ML20052A959
| ML20052A959 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 04/09/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20052A954 | List: |
| References | |
| NUDOCS 8204290468 | |
| Download: ML20052A959 (18) | |
Text
dps at og'o fiNITED STATES
!], g[,i NUCLEAR REGULATORY COMMISSION WASmNGTON, D. C. 20555 j
N..A...f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE MODIFICATION OF THE SPENT FUEL STORAGE POOL FACILITY OPERATING LICENSE NO. DPR-29 AND FACILITY OPERATING LICENSE NO. DPR-30_
COMMONWEALTH EDISON COMPANY OUAD CITIES NUCLEAR POWER STATION UNIT NOS.1 AND 2 DOCKET NOS. 50-254 AND 50-265 1.0 INTRODUCTI0ff By letter dated Marph 26, 1981, and supplemented by letters dated May 24, July 24,.Aug,ust 19, September 21, October 19, November 2, and
- December 29,198177anuary,27 and March 12, 1982, Commonwealth Edison Company (CECO, the licensee) requested amendments to Facility Operating Licenses DPR-29. and DPR-30 for Quad Cities Station, Units 1 and 2, res pectively. The request is to authorize increased storage capability in the spent fuel pools (SPF) for the two nuclear units. The proposed modi-fications would increase the SFP storage spaces from the currently licensed 2920 spaces to 7684 spaces combined total for the two pools.
This expanded storage capacity will allow the continued operation of the two nuclear units with onsite storage of spent fuel to past the year 2000. The licensees basic supporting document for this action is a,,. ;
report, Spent Fuel Pool Modification for Increased Storage Capacity, Quad Cities Nuclear Unit 1, Docket No. 50-254, and Quad Cities Nuclear Unit i
No. 2, Docket No. 50-265, Rev.1, dated June,1981.
2.0 DISCUSSION The licensee's proposal would increase the SFP storage capacity by replacing the existing spent fuel storage racks with new high density The new racks will contain neutron absorber material in storage racks.
l the rack wa'is so that spacing between stored assemblies can be reduced -
I while maintaining adequate criticality margin.
The high density racks are made up of modules, each module being composed l
of six-inch square cells, each cell accommodating a single BWR fuel i
a ssembly. The cell ' walls contain a neutron absorber material sandwiched between sheets of stainless steel. The cells making up the module have 6.22-inch center-to-center spacing. The general arrangement of the modules-in the pools is shown in Figures 2.1 and 2.2 of the licensee's application and basic supporting document. The general details of 820429 D %
J
. design and construction of the racks are contained in Figures 3.1 through 3.8 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 interconnec ted.
The applicable codes, standards, and practices for this modification are set forth in Section 3.2 of the licensee's basic supporting l
document. A detailed structural analysis is described in Section 6 l
of the document to show the adequacy of the racks 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 structural acceptance requirements when conservatively analyzed.
The safety considerations associated with this proposed action are addressed below. A separate environmental impact appraisal has been prepared for this action.
'.i\\
l 3.0 ' EVAL'ATION
' ~ ~ '
U j
3.1 Structural and'Hechanical Design Considerations Description Quad Cities Units 1.and 2 each have fuel storage 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 3970 storage cells arranged in 20 racks with six :
different module sizes in this pool.
l 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 and nine inches 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.
\\
The racks will be constructed from ASTM 240 - 304, austenitic steel sheet material, ASTM 204-304 austenitic steel plate material, and ASTM 182 - 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 in the sub-elements which make up the fuel racks.
l
_3._
The rack assembly is typically supported on four plate-type supports.
The supports elevate the module base plate 6.5 inches above the pool floor level, 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 licensee's proposed action, established codes, standards and criteria were applied, consistent with the NRC's guidance, "0T Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Application,"
dated April,1978 and revised January,1979.
Accor.dingly, the design of the racks, fabrication, and installation criteria; the structural design and analys.is procedures for all loadings, including seismic and impact.
loadings; the load combinations; the structural acceptance criteria; the quality assurance requirements for design, and applicable industry codes were all reviewed in accordance with the applicable portions of that NRC guidance.
s For the design of the spent fuel modules, two sets of criteria were to be satisfied. The first
- establishes requirements to ensure that. adjacent racks will not impact;during the Safe Shutdown Earthquake (SSE), assuming the lower bound value of the potrl surface friction coefficient.
