ML15112A983
| ML15112A983 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Vermont Yankee  |
| Issue date: | 12/24/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML15112A982 | List: |
| References | |
| NUDOCS 8101100462 | |
| Download: ML15112A983 (9) | |
Text
1 o5 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.90 TO FACILITY OPERATING LICENSE NO. DPR-38 AMENDMENT NO.90 TO FACILITY OPERATING LICENSE NO. DPR-47 AMENDMENT NO. 87 TO FACILITY OPERATING LICENSE NO.
DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 DOCKETS NOS. 50-269, 50-270 AND 50-287 INTRODUCTION By letter dated July 25, 1980, as supplemented July 1, August 7 and 14, October 15 and 31, November 3, and December
, 1980, Duke Power Company (DPC or the licensee) requested an amendment to Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55 for the Oconee Nuclear Station, Units Nos. 1, 2 and 3. The request would revise the provisions in the Station's common Technical Specifications (TSs) to allow an increase in Units Nos. 1 and.2 common spent fuel pool (SFP) storage capacity from 750 to a maximum of 1312 fuel assemblies through the use of neutron absorbing "poison" spent fuel storage racks.
The expanded storage capacity would allow the Oconee units to operate until about 1986 while still maintaining the 'capability for a full core discharge.
The major safety considerations associated with the proposed expansion of the SFP storage capacity for the Oconee Station are addressed below. A separate Environmental Impact Appraisal has been prepared as part of this licensing action.
DISCUSSION AND EVALUATION Criticality Considerations The licensee has provided an analysis of the criticality of the proposed storage racks. The analysis was performed by DPC with the KENO-IV code - a three dimensional Monte-Carlo program designed for reactivity calculations.
Cross-section input to the code is from the ENDF/B-IV compilation which is pro cessed by the AMPX system of codes. This analysis procedure'has been verified by using it to calculate a series of 27 critical, experiments. These experi ments spanned the enrichment range of interest to the Oconee racks and included experiments with separated fuel assemblies having stainless steel and boral absorbers interposed. From this comparison a calculational bias and variability were determined.
In addition to the base case calculation, the effect of mechanical uncertainties on biases and uncertainties was examined.
These included the pile-up of mechanical (D 310 0'1p
-2 tolerances, particle self-shielding in the boron, the effect of bowing in the cans, etc. The calculational uncertainty and mechanical uncertainties were. summed' to obtain a total uncertainty. The result of the analysis is an effective multiplication factor for the racks of 0.9475 with all uncer tainties included.
The effect of accidents on the reactivity of the racks has been analyzed.
Storage of an assembly in a location other than analyzed is precluded by rack design. The effect of other accidents is dominated by the presence of a large boron concentration in the water so that the value of the effective multiplication factor is smaller for the accident configurations than the design value.
Conclusion on Criticality We have reviewed the submittal and conclude that the rack design is acceptable from the criticality point of view. This conclusion is based on the following:
- 1. The analysis methods as used by DPC are state-of-the-art and have been, verified by comparison with critical experiments which incorporate the main features of the rack design.
- 2. The uncertainties evaluated encompass those expected to be encountered. For some effects, the limiting conservative value has been used in the analyses.
For others, sensitivity studies have been used to obtain an uncertainty in the rack multiplication factor.
- 3. Credible accidents have been considered and shown to have acceptable con sequences.
- 4. The value of the effecttve multiplication factor meets our acceptance cri terion, less than or equal to 0.95, when all uncertainties have been added.
Thus we conclude that any number of fuel assembliesof Babcock & AWilcox (B&W) 15 X 15 design can be stored in the racks provided that the uranium in the fuel has an enrichment no larger than 4.3 weight percent U-235.
Spent Fuel Cooling The licensed thermal power for Oconee Units Nos. 1 and 2 is 2568 MWt each.
DPC plans to refuel these reactors every 18 months at which times about 70 of the 177 fuel assemblies in the cores will be replaced. To calculate the maximum heat loads in the'SFP, DPC assumed a 168-hour time interval between reactor shutdown and the time when either the 70 fuel assemblies in the normal refueling or the 177 fuel assemblies in the full core offload are placed in the SFP. For this cooling time, DPC used the method given in NRC Branch Technical Position APCSB 9-2 (BTP) to calculate maximum heat loads of 21.9 x 106 BTU/hr for a normal refueling and 34 x 106 BTU/hr for a full core offload.
