ML17318A382
| ML17318A382 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/16/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17318A378 | List: |
| References | |
| NUDOCS 7911020269 | |
| Download: ML17318A382 (24) | |
Text
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE MODIFICATION OF THE SPENT FUEL STORAGE POOL FACILITY OPERATING LICENSE NOS.
DPR-58 AND DPR-74 INDIANA AND MICHIGAN ELECTRIC COMPANY INDIANA AND MICHIGAN POWER COMPANY DONALD C.
COOK NUCLEAR PLANT, UNIT NOS.
1 AND 2 DOCKET NOS.
50-315 AND 50-316 October 16, 1979
1.0 INTRODUCTION
By letters dated November 22, 1978 and January 22, 24, April 16, June-29, July 27 and September ll, September 26, and September 27, 1979, the licensees, Indiana and Michigan Electric Company and Indiana and Michigan Power Company (IBMPC), requested an amendment to Facility Operating License Nos.
DPR-58 and DPR-74 for the Donald C.
Cook Nuclear Plant, Unit Nos.
1 and 2.
The request was made to obtain authorization to provide additional storage capacity in the shared spent fuel pool (SFP).
The proposed modi-fications would increase the capacity of the SFP from the present design capacity of 500 spent or irradiated fuel assemblies to a capacity of 2050 fuel assemblies.
The increased SFP capacity would be achieved by installing new racks with a decreased spacing between fuel storage cavities.
The present rack design has a nominal center-to-center spacing between fuel storage cavities of 21 inches.
The proposed new spent fuel racks would be modular stainless steel structures with individual storage cavities to provide a nominal center-to-center spacing of 10.5 inches.
Each stainless steel wall of the individual cavities would contain sheets of Boral (Boron Carbide in an aluminum matrix) to provide for neutron absorption.
The SFP is located in the auxiliary building located between the Unit Nos.
1 and 2 reactor containment buildings.
The pool structure is located adjacent to the fuel cask handling and new fuel storage areas and is above the spray additive tank room, the sampling
- room, the evaporator waste tank rooms, and the hallway in the basement of the auxiliary building.
The-general arrangement and details of the proposed new spent fuel storage racks are shown in Figures l-l through 1-3 of the licensee's submittal of November 22, 1978.
The expanded storage capacity would allow for the continued operation of both Unit Nos.
1 and 2 until about the first part of 1992 while still maintaining the capacity for a full core discharge reserve of 193 locations.
The major safety considerations associated with the proposed expansion of the SFP storage capacity are addressed below.
A separate environmental impact appraisal has been prepared for this proposed action.
2.0 DISCUSSION AND EVALUATION The proposed modification for the spent fuel storage capacity expansion has been reviewed in accordance with the NRC report "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," April 1978.
- 2. 1 Criticalit Considerations After irradiation, spent fuel is removed from the core, transferred from the reactor containment building and placed in the common spent fuel pool.
The proposed spent fuel storage racks are to be made up of double-walled stainless steel containers.
These will be about 14 feet long and will have a square cross section with an inner dimension of 8.97 inches.
The nominal distance between the centers of the stored fuel assemblies, i.e., the lattice pitch, is 10.5 inches.
The effective dimension of the square 'fuel assembly, which was used in the criticality calculations, is 8.432 inches.
This results in an overall fuel region volume fraction of 0.645 in the nominal storage lattice cell.
Boral plates are to be press fit and seal welded in the cavities between-the double stainless steel walls.
In its January 14, 1979 submittal IEMPC states that stringent in-process inspection and process controls are imposed during manufacturing of the Boral to assure that the Boral plates contain at least 0.020 grams of the boron ten isotope per square centimeter of plate.
In this full array of storage containers there will be two Boral plates between adjacent fuel assemblies.
This makes the mjnimum areal density of boron between fuel assemblies 2.41 x 10 'oron ten atoms per square centimeter.
As stated in IEMPC's November 22, 1978 submittal, the fuel pool criticality calculations are based on unirradiated fuel assemblies with no burnable poison and a fuel loading of 44.2 grams of Uranium-235 per axial centimeter of fuel assembly.
The Exxon Nuclear Company performed the criticality analyses for IEMPC.
The basic method was to use the NITAWL and XSDRNPM computer programs with the XSDRN 123 group microscopic cross section data to generate multigroup cross sections for the KENO IV Monte Carlo program.
These programs were used to calculate the effective neutron multiplication
factors (keff)* for the fuel pool.
Also, CCELL, which is a fuel pin cell program, was used to determine the effects of U02 pellet density, moderator temperature, fuel temperature, and U-235 enrich-ment on the neutron multiplication factor.
