ML20125D223
| ML20125D223 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 12/07/1992 |
| From: | OMAHA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20125D184 | List: |
| References | |
| NUDOCS 9212150057 | |
| Download: ML20125D223 (28) | |
Text
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4 References 1
(1) USAR, Section 9.5 (2) USAR, Section 9.5.1.2 2-39 Amendment No. 24,75,403,4+7,443,444, i
i a
i e
i l
t 9212150057 921207 PDR ADOCK 05000285 i P PDR
TAItLE 3-5 (Continued) USAR Section Test Frecuency Reference 10c. (Continted) 4. Automatic and/or manual initiation At least once per plant operating cycle.
of the system sha*,1 be demonstrated.
- 11. Containment Cooling and 1. Demonstrate damper action. I year, 2 years, 5 years, and every 5 9.10 Iodine Rerroval FuadkFisible years thereafter.
Linked Dampers 2. Test a spare fu enMefusible link.
- 12. Diesel Generator Under- Calibrate During each refueling outage. 8.4.3 Voltage Relays
- 13. Motor Operated Safety Verify the contactor pickup value at During each refueling outage.
Injection loop Valve f,85% of 460 V.
Motor Starters (MCV-311, 314, 317, 320, 327, 329, 331, 333, 312, 315, 318, 321)
- 14. Pressurizer Heaters Verify control circuits operation During each refueling outage.
for post-accident beater use.
- 15. Spent Fuel Pool Test neutron poison samples for ' c c ! c' 1, 2, 4, 7, 44 and 10,15, 20, an '
Regr- ! Racks dimensional change, hardness change,and 25 years after installation!and eveiy 5.l years [thereaRef.
Esigh6 neutron attenuation change, and s' pecific~jplivny)harige.
- 16. Reacto Coolant 1. Verify all manual isolation valves During each refueling outage just Gas Vent System in each vent path are in se open prior to plant start-up.
position.
- 2. Cycle each automatic valve in the During each refueling outage.
vent path through at least one complete cycle of full travel from the c(mtrol room. Verification of valve cycling may be determined by observation of position indicating lights.
- 3. Verify fkwv through the reactor During each refueling outage.
coolant vent system vent paths.
3-20d Amendment No. 41,54,60,75,W,40
4.0 DESIGN FEATURF3 4.4 Fuel Storace 4.4.1 New Fuel Storagg The new unitradiated fuel bundles will normally be stored in the dry new fuel storage rack with an effective multiplication factor of less than 0.9. The open grating Door below the rack and the covers above the racks, along with generous provision for drainage, precludes flooding of the new fuel storage rack.
New fuel may also be stored in shipping containers or in the spent fuel pool racks which have a maximum effective multiplication factor of 0.95 with Fort Calhoun Type C fuel and unborated water.
The new fuel storage racks are designed as a Class I structure.
4.4.2 Spent Fuel Storage Irradiated fuel bundles will be stored prior to off-site shipment in the stainless steel lined spent fuel pool. The spent fuel pool is normally filled with borated water with 4
a concentration of at least the refueling boron concentration.
The spent fuel racks are designed as a Class I structure.
Normally the spent fuel pool cooling system will maintain the bulk water temperature of the pool below 120'F. Under other conditions of fuel discharge, the fuel pool water temperature is maintained below 140'F.
The spent fuel racks are designed and will be maintained such that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertainties) assuming the pool is flooded with unborated water. The racks are
! divided into 2 regions. Region 1 ind]25114 raeks are surrounded by Borefleg Regica 2 raeb have ac poison B6fal. Acceptance criteria for fuel storage in Regions 1 and 2 are delineated in Section 2.8 of these Technical Specifications.
4-4 Amendment No. M,43,75, i M3,G3,444, l
l
i 5.0 ADMINISTRATIVE CONTROLS i 5.10.2 The following records shall be retained for the duration of the Facility Operating
! License:
t-4 j a. Records of drawing changes reflecting facility design modifications made to i systems and equipment described in the Final Safety Analysis Report.
j b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup
! histories.
- c. Records of facility radiation and contamination surveys. 1 i
i d. Records of radiation exposure for all individuals entering radiation control areas.
j e. Records of gaseous and liquid radioactive material released to the environs. .
L f. Records of transient or operational cycles for those facility components designed j for a limited number of transients or cycles, j g. Records of training and qualification for current members of the plant staff.