-3 The spent fuel cooling system presently consists of two pumps and two heat exchangers.
Each pump is designed to pump 1000 gpm (5.0 105 lbs./hr.),
and each heat exchanger is.designed to transfer 7.75 x 10 BTU/hr from 125 0 F fuel pool water to 90oF Recirculating Cooling Water (RCW),
which is flowing through the heat exchanger at a rate of 5.0 x 10 lbs./hr.
DPC states that this system will be sufficient to keep the SFP water temperature below 150 0F, the pool design temperature, until April 1981, prior to the Oconee Unit 1 refueling in 1981, when an additional SFP cooling pump and heat exchanger of the same capacity will be installed.
We find this acceptable.
Using the method given on pages 9.2.5-8 through 14 of the November 24, 1975, version of the NRC Standard Review Plan, with the uncertainty factor, k, equal to 0.1 for decay times longer than 107,seconds, we calculate that the maximum peak heat load during the refueling which would fill the pool could be 22 x 10 BTU/hr and that the maximum peak heat loads for a full core offload that essen tially fills the pool could be 34 x 106 BTU/hr. This full core offload was assumed to be a fully irradiated core which was taken out of its reactor vessel 35 days after the other Oconce unit, which shares this SFP, had beGn refueled.
We also find that the maximum incremental heat load that could be added by in creasing the number of spent fuel assemblies in the pool from 750 to 1312 is 1.9 x 106 BTU/hr. This is the difference in peak heat loads for the present and the modified pools.
We conclude that with the three pumps operating, as DPC has committed to provide by Aprif 1981, the cooling system can maintain the fuel pool outlet water tempera ture below 125 0F for the normal refueling offload that fills the pool and below 136 0F for the full core offload that fills the ppol.
In the highly unlikely event that all three SFP cooling systems were to fail at the time when there was a peak heat load from a full core in the pool, we calculate that the maximum heatup rate of the SFP water would be 9.00F/hr.
Thus, if the water were initially at an average temperature of 125 0F/hr, it would be more than nine hours before boilino would start. We also calculate that after boiling starts the required water makeup rate will be less than 70 gpm. We conclude that nine hours will be sufficient time to establish a 70 gpm makeup rate.
Conclusion on Spent Fuel Cooling We conclude that the present two loop cooling system is adequate to handle the heat load of 342 spent fuel assemblies. The licensee has committed, in his June 24, 1980 letter, not to exceed this number of spent fuel assemblies in the Units 1 and 2 SFP until the additional cooling train is in service.
We conclude that the cooling capacity of the three loop system proposed by DPC for the Oconee Nuclear Station Units 1 and 2 SFP cooling system will be suffi cient to handle the heat load that will be added.by the proposed modifications.
We also conclude that the incremental heat load due to this modification will not alter the safety considerations of spent fuel cooling from that which we pre viously reviewed and found to be acceptable.
-4 Installation of Racks and Fuel Handling In their July 1, 1980 proposal, DPC states that at the time of the installation of the new racks there will be 342 spent fuel assemblies in the pool.
The licensee's installation plan is to both remove old racks and install new racks, under water, from the north end of the pool, making use of the existing Cask Storage Platform. The plan is to move the racks in the pool at an elevation which is lower than the top of any stored fuel assemblies, such that there will be no movement of racks over stored fuel.
Conclusion on Fuel and Rack Handling We conclude that DPC's plan will insure that no racks will be moved over the spent fuel assemblies in the pool.
After the racks are installed in the pool, the fuel handling procedures in and around the pool will be the same as those procedures that were in effect prior to the proposed modifications. On this basis, we conclude that the fuel and rack handling procedures are acceptable.
Structural and Seismic Loadings The spent fuel storage rack is composed of individual storage cells made of stainless steel.
Each cell has a lead-in opening which is symmetrical and is blended smooth. These racks utilize a neutron absorbing material, Boraflex, which is attached to each cell.