Exxon calculated the worst case keff by assuming a minimum water gap thickness between adjacent storage cells and a maximum water gap temperature of 100'C.
IBMPC's November 22, 1978 submittal states that the assumption of' minimized water gap thickness of 0.953 inches, rather than the nominal thickness of 1.118 inches, accounts for all of the tolerances and the possible deformations due to the design and seismic loads.
Exxon's calculated value for this worst case keff is 0.923.
Exxon checked the accuracy of this KENO IV method by calculating several sets of critical experiments.
Two of these had Boral plates in them.
These were the experiment done by E.
B. Johnson and G. E. Whitesides at the Oak Ridge National Laboratory and the one done by S.
R. Bierman et al. at the Pacific Northwest Laboratories (PNL-2438).
Exxon's calculated values of the keff s of those two experiments agreed with the experimental values of 1.00 within the limits of the statistical uncertainty in the Monte Carlo program.
In its January 24, 1979 response to our request for additional information IBMPC stated the following:
1.
There are steel structural members on the periphery of the rack modules that will provide an additional two inches of water between any fuel assembly outside of the rack and those in the racks.
This two inches of water will prevent an increase in the keff.
2.
Neutron attenuation tests will be performed on the rack modules at the Cook site to verify the presence of the boron.
eff e
ect> ve mul tiplication factor, i s the ratio of neutrons from fissions in each generation to the total number lost by absorption and leakage in the preceding generations.
To achieve criticality in a finite system, k ff must equal 1.0.
3.
Sufficient prototypical surveillance specimens will b'e provided which will permit inspection of both leaking and leak tight Boral cells.
2.1.1 Evaluation The above results compare favorably with the results of calculations made with other methods for similar fuel pool storage lattices.
By-assuming new, unirradiated fuel with no burnable poison or control
- rods, these calculations yield the maximum neutron multiplication factor that could be obtained throughout the life of the fuel assemblies.
This includes the effect of the plutonium which is generated during the fuel cycle.
The NRC acceptance criterion for the criticality aspects of high density fuel storage racks is that-the neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions throughout the life of the racks.
This 0.95 acceptance criterion is based on the overall uncertainties associated with the calculational
- methods, and it is our judgment that this provides sufficient margin to preclude critical ity in fuel pool s.
Accordingly, Technical Specifications l.imit the neutron multiplication factor, keff, in spent fuel pools to 0.95.
To preclude any unreviewed
- increase, or increased uncertainty, in the calculated value of the neutron multiplication factor which could raise the actual k ff in the fuel pool above 0.95 without being detected, a limit 5n the maximum fuel loading is also required.
Accordingly, we find that the proposed storage racks will meet the NRC criterion when the fuel loading in the assemblies described in these submittals is limited to 44.2 grams or less of Uranium-235 per axial centimeter of fuel assembly.
We find that IEHPC's proposed neutron attenuation test for the verification of the boron in the racks proposed for the Cook site is satisfactory.
However, in this test, if any Boral plates are found to be missing the NRC shall be notified and a complete test on every storage location shall be performed.
The licensee has agreed to these conditions.
We find that I8HPC's proposed boron surveillance program as described above is satisfactory for monitoring the condition of the Boral plates.
2.1.2 Conclusion We find that when any number of fuel assemblies, which IBMPC described in these submittals, which have no more than 44.2 grams of Uranium-235 per axial centimeter of fuel assembly, are loaded into the proposed racks,.the kerf in the fuel pool will be less than the 0.95 limit.
We also find that in order to preclude the possibility of the keff in the fuel pool from exceeding this 0.95 limit without being detected, it is necessary, pending NRC review, to prohibit the use of the proposed storage racks for fuel assemblies that contain more than 44.2 grams of Uranium-235 per axial centimeter of fuel assembly.
On the basis of the information submitted, and the k ff and fuel loading limits stated above we conclude that the health eff and safety of the public will not be endangered by the use of the proposed racks.
2.2 S ent Fuel Cool'in The licensed thermal power for D.
C.
Cook Unit No.
1 is 3,250 MWth and 3,391 MWth for Unit No. 2.
A full core in each of the Units consists of 193 fuel assemblies.
In its January 22, 1978 submittal ISMPC states that the evaluation of the spent fuel pool cooling system assumes that 65 fuel assemblies will be discharged annually from Unit No.
1 and 88 fuel assemblies will be discharged every 18 months from Unit No. 2.
IEMPC assumed a cooling time of 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> after the reactor is shut down following 1,080 full power days of operation to calculate the maximum heat load by the method given in NRC Branch Technical Position APCSB 9-2.