It is required u
by this criterion that the factors of safety against tilting be 1.5 for the OBE and 1.1 for the 'SSE.
The second set of criteria establishes requirements to ensure that loading combinations and stress allowables are in accordance with Section III. Subsection NF of the ASME 1980 Edition. The basic material allowables, fabrications, installations and quality control of the modules also conform with the same code. The loading considered in the analysis involves dead loads, live loads, thermal loading, and seismic loadings (OBE or SSE).
Additional analyses were perforned to evaluate the effects of a postulated accident involving the dropping of a fuel assembly on tne racks and on the fuel pool liner, and the fuel handling crane uplift accident.
A dynamic analytical model, consisting of beams, gaps, springs, dampers and inertia coupling representing fluid coupling between rack and assemblies, and between rack and adjacent racks, was used to predict the maximum sliding.
distance and seismic forces resulting from the SSE. These forces were then used to predict the seismic stresses and displacements. The. coefficient of friction between the stainless steel liner and the leveling legs of the 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 Steel for a Spent Fuel Rack Facility" dated November 5,1976. The result of 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 and the modules and the pool walls such that the inter-rack impact, or the rack collision with the pool walls, is precluded.
~
e
~
_4_
The analysis, design, fabrication, and criteria for estab'lishing installation procedures 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 +
r 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 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 theStandardReviewPlayection3.8.4.
~
N '.
ResuT s of the seismi6 ihd structural analyses indicate that the racks are capable of withstanding the loads associated with all design loading conditions. Also, fmpact due to fuel assembly / cell interaction has been considered, and will result in no damage to the racks or fuel assemblies.
. Two types of postulated fuel assembly drops onto the racks were analyzed by the licensee 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 l
location and impacting the base. The second drop involves a fuel assembly dropping from a maximum of 36 inches above the rack and hitting the top'
- ~~
of the rack.
In both cases, the impact energy is dissipated by local yielding; however, the sub-criticality of the fuel arrays ic not violated.
The dropping of a heavy load onto the protective pool liner of the pool floor was also analyzed. Although local damage and plastic deformation may occur, the overall structural integrity of the liner is maintained, The effect of postulated stuck fuel assembly due to the attempted withdrawal s
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 analysis show that the stress is below that allowed for the applicable loading combinations.
We find that with respect to structural and mechanical design the subject modification proposed by the licensee satisfies the applicable requirements of General Design Criteria 2, 4, 61, and 62 of 10 CFR, Part 50, Appendix A and is acceptable.
- ~
9 h
e r
3.2 Materials Considerations Discussion and Evaluation We have reviewed the compatibility 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 Boraflex neutron absorber material used in the high density storage 1.ocations for environmental stability.
There will be both the old and the new types of spent fuel storage 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 racks are of welded stainless steel constr'uction with a "Boraflex" neutron absorber sandwiched between the stainless steel sheets.
The neutron absorber is composed of boron carbide powder in. a,., rubber-like silicone polymeric matrix.
The old low densi y fuel storage tubes provide for the interim storage of fuel assemblies and are constructed of aluminum without neutron absorber materfal.
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 corro: ion 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 corrosion of 1.1 mils in twenty years.
Since the oxidation rate will continue to decrease slightly over this period, this estimate is considered to be conservative., ;
The inherent high corrosion resistance of aluminum and stainless steel makes them well suited for use in demineralized water. Aluminum and stainless steel fuel storage racks submerged in water have been in use for 10 years with no deterioration evident.
\\
W OS e
Aluminum and 300-series stainless steel are very similar insofar as their coupled potential is concerned. Because the pool water has very low-conductivity, galvanic corrosion should not occur. The use of stainless steel fasteners in aluminum to avoid detrimental galvanic corrosion is a recommended practice and has been used successfully for many years by the aluminum industry.
The pool liner, rack lattice structure and the high density fuel storage tubes are stainless steel which is compatible with the storage pool environment.
In this environment of oxygen-saturated high purity water, the corrosive deterioration of the type 304 stainless steel should not x 10-5 nches in 100 years, which is negligible t
exceed a depth of 6.0 relative to the initial thickness.