The cells within a module are interconnected to form an integral structure. Each rack module is provided with leveling pads which contact the SFP floor and are remotely adjustable from above through the cells at installation. The modules are neither anchored to the floor nor braced to the pool walls.
The SFP is constructed of-reinforced concrete lined with stainless steel clad plate.
No alteration was made to the pool design to accommodate more spent fuel.
Rather, more fuel assemblies are fitted into the existing pool configuration by reducing spacing between the fuel assemblies and installing a neutron absorbing material.
The proposed modification for the spent fuel storage capacity expansion program has been reviewed in accordance with the NRC report "Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", issued April 1978 and revised January 18, 1979. The structural review consisted of an examination of the following areas:
the proposed design criteria, the design loads and load combinations, methods of analysis, the dropped fuel accident, the material pro perties, the hydrodynamic effects, and the effect of increased loads on the floor slab and liner.
The spent fuel rack is made of stainless steel.
The material properties for structural components of the spent fuel racks used in the analyses were taken from Section III of the ASME Code. Load combinations and acceptance limits are in conformance with the NRC Standard Review Plan,.Section 3.8.4, and ASME Section III, Subsection NF.
The Oconee Units 1 and 2 SFP poison racks have been designed to meet the require ment for Seismic Category I Structures. The dynamic response of the fuel rack
-5 assembly during a seismic event produced the largest stresses among the load ing conditions considered.
The dynamic response and internal stresses and loads are obtained from a seismic analysis which is performed in two phases. The first phase is a time history analysis on a simplified nonlinear finite element model.
The second phase is a response spectrum analysis of a detail rack assembly finite element model.
The damping values used in the seismic analysis are two per cent damping for an operating basis earthquake (OBE) and four percent damping for a safe shutdown earthquake (SSE) as specified in Regulatory Guide 1.61.
The responses of the model from accelerations in three directions are combined by the SRSS method in the structural analysis. The loads in four major com ponents (support pad assembly, botton grid, top grid, and fuel cell ) are examined, and the maximum loaded section of each of these components is found.
These maximum loads from the detail model are used in the structural analysis to obtain the stresses within the rack assembly.
The licensee has shown that, during a postulated earthquake, the fuel rack modules may slide laterally along the bottom of the pool.
However, the mag nitude of sliding was small -enough so that the modules will not collide with the pool wall nor with each other. The calculation was performed using a non-linear code, WECAN. Certain aspects of the non-linear code, such as the sliding friction element, have not been fully reviewed by NRC.
The licen see, therefore, supplied by letter dated December
, 1980, a simplified analysis based on a linear response. The reanalysis showed the magnitude of sliding to be small when compared with gap spaces available between the rack modules and the pool wall.
We find the gap spaces large enough to accommodate lateral module motion due to earthquake forces.
Two accident loading conditions are postulated for fuel handling crane uplift analysis. The first condition assumes that the uplift load is applied to a fuel cell.
The second condition assumes that the load is applied to the top grid.
Calculations show that for eithier condition, the resulting stresses are within acceptable stress limits. In order to ensure that the SFP liner will not be perforated, two accident conditions are evaluated. The first accident condition assumes that the weight of a fuel assembly, control rod assembly and handling mechanism (3,000 lbs.) impacts on the top of the rack.
Calculations show that the impact energy is absorbed by the dropped fuel assembly, the stored fuel assembly, the cell funnels and the section of cell above the upper grid structure and the rack base plate/lower grid assembly.- The second accident condition assumes that the fuel assembly falls straight through an empty cell and impacts the rack base plate from a drop height of 234 inches. The results of this analysis show that the impact energy is absorbed by the fuel assembly and the rack base plate. The SFP liner will not be perforated and the margin of safety is positive.
No alteration was made to the pool itself. The fuel pool concrete reinforcing steel, linear plate and welds connecting the inner plate to the fuel pool floor concrete embedments were analyzed based on consideration of the new racks and additional fuel. The results-of the analysis were found to be acceptable and within the criteria given in the Final Safety Analysis Report (FSAR).
