ML19253B987
| ML19253B987 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/19/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19253B981 | List: |
| References | |
| NUDOCS 7911020525 | |
| Download: ML19253B987 (9) | |
Text
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UNITED ST ATES NUCLEAR REGULATORY COMMISslON
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SAFETY CVALUATION BY THE OFFICE OF NUCLEAR REACTOR R2GULATIC!'
SUPPORTING AttENDf'ENT NO. 34 TO FACILITY OPERATIt!G LICENSE f:0. CPR-50 METROPOLITAN EDISCt' C0f*PA!'Y JERSEY CENTRAL POPER A!!D LIGHT CCFPAT.Y PEN!!SYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATI0f!, UNIT NO.1 DOCKET NO. 50-289 1.0 Infroduction By letter datej February 3,1977, Metropolitan Edison Company
( MEC ) proposed to modify the spent fuel pool (SFP) storage arrangenent for SFP "B" at the Three Mile Island Nuclear Station Unit No. 1 (TMI-1) from the design which was reviewed and approved ouring the operating license review and which is de-scribed in the TMI-1 Final Safety Analysis Report and Technical Specifications. The proposed mooification would replace the storage racks presently approved for SFP "B", which provide storage capacity for 174 fuel assemblies, with new racks which With would provide storage capacity for 496 fuel assemblies.
this modification the total storage capacity for SFP's "A" and "B" would be increased frou 430 asseymblies to 752 assemblies.
This modification was reauested by tiEC based on its projections of the ncnavailability of offsite spent fuel storage or reproc-essing facilities prior to filling the presently authorized storage capacity.
The new storage racks will be constructed from stainless steel and are designed to seismic Category I criteria.
The new racks consist of a rectangular array of storage cells welded to lattices of structural stainless steel channel located near the top and bottom of the cells. The lattice forred from the stainless steel channels limits structural deformation and maintains a nominal center-to-center spacing of 13.625 inches between adjacent storage cells.
The cells have a square cross section with a 9.l?
inch 1.D. and a.187 inch wall thickness. The new racks are supcortec by the existing "B" pool floors and walls and utilize compressicn type restraints with pads at the points of contact with the p001 liner. The racks wili be fabricatec in nodules consisting of 5X5, 5X4, and 8X2 cells.
1505 030 d
ygn o 2
. We have reviewed the proposed modification and by letter dated April 8,1977, reouested additional informtion.
This additicr.al information was provided by MEC in a letter dated llay 2?, 1977, and by GPU Service Corporation (consultant to fiEC
) in a letter dated July 21, 1977.
Our review addressed the following considerations:
criticality, fuel pool cooling, structural and mechanical considerations, material considerations, fuel handling, rack installation, occupational radiation exposure and radioactive waste treatment.
2.0 Evaluation 2.1 Criticality Analysis In its February 3,1977, submittal f1EC states that its criticality calculations are based on fuel assemblics with fresh (i.e., unirradiated) fuel with a nominal 3.5 weight per-cent uranium-235 content and containing no burnable poison or control rods.
MEC also states that the 3.5 percent enrich-ment corresponds to a fuel loading of 45.9 grans of uranium-235 per axial centimeter of fuel assembly.
NUS Corporation, a consultant to MEC, perforned the criticality analyses assuming the racks to be fully loaded.
For parametric calculations, NUS used their version of the LEOPARD computer pro-gram, called NUMICE, to obtain four group cross sections for PDQ-7 diffusion theory calculations.
The accuracy of this rethod was checked by using it to calculate water-moderated, uranium lattice experiments.
NUS states that the calculated neutron multiplication factors obtained from NullICE/PDQ-7 deviated from the experimental values by an average of + 0.009.
In order to ensure that the results of these four group calculations for the storage lattice were accurate, NUS used the KEN 0 Monte Carlo program with 123 group cross sections from the XSDRM progran with the GAM-THERMOS library to check selected cases and to verify the neutron multiplication factor of the final design.
This method was checked by using it to calculate critical experirents of shipping cask configurations. This series of calculations showed that this GAM-THERMOS / KEN 0 nethod yielded neutron multi-plication factors that are within + 0.008 of the experimental values.
However, there is an additional statistical uncertainty af +.008 in these calculations.