- h. Records of in-service inspections performed pursuant. to these Technical l Specifications.
- i. Records of Quality Assurance activities required by the QA Manual.
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of meetings of the Plant Review Committee and the Safety Audit and l Review Committee.
l 1. Records of Environmental Qualification of Electric Equipment pursuant to.10
- CFR 50.49.
i m. Records of the service lives of all hydraulic and mechanical snubbers which are
[ covered under the provisions of Section 2.18 of the Technical Specifications, i including the date at which the service life commences and associated installation
! and maintenance records.
[ n. Records of analyses required by the Radiological Environmental Monitoring l Program.
! 5.10.3 A complete record of the analysis employed in the selection of any fuel assembly to be p placed in Region 2 of the spent fuel racks will be retained as long as that bundles'sssssblj i remains in Region 2 (reference Technical Specifications 2.8(42) and 4.&4).
1-5.11 Radiation Protection Program i
l Procedures for personnel radiation protection shall be prepared consistent with the I requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
4 4
E E
y 5-19 Ordef 70/24/80
} Amendmer,t No. 59,86,93,99,4 45,.
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2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Ooerations (Continued)
(6) Direct communication between personnel in the control room and at the refueling machine shall be available whenever changes in core geometry are taking place.
(7) When irradiated fur' is being handled in the auxiliary building, the exhaust ventilation from the spent fuel pool area will be diverted through the charcoal filter.
(8) Prior to initial core loading and prior to refueling operations, a complete check out, including a load test, shall be conducted on fuel handling cranes that will be required during the refueling operation to handle spent fuel assemblies.
(9) A minimum of 23 feet of water above the top of the core shall be maintained whenever irradiated fuel is being handled.
(10) Storage in Region 1 and Region 2 of the spent fuel racks shall be restricted to fuel assemblies having initial enrichment less or equal to 4.2 weight percent of l U-235.
(11) Storage in Region 2 of the spent fuel racks shall be restricted to those assemblics whose parameters fall within the " acceptable" area of Figure 2-10. Storage in the peripheral cells of Rcgion 2 shall be restricted to those assemblics whose parameters fall within the noted area of Figure 2-10.
(12) A minimum boron concentration of 100 ppm shall be maintained in the Spent Fuel Pool whenever storing unirradiated fuel in the Spent Fuel Pool.
If any of the above conditions are not met, all refueling operations shall cease immediately, work shall be initiated to satisfy the required conditions, and no operations that may change the reactivity of the core shall be made. However, refueling operations l may commence and continue with less than 5 containment atmosphere and plant l ventilation duct radiation monitors provided that gross, particulate and iodine monitors
! are monitoring the stack effluent. These three plant ventilation duct radiation monitors will initiate closure of the containment pressure relief, air sample and purge system valves and shall employ a one-out-of-three logic for the initiation of VIAS.
The spent fuel assembly may be transferred directly from the reactor core to the spent fuel pool Region 2 provided the independent verification of assembly burnups as defined in Special Procedure SP-BURNUP-1 has been completed ud the assembly burnup meets l the acceptance criteria identified in Technical Specification Figure 2-10.
l Irradiated fuel movement shall not be initiated before the reactor core has decayed for l a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if the reactor has been operated at power levels in excess of 2%
l rated power.
2-38 Amendment No. 5,24,25,43,-75,G3,
' FIGURE 2-10 37500 -
35000 :
32500 :
30000':
/
27500 3 ACCEPTABLE /
3 BURNUP DOMAIN / ,
B / '
r 25000 :
s f
- g/ ,
22500 : ,s i
20000 :
Alsl- 'G'#. '
@ ${ & ,
a- C2
.i g 17500 :
,e '
-8 w 15000 5 '8 '['
A ! d ,
[
12500 . <
l
[, .Y' 10000 : s i UNACCEPTnBLE 7500 ; ,
.BUFNUP DOMAIN
, ( Re autres Region 1 Storege) 5000-!
l
= /' . , '
2500 '
0 . /... .... .... .... .... . . . , _ .
-1.5- 2.0 '2. 5 . '3.0 3.5 14.0 4.5 '5. 0 -
INITIAL ENRICHMENT, wt% U235 -
- LIMITING BURNUP CRITERIA FOR ACCEPTABLE STORAGE-IN REGION 2 NOTