ML17192A771

From kanterella
Jump to navigation Jump to search
Safety Evaluation Supporting Proposed Facility Mods to Increase Spent Fuel Storage Pool Capacity
ML17192A771
Person / Time
Site: Dresden  
Issue date: 06/06/1980
From:
Office of Nuclear Reactor Regulation
To:
Shared Package
ML17192A770 List:
References
TASK-09-01, TASK-9-1, TASK-RR NUDOCS 8007010008
Download: ML17192A771 (14)
v • d • e


Text

~

.*. t UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATICN 3Y T:iE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE WJDIFICATION OF THE SPENT FUEL STORAGE POOL PROVISIONAL OPERATING ucrnsE t\\O. DPR-19 AND FACILITY OPERATING LICENSE NO. OPR-25 CO~O~IWEAL TH ED ISON COMPANY DRESDEN

~~UCLE.~R POWER STATION UNIT NOS. 2 ArlD 3 DCCKET NOS. 50-237 ANO 50-249 DATE:

June 6, 1980

})tAfL ~

100101ooor I

1 ~

!~TRODUCTION 3y appl icaticn da:e:::! !*'.ay 11, 1978, and supplements dated January 12, January 24, ~ay 3J, June 12, August 17, and October 19, 1979, the Cc~onwealth Edis:n Cc~pany (CECo) (the licensee), requested an a~er.d~ent to Prov:sional Operating License No. DPR-19 and Facility Operating License ~o. DPR-25 for the Dresden Nuclear Power Station, Unit r:os. 2 and 3, respectively.

The request 1*1as made to obtain authorization to increase the storage capacity of each of the spent fuel pools (SrP) at Dresden Unit Nos. 2 and 3.

The proposed modifi-cations would increase the available SFP storage spaces from the present 1,420 spa~es per spent fuel pool to 3,780 spaces per pool.

2.C DISCUSSION The proposed increase in SFP capacity would be accomplished by replacing the existing spent fuel storage racks at both Dresden 2 ar.d 3 with new, higher capacity, neutron absorbing soent fuel storage racks.

The new racks are to be constructed of rectangular stainless steel t~bes welded together in a checkerboard design to form a module of fuel storage cavities.

The tJbes forming the cavity walls would contain two stainless steel shrouds with a layer of neutron absorbing Boral (boron carbide in an aluminum matrix) in between.

The present rack design has a nominal center-to-center spacing between fuel storage cavities of 6.5 x 12 inches.

The proposed rack design would reduce the nominal center-to-center spacing to 6.3 x 6.3 inches, and add a neutron absorbing container wall between adjacent fuel storage cavities.

The general arranqement and details of the proposed new spent fuel storage racks are presented in Figures 3.1-2, 3.2-1 and 3.2-2 of the report dated February 3, 1978, prepared by the lice'nsee's consultant, Nuclear Services Corporation, entitled "Licensing Report Dresden Nuclear Power Plant Units 2 and 3 Spent Fuel Rack ~odification". (This report was sutmitted to the NRC by the licensee's application dated May 11, 1978.)

The modifications to the SFP would extend the spent fuel storage capability at Dresden 2 and 3 through 1995, at which time the ability to accorranodate a full core discharge would be lost.

The major safety considerations associated with the proposed expansion of the spent fuel pool storage capacity for Dresden Station are addressed below.

~.'"\\

~'.'ALUATION

3. l Criticality Considerations CECo
  • s fuel pool criticality calculations are based on unirradiated fuel assemblies with no burnable poison and a maximum fuel loading of 14.8 grams of ~ranium-235 per axial centimeter of fuel assembly.

These ca1c~lations were made by the Nuclear Services Corporation (NSC)

r)

  • for CECo.

The basic method used by NSC in the criticality ca1~ulations was to generate four energy group cross sections, using the CHEEihH ad XSDRN computer programs.

This data was then utilized in the CITATION diffusion progra~.

XSDRN, a one-dimensional, discrete-ordinates, spectral-averaging program, was used to calculate the cross sections for the Baral core regions. Moreover, the internal black boundary condition in the CITATION program ~as used to calculate the neutron flux in the the:.:ial energy grcup in the Baral plates.