This gave a maximum in-pool heat generation rate of 54.1 kW per fuel assembly.
The spent fuel pool cooling system as described in Chapter 9 of the D.
C.
Cook FSAR, consists of two pumps and two heat exchangers.
Each pump is designed to pump 2,300 gpm (1.15 x 106 pounds/hr) and each heat exchanger is designed to transfer 14.9 x 10 BTU/hr from 120'F fuel pool water to 95'F component cooling water, which is flowing through the heat exchanger at a rate of 1.49 x 10 pounds per hour.
As shown in Figure 9.4-1 and 9.2-4 of the FSAR the Seismic Category I
source of makup water for the soent fue]
oool is the chemical aced volume control system hold-up tanKs.
frere are tnree pairs of ~nese
- tanks, and each pair has a capacity of 128,000 gallons.
The hold-up tank recirculation pump, which is rated at 500 gpm, can be used to pump water from these tanks to the spent fuel pool.
e
The spent fuel pool is monitored both locally during each shift and continuously in the control room.
There are both high and low water level alarms (one foot of water difference) in the control room and an alarm for the SFP cooling and purification system pump failure.
The SFP temperature is alarmed when the water reaches 180'F and the radiation 'level is alarmed when the level above the pool reaches 10 mrem.
The temperature readings and radiation area monitor alarm are also local to the SFP.
2.2.1 Evaluation Using the method given on pages 9.3.5-8 through 14 of the NRC Standard Review Plan, with the unc~rtainty factor, K, equal to O.l for decay times longer than 10 seconds (about 116 days),
we calculate that the maximum peak heat load during the 28th refueling, which would fill the pool, could be 22.6 x 106 BTU/hr and that the maximum peak heat load for a full core off-load that essentially fills the pool could be 41 x 10 BTU/hr.
This full core offload was assumed to take place one year after the 25th refueling.
We also find that the maximum incremental heat load due to increasing the number of spent fuel assemblies in the pool from 500 to 2,050 would be 6.9 x 10 BTU/hr.
(This is the difference in peak heat loads for full core offloads that essentially fill the present and the modified pools.)
We calculate that with both pumps operating, the spent fuel pool cooling system can maintain the fuel pool outlet water temperature below 120'F for the normal refueling offload that fills the pool (2,050 assemblies) and below 130'F for the full core offload that fills the pool (2050 assemblies).
With a full core offload that fills the pool (2050 assembl'ies) and with only one cooling pump operating, the pool temperature can be maintained below 165'F.
In the highly unlikely event that both spent fuel pool cooling systems were to fail at the time when there was a peak heat load from a full core in the pool and the water was at its maximum temperature and minimum possible elevation (i.e.,
23 feet of water);
we calculate that boiling could commence in about 6-1/2 hours.
This elevation of water corresponds to the level the water would reach if the SFP drain line were inadvertently broken at the wall of the SFP in the cask handling area.
On September 26, 1979 the licensee committed to remove this line and seal its SFP penetration so that the minimum water elevation would be about 37 feet.
We also calculate that after boiling commences the
required water makeup rate will be 85 gpm.
We find that 6-1/2 hours will be sufficient time to establish a 85 gpm makeup rate.
By sealing the SFP drain line, this time will be increased to about 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.
2.2.2 Conclusion We find that the present cooling capacity for the spent fuel pool at D.
C.
Cook Nuclear Plant, Unit Nos.
1 and 2 will be sufficient to handle the incremental heat load that will be added by the proposed modifications.
We also find that this inc)cmental heat load will not alter the safety considerations of spent fuel cooling which we previously reviewed and found to be acceptable.
2.3 Installation of Racks and Fuel Handlin The new fuel racks will be installed without removing from the pool the fuel assemblies presently stored there.
This involves moving 'loads over and within the SFP.
These loads include the fuel assemblies that must be transferred underwater from-existing locations, the removal of the existing rack modules from the SFP and the introduction of the new rack modules.
Using existing procedures, several of the presently stored fuel assemblies will be relocated to other positions in existing rack modules to empty a
number of the rack modules.
These empty rack modules will be uncoupled and lifted from the pool so as not to be carried over spent fuel stored in the pool.
Several modules of the new racks will then be lifted and lowered into the pool in the space thus vacated,
- again, so as not to be carried or lifted over spent fuel stored in the pool.
After securing these modules all of the stored fuel will then be transferred underwater, using existing procedures, to these new modules.
The now empty remainder of the present rack modules will then be removed from the pool and the remaining new rack modules will be lifted into the pool and secured in place.
All movement of the rack modules will be done in conformance with existing and new Technical Specifications on crane travel over spent fuel (T.S. 3.9.7).