Conclusion on Structural and Seismic Loadings The structural aspects of the spent fuel storage racks have been evaluated based upon NRC guidance provided in the report entitled, "Position for Review and Acceptance of Spent Fuel Storage and Handling Applications,,
issued April 1978 and revised January 18, 1979. Based upon our review of the analyses and the design done by the 'licensee, we conclude that the rack structure itself, the supporting pool liner and slab, are capable of supporting the applied loads without exceeding relevant stresses of Subsection NF of ASME Section III or the FSAR design criteria. The proposed modifications to the Oconee spent fuel storage are in conformance with NRC requirements.
Fuel Cask Drpp Accident Evaluation We previously evaluated this accident in the June 19)3 Safety Evaluation (SE) for the Oconee Nuclear Station for the original Units 1 and 2 SFP with 336 stoiage spaces, as described in the FSAR. In our SE dated June 19, 1979, related to the increase in capacity of this SFP from 336 to 750 storage spaces, we reevaluated the effects of a cask drop due to the closer spacing of spent fuel assemblies near the cask loading platform area. We concluded in our June 19, 1979 SE that the radiological consequences were mitigated by limiting the age of the spent fuel stored in the first 28 rows closest to the loading platform to a minimum decay time of 55 days.
To provide qquivalent mitigation of such an accident for the SFP increase in capacity from 750 to 1312 storage spaces, the licensee has proposed to limit the age of fuel stored in the first 36 rows closest to the loading platform to a minimum decay time of 55 days. We find the 36-row limit equivalent to the previous 28-row limit.
The 36-row limit and 55-day minimum decay inter val are provided in proposed TS 3.8.13. We conclude that the consequences of a cask drop accident in the Units 1 and 2 SFP are not changed from those pre sented in our June 1973 and June 19, 1979 SEs with the implementation of the limits prescribed in TS 3.8.13 and are thus acceptable.
Materials Evaluation The spent fuel racks in the proposed expansion will be constructed entirely of type 304 stainless steel, except for the nuclear poison material.
The existing SFP liner is constructed of stainless steel.
The high density spent fuel storage racks will utilize Boraflex sheets as a neutron absorber. Bora flex consists of 42 weight percent of boron carbide powder in a rubber-like silicone polymetric matrix. The spent fuel storage rack configuration is com posed of individual storage cells interconnected to'form an integral structure.
The major components of the assembly are the fuel assembly cells,- the Boraflex material, the wrapper and the upper and lower spacer plates.
The upper end of the cell has a funnel shape flare for easylinsertion of the fuel assembly. The wrapper surrounds the Boraflex material, but is open at the top and bottom to provide for venting of any gases that are generated. The Boraflex sheets sit in a square annular cavity formed by the square inner stainless steel tube and the outer wrapper. Each sheet is supported by lower spacer plate.
The pool contains oxygen-saturated demineralized water containing boric acid, controlled to a temperature below 150 0F.
-7 The pool liner, rack lattice structure and fuel storage tubes are stainless steel which is compatible with the storage pool environment. In this environment of oxygen-saturated borated water, the corrosive deterioratlon of the type 304 stainless steel should not exceed a depth of 6.00 X 10 inches in 100 years, which is negligible 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 passi vating oxide films and are therefore at similar potentials. The Boraflex is composed of non-metallic materials and therefore will not develop a galvanic potential in contact with the metal components. Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments, and to verify its structural integrity and suitability as a neutron absorbing material.
The evaluation tests have shown that the Boraflex is unaffected by the pool water environment and will not be degraded by corrosion. Tests were performed at the University of Michigan, exposing Boraflex to 1.03 X 10" rads of gamma radiation with substantial concurrent neutron flux in borated water. These tests indicate that Boraflex maintains its neutron attenuation capabilities after being subjected to an environment of borated water and gamma irradiation. Irradiation will cause some loss of flexibility, but will not lead to break up of the Boraflex. Long term borated water soak tests at high temperatures were also conducted. The tests show that Boraflex withstands a orated water immersion of 240oF for 260 days without visible distortion or softening. The Boraflex showed no evidence of swelling or loss of ability, to maintaina uniform distribution of boron carbide.
The annulus space which contains the Boraflex is vented to the pool at each corner storage tube assembly.
Venting of the annulus 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 inner stainless steel tube.