Dissimilar metal contact corrosion (galvanic attack) between the stainless steel of the pool liner, rack lattice structure, fuel storage tubes, and the Inconel and the Zircaloy ~
in the spent fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are therefore ab. similar galvanic potentials. The Boraflex poison material is composed of non-conductive materials and therefore will not develop a galvanic potential in contact with the metal components. Boraflex has undergone extensi'vdi, testing to study the effects of gamma irradiation
_,in various envird,n@ents, and to verify its structural integrity and suitability as a neutron absorbing material.
~
The space whic$ contains the Boraflex is vented to the pool. Venting will allow gas generated by the chemical degradation of the silicone polymer binder during heating and irradiation to escape, and will prevent bulging or swelling of the stainless steel tube.
To provide added assurance that no unexpected corrosion or degradation of the materials will compromise the integrity of the racks, the licensee
~~
has committed to conduct a long term fuel storage cell surveillance program. Surveillance samples are in the form of removable stainless steel clad Boraflex sheets, which are proto-typical of the fuel storage cell walls. These specimens will be removed and examined periodically.
Conclusions From our evaluation as discussed above we conclude t' hat the corrosion that will occur in the spent fuel storage pool environment should be of little significance during the remaining life of the plant.
Components in the spent fuel storage pool are constructed of alloys which have a low differential galvanic potential between them and have a high resis -
tance tc general corrosion, localized corrosion, and galvanic corrosion.
Tests under irradiation and at elevated temperatures in water indicate that the Boraflex material will not undergo significant degradation during the expected service life of 40 years.
e 9
w
- 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 licensee, will provide reasonable assurance that the Boraflex material will continue to perform its function for the design life of the pool. We therefore find that the implementation of a monitoring 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 permit appropriate periodic inspe9 tion and testing of components, and Criterion 62, by preventing criticality by maintainini. structural integrity of components ~and of the boron poison.
3.3 Installation and Heavy Load Handling Considerations The results of the staff's generic review of handling heavy loads at nuclear power planty,"i.e., NUREG-0612, " Control of Heayy Loads at Nuclear Power Plants, is ongoing and will not be ccipleted before the
-spent fuel pool m6dTficatior.s are to commence. Therefore, we have limited this review and evaluation to the heavy load handling operations 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 storage rack weighing 161/2 tons, whereas the main-haist on the reactor building 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 has 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 adjacent to the-enpty rack being moved.
Consequently, a load handling mishap will not impact on stored fuel.
Based on these s
considerations, we conclude the procedures are acceptable.
I 9"u..
n e -
~ The June 22, 1981 Commonwealth Edison response to our December 22, 1980 generic letter on control of heavy loads states that operator training 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 inspection, testing and maintenance related to
'~ '
Quad Cities cranes comply with ANSI B30.2-1976.
From this we conclude that adequate measures will be taken to assure the operability of the cranes used in handling the spent fuel pool modifications loads, and
'are therefore, in this respect acceptable.
A lifting yoke has been designed to handle the new storage racks.
It 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 permit the lifting rod spacing to be adjusted so as to pemit them to i
remain vertical and yet accommodate the seven different sized racks.
Figure.3-8 of 'the licensee's submittal indicates the lifting yoke is rated for 22.7 tons while the heaviest storage rack is 161/2 tons. Based on the above, we.c9nclude that the lifting yoke is ade,quate for handling the new storage racks, and therefore, acceptable.
-~
The existing aluminum open lattice storage 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 that rigging interposed between the crane hook and.the load is acceptable for handling the old storage racks, and that the crane meets the objectives of APCSB BTP 9-1 and has sufficient capacity for the described operatio.nss -
The travel paths, procedures, operator training and crane maintenance are adequate to accomplish the heavy load handling operations associated l
with spent fuel pool modifications and are therefore acceptable.
In regard 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 keff remains equal to or less than s
0.95, 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.
