ML20054E124
| ML20054E124 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/30/1982 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20054E122 | List: |
| References | |
| TAC-48219, NUDOCS 8204260163 | |
| Download: ML20054E124 (14) | |
Text
FARLEY NUCLEAR PLANT UNIT 1 SPENT FUEL POOL MODIFICATION AMENDMENT 1 REVISION INSERTION INSTRUCTIONS Section Page Instruction III III-2 Replace III III-3 Replace III III-4 Replace III III-5 Replace III III-6 Replace III III-7 Replace III III-8 Replace III III-9 Replace..,
i III III-10 Add III III-ll Add III III-12 Add III III-13 Add III III-14 Add i
B204260 b>
)
10 in.) existing between the top of the assemblies already inside the cavities and the dropped assembly resting on top of the rack.
Moreover, the calculational model assumes an infinite fuel length in the a:tial direction.
b)
Single assembly next to rack The dropping of an assembly outside the rack is a possible event because of the unobstructed water area existing between the periphery of the storage racks and the side walls of the pool.
A conservative analysis was performed to evaluate this situation.
The results indicate that with the presence of soluble boron in the pool water, as permitted by the double contingency rule, the dropping of fuel assembly next to the rack proper does not raise the k ef f value of the racks, with all uncertainties and biases included, to above 0.950.
III.1.2.(2)
Protection against a cask drop is assured by the Seismic Category 1, Crane Manufacturers Association of American (CMAA)
Specification No. 70, Class A1, single failure-proof outdoor spent fuel cask crane, by the single failure-proof lifting device, and by the interlocks and administrative controls described in the Farley FSAR subsection 9.1.4.
A cask drop or tip into the spent fuel pool is also prevented by permanently installed rail stops and mechanical bumpers which prohibit cask crane travel over or into the vicinity of the spent fuel pool.
The cask crane hook approach to the spent fuel pool, as shown in Farley FSAR figures 1.2-2 and 1.2-10, is limited to approximately 12 feet.
Since the cask will not be handled in the vicinity of the spent fuel pool, the consequences of the cask drop are not affected by the increased storage of the spent fuel pool.
The spent fuel bridge crane, located inside the spent fuel pool building, is used for refueling operations.
The spent fuel bridge crane is the only crane capable of handling objects over the spent fuel pool.
When fuel assembies are stored in the spent fuel pool, the size of the load that can be handled over the spent fuel pool is limited to 3000 lb by Farley Unit 1 Technical Specification, Section 3.9.7.1.
Use of the spent fuel bridge crane is discussed in Farley FSAR subsections 9.1.2 and 9.1.4.
The 3000-lb load bounds the spent fuel assembly with control rod anu handling fixture (1992 lb).
The fuel assembly drop is analyzed at a drop height of 42 in.,
which bounds the actual drop of 39.5 in.
The drop height is limited by a high level limit switch on the lift in combination with a long handling tool.
III-2 AMEND 14/82
Administrative controls prohibit a gate (3600 lb) from being carried over spent fuel.
However, the racks are designed to withstand a gate drop from 9 in.,
which bounds the actual drop of 6 1/2 in.
The drop height is limited by a physical limitation in lifting capability.
The fuel assembly drop assuming a load of 3000 lb at 42 in. is the worst impact load condition, but is very conservative considering an actual drop of 1992 lb at 39.5 in.
Therefore, the load drops identified in Section IV.(1).b of this report do not have a higher kinetic energy than that assumed in the accident analysis.
The loads associated with a fuel assembly with control rod (1650 lb) and handling fixture (376 lb) are bounded by the 3000-lb load specification, which was used in the drop accident analysis.
The handling fixture is not used with the 3600-lb gate.
The gate is designed with a clevis at top center, which the crane hook fits into for lifting and relocating.
The analyzed fuel assembly drop load of 3000 lb at 42 in.
reprssents the worst fuel rack impact load condition; i.e.,
highest kinetic energy; of all loads that could be moved _over the spent fuel pool.
The kinetic energy of loads lighter than one fuel assembly will be less than that calculated for the analyzed fuel assembly drop load described above even if the lighter loads are released from the maximum possible elevation.
Therefore, limitations on lift height of these lighter loads are not required.
Subsection III.1.2.(1) of this submittal discusses the dropping of a fuel element on top of the racks or any other achievable abnormal tocation of a fuel assembly in the pool.
l l
III.1.2.(3)
The exterior walls and roof slab of the spent fuel area are designed for tornado wind, differential pressure, and missile loadings on the basis that the area is fully enclosed.