NSC checked the accuracy of this ca1culational method by using it to calculate two critical experir.ents which had Baral plates in them.

As shown in Table 3.3-3 of the licensee's submittal, the resulting neutron multiplication factors were more than one percent higher than the experir.:er.tally determined values.

Thus,

~;sc assumed that this calculational r.ethod gives conservative results for the neutron multiplication factor in the spent fuel pcol.

NSC first used these progra~s to calculate a neutron multip1~cation factor, k=, of 0.91 for the nominal proposed storage rack :attice. For the purpose of this calculation, the density of the boron ten isotope (the neutron absorber) in the Baral was assumed to be at its minimum value of 0.02 grams per square centimeter of Baral plate and the pool temperature was assumed to be 40°F.

NSC then calculated the neutron multiplication factor for each of the following conditions:

(1) increasing the temperature to 212°F, (2) increasing the lattice pitch, (3) eccentrically positioning the fuel assemblies in the racks, and (4) taking the Zircaloy channels off of the fuel asse:iblies which are placed in these racks.

NSC found that all of these changes decreased the neutron multiplication factor in the pool.

NSC then calculated the following possible increases in the neutron multiplication factor (tk):

l. One extra fuel assembly at the outer periphery of the rack --+.002
2.

All of the racks pushed as close together as ~ossible -------+.018

3.

One out of every 32 Bora 1 plates missing ------------------+. 015 3.1. 1 Evaluation A comparison of the above results with the results of other calculations which were made for high density, spent fuel, storage lattices, with boron plates, shows them to be acceptably accurate.

Sy assuming new, unirradiated fuel with no burnable poison or control rods, these calcu1ations yield the maximum neutron multiplication factor that could be obtained throughout the life of the fuel assemblies.

This includes the e~fect of the plutonium whi~h is generated during the fue 1 cycle.

3. l. 2
  • The NRC ~ccepta~ce criteria for the criticality aspects of high den~ity fuel st~rage racks is that the neutron multiplication factor in spent f~el 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 criticality in fuel pools.

Accordingly, there is a technical specification which limits the neutron multiplication factor, k :* in spent fuel pools to 0.95.

Since the neutron multiplicalion factor in spent fuel pools is not a quantity which is measured with good accuracy, the only available value is a calculated one.

To preclude any unreviewed increase, or increased uncertainty, in the calculated value of the neutron multiplicaton factor which co~ld raise the actual k ff in the fuel pool above 0.95 without being detected, a limit On the maximum fuel loadinq is also required. Therefore, we find that the proposed high density storage racks will meet the NRC criteria when the fuel loading in the asse11blies, described in these submittals, is limited to 14.8 grar.is or less of uranium-235 per axial centimeter of fuel assembly..

In its response to our request for additional infonnation, CECo stated that in addition to the usual quality assurance program, a neutron poison verification test will be conducted at the Dresden plant after the racks are installed in the pool.

This will be a test to statistically show, with 95 percent confidence, that the boron is not missing from more than one out of every thirty-two Baral plates.

We find that the limits on the amount of missing boron, set by this test, will ensure that the neutron multiplication factor in the fuel pool is not above 0.95.

However, in.this test, if any Baral plates are found to be missing, the NRC shall be notified and a complete test on every storage location shall be perfonned.

Conclusion We find that when any number of the fuel assemblies described in the CECo submittals, which have no more than 14.8 grams of uranium-235 per axial centimeter of fuel assembly, are loaded into the proposed r:cks, the k : in the fuel pool will be less than the 0.95 limit.

Pending NRC F'~iew, we will prohibit the use of these high density storage racks for fuel assenblies that contain more than 14.8 grams of uranium-235 ?er axial centimeter of fuel assembly.

On the basis of the information suf:rnitted, and the kliff and fuel loading limits stated above, we conclude that the use of t e proposed racks is acceptable.

  • I>

I

.*. \\.

  • 3.2.

S~ent Fuel Coolino

3. 2. 1 The licensed thell.lal power for each unit of Dresden 2 and 3 is 2527 Mlo/T.

CECo has been refueling these units annually,* but in the future it plans to extend the refueling periods to eighteen months.