I The proposed modifications meet the guidelines of the applicable portions i
of the following:
Regulatory Guides 1.13, 1.29 and.1.71, 1.85, 1.92 and l.124; and 10 CFR Part 50, Appendix A, General Design Criteria 1, 2, 61, 62 and 63; Standard Review Plan Sections 3.8.3 and 3.8.4 and industry standgds ANSI N210-1976, ACI 318-77. AISC, ASTM, ASME Section III
,Difisioh 'I' Subsection NF 1980 and ASME Section IX-1980.
e w-
-y,---
s--e----
-r ry-.----.--%
-- --,,, r
--_--,,.e---.
w-1 -
. l 3.4 Criticality Considerations
~
Discussion and Evaluation The boron content in the neutron absorber material in the rack walls is equivalent to a B-10 areal density of 0.01728 grams per square centimeter. The multiplication factor of the racks is calculated for an 8 x 8 assembly having a uniform enrichment of 3.2 weight percent U-235. The infinite multiplication factor for.this assembly.in the standard reactor configuration at cold clean conditions ~is 1.362; 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).
s The rack design i*s.4hus conservative for assemblies which are anticipated
_tp be stored in th. racks. Other conservatisms present in the analysis s
include the use of the mintmum (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-KENO computer code package using the 123 group XSDRN cross-section set with the NITAWL subroutine for U-238 resonance shielding effects. This code has 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
- ~~
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 percen.t confidence level. Trend analyses were performed to obtain an estimate 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 for this over' prediction in the analysis.
\\
Sensitivity analyses were performed to obtain 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 0.97 percent reactivity change due to these effects.
e
. The cai ulated value of the nominal case multiplication factor was 0.9155 +.0067 where the uncertainty is the statistical uncertainty in the IIonte-Carlo (KENO) calculation only.
To thir 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 value 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'nd postulated accident conditions have been investigated. This includes improper positionina of an assembly in its storage rack, bowing of the channel, variations in pool temperature, a dropped fuel assembly, and a missing absorber plate in the racks. These analyses show that the criticality acceptance criterion 14 not v'iolated when not more than 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 the,, boron absorber material as designe,d.
=
In the course of our revieF, we have found that:
1.
State-of-th'e-art calculation methods which have been benchmarked against critical experiments have been used, 2.
Credible abnormal configurations have been invr _ eigated, 3.
Uncertainties and biases have been treated, and'
~
4.
The result, including all uncertainties, meets our acceptance criteria for the nominal case and for abnormal and postulated accident conditions.
From the above considerations, we find that fuel assemblies of the 8 x 8 two-water rod design, having. average enrichments less than or equal to 3.2 weight percent U-235, other fuel designs containing less than 15.49 grams of U-235 per axial centimeter, or BWR assemblies having cold clean infinite multiplication factors in the Quad Cities. reactor geometry s
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 fuel racks with adequate criticality margin.
- - -. ~..
9 0
m
. 3.5 Spent Fuel Pool Cooling Considerations Description and Evaluation Quad Cities Units 1 and 2 each has a stainless steel lined reinforced concrete spent fuel storage pool.
The two pools are joined by a transfer canal.
Fuel can be transferred between the two pools via the transfer canal after opening the two gates, located at the sides of the respective pools. A normal fuel discharge, i.e., about 200 assemblies, 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,ll not occur simultaneously.
Separite 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 res,pective pools below 125 degrees F during maximum normal discharges, when the reactor building closed cooling water system
)
is. at its maximum temperature of 105 degrees F.
This assures that a comfortable working environment can be maintained during normal conditions.
Further, on those infrequent off normal conditions where, i
--for example, a fu1TTcore discharge occurs,the pool wat.er. temperature will not exceed 150 degrees F.
Analyses of the pool water temperatures 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 degrees F limit given in the Standard Review Plan Section 9 l.3 - Spent Fuel Pool Ccoling and Cleanup System and is acceptable.
Further, the analysis of the l
i 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 degrees F.
This is less than the 150 degrees F stated in the FSAR, and' 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 temgerature is at.125 degrees F each heat exchanger is rated at s
3.65 x 10 BTU /hr. Therefore each unit's spent fuel pool cooling system l
has a total design flow of 700 000 pounds per hour and a total heat 0 BTU /hr at a pool water temperature 'of removal capability of 7.3 x 10 125 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 BTU /hr.
cooling system increases to approximately 10.9 x 10
- - =.
I i
In addition to the above spent fuel pool cooling system, provisions have been made to cross tie the spent fuel pool cooling system to the residual heat removal (RHR) system. This is. accomplished by installing two 6 inch pipe size spool pieces in the two legs of the scent fuel pool cooling loop. The six inch RHR tie-in line will provide an additional spent fuel pool cooling water flow of 1,000 gpm f.e., 500,000 pounds 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 y
interchange-of water Setween the two pools.