The manufacturer's tests have shown that neither irradiation, environment nor Boraflex composition has a discernible effect on the neutron transmission of the Boraflex material. The tests also show that Boraflex does.not possess leachable halogens that might be released into the pool environment in the presence of radiation. Similar conclusions are reached regarding the leaching of elemental boron from the Boraflex. Boron carbide of the grade normally in the Boraflex will typically contain 0.1 weight percent of soluable boron. The test results have confirmed the encapsulation function of the silicone polymer matrix in preventing the leaching of soluble specie from the boron carbide.
To provide added assurance that no unexpected corrosion or degradation of the materials will compromise the integrity of the racks,.the licensee has.commited 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.
Conclusion on Materials From our evaluation as discussed above we conclude that the corrosion that will occur in the Oconee SFP environment should be of little signifi cance during the 40-year life of the plant. Components in the SFP are constructed of alloys which have a low differential galvanic potential between them and have a high resistance of general corrosion, localized
corrosion, and galvanic corrosion. Tests under irradiation ard at elevated temperatures in borated water indicate that the Boraflex material will not undergo significant degradation during the expected service'life of 40 years.
We further conclude that the environmental compatibility and stability of the materials used in the Oconee expanded SFP is adequate based on the test data cited above and actual service experience in operating reactors.
We have reviewed the surveillance Droaram, and we conclude that the moni toring of the materials in the SIV, as proposed by the licensee, will pro vide reasonable assurance that the Boraflex material will continue to per form 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, having a capability to permit appropriate periodic inspection and testing of components, and Criterion 62, preventing criticality by maintaining structural integrity of components and of the boron poison.
Occupational Radiation Exposure We have reviewed the licensee's plan for the removal and disposal of the existing racks that were installed during a previous modification in 1979 and the installation of the new racks with respect to occupational radiation exposure. The occupational exposure for this operation is estimated by the licensee to be about 23 man-res. We consider this to be a reasonable estimate because it is based on the licensee's detailed breakdown of occupational expo sure for each phase of the modification based on task comparisons with the pre vious re-racking. The licensee considered the number of individuals performing a specific job, their occupancy time while performing this job, and the average dose rate in the area where the job is being performed. Although divers will be required during the modification, their expected cumulative dose equivalent will be about 10 mai-rems.
The-modification will be performed by arranging the spent fuel elements stored in the pool in such manner as to yield the lowest dose rates in the area to be occupied by divers while at the same time minimizing spent fuel movements or rearrangements of the assemblies which cause additional exposure to personnel performing and monitoring this operation. The existing spent fuel racks that will be removed from the SFP will be washed down and crated for disposal as low level radwaste at a licensed disposal site. The work to be performed will.be performed in a manner consistent with "as low as is reasonably achievable" (ALARA) obcupational exposures. All work will be performed in accordance with a radiation pre-plan to identify all protection requirements. Health physics personnel will be available to assure that ALARA radiation exposures prevail.
We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies on the basis of information supplied by the licensee for dose rates in the spent fuel area from radionuclide con centrations in the SFP water,and deposited on the SFP walls. The spent fuel assemblies themselves will contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel.
The occupational radiation exposure resulting from-the additional spent fuel in the pool repre sents a negligible impact. Based on present and projected operations in the SFP
-9 area, we estimate that the proposed modification should add less than one percent to the total annual occupational radiation exposure burden at this facility. The small increase in additional exposure will not affect the licensee's ability to maintain individual occupational doses to ALARA and within the limits of 10 CFR Part 20. Thus, we conclude that storing addi tional fuel in the SFP will not result in any significant increase in doses received by occupational workers.
Radioactive Waste Treatment The station contains waste treatment systems designed to collect and pro cess the gaseous, liquid and solid wastes that might contain radioactive material. The waste treatment systems were evaluated in the SE dated December 1970 for Oconee Unit 1 and in the SE dated July 1973 for Oconee Unit 2. There will be no change in the waste treatment systems or in the conclusions of the evaluations of these systems because of the proposed modification.
CONCLUSION ON SAFETY We 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 activi ties will be conducted in compliance with the Commission's regulations and that the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
December 24, 1980