Farley l
FSAR subsection 3.3.2.1 specifies tornado design criteria for fully enclosed Category I structures.
A tornado will not have an effect on the deformation and relative position of the fuel racks since the building structure will not be modified for the l
rerack program, and all existing building analyses will remain l
valid and unchanged.
The design of the racks is seismically qualified, therefore, the earthquake effect on criticality is of no concern due to l
the structural acceptance of the rack.
i l
l III-3 AMEND 14/82 l
1 A%,Y
III.l.2.(4)
Loss of all cooling systems will not result in criticality.
In addition, the spent fuel pool cooling system is Seismic Category I and meets the single failure criteria.
III. 1.3 Calculation Methods The criticality calculations are performed by Nuclear Associates International (NAI), a consulting service of the Control Data Corporation.
Both the KENO-IV and PDQ-7 code s
packages have been installed and in use on the CDC CYBER systems for a number of years.
The AMPX/ KENO-IV model, which is the basic code package used in the calculation of the effective multiplication factor of the infinite rack array, has been benchmarked by NAI against earlier Oak Ridge National Laboratory (ORNL) critical experiments measured by Dr. E.
Johnson, at La Crosse boiling water reactor (BWR) measured cold critical; and the more recent critical experiments conducted by S.
R.
Bierman, et al, at the Batelle Pacific Northwest Laboratory as reported in PNL-2438.
Although the PDQ-7 model is mainly used to provide relative reactivity results for the several sensitivity studies conducted for the Farley analysis, it is also used to calculate the infinite array reactivity of the nominal configuration in part to cross check and verify the results of the Monte Carlo KENO-IV calculations mentioned above.
The PDQ-7 model has been in use at NAI for core physics and fuel management work since the late 1960's when the code was first released for industrial use.
This model has been used to calculate cold critical measured at a dozen pressurized water reactor (PWR) and BWR power plants in the United States.
Furthermore, it has also been used in calculating the La Crosse BWR cold critical and the ORNL/ Johnson critical experiments.
The criticality analysis employed two independent models to validate the results of the nominal geometry calculations.
The base calculational method employs the KENO-IV/AMPX model.
The basic neutron cross-section data comes from the master library of AMPX - a 123 group GAM-THERMOS neutron library prepared from ENDF/B version II data.
The NITAWL model of the AMPX program is used to perform a Nordheim integral treatment of the U-238 resonsances accounting for self-shielding effect.
The working library produced by this process retains the 123 group energy structure and is used directly by KENO-IV.
The KENO-IV/AMPX model has been benchmarked against the critical experiment data measured by Battelle Pacific Northwest Laboratories.
The diffusion theory model, which uses three codes, namely, CHEETAH-P, CORC-BLADE, and PDQ-7 is used for the validation III-4 AMEND 14/82
process and is also used to perform all the sensitivity calculations.
The model has been extensively tested through benchmarking calculations of measured criticals, as well as through core physics calculations for several operating reactors.
The final keff value for the Farley spent fuel racks is obtained by summing the k eff value calculated by the KENO-IV/AMPX model, the calculational bias which was obtained from the benchmark work and the total uncertainty which was obtained by a statistical combination of the calculational and mechanical uncertainties.
The calculational uncertainity is such that the true multiplication factor (keff) will be less than the calculated value with a 95-percent probability at a 95-percent confidence level.
The nominal reactivity of the new high density spent fuel pool (HDSFP) is determined using the Monte Carlo Code KENO-IV to be 0.9217 0.0044 with the 95-percent confidence interval ranging from 0.9127 to 0.9305.
The results have been adiusted to include the effects due to the presence of U-234 and Inconel spacer grids.
The following summarizes the reactivity inventory of the Farley HDSFP.
Nominal K 0.9217 0.0044 eff Calculational bias
+0.0027 Mechnical uncertainty (worst case
+0.0159 geometry)
Final Keff 0.9403iO.0044 95-percent confidence interval 0.9315-0.9491 A temperature reactivity sensitivity study covering the range from 68 F to 212 F shows that the effective multiplication factor has the highest value at the lowest reasonable pool temperature (68 F).
The nominal effective multiplication factor is for the 68 F condition.
III.1.4 Rack Modification The spent fuel storage racks being supplied to Alabama Power Company are of new construction with no modifications to the existing racks or pool liner being required.
The existing racks will be partially removed prior to installation of the new racks as described in section III.1.5.(4).