~
{ [-
~'
Decay Heat The licensee has analyzed five different. cases of spent fuel' ~ pool: decay heat loads and the resultant pool water temperatures with and without'-
the additional cooling provided by the residual heat removal system (RHR).
The cases investigated are as follows:
(.1) The pool is filled with normal discharges of 240 fuel assemblies
.and cooligg %only provided b'y the SFPCS (decay heat equals 11.2 x 10 BT07hr)..
~
(2) The pool is filled with normal discharges of 240 fuel assemblies and cooling is provided by the SFPCS and the RHR system (decay 6 BTU /hr).
heat equals 11.2 x 10
(;3) The pool is fil. led with normal discharges of 200 fuel assemblies and cooling is provided only by the SFPCS (decay heat equals 9.65 x 106 BTU /hr)..
(.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).
l l
(5) The pool is filled, with normal discharges plus a recently discharged full core and cooling is provided by,the SFPCS and RHR 6 BTU /hr).
system (decay heat equals 24.7 x 10 N
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 assemblies can be accomplished in two days.
In the case of a full core discharge, six days will be required to transfer the fuel to the storage pool.
l t
w
. 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 i
fuel is placed in the pool. Although no safety problem is created by l
a somewhat higher pool temperature, the' higher temperature encroaches upon marain assumed in our 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 licensees analysis shows that the pool temperature will not. exceed i
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.
Makeup Water.
The spent fuel po,ol, system is designed to minimize the lo'ss of water from the pool and to prdv.ent the water level from falling below a safe level
_abovethestoredfiel.
For example all penetrations into the pool,
~
except for valved drains, are located at a height such~th,at 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 of makeup water, the condensate storage tanks and the fire system.
Approximately 550 gpm of condensate water can be delivered to the 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 4
,~
l 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. A5out three hours would be required to install the spool piece.
In the event that the above identified sources of water become unavailable, l
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 o.f any reasonable need.
l
~m 9
. la.
We conclude the makeup water system is adequate and acceptable because makeup water is available from the condensate storage tanks and river via the fire systen,and their respective makeup rates exceed the boil off rate descriEed helow.
Further, this makeup water can 6e
.made available before b' oiling 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 li 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 no exchange of water between 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 above conditions about 71/2 hours would elapse before bulk boiling would occur. The maximum boiloff rate would be 51 gpm.
Based on the above, 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 boili,ng,and the makeup rates from both makapp sources exceed the Boil off rate, and therefore
~the provisions for makeup wa.ter are acceptable.
Local Boiling Using a conservative thermal hydraulic circulation model of poo.1 water flowing down along the walls, laterally across the pool flocr in the water plenum and up.through the stored fuel assemblies, the maximum calculated water temperature at the outlet of the fuel assemblies was shown not to exceed 167 degrees Fahrenheit.
The sat'uration temperature at this point is 240 degrees F.
Due to the margin between these two temperatures we conclude that nucleate boiling will not occur and in this respect the design is acceptabl e.
Conclusion Cooling capability for the spent fuel pools for the two nuclear units s
has been evaluated for the maximum expected loading conditions for the new racks. We conclude that the presently installed pool cooling capability is adequate to handle the heat load under any reasonably expected conditions of operation.
--e i
,e 3.6 Spent Fuel Pool Cleanup System Description and Evaluation The spent fuel pool cleanup system consists of a filter demineralizer (precoat filter material and powdered anion and cation resin), 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 monitored by a continuous. conductivity meter installed on the inlet to the fuel pool demineralize'rs, and by periodic grab samples for laScratory analysis' Once a week a repre-sentative grab sample is obtat'ned from the fuel pool demineralizer inlet line 'for pH, for chloride, stitca, and turbidity analysis. Weekly activity checks are made for gross Beta and gross alpha activity. Once
'a month a sample from the same location is obtained for a gamma isotopic analysis. All, peaks are identifted. All identified isotopes are quantified, 'and an LLD is determined for Kr-85.