For each eighteen month refueling, CECo assumed that about forty-two percent, or 306 of the 724 assemblies in the core, would be moved to the spent fuel pool in the ten-day time period following the shutdown of the reactor.

For the purpose of determining the maximum possible heat load, CECo alsc assumed that a full core, i.e., 724 asse~blies, could ~e moved to the spent fuel pool in the ten-day time period following the shutdo\\om of the reactor.

For the power history prior to refueling, CECo assumed an energy production of 19,000 MWD/MTIJ obtained at a *continuous energy density of 20 MW/MTIJ.

7be spent fuel pool cooling system, as described in Chapter 10 of the FSAR, consists of two pumps and twg heat exchangers.

Each pump is

  • designed to pump 700 gpm (3.5 x 10 pounds ger hour) and each heat exchanger is designed to transfer 3.65 x 10 BiU/hr from the fuel pcol water to the reactor building cooling *water5.,.,hich is flowing through the heat exchanger at a rate of 7.5 x 10 pounds per hour.

When a full core is discharged to the spent fuel pool, one of the three loops of the shutdown reactor cooling system will be connected in parallel with the fuel pool cooling system. This gonnection will provide an additional flow of 3000 gpn (1.5 x 10 pounds per hour) of fuel pool water which will be cooled by the shutdown reactor.

cooling system.

Evaluation Using the method given on pages 9.2.5-8 through 14 of the NRC Standard Review Plan, with the3uncertainty factor, k, equal to 0.1 for decay times longer than 10 seconds, we calcu1ate tha~ the maximu~ peak heat load for a discharge of spent fuel would be 13.2 x 106 Bi1J/hr and that the m~ximum* peak geat load for a full core offload that fills the pool would be 26. 2 x 10 BTU/hr.

In both cases the spent fuel from previous refuel ings contributes approximately 3. 5 x 105 BTU/hr to the heat load.

We also find that the maximum incremental heat load that would be added by increasing the number of 6spent fuel assc~blies in the pool from 1420 to 3780 will be 2.1 x 10 BTIJ/hr for a fuily loaded pool.

This is the difference in peak heat loads for full core offioads that essentially fill the presen~ and the

iod if i ed pools.

Based on our calculations, we have determined that with both pumps operating, the spent fuel pool cooling system can maintain the fuel pool outlet water at an acceptable teuperature belcw l41°F for a peak refueling heat load of 13.2 x 10° BTIJ/hr.

In addition, we find that when the shutdown reactor cooling system is connected in,

parallel with the spent fuel pooi cooling syst~~, the combined system will have sufficient capacity to keep the spent fuel pool outlet water below an ~cceptable tenperature of 145°F for a full core heat load of 26.2 x 10° BTIJ/hr..

3.2.2 3.3 3.3. 1 Assl:.DTling a maximum fuel pool temperature of 145°F for accident analysis, the minimum possible time to achieve bulk pool boiling after any credible accident would be about eight hours.

After bulk boiling conunences, the maximum evaporation rate would be 54 gpm.

We find that eight hours would be sufficient time for CECo to establish a 54 gpm make up rate.

We also find that under bulk boiling conditions, the temperature of the fuel would not exceed 350°F.

This fs an acceptable temperature from the standpoint of fuel element integrity and surface corrosion.

Conclusion We find that the present cooling capacities of the Dresden 2 and Dresden 3 systems are sufficient to handle the incremental heat loads that will be added by the proposed modifications.

We also find that these incremental heat loads will not alter the safety considera-tions of spent fuel pool cooling from that which we previously reviewed and found to be acceptable, and therefore the proposed design is acceptable.

/

Installation of Racks and Fuel Handling The fuel building crane, which will be used to remove the present racks and install the new ones,.is rated for a 125 ton load.

The heaviest rack* will weigh about six tons.

Also, the overhead crane is inspected periodicaily, including a dye-penetrant inspection of the lifting hook prior to each refueling.

In the Safety Evaluation dated June 1976, we concluded that the overhead crane handlfng system and the spent fuel cask handling Technical Specifications meet our requirements and are acceptable for handling spent fuel casks weighing up to 100 tons.

CECo states in their submittal that the racks will not be carried over stored fuel assemblies in order to preclude a heavy load falling onto stored spent fuel.