III-5 AMEND 14/82
III.1.4.(a)
The overall fuel assembly parameters for the Earley PWR 17 x 17 fuel assemblies are as follows.
Fuel assembly dimension 8.426 in. x 8.426 in.
Storage cell pitch 10.75 in.
Percent of total cell area 61.4 occupied by a fuel assembly III.l.4.(b)
This section of the NRC position paper is not applicable since the new high-density racks utilize a neutron absorbing poison rather than stainless steel flux traps.
III.1.4.(c)
Refer to III.1.4.(b) above.
III.1.4.d.(1)
There are two poison plates separating every two adjacent fuel assemblies stored in.the racks.
The poison 2
plates have a B-10 loading of 0.015 g/cm,
a, III.1.4.d.(2).a The analysis is based on an average enrichment of 4.3 weight percent of U-235.
The fuel loading is calculated to be 54.25 gram of U-235 per axial centimeter of fuel assembly.
The reactivity sensitivity of enrichment is calculated to be 0.0031 delta k per gram U-235 change around the nominal fuel loading.
I I I ~.1. 4. 3. ( 2 ). b The nominal storage lattice pitch is 10.75 in.
The pitch reactivity sensitivity is calculated to be 0.010 delta k per O.1 in, change in pitch around the nominal value.
III.1.4.d.(2).c The B-10 loading in the poison plates is 0.015 2
g/cm.
The reactivity sensitivity due to B-10 loading 2
variation is calculated to be 0.0109 delta k per 0.005 g/cm change around the nominal loading.
III.1.5 Acceptance Criteria For Criticality The acceptance criteria for criticality calculations is that kegg be less than or equal to 0.95 including all uncertainties.
III.1.5(1)
Neutron Absorber Verification par Systems, operating under a Quality Assurance Program which meets 10 CFR 50 Appendix.B, requires the poison manufacturer to produce his product under a program which also meets 10 CFR 50 III-6 AMEND 14/82
Appendix B.
A detailed specification is part of the purchase order.
This specification covers the neutron absorber sheet requirements, material requirements, quality assurance program requirements, documentation requirements, etc.
par requires the manufacturer to submit his quality assurance program manual and operating procedures for approval before the start of production.
par also audits the quality assurance program at the manufacturing facility at least once a year (or before the first order).
After receipt of material, par reviews all documentation for conformance before incorporating the poison into the spent fuel racks.
par maintains traceability of the poison material throughout the rack manufacturing process.
Alabama Power Company or its agent is committed to periodically perform quality audits and inspection of the above described quality program.
III.1.5(2)
Decay Heat Calculation for the Spent Fuel The calculations for the amount of thermal energy that will have to be removed by the spent fuel pool cooling system are made in accordance with Branch Technical Position APCSB 9-2 entitled, " Residual Decay Energy for Light Water Reactors for Long Term Cooling." This Branch Technical Position is part of the Standard Review Plan (NUREG 78/087).
III.1.5.(3)
Thermal Hydraulic Analysis of Spent Fuel Cooling The computer code HPOOL is used to analyze the natural circulation cooling of the spent fuel under normal cooling conditions.
HPOOL is a proprietary program of Nuclear Associates Incorporated (NAI).
HPOOL calculates the pressure loss through a fuel assembly for a given flowrate.
This pressure loss is compared with the buoyant head resulting from the difference between the average density of the fluid in the fuel channel and the average density of the fluid in the downcomer.
The downcomer is the space between the wall of the pool and the racks.
If the density difference results in a buoyant head greater than the pressure loss, the flowrate through the fuel assembly is increased and a new average density of the fluid is determined.
This iterative process is continued until the buoyant head and pressure loss in the fuel assembly are equal.
Using this flowrate, HPOOL determines the fuel temperature.
HPOOL was used to determine the bulk spent fuel pool water temperature for the normal refueling case and the emergency full-core offlead case.
The full-core offload case represents the worst case heat loading condition which provides the maximum bulk spent fuel pool water temperature.
The analysis and assumptions results fer these two cases are presented below.
III-7 AMEND 14/82
A.
Normal. Refueling:
Assumptions:
1.
All storage cells filled.
2.
28 years - 1/3 of the core removed each year.
29th year - 0.40 of the core removed.
(Envelope's largest normal refueling discharge).
3.
105*F component cooling water temperature (Influent to the spent fuel pool heat exchangers.)
Results:
Total Spent Fuel No. of Cooling Hours After Heat Load Pool Bulk Trains Operational Shutdown (10s Btu /hr)
Temp. (*F) 1 100 19.755 139 2
100 19.755 122 1
150 17.579 135 2
150 17.579 120 B.