1505 031
e The use of these computer programs gives a neutron multiplication factor of 0.89 for an infinite array of these spent fuel assemblies located in the nominal storage lattice, which is assured to be at a temperature of 20 C, with no soluble poison present.
Because the rack design allows a free space of 0.3 inches between a centered fuel assembly and each of the container walls, it would be possible for assemblies to be located off center (i.e.,
eccentrically) in the storage containers.
Eccentric loading will increase the neutron multiplication factor.
Other factors that could increase the neutron multiplication factor in the spent fuel storage pool are:
(1) mechanical desicn tolerances; (2) increased U-235 content (assumed 102% of nominal); (3) possible variations in stainless steel composition; and (4) increased water temperature.
flVS calculated that all of these factors acting together could increase the neutron multiplicathn factor by 0.024.
f1EC states that it will not be possible to inadvertently bring a transient fuel assembly up to the outside of a fully loaded rack because:
(1) all of the racks will be installed before fuel storage comences; (2) after the racks are installed, there will not be any open water regions except between the racks and pool walls; and (3) a permanent barrier will be installed in each gap between the racks and the pool walls, as necessary, to prevent the insertion of an assembly.
By summing the maximun calculaticnal deviation of 0.008 from experiment, the statistical uncertainty of 0.003 in the KEf10 calculations, and the 0.024 effect of the worst tolerances and conditicns, NUS finds the taximum neutron maltiplication factor for this storage lattice to be 0.93.
The staff has reviewed the analytical model used by NUS to perform their calculations and has concluded that the model is capable of accurately predicting the maximum neutron multiplication factor.
In addition, the above results compare favorably with the results of parametric calculations made with another rethod for a similar fuel pool storage lattice.
Accordingly, we conclude that the NUS calculations are substantially accurate.
By assuming neu unirradiated ft al with no burnable or soluble poison, the NUS calculations give the maximun neutron multiplication facter that could be obtained throughout the core life of the nominal fuel assenbly.
This includes the effect of the plutoniun which is generated during the fuel cy.cle.
Therefore, we find the maximun neutron multiplication factor in the pool to be 0.93.
1505 032
. To conform with the assunptions in the criticality analysis, MEC has agreed that the station's Technical Soecifications should be modified to prohibit the storage of fuel assemblies that contain more than 46.8 grams of uraniun-235 per axial centi-meter of assembly. This corresponds to 102S of the nominal U-235 loading and is considered in the calculations cited above.
We find that when any number of fuel assemblies, which have no more than 46.8 grans of uranium-235 per axial centimeter of fuel assembly, are loaded into tha proposed racks, the neutron nulti-plication factor will be less than 0.93.
Since this factor is less than our acceptance criterion of 0.95, we conclude that b'ased on criticality considerations, the proposed design is acceptable.
2.2 Spent Fuel Cooling MEC plans to refuel annually. This will require the replace-ment of about 52 of the 177 fuel assemblies in the core every year.
In its February 3,1977 submittal fiEC assumed a 150 hour0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> time interval af ter the reactor is shutdown prior to moving fuel durina the annual refueling and during any full core off-loading into the spent fuel pml.
For this cooling time,
!!EC stated that the heat load on che SFP cooling systen for any annual refueling will notexceed 9.7X106 BTU /hr (2.8 MWt) and that the heat load for the full core of f-load,6which fills the capacity of the racks, will BTV/hr (7.5 MWt).
MEC stated that these not exceed 27.7X10 heat loads were calculated with the ORIGErl point depletion pro-gram which was developed at the Oak Ridge flational Laboratory.
In Section 9 of the Three Mile Isli to FSAR, MEC stated that the SFP cooling system consists of two pumps,each rated for a flow of 1000 gallons per minute,and two heat exchangers each rated for reoving 6.0X106 BTU /hr. At these rated conditions, the spent fuel cooling system will reduce the SFP outlet water temperature by 12 F.
In addition, Mir stated in its February 3,1977 submittal that the seisnic Category I Decay Heat Removal System (DHRS) is connected to the SFP cooling systen so that it can be used to cool the pool during reactor shutdown periods when there is an excess core cooling capability.
In its February 3,1977 subnittal, MEC states that. the SFP outlet water temperature will be maintained at or below 135'F during any annual refueling and at or below 147'F during a full
. core off-load.
1505 033
. By using the cnnservative method given on pages 9.2.5-8 tnrough 14 of the NRC Standard Review Plan, we find that about ten days of cooling, rather ti.an MEC's 159 hours0.00184 days <br />0.0442 hours <br />2.628968e-4 weeks <br />6.04995e-5 months <br /> (6.25 days), would be required for the heat loads to decrease to those stated in MEC's February 3, 1977 sub-mittal.
However, we find that t'EC's 147'F value for the maximum fuel pool outlet water temperature to be sufficiently conservative for the rated flow rates, and that the 147'F outlet water temperature will not be exceeded even with only 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of cooling time.
If, shortly af ter placing the 52 fuel assemblies from an annual refueling in the SFP, one of the two SFP cooling pumps were to fail, the fuel pool outlet water temperature would not exceed 147 F.
If both pumps were to fail, the excess capacity of the seismic Category I DHRS could be used to keep the outlet water temperature below 147 F, as long as the reactor was shutdown.
By the time the reactor is started up after a refueling operation, only one SFP cooling pump will be needed to maintain the outlet water temperature below the stated i35 F so the other pump will provide for redundancy.
When a full core is off-loaded into the fuel pool, the DHRS, which is designed to engineered safety feature criteria and seismic Category I criteria, will be available for cooling the SFP if it is needed. We find that a single failure in this system will not cause the SFP outlet water temperature to increase above 147 F.
We therefore conclude that the present cooling capacity in TMI-1 will be sufficient to accommodate the increrental heat load that will be added by the proposed modifications.
We also find that this incremental heat load will not alter the afety con-siderations of SFP cooling from that which we pre.iously reviewed and found to be acceptable.
2.3 Installation of Racks and Fuel Handling
!1EC states that the proposed fuel rack modifications will be made in a dry, erpty pool which has not previously contained spent fuel assemblies.
It is further stated that the installation will not require movement of the new racks over the other $FP or over the storage area for new fuel.
1505 034
. Since there will be no fuel assemblies in SFP "R" while it is being modified, it will not be possible for an accident in this pool to result in any increased ne'utron multiplication factor.
The NRC staff has underway a generic review of load handling operations in the vicinity of SFPs to deternine, among other considerations, the likelihood of a heavy load impacting fuel in the pool. As an interim measure pending completion of this review and to facilitate installation of the modified fuel storage racks, fiEC has agreed to anendment of the TM1-1 Technical Specifications to provide administrative limits on the handling of loads weighing in excess of 3000 pounds.
These limits have been selected to prohibit handling such loads over irradiated fuel or in such a manner that a dropped load which tipped cver ceuld damage spent fuel. The limits also require that such loads be handled at the minimum practicable height.
ItEC has also agreed to an amendment of the TMI-l Technical Specifications which would prohibit the presence of the Spe.
Fuel Cask in the Unit 1 Fuel Handling Building pending completion of tae review cf load handling operations.
v!e conclude that with these additional limitations, as set forth in the imended Technical Specifications, installation of the modifiej racks will not significantly affect the probability or con-sequences of the design basis accident for the SFP, i.e.,
the rupture of a fuel assembly and subsequent release of the assembly's radicactive inventory.
2.4 Structural and Mechanical The new spent fuel storage rack designs are designated Seismic Category I and were reviewed for the following in accordance with the applicable portions of Sections 3.7 and 3.8 of the Standard Review Plan:
structural design and analysis pro-cedures for all loads including seismic and impact loadings; supporting arrangenents for the racks including their restraints; loading combinations and structural acceptance criteria and quality control for design, fabrication and installation.
Seismic analyses of the fuel storage racks were perforned using a response spectrum modal dynamic analysis, envelopej over the elevation changes, in the two horizontal directions and a static seismic analysis in the vertical direction in acce dance with 1505 035
. Section 3.7 of the Standard Review Plan.
The rodal responses for each horizontal direction and the combination of each of the independent direction results were arrived at in accordance with Regulatory Guide 1.92.
The effective mass of the water and the fuel-cell interaction were also included in the seismic analyses.
The existing poci structure was analyzed, using a finite elenent model, for the increased loading conditions imposed by the new high density storage racks and all loadings and load com-binations were in accordance with Section 3.8.4 of the Standard Review Plan. Welding is to be performed in accor-dance with Section IX of the ASME Boiler and Pressure Vessel Code.