To prevent movement over fuel assemblies, CECo plans to move the fuel assemblies which are now in the pool, as* far away as possible from the location where the racks are being changed, i.e., to the other end of the pool.

The entire transfer operation will be supervised by licensed fuel handling foremen.

Evaluation At the beginning of 1979 there were 1069 fuel assemblies stored in the Dresden 2 and Dresden 3 pools.

Since the present combined capacity is 2840 assemblies, this means that fuel assemblies can be rE!lloved from over one half of the racks in the pools. After the 1979 refuelings there will be less room, but we find that there will still be enough room to allow the replacement of the racks without having to move them over fuel assemblies~

3.3.2 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 proposed increase in spent fuel pool storage capacity does not change the consequences of fuel handling accidents in the spent fuel pool from those presented in the Dresden 2 Safety Evaluation dated October 1969 and in the Dresden 3 Safety Evaiuatjon dated November 1970.

Conclusion We conclude that removal of the present racks and installation and use of the proposed racks can be perfonned safely and is therefore acceptable.

3.4 Structural and Mechanical Design and Material Considerations ihe proposed modification of the spent fuel storage capacity will involve the replacement of existing spent fuel storage racks with new, higher capacity, neutron absorbing spent fuel storage racks.

This will increase the storage capacity from 1420 storage spaces -

per pool for Dresden 2 and 3 to 3780 storage spaces for each pool.

The new storage rack, base plates and legs will be constructed from structural stainless steel, Type 304, and are designed to seismic Category I criteria. The new racks consist of a number of rectangular tubes welded together in a checkerboard design forming an array of cells.

The tubes are made of two stainless steel shrouds with a layer of neutron absorbing material (BORAL) sandwiched in between.

The Baral is a composite panel* of boron carbide (B4C)/aluminum matrix clad with aluminum.

The tubes are interwelded along both their entire lengths and at the bottom to a base plate which is elevated above the pool floor.

The whole rack assenbly is supported by six legs which transmit the loads to the pool floor.

3.cl. l Evaluation 3.4.1.1 Structural and Mechanical The new spent fuel storage rack designs were revie1*ied in accordance with the applicable parts of Sections 3.7 and 3.8 of the Standard Review Plan dated Novenber 1975 and the Branch Technical ?osition entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Appl ications 11 (forwarded to all 1 icensees in April 1978).

The following structural and mechanical aspects cf the new rack design were.reviewed:

structural design, material aspects and analysis procedures for all loads including seismic and impact loadings, supporting arrangements for the racks, loading combinations and structural acceptance criteria, and quality control for fabrication and installation. Structural adequacy of the racks were verified by using stress allowables and service limit designation of ASME Code Section III, Division 1, Subsection NF 3300.

The seismic model utilized was a lumped mass stick model with stiffness properties determined from a detailed finite element model of the actual rack.

In accordance with the plant FSAR, one percent and two percent damping values were used for OBE and SSE., respectively.

lhe mass of fuel bundles and water contained within the racks cells

\\'/as lumped together with the rack mass.

The hydrodynamic mass effects and the effect of the gap between the rack cell and the fuel bundle were included in this analysis. The resulting modal responses were combined in accordance with Regulatory Guide 1.92.

Time history analysss were performed to judge the effects of potential sliding of the rack modules on the pool floor.

These analyses were necessary in order to ensure that under the action of SSE vibratory motion, the racks would not impact with each other or with the adjacent pool walls.

Each module was modeled as a lump mass stick model with friction el enents provided to account for the racks sliding on the floor.

The friction factors used are consistent with the values contained in a report entitled, "Friction Coefficients of Water Lubricated Stainless Steel for a Spent Fuel Rack Facilityu, by Professor Ernest Robinawicz of the Massachusetts Institute of Technology.

To ensure that the proposed racks are laterally stable against overturning loads during an SSE, stability analyses were performed using an energy method.

Safety factors against overturning were calculated based on the worst case rack*configuration, with overturning occurring at the edge closest to the center of gravity of the racks-.

A thermal analysis was also performed which considered both the increase in pool temperature and the thennal gradient resulting from a "hot" fuel bundle adjacent to empty cells. The racks were also designed to withstand the local as**

well as gross effects of the impact of a fuel assembly dropped from a height of 12" above the rack (the maximum height limitation restricted by the fuel handling tool) and the effects of a fuel assembly dropping into an empty cell and impacting the botton rack support plate.

An elasto-plastic analysis of the rack: was performed to determine the maximum length of the racks stressed beyond the elastic limit and safety factors against having Keff > 0.95 were calculated. Objects other than the fuel assenblies were also considered from the maximum kinetic energy with which they might impact the rack thereby causing similar damage to the rack.

The fuel pool struc:ure consists of concrete i*1alls and floor, lined with a Type 304 stain1ess steel liner plate. Using the working stress and ultimate strength methods, analyses in accordance with ACI Building Code 318-71 were perfonned to detennine the section strength and stiffness properties of pool floor and wall slabs.

The increased pool loads, pool temperature, maximum thennal gradient, (in accordance with ACE 307-69) operating and safe shutdown earthquakes were considered in the ana1yses.

The resulting loads were evaluated in ~ccordance with the applicable parts of Section 3.8.4 of the Standard Review Plan, dated November 1975.

In August 1978, the staff was made aware of a problem at the Monticello Nuclear facility with regard to spent fuel storage racks similar in design tothoseproposed for use at Dresden.

The problem involved in-leakage of water into the stainless steel cans.

As a result, hydrogen gas was generated by oxidation of the exposed aluminum material.

This gas caused a pressure buildup and resultant swelling of the stainless steel cans, such that the removal of a fuel assembly, if located at an affected storage location, could not be performed.

The Dresden fuel rack cells will be vented to release any off-gases from the 3oral and therefore swelling due to gas buildup will be prevented.

The venting is provided by four 1/4 inch holes punched into the inside walls of the top of the tube and by not ;4elding the bottom four corners of each tube.

The venting allows the spent fuel pool water to come into contact with the Baral material and hydrogen gas from oxidation of the aluminum to escap~. Venting of the tubes will not compromise

  • the structura 1 *integrity of the rack.

3.4. 1.2 Material The Type 304 stainless steel used in the new storage racks is compatible with the storage pool environment, which is oxygen-saturated, high purity, demineralized water, controlled to a maximum 145°F temperature.

In this pool water environment, the corrosive det5rioration of the 304 alloy should not exceed a depth of 6.00 x 10-inches in 100 years, which is minute relative to the initial thickness.

Dissimilar alloy interaction (electrolytic or galvanic corrosion) between the 304 stainless steel storage rack, Inconel and Zircaloy in the,spent fuel assemblies, and the 304 stainless steel pool liner, will be of no significance because of the minute electrical potential differential and the high resistivity of the water.

The aluminum in the Baral neutron absorber plates is more reactive than stainless steel. tubes encapsulating the Baral plates since the Baral is being vented to the pool water environment.

The high resistivity pure water environment, however, will minimize any galvanic action.

The more noble stainless steel would not be affected by any galvanic attack when contacted with aluminum.

Although slight galvanic corrosion may occur in the unanodized aluminum of the Boral plates, it should not have any significance on the neutron absorption capability of the Baral, and certainly no effect on storage rack structural integrity for a period far in excess of 40 years.

-*.. 3.4.2 Conclusion Analyses, design, fabrication and installation of the proposed fuel rack storage system are in accordance with accepted criteria.

The new storage racks were designed as seismic Category I equipment.

The analysis of the structural loads imposed by dynamic, static, seismic and thennal loadings; and the acceptance criteria for the appropriate loading conditions are in accordance with the appropriate portions of the "NRC OT Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications," April 1978.

The effects of additional loads on the existing pool structure du.e to high density storage racks have been examined.

Comparison of computed loads and stresses versus pennissible ones shows adequacy of the pool structure to withstand the added loads. Although acknowledgement has been made that corrosion will occur in the Dresden spent fuel storage pool enviror.rnent, it should be of little significance for at least 40 years.

All of the components in the Dresden spent fuel storage pool, excluding the aluminum in the Boral neutron absorber plates, are constructed of alloys with the same electrical potential (or a minute differential) that have a high resistance to general chemical corrosion*, electrolytic corrosion, and galvanic corrosion.

