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scent fuel storage system.
3.7 Occuoational Radiation Exoosure Descriotion and Evaluation reviewed the licensee's plan for the removal and disposal of the low density racks, and installation of the high density racks, We have The occupational with respect to occupational radiation exposure.
3x;ns,ure, for this operation is estimated by the licensee to range fr e
m 18 to 39 man-rem. This estimate is based on the licensee's detailed breakdown of occupational exposure for each phase of the modification.
The licensee considered the num5er of individuals perfonning a specific job, their occupancy time while performing this jo5, and the average dose rate in the area where the joS is being performed. The spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area Secause of the depth of water shielding the fuel.
One potential source of radiation is radioactive activation or corrosion products called crud. Crud may be released to the pool water because of fuel movements during the proposed modification. This could increase radiation levels in the vicinity of the pool.
During refuelings, when the spent fuel is first moved into the fuel pool, the additten of crud to the pool water from the fuel assembly and from the intro-duction of primary coolant to the pool water is greatest; However, the licensee does not expect to have significant releases of crud to the
. pool water during modification of the pools. The purtfication system for the pool, which has Rept radiati'on levels in the vicinity of the pool to lowslevels, includes a filter to remove crud and will. be operating during the modification of the pool.
The licensee has ' presented three alternative plans for? removal and
_ disposal of the did-racks. These are (1) to crate and ship intact racks to '
a licensed burial facilityr(2) to cut the racks in'to 'small pieces with a shredder and pack the pieces into drums for burial at a licensed burial facility; and (3) to have an outside vendor chemically decontaminate the intact racks.
If the decontamination option is selected,-the decontamination chemicals would be reduced in volume, solidified and buried. The bulk'of the decontaminated racks could be disposed of as clean scrap. This last altbrnative is to be tested at the Dresden station and results of that work will be influential in the final decision.
In any event, the disposal methodology will follow "as low as reasonably achievable" ( ALARA)* guide
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lines for each of the alternatives.
It should be noted that the precedure's for removal of old racks from the pool will be performed independent of the aforementioned disposal alternatives.
The racks will be individually.
lifted from the pool water and rinsed by hydrolasing to remove any loose radioactivity that will drip back into the pool water prior to movement to l
a receiving area for preparation for disposal.
Divers will be used for setting and shimming the high density racks.
Related experience from the Dresden SFP modification indicates that the diver exposure should be' less than 2 man-rem for rack installation l
including clean-up and diver work.
Conclusion l
l Based on our review of the manner in which the licensee will perform l
their modification, and related experience from other operating reactors that have performed similar spent fuel pool modifications, we conclude that the Quad City spent fuel pool modification can be performed in a manner that will enstere as~ low as is reasonably achievable (ALARA) exposures to workers.
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4.0 CONCLUSION
We have perfonned an evaluation of the licensee's proposed modifications based primarily on' information provided to us in the-licensee's ba~ ic s
. supporting document. This document has been revised ahd supplemented during the course of our review in response to staff questions, and from meetings and discussions with the licensee, and to address new or more refined information regarding the proposed modification.
Our evaluation concludes that the proposed. modification of the Quad Cities Station Units 1 and 2 spent fuel storage is acceptable because:
(1)
The structural design and the materials if construction are gcceptabl e.
The installation and use of the proposed fuel handling racks can (2) be accomplished safely.
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(3)
The likelihood of an accident involving heavy loads in the vicinity of the spent fuel pool is sufficiently small that no additional restrictions on load movement. are necessary while,our generic review of thq'fissues is underway.
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(4) The installation and use of the new fuel racks does.not alter the potential consequences of the design basis accident for the SFP, i.e., the rupture of all the fuel pins in the equivalent of a single fuel assem51y and the subsequent releas'e of the radioactive inventory within the gap of each fuel pin, as already g reviewed and approved in the FSAR for Quad Cities Station'.
(5) The physica1' design of the new storage racks will preclude
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criticality for any cr,edib1'e moderating condition.
(6) The cooling system for each of the spent fuel pools has acceptable cooling capacity.
(7) The conclusi.ons of the evaluation of the waste treatment systems are unchanged by the modification of the spent, fuel pool'.
(8) The' increase in occupational radiation exposure to individuals s
due to the storage of additional fuel in the spent ' fuel pool would be negligible.
We conclude, then, based on the considerations discussed aSove, that:
(il there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will 5e conducted in compliance.with the Commission's regulations and the proposed license amendments will not be inimical t,o the comort defense and security or to the health and safety of the publ ic.
Dated:
April 9,1982
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