Th'e criterion for. a,demineralizer backwash and precoat.fs a consistent excursion from the'.fchemistry limits, or high differential pressure 4 25 psid} acrossJthe demingtaltzer. We agree with the; licensee that 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 amaunt of core to be replaced for each fuel cycle, and frequency of operating 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 l
proposed high density fuel storage. Past experience also shows that no significant leakage of fission products from spent fuel stored in pools occurs after the fuel has cooled for several months. To maintain water quality, the licensee has established the frequency of chemical and radionuclide analysis that will be performed to monitor the water quality and the need for spent fuel pool cleanup system demineralizer l
resin and filter replacement.
In addition, the licensee has also set s
the. chemical and radiochemical limits to be used in monitoring the spent fuel pool water quality and initiating corrective action.
We agree with the licensee that the increased quantity of spent fuel to be stored will not contribute significantly to the I
amount of radioactivity from fission products in the spent fuel pool water.
- ~...
- The proposed expansion of the spent fuel pool will not More frequent replacements of filters or fuel pool cleanup system.
d
. demineralizer resin, required when the differential pressure excee s 25 psid or decontamination effectiveness is reduced, impurities.in the pool water as a result of the expansion o system with the proposed high density fuel storage (1) provides the spent fuel.
capability and capacity of removing radioactive materials, corrosion products, and impurities from the pool and thus meets the requirement of General Design Criterion 61 in Appendix A of 10 CFR Part 50 as it relat.es to appropriate fuel s'torage systems, (2) is capable of reducing
~
occupational exposures to radiation by removing radioactive products from the pool water, and thus meet the requirements of Section 20.l(c of 10 CFR Part 20, as it relates to maintaining radiation exposures as low as reas,onably achievable; (3) confines radioactive materials in the pool water into the filters and demineralizers, and thus meets Regulatory Position C.2.f(c) of Regulatory Guide 8.8, as it relates to the source; and (4) removes reducing the spneaf of contaminants from suspended impurit.ies from pool water by filters, and. hus meets t
" Reg'ulatory PositT66 C.2f(,3) of Regulatory Guide 8.8, as it relates to removing crud from flui'ds through physical action.~
Conclusion _
On the basis of the above evaluation, we conclude that:
(1) The existing s' pent fuel pool cleanup system meets General Des Criterion 61 of 10 CFR Part 50, Appendix A, Section 20.l(c) of 10 CFR Part 20 and the appropriate Sections i
(2) The existing spent fuel pool cleanup system is adequate for the.
proposed modification.
(3) The' conclusions of the, evaluation of the waste Report (August 25, 1971], are unchanged by the modification of t s
spent fuel storage system.
3.7 Occupational Radiation Exposure Description 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. Exposur 8
+
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 tfie number of individuals performing a specific job, their occupancy time while performing this jo6, 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 because 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 firstt moved into the fuel pool, the addition of crud to the pool water from the fuel assembly and from tfie 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 purification system for the pool, which has Pept 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 'pdesented three alternative plans for;' removal and
_ disposal of the dhi-racks. These are (1) to crate and ship intact racks to a licensed burial facility 1*(2) to cut the racks into 3 mall 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 alternative 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
, ~.
lines for each of the alternatives.
It should be noted that the procedure ~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 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 including clean-up and diver work.
Conclusion Based on our review of the manner in which the licensee will perform 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 ensure as' low as is reasonably achievable (ALARA) exposures to workers.
~~
~.
e
4.0 CONCLUSION
We have performed an evaluation of the licensee's proposed modifications based primarily on information provided to us in the licensee's basic
. supporting document. This document has been revised and supplemented during the course of our review in response to staff questions, and from e
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 'o'f construction are pcceptabl e.
(2)
The installation and use of the proposed fuel handling racks can be accomplished safely.
(3) The lik'elihood 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 whileyour generic review of th6fssues is underway.
.w (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 assembly and the subsequent release of the radioactive inventory within the gap of each fuel pin, as already reviewed and approved in the FSAR for Quad Cities Station.
(5) The physical design of the new storage racks will preclude criticality for any credible moderating condition.
(6) The cooling system for each of ^he spent fuel pools has acceptable cooling capacity.
(7) The conclusions 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 above, that:
(1) 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 be conducted in compliance with the Commission's regulations and the proposed license amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated: April 9, 1982
-