Based on the above, se find that the analysis, design, fabrication, and installation of the proposed racks are in accordance with accepted criteria, and are in conformance with the rules of ASME Boiler and Pressure Vessel Code and AISC " Specification for Design, Fabrication and Erection of Structural Steel for Buildings" including supplements 1, 2, and 3.
Accordingly, we renclude that the effect of the additional loads imposed or the existing pool structure by the proposed modification ar e within acceptable limits and therefore that the proposed modification is acceptable with respect to structural and mechanical considerations.
2.5 tiater'al The SFP racks, their associated hardware, the seismic restraints, and the pool liner are all constructed of stainless steel.
Based on our review and operating experience to date, we conclude that, considering the pool temperature and the quality of the demineralized pool water, and taking no credit for inser-vice inspection, there is reasonable assurance that no significant corrosion of the racks, fuel cladding or pool liner will occur o_
1505 036
.e
. over the lifetime of the plant.
TI)is issue, however, is under generic review by the NRC Staff.
If the future results of this investigation indicate that additional pro-tective measures are needed, we will at that time require implementation of appropriate measures.
2.6 Occupational Radiation Exposure We have estimated the increment in onsite occupational dose resulting from the proposed increase in the number of stored fuel assemblies. This estimate was developed on the basis of information supplied by MEC and by utilizing realistic assumptions for occupancy times and for dose rates in the spent fuel area from radionuclide concentrations in the SFP water. 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.
Based on present and projected operaticns in the SFP area, we estimate that the proposed modification will add less than one percent to the total annual occupational radiation exposure burden at this facility. The small increase in radiation exposure will not affect MEC's ability to maintain individual occupational doses to as low as is reasonably achievable and within the limits of 10 CFR 20. Thus, we conclude that storing additional fuel in the SFP will not result in any significant increase in doses received by occupational workers.
2.7 Radioactive Waste Treatrent The station contains waste treatrent systems designed to collect and process the gaseous, liquid and solid wastes that might contain radicactive naterial. The waste treatrent systens are evaluated in the TMI-l Safety Evaluation Report (SER) dated July 1973.
There will be no change in the waste treatment systems dcscribed in Section 11.0 of the SER and r.o change in the conclusions of the evaluation of these systems in Section 11.0 of the SER because of the proposed rodifications.
3.0 Technical Soecifications By :_tter dated February 3,1977, "C proposed an amendmen+.
to the TMI-l Technical Specificati=.
Section 5.4.2.d, which would revise the description of the facility's design features to reflect the proposed increased spent fuel storage cacacity.
During our. review we found it necessary to include three additional provisions to the Technical Spccifications.
Thera 1505 037
/
9-were:
(1) a limit of 46.8 grams per axial centimeter of fuel element stored in the SFP, (2) limits on handling loads weighing in excess of 3000 pounds, and (3) a prohibition on the presence
,ent fuel handling casks in the TMI-l Fuel liandlin-Bu1 _ a. ]
' ding completion of our review of load hsnd' S. peratS. s n that building.
These additional revie'cas hace been discussed with and accepted by MEC.
4.0 Summ,
Our evaluation supports the conclusion that the proposed modification to the SFP at TMI-1 is acceptable because:
(1) The physical design of the new storage racks will preclude criticality for any credible moderating condition with the limits to be stated in the Technical Specifications.
(2) The SFP cooling system has adequate cooling capacity.
(3) No shielded cask movement will be permitted within the Fuel Storage Building prior to the conpletion of the cask drop analysis reviaw and no movement of loads in excess of 3000 pounds will be allowed over or near irradiated fuel assemblies in the SFP's.
(4) The structural design and the materials of construction are adequate to function normally for the duration of the plant lifetime and to withstand the seismic loading of the design basis earthquake.
(5) The increase in occupational radiation exposure to individuals due to ' '
storage of additional fuel in the SFP would be neg.igible.
(6) The installation and use of the new fuel racks does not alter the probability or consequences of the design basis accident for the SFP, i.e., the rupture of a fuel assembly and subsecuent release of the assembly's radioactive inventory.
5.0 Conclusion lie have concluded, 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 Comnission's regulations and the issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public.
Dated: December 19, 1977 1505 038