Full-Core Offload:
Assumptions:
1.
All storage cells filled.
2.
25 years - 1/3 of core removed each year.
26th year - 0.40 of core removed.
Emergency full-core offload is necessary 180 hr after normal refueling in 26th year.
Full core is discharged i
l to the spent fuel pool 10 days after the emergency shutdown.
3.
105 F component cooling water temperature.
l Results:
1 j
No. of Total Spent Fuel l
Ccoling Trains Heat Load Pool Bulk l
Operation (108 Btu /hr)
Temperature (*F) 1 30.384 158 l
2 30.384 131 The computer code BPOOL is used to analyze the natural l
circulation cooling of the spent fuel in the event of a loss of all external means of cooling for the spent fuel pool.
BPOOL is a proprietary program of NAI.
The code is based on the III-8 AMEND 14/82 1
assumption that boiling takes place near the top of the fuel channel.
BPOOL evaluates the saturation properties of the coolant on the basis of the static pressure at the top of the storage racks.
These properties include water density, temperature, and steam density.
The steam is assumed to separate and flow out of the pool.
The water at the saturation temperature corresponding to the pressure at the top of th'e racks flows downward to the inlet of the storage racks.
The static pressure at this location is higher than the pressure at the top of the storage racks and as a result the fluid is subcooled as it enters the fuel assembly.
The fluid becomes less dense as it passes up the fuel channel.
Near the top of o
the fuel channel the fluid reaches saturation conditions and net boiling occurs.
The computer code, BPOOL, assumes a loss of all external means of cooling, but it should be noted that the Farley spent fuel pool cooling system is redundant and single failure-proof.
Under normal conditions, voiding between fuel assemblies is highly unlikely because these spaces are not sealed to keep out water.
Holes are provided at the top and bottom of each inner can to permit a definitive flowpath for circulation of water in these spaces.
m No modifications to the spent fuel pool cooling and cleanup system will be made as a result of the installation of high density poison spent fuel storage racks.
The Farley spent fuel pool cleanup system is described in FSAR subsection 9.1.3.
The system's demineralizers and filters are designed to provide adequate purification to permit unrestricted access for plant personnel to the spent fuel storage area and maintain optical clarity of the spent fuel pool water.
The optical clarity of the spent fuel pool water surface is maintained by use of the system's skimmers, strainer, and skimmer filter.
Operating plant experience has indicated that crud release occurs immediately after refueling with the fuel pool water returning to normal conditions within a few days.
The increase in spent fuel assembly capacity of the spent fuel pool will not affect the optical clarity of the spent fuel pool water.
A maximum radiation level and minimum decontamination factor (DF) will be established by procedure for the spent fuel pool (SFP) demineralizer.
Accordingly the demineralizer resins may be exchanged at a more frequent interval than would be required for the original rack configuration.
The SFP filters may also be exchanged at a more frequent rate according to AP limitations in table III-1.
The SEPCCS monitoring parameters, sampling frequency, limits to be maintained and their bases are shown in table III-1.
AMEND 14/82 7 7 gg
III.1.5.(4)
Potential Fuel and Rack Handling Accidents The high-density poison racks are of a free-standing design, utilizing bottom support pads, resting on the floor of the spent fuel pool.
The installation of the high-density racks will include removal of the existing 13-in. center storage racks.
The high-density racks will be installed wet since there is spent fuel in the storage pool.
The following is a sequence of events for installing the high-density poison racks.
Phase I Install and test a temporary crane for handling the existing racks and the high-density racks.
The spent fuel bridge crane is a 4,000-pound capacity crane and is not of adequate capacity for the re-rack modification.
Phase II Remove and decontaminate a portion of the existing 13-in. center racks, leaving enough racks intact for existing spent fuel assemblies and one emergency full-core offload and those fuel assemblies, which are stored in the pool from three previous refuelings.
At no time during this phase of work will the 13-in. spent fuel rack modules be moved over spent fuel assemblies.
Phase III Remove all adapter plates.
Phase IV Install the high-density poison racks into the pool areas vacated by the removal of the 13-inch center-to-center racks.
This work will be done systematically to assure that at no time will the new spent fuel racks be moved over spent fuel assemblies.
Storage capability for one emergency full-core offload will be available at all times j
during the spent fuel pool rerack program.