The only spent fuel pool components of concern are the storage rack modules, which have a galvanic coupling between the stainless steel tubes and the unanodized aluminum in the Boral.

The deteriora-tion of the aluminum in the Boral should be minimal and should not affect the neutron shielding properties of the Boral.

The B4C neutron absorber particles are inert to the pool water environment.

Based on our review of the design, fabrication, installation and analyses of the proposed fuel racks, we conclude that these aspects of the proposed modifications to the.Dresden 2 and 3 spent fuel' pool are in confonnance with NRC requirements.

To aid in verifying the above conclusions, the licensee has c01T111itted to conduct a long-tenn fuel storage surveillance program to verify that the spent fuel storage cell retains the material suitability and mechanical integrity over the life of the spent fuel storage racks under actual spent fuel pool service conditions.

Sample flat plate sandwich coupons and short fuel storage cell sections will be placed adjacent to the fuel storage racks and removed for visual and weight analysis on a schedule extending over a.40 year period.

We conclude that the proposed structural, mechanical and material design of the modified spent fuel pool is acceptable.

~-

. :'~ 3.5 Radiological' Considerations Our radiological review addressed the following major considerations:

fuel and heavy load handling, occupational radiation exposure, and radioactive waste treatment.

3.5.1 Evaluation 3.5.1.1 Fuel and Heavy Load Handling The consequences of fuel handling accidents in the spent fu~l pool are not changed from those presented in the Dresden 2 S-af ety Evaluation dated October 1969 and in the Dresden 3 Safety Evaluation dated November 1970, and are therefore acceptable.

The NRC staff has under way a generic review of load handling operations in the vicinity of spent fuel pools to detennine the likelihood of a heavy load impacting fuel in the pool and, if necessary, the radio-logical consequences of such an event.

Because Dresden 2/3 will be prohibited from the movement of loads, other than a spent fuel shipping cask (rn*the Safety Evaluation dated June 1976, we concluded that the overhead crane handling system and the spent fuel cask handling Technical Specifications meet our requirements and are acceptable for handling spent fuel casks weighing* up to 100 tons.}, with weight in excess of the nominal weight of a fuel assembly and handling tool over spent fuel assemblies in the SFP, we have concluded that the likelihood of a heavy load handling accident is sufficiently small that the proposed modification is acceptable and that no additional restrictions on load handling operations in the vicinity of the SFP are necessary while our review is under way.

3.5.1.2 Occupational Radiation Exposure We have reviewed CECo's plan for the removal and disposal of the low density racks and the installation of the high density racks with respect to occupational radiation exposure.

The occupational exposure for this operation is estimated to range from 18 to 47 man-rem.

This estimate is based on the licensee's detailed break-down of occupational exposure for each phase of the modification.

CECo considered the number of individuals performing a specific job, their occupancy time while perfonning this job, and the average dose rate in the area where the job was being performed.

In the calculations CECo uses conservative estimates of dose-rate and man-hours to perfonn a specific operation.

Crud may be released to the pool water because. of fuel moveme.nts during the proposed SFP modification. This could increase radiation levels in the vicinity of the pool and decrease the clarity of the water.

There will be a number of fuel movements in each pool during the modification equivalent to less than eight refuelings for each pool (i.e., the modification is done after the 1979 refuelinqs

llt... for* each pool). Most of the addition of crud to the pool water is from the fuel assemblies and the primary coolant introduced to the spent fuel pool during refueling.

However, the plants have not experienced significant releases of crud to the pool water during refuelings. The licensee does not expect to have signifi-cant releases of crud to the pool water during the modification of the pools.

The purification system for each pool, which has kept radiation levels in the vicinity of the pool to low levels, includes a filter to remove crud and will be operating during the modification of the pools.

The pool floor will be vacuumed during the modification to remove particles which fall to the floor.

The staff concludes that the SFP modification can be performed in a manner that will ensure as low as is reasonably achievable (ALARA} exposures to occupational workers.

CECo has presented the following alternative plans for the disposal of the old racks:

(l} cutting the old racks into small sections to significantly reduce the volume to be shipped to the burial site or (2} crating the racks whole which will reduce the man-rem exposure involved with disposing of these racks. Cutting the old racks into small sections will permit more efficient packaging in the shipping containers. This will result in a smaller volume of radioactive waste to be disposed of with resulting economic and environmental benefits, e.g., fewer waste shipments and conser-vation of low level waste burial site space.

This also requires that CECo commit personnel to cut up the old racks, resulting in an increase in occupational exposure.

CECo has stated that estimates of the exposures associated with the different ways to dispose of the old racks will be made from measurements of the dose rate from the old racks when the racks are removed from the SFP, decontaminated and ready for disposal. At this time, taking into account alternative disposal costs and exposures, CECo will make the final decision as to the choice of method of disassembly and disposal of the old racks so that exposures will be kept to levels that are as low as reasonably achievable (ALARA).

We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies at both units. This estimate is based on information supplied by the licensee and by utilizing relevant 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 cool area because of the depth of water shielding the fuel.

The occupational radiation exposure resulting from the proposed action represents a negligible burden.

Based on' present and projected operations in the spent fuel pool area, we estimate that the proposed modification should add less than one percent to the total annual occupational radiation exposure burden at both units.

The small increase in radiation exposure should not affect the licensee's ability to maintain individual occupational doses to as low as is reasonably achievable levels and within.the limits of 10 CFR Part 20.

Thus, we conclude that storing additfonal fuel in the two pools will not result in any significant increase in doses received by occupational workers.

  • **I
  • 3.5.1.3 Radfoactive Waste Treatment 3.5.2 The station contains waste treatment systems designed to collect and process the gaseous, liquid and solid wastes that might contain radioactive material. The waste treatment systems were evaluated in the Dresden 2 Safety Evaluation (SER) dated October 1969 and in the Dresden 3 SER dated November 1970.

There will be no change in the waste treatment systems or in the conclusions of the evaluation of these systems in Section 3.6.3 of the Dresden 2 SER and Section 3.7.2 of the Dresden 3 SER because of the proposed modification.

Conclusions Our evaluation of the radiological considerations supports *the conclusion that the proposed modification to the Dresden 2/3 spent fuel pools is acceptable because:

(1)

The Jikelihood of an accident involving heavy loads in the vicinity of the spent fuel pool is*sufficiently small that no additional' restrictions on load movement are necessary while our generic review of the issues is underway.

(2)

The insta1lation and use of the new fuel racks does not alter the potential consequences of the design basis accident for the SFP, i.e., the rupture of all the fuel pins in the equivalent of a single fuel assembly and the subsequent release of the radioactive inventory within the gap of each fuel pin.

(3}

The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the SFP would be negligible.

(4)

The conclusions of the evaluation of the waste treatment systems are unchanged by the modification of the SFP.

4.0

SUMMARY

Our evaluation concludes that the proposed modification of the Dresden 2 and 3 spent fuel pools is acceptable because:

(1) The physical design of the new storage racks will preclude criticality for any credible moderating condition.

(2}

The cooling system for each of the spent fuel pools has adequate cooling capacity.

(3)

The installation and use of the proposed fuel handling racks can be accomplished safely.

(4)

The structural design and the materials of construction are adequate.

( 5 )*"

( 6)

(7)

(8) The likelihood of an accident involving heavy loads in the vicinity of the spent fuel poo1 is sufficiently small that no additional restrictions on load movement are necessary while our generic review of the issues is underway.

The installation and use of the new fuel racks does not alter the potential consequences of the design basis accident for the SFP, i.e., the rupture of all the fuel pins in the equivalent of a single fuel assembly and the subsequent release of the radioactive inventory within the gap of each fuel pin.

The increase in occupational radiation exposure to.individuals due to ~he storage of additional fuel

~n the SFP would be negligible.

The conclusions' of the evaluation of the waste treatment systems are unchanged by the modification of the SFP.

5.0 CONCLUSION

We have concluded, based on the cons~derations discussed above, that:

(1) there is reasonable assurance that the health and*

safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the proposed license amendments will not be inimical to the common defense and security or to the health and safety of the public.

Retrieved from "https://kanterella.com/w/index.php?title=ML17192A771&oldid=5130875"