The licensee's procedures for controlling the movement of portions of the present and new racks which will prevent their movement over spent fuel, are described in Exxon Procedure XN-NS-IP012, which like other detailed procedures, is available for inspection at the D.
C.
Cook Nuclear Plant.
The NRC staff has underway a generic review of load handling operations in the vicinity of spent fuel pools to determine the likelihood of a heavy load impacting fuel in the pool, and if necessary, the radiological consequences of such an event.
Because the Cook 1
and 2 Technical Specifications (3.9.7) prohibit loads greater than 2,500 pounds (the nominal weight of a fuel assembly and handling tool) to be transported over spent fuel in the
- SFP, the movement of storage racks over spent fuel assemblies will be prohibited.
There are other lighter loads, however, that are handled over stored fuel assemblies.
These loads are the Hew Fuel Assembly Fuel Handling Tool, Thimble Plug Handling Tool, Spent Fuel Assembly Handling Tool, and Burnable Poison Rod Assembly Handing Too'l..
A set of these tools are used for both 15 x 15 fuel and 17 x 17 fuel.
The weights of tools range from 72 pounds to 800 'pounds.
Although lighter than a single fuel
- assembly, these could develop greater kinetic energy should they be dropped because of greater potential drop heights.
This larger kinetic energy could theoretically cause more damage to stored fuel assemblies than that calculated assuming a single dropped fuel assembly.
The licensee has therefore examined the use of these loads and has committed to restricting the height of. tool movement over the spent fuel assemblies such that if a tool were to drop, the impact energy would not exceed that of an analyzed spent fuel drop accident, i.e., 24,240 in-lbs.
In addition, to insure that the handling tools will not drop the licensee will install a
backup cable sling.
2.3.1 Evaluation Since over one half of the fuel assembly positions will be vacant during reracking, I8NPC should have no difficulty in keeping the rack portion being removed and installed away from the spent fuel that is 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.
The consequences of fuel handling accidents in the spent fuel pool area are not changed from those presented in the Safety Evaluation Report dated September 1973.
This design basis accident is independent of the number of fuel assemblies in the pool and is defined for fuel with the least decay after shutdown for refueling.
The accident is assumed to occur at the time after shutdown identified in the Technical Specifications as the earliest time fuel handling operations may begin.
The Technical Specifications which prohibit loads greater than 2,500 pounds or loads at heights which would exceed a kinetic energy impact of 24,240 in-lbs. allow flexibility in the movements of fuel and other relatively light loads, while providing reasonable assurance that the consequences of the design basis accident will not be exceeded.
2.3.2 Conclusion We conclude that there is reasonable assurance that the health and safety of the public will not be endangered by the removal of the present racks and the installation and use of the proposed racks.
2.4 Structural and Mechanical Desi n and Materials Considerations The present spent fuel pool storage capacity is 500 fuel assemblies.
The proposed replacement racks will permit the storage of up to 2,050 fuel assemblies.
These new racks would allow for the continued operation of Unit Hos.
1 and 2 until 1992 while still maintaining the capacity for a full core discharge reserve.
The new racks are to be fabricated primarily from type 304 stainless steel.
The individual fuel assemblies will be stored in square fuel storage cells formed from a thicker (0.075") inner shroud of stainless
- steel, a center sheet of aluminum clad boron carbide, and a thinner (0.030") outer shroud of stainless steel.
A flared guide and transition section is provided-at the top of each storage cell.
The boral sheets provide neutron absorption to allow spacing
'of fuel assemblies in a 10.5 inch by 10.,5 inch cell array.
The cells are welded to a stiffened module base which carries vertical loads and an upper box structure consisting of plate diaphragms and a top grid.
Horizontal seismic loads are carried to the module base through the plate diaphragms.
Tipping is prevented by coupling adjacent racks through a bolted connection at the top grid level.
2.4.1 Evaluation Structural and Mechanical The structural and mechanical review consisted of an examination of the following areas:
the proposed design criteria, the design loads and load combinations, methods of analysis, and a dropped fuel assembly accident; the material properties and allowable stresses of type 304 stainless steel; the hydrodynamic effects; the fabrication and installation provisions; and the effect of increased loads on the floor slab and liner and on the new support structure to be added beneath the floor slab of the pool.
The material properties for structural components of the spent fuel racks used in the analyses were taken from Appendix I of Section III of the ASME code, i.e., yield and ultimate strength, modulus of elasticity and thermal expansion.