Phase V When one full-core offload storage capacity is achieved with new racks and existing fuel assemblies are transferred to new racks, then remove any remaining 13-in. center-to-center racks and adapter plates from the spent fuel pool.
Phase VI Install the balance of high-density racks into the spent fuel pool to complete rack installation.
l Phase VII Remove temporary crane from the spent fuel pool area.
Throughout the above phases of work the spent fuel assemblies in the spent fuel pool vill be moved and located so that no heavy loads will be carried over them.
Strict administrative procedures will be imposed throughout the rerack program to AMEND 14/82 III-10
assure protection against a rack module dropping on the spent fuel assemblies.
These phases of work will require support work; i.e.
leveling of new racks, testing, etc.; to complete the reracking program.
The* outdoor spent fuel cask crane will be used to bring the high-density poison racks from the delivery vehicle into the spent fuel cask area.
Physical stops prohibit this crane from carrying loads over the spent fuel pool.
The racks will then be moved from the spent fuel cask area, by the temporary crane, into the spent fuel pool.
The reverse sequence will be a
performed to remove the existing 13-inch center storage racks from the spent fuel pool.
The installation of the high-density poison racks will not increase the potential for a fuel and racP. handling accident for the following reasons:
No heavy loads will be carried over any spent fuel e
assemblies during the rerack program.
The spent fuel assemblies will be moved and located so that no heavy loads will be carried over them.
e The temporary crane, as with the spent fuel bridge crane, can carry loads over the spent fuel cask area and the spent fuel pool only.
There is no safe shutdown equipment located in these areas.
Therefore, there will not be any damage to safe shutdown equipment should a rack drop into these areas.
e The spent fuel cask crane, used to bring the racks into and out of the spent fuel cask area, is a single failure proof crane as described in subsection III.1.2.(2).
Physical stops prohibit this crane from carrying loads over the spent fuel pool.
Protection against a rack drop is assured since the cask crane is single failure-proof, and a dual point attachment will be used between the spent fuel pool cask crane main hook and the lifted spent fuel rack module.
III.1.5.(5)
Technical Specifications To ensure against criticality, the following technical specifications are proposed in figure III-1 on spent fuel storage in the high-density poison racks.
III.1.5.(5).1 Paragraph 5.6.1.1 of the proposed revicion to the Farley Unit 1 Technical Specifications requires that the spent fuel storage racks be designed and maintained such that the neutron multiplication factor (keff) in the fuel pool shall III-11 AMEND 14/82
be less than or equal to 0.95 when flooded with unborated water.
This represents the most conservative pool condition from a criticality standpoint.
III.l.5.(5).2.In addition, paragraph 5.6.1.1 of the proposed revision to the Farley Unit 1 Technical Specifications also specifies a maximum enrichment of 4.3 weight percent U-235 (which equates to 54.25 grams per axial centimeter of the fuel assembly) for fuel loading in the fuel assemblies.
This limit is consistent with the design of the high-density poison racks to preclude criticality in the fuel pool.
e l
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r III-12 AMEND 14/82
TABLE III-1 (SHEET 1 OF 2)
PURITY MONITORING A.
Samples Parameter Frequency Limit Basis Boron Weekly 2000-The limitations on 3000 reactivity condi-ppm tions during re-fueling ensure that:
1.
The reactor will remain subcritical during core alter-nations.
2.
A uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
These limitations are consistent with the initial condi-tions assumed for the boron dilution incident in the ac-cident analyses.
1 pH Weekly 4.0-To be consistent 4.9' with the boron concentration.
l' Cl-Weekly 50.15 Consistent with the ppm acceptable limit for stainless steel systems.
F-Weekly 50.15 Consistent with the ppm acceptable limit for stainless steel systems.
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AMEND 14/82 III-13
TABLE III-1 (SHEET 2 OF 2)
Demineralizer DF Monthly ca>
giocb>
Iodine specific DF to ensure the re-moval of 99 percent of the assumed 10-percent iodine gap activity released from the rupture of an irradiated fuel assembly.
B.
Instrument Parameter Frequency Limit Basis Area Monitor Continuous 15 N/A RE-5 hr Filter AP Once per 20 psid Manufacturer's 8 hr recommendation.
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l t
a.
This will be done daily during forced oxygenation cleanup prior to refueling.
b.
If iodine inventory in the spent fuel pool is low; i.e.,
I-131 s 1 x 10'* pCi/g; the demineralizer may be allowed to stay in service longer.
This may be at the discretion of the C&HP supervisor or C&HP section supervisor.
III-14 AMEND 14/82.
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