Load combinations and acceptance limits are in conformance with the NRC position paper referred to in 2.0 above which referenced acceptance limits of the ASHE Section III
- Code, Subsection NF 3231.1 and 3321.2 as supplemented by paragraphs C.2, C.3, and C.4 of Regulatory Guide 1.124 entitled, "Design Limits and Load Combinations for Class I Linear Type Component Supports."
The seismic contribution to the load combinations was obtained by using both response spectra and time-history analyses.
The spent fuel pool is located in the auxiliary building.
The response spectra used for analysis of the fuel storage racks were the floor response spectra developed from the seismic analysis of the auxiliary building at the elevation of the spent fuel storage pool floor.
Seismic stresses due to the three orthogonal earthquake components were combined by the square root of the sum of the squares (SRSS) method in conformance with Regulatory Guide 1.92 entitled, "Combining Modal Responses and Spatial Components in Seismic
Response
Analysis."
Two percent of critical damping was used for both operating basis earthquake (OBE) and safe shutdown earthquake (SSE) loading consistent with the values used in.the FSAR.
No credit was taken for additional damping due to submergence in water.
The total mass of the water enclosed in the proposed new storage rack was lumped together with the masses of the fuel assembly and the rack structure in the horizontal direction (for both the response spectra and the history analyses).
At the east end of the pool, there is a
12 foot clearance between the new rack structure and the pool wall.
The hydrodynamic effects of water on the new rack structure were analyzed in accordance with the methods given in report TID-7024, "Nuclear Reactors and Earthquakes,"
August 1963.
The added hydrodynamic loads increased the horizontal loads in the east-west direction by 87.
Seismic loads obtained from the response spectra analysis were increased by impact factors of 1.54 and 1.94 for SSE and OBE loads, respectively.
These values were obtained by performing time history analyses using the ANSYS computer program to account for the effects of the clearance gap between the storage cells and the fuel assemblies contained herein plus the effects of the new rack structure rocking and sliding on the pool floor.
Two different but equivalent models were employed by the licensee for the two types of seismic analysis.
The model used for the
response spectra analysis was a three dimensional model consisting of 251
- modes, and 460 structural elements comprised of beams, trusses and membrane elements.
The SAP IV computer program was employed for the analysis.
The model used for time-history analysis was a lumped mass-spring model which was based on the assumption that, all fuel assemblies were moving in phase with each other.
Spring stiffnesses of the rack structure in this model duplicated the frequencies and mode shapes of the primary horizontal and vertical modes as determined from the three dimensional model.
Friction coefficients for the sliding gap elements were taken from two reports:
"Friction Coefficients of Water-Lubricated Stainless Steels for a Spent Fuel Rack Facility," by Professor E. Rabinowicz, NIT, November 5, 1976, and G.E. Report No. 60GL20, "Investigation of the Sliding Behavior of a Number of Alloys Under Dry and Water Lubricated, Conditions," by R.
E. Lee, Jr., January 22, 1960.
The model gas subjected to simultaneous, statistically independent horizontal and vertical time-histories at the pool floor whose response spectra enveloped the floor response spectra.
To determine the impact factors, the licensee performed two time-history investigations, one with non-linear effects of the cell to fuel gaps with rack feet sliding taken into consideration and one linear analysis with these gaps closed and the rack feet fixed.
The maximum ratios of the non-linear response to the linear response for OBE and SSE are the impact factors which were then applied to all seismic loads obtained from the response spectra analysis.
The licensee performed an evaluation of the effects of two postulated cases of a dropped fuel assembly.
The first case was that of an assembly falling vertically directly on one cell but rotated 45'uch that the corners of the assembly hit the sides of the cell in a diamond pattern.
This case produced maximum force and deflection of an individual cell.
The second case was that of an assembly falling vertically at the center of a group of four cells, resulting in the maximum force applied to the rack structure.
The dynamic response of the rack structure to the two drop accidents, indicated:
(1) inelastic deformations were limited to the immediate areas of assembly
- impact, and (2) all other rack member stresses were within the limits specified by the load combinations and allowable stresses permitted by the NRC.
I The spent fuel storage rack modules are designed in conformance with the fabrication, installation, and examination criteria of the 1977 ASflE Code,Section III, Subsection NF, Articles NF-2000, NF-4000, and NF-5000, with the following two exceptions.
- First, documentation and certificatioh programs for code qualified components were in conformance with Exxon Nuclear Company's guality Assurance/guality Control
- program, which is in compliance with ANSI N45.2 and, 10 CFR 50, Appendix B, rather than being stamped as per NF-4120.
- Second, except for neutron poison material, identification markings as per NF-4122 were not provided for components of the fuel rack.
However, the fabricator is required to demonstrate a material control program which will insure that only certified material is used.
We find these provisions acceptable.
The floor slab modifications were done in conformance with the FSAR Design Criteria, specifically the ACI-318-63 Code (American Concrete Institute).
Hand calculations were performed to determine the modifi-cation necessary to support the additional load of approximately 2.7 million pounds, resulting from the increased rack weight of 518,000 lbs.
and increased fuel weight of 2,170,000 lbs.
The modification consisted of adding steel, wide flange sections and cover plated tubular columns under the supporting slab.
The modifications then allows for a limit capacity of the slab of 415 kip-feet per foot slab width and 272 kip-feet per foot width for positive and negative ultimate bending moment capacity, respectively, as well as 71 kips per foot width ultimate shear capacity.
By comparison, the analysis deter-mined the minimum load bearing capacities required to be 316 kip per foot, 177 kip per foot and 53 kip per foot for positive ultimate moment, capacity, negative ultimate moment and ultimate shear, respectively.
We find the proposed modification acceptable.
~Swei 1 1 n In August 1978, the staff was made aware of a problem at the Monticello facility regarding spent fuel storage racks similar in design to those proposed here.
The problem there involved inleakage of water into the stainless steel cells, such that hydrogen gas was generated due to oxidation of exposed aluminum material.
This gas caused a pressure buildup and resultant swelling of those stainless steel cells such that a fuel assembly located at an affected storage location could not be readily removed.
A discussion of how this potential problem has been considered at D.
C.
Cook is provided below.
The proposed D.
C.
Cook cells use boral material sealed between an inner and outer stainless steel shroud.
The cells are supplied to Exxon Nuclear Company by Brooks and Perkins, Incorporated.
The stainless steel shroud (or cladding) is type 304.
The boral consists of an 1100 series aluminum and boron carbide matrix core sandwiched between two layers of 1100 series aluminum cladding.
The stainless steel shrouds are seal-welded together at both ends such that the annulus between the shrouds is leaktight.
The inner shroud is thicker (0.075") than the the outer shroud (0.030") to provide protection against inward swelling of the cell and binding of stored fuel assembly.
In the event that there are leaks allowing water to enter the annulus, there will be corrosion of the aluminum with hydrogen gas as an off product.
Once the pressure buildup within the composite exceeds the confining pressure of the cell materiyl and hydostatic head, the outer shroud will bulge outward and will not contact the fuel bundle.
In an effort to minimize the consequences of water leakage into the cell annulus, the licensee will impose strict welding procedures, welding operations and qualifications of welders in accordance with the requirements of the ASME Code,Section IX, and nondestructive examination requirements, in accordance with ASME Section V.
In addition, leaktightness tests will be conducted by immersing the fuel storage cell in water while pressurizing the cell annulus with helium gas.
Leaks are detected by helium gas bubbles escaping to the surface.
Corrosion In the controlled environment of the spent fuel pool water, which is high quality demineralized water (with dissolved boric acid) maintained at relatively low temperatures by a cooling and cleanup
- system, the spent fuel pool liner, the proposed storage racks and the spent fuel itself have an extremely low potential for corrosion.
Weekly chemical analyses are performed for, among other things, flourides and chlorides in the SFP water.
The results have always been below the licensee's detectibility level.
Crackin of Stainless Steel Com onents On July 26, 1979, the Office of Inspection and Enforcement issued IE Bulletin No. 79-17 on Pipe Cracks in Stagnant Borated Water Systems at PWR Plants..
This Bulletin was concerned with 'pipe cracking incidents, the latest at Three Mile Island Unit No. 1, in safety-related stainless steel piping systems and portions of systems which contain oxygenated, stagnant or essentially stagnant borated water.
The cracks at the Three Mile Island Unit No.
1 occurred in heat affected zones of Type 304 stainless steel and appeared to be intergranual stress corrosion cracks (IGSCC).
In the Zion spent fuel pool expansion
- hearing, questions were raised on the applicability of Bulletin 79-17 to the materials and conditions in the Zion SFP.
As a result of staff questions on the D.
C.
Cook proposed amendment, the licensee submitted on September 27, 1979, his responses to similar questions asked by the Zion Board.
The piping and liner are Type 304 stain'less steel.
The pool circulation is maintained by the cooling and purification system and by natural circulation around the spent fuel in the pool.
Special care has. been taken in the fabrication of the racks to minimize any sensitization in the heat affected zones which is a necessary requirement for IGSCC in weldments.
The SFP liner was fabricated from the low carbon version of 304 and all welding was controlled to minimize or eliminate carbide precipitation in the heat affected zone.
Piping in the pool cooling and puri-fication system has a low potential for cracking since flow is not stagnant, the water purity is maintained, and controlled welding procedures were used in fabrication to minimize any problems.
2.4.2 Evaluation The analyses, the design, the fabrication and the planned installation of the proposed fuel rack storage system including the addition of a support wall under the pool floor, and the analysis of the structural loads imposed by dynamic, static, seismic and thermal loadings are in accordance with accepted criteria, and the acceptance criteria for the appropriate portions of the NRC OT Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications, April 1978.
The mechanical properties for the materials used in the rack design were those consistent with a pool normal operating temperature of 150'F and maximum temperature of 240'F.
The quality assurance-procedures for the materials, the fabrication, the installation and the examination of the new rack structures are in acceptable general conformance with the accepted requirments of ASME Code,Section III, Subsection NF, Articles NF-2000, NF-4000 and NF-5000.
E The effects of the additional loads on the existing pool structure due to the proposed racks have been examined.
The pool structural integrity including the addition of a support wall beneath the pool in the spray additive tank room is assured by conformance with the original FSAR acceptance criteria.
In turn, this provides adequate assurance that the pool will remain leaktight.
There is no evidence at this time to indicate that corrosion of the fuel assemblies, the stainless steel rack structures or the fuel pool liner will occur at the temperatures and qual,ity of the demineralied water (with dissolved boric acid) to be maintained in the pool.
The welding techniques and procedures and the non-destructive examination techniques provide a high level of confidence that the annuli containing the boral in the installed cells will be leaktight..
Although no cell inleakage is likely to occur, tests were conducted which demonstrated that if isolated cases of leakage should occur in service, any swelling of the cells would not represent a safety hazard.
If the boral plates (BgC/A'l matrix) in a cell should, through inleakage, be exposed to the spent fuel pool water, galvanic coupling between the aluminum-boral liner, aluminum binder and the stainless steel shroud of the cell could occur.
Deterioration of the boral would, be limited to edge attack by general corrosion and pitting corrosion of the cell's aluminum liner and binder in the general area of the leak path.
The B<C neutron adsorption particles are inert to the pool water and would become embedded in corrosion products preventing loss of the B~C particles.
Thus, this small amount of deterioration would have no effect on neutron shielding, attenuation properties or criticality safety.
To aid in verifying the above conclusions, the licensee has committed to conduct a long-term fuel storage surveillance program to verify that the spent fuel storage cells retain the material stability and mechanical integrity over the life of the spent fuel storage racks under actual spent fuel pool service conditions.
Sample specimens will be placed in the SFP and will be periodically examined visually and by weight analysis.
The licensee has also agreed to apply the results of Bulletin 79-17 investigations on stress corrosion cracking to the SFP and other systems in the D.
C.
Cook facility in order to help prevent the development of cracks.
Based upon our review to date of the corrosion potential in spent fuel pool environments and previous operating experience, we have concluded that, for the temperature and the quality of the pool demineralized water (with dissolved boric acid), there is reasonable assurance that no corrosion of the stainless steel in the racks, the fuel cladding or the pool liner will occur over the lifetime of the plant, that would significantly impact the structural integrity of the racks.
Since the possibility of long-term storage of spent fuel exists, the effects of the-pool environment on the racks, fuel cladding and pool liner are under continued surveillance.
2.4.3 Conclusion The structural and mechanical aspects of the proposed racks have
-been evaluated based upon NRC guidance provided in the report
- entitled, "OT Position for Review and Acceptance of Spent Fuel Storage and HandIing Applications," April 1978.
Based upon our review of the analyses, the design done by the licensee, and the commitments to apply Bulletin 79-17 results to the D.
C.
Cook facility, we conclude that the rack structure itself, the supporting pool liner and slab when strengthened by the proposed additiona1 support wall under the pool slab are capable of supporting the applied loads without exceeding relevant stresses of subsection NF and Regulatory Guide 1.124 or the FSAR Design Criteria.
As previously stated, we find the material, fabrication, installation and examination criteria acceptable.
We conclude that these aspects of the proposed modifications to the D.
C.
Cook spent fuel storage are in conformance with NRC requirements.
2.5 Occu ationaI Radiation Ex osure Rack Removal and Dis osaI We have reviewed the licensee's plans for the removal and disposal of the present racks and the installation of the proposed racks with respect to occupational radiation exposure.
The rer oval of the old racks and installation of new racks will be done with remote handling tools.
No divers will be required.
Decontamination of the old racks will be done underwater in the SFP using high pressure water jets.
Such exposure for this oper ation is estimated by the licensee to range from about 18 to 20 man-rem.
We consider this to be a conservative estimate.
This estimate represents a small fraction of the total man-rem burden from occupational exposure at the plant.
The licensee has presented alternative plans for the disposal of the present rack modules which considered removing and crating the modules modul e intact versus removing, cutting and then crating th d'ied s.
It is unlikely that the licensee will dispose of the e
isman present rack modules intact because this will preclude the use of standard shipping packages and increase the cost and time to dispose of the racks.
The licensee is considering two methods of disposal:
(I) crating the modules semi-whole which will reduce the man-rem exposure involv'ed with a major cut up of these racks or (2} cutting the modules into small sections which would permit more efficient packaging in the shipping containers.
This second to be di alternative would result in a smaller volume of radioa t t
ioac ive waste i.e.
fewe e
isposed of with resulting economic and environmental b
f t r waste shipments and conservation of Iow level waste ene i s, to ex end x
burial site space.
However, it would also-require th I '
e tr a effort to cut the old racks with some increase in occupational exposure.
The licensee has estimated that the of them sem'-
occupational exposure to decontaminate the old racks d
d'n ispose a
d
,i-whole would be about 4 man-rem while to decontam nat n
cut into small sections would be about 6 man-rem.
The licensee has not estimated the occupational exposure to decontaminate and crate the nodules whole but the exposure for this is estimated to be lest than that for semi-whole disposition.
The licensee has not yet quantified a cost-benefit analysis of the alternatives so that selection of a disassembly and disposal method has not been finalized.
The licensee will estimate the exposures associated with the different ways to dispose of the present rack modules from measurements of the activity levels on them when they are dis osal.
removed from the pool, decontaminated and otherwis d f At that time, taking into account alternative disposal rea y or costs and exposures, the licensee will select the method of disassembly for disposal so that exposures will be kept to levels that are as low as is reasonably achievable.
Use of the Pro osed Racks After Installation 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 pool area from radionuclide concentrations in the pool water and the spent fuel assemblies.
The spent fuel assemblies ool will contribute a negligible fraction of the dose rates th p
area because of the depth of water shielding the fuel.
Consequently, the occupational radiation exposure resulting from the additional spent fuel in the pool represents a negligible burden.
A site visit review of the SFP area revealed a 4" line penetrating a wall of the pool. If this line should break at the wall, the pool water would drain to 23 feet above the floor.
Our calcu-lations show that the pool surface radiation level with only 23 feet of water in the pool would be between 1
and 10 R per hour.
The licensee, on September 26,
- 1979, committed to remove the line and seal the penetration to remove this consideration.
We find that removal of this line is both prudent and acceptable.
Based on present and projected operations in the pool 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 radiation exposure will not affect the licensee's ability to maintain individual occupational doses to as low as is reasonably achievable and within the limits of 10 CFR Part 20.
- Thus, we conclude that storing additional fuel in the SFP will not result in any significant increase in doses received by occupational workers.
2.6 Radioactive Waste Treatment The plant at present has radioactive waste treatment systems designed to collect and process the gaseous, liquid and solid wastes that contain radioactive material.
These waste treatment systems were evaluated in the staff' D.
C.
Cook SER dated September 1973.
The licensee has proposed no chahge in these waste treatment systems because of the proposed modification.
Based on our review, we find no need for changes in these systems because of the proposed modifications.
There will be no change in the conclusions of the evaluation of these systems as described in Section ll of the D. C.
Cook SER because of the proposed pool modification.
3.0
SUMMARY
Our evaluation supports the conclusion that the proposed modification to the D.
C.
Cook Unit Nos.
1 and 2
SFP are acceptable because:
1.
The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the SFP would be negligible.
2.
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 a single fuel assembly and the subsequent release of the assembly's radioactive inventory-within the gap.
3.
The likelihood of an accident involving heavy loads in the vicinity of the spent fuel pools is sufficiently small that, with the additional kinetic energy limit for lighter loads, no other restrictions on load movement are necessary while our generic review of the issues is underway.
4.
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.
5.
The SFP has adequate cooling with existing systems.
,. 6.
The capacity of the existing radioactive waste treatment systems remains adequate.
7.
The structural design and the materials of construction are adequate to assure safe storage of D.
C.
Cook generated spent fuel in the pool environment for the duration of plant lifetime and to withstand the seismic loading of the design earthquakes.
4.0 CONCLUSION
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 ihe proposed
- manner, and (2) such activities wi'll be conducted in compliance with the Commission's regulations and that the proposed action to permit installation and use of high density spent fuel storage racks in the spent fuel pool at the D.
C.
Cook Nuclear Power Station will not be inimical to the common defense and security or to the health and safety of the public.
Date:
October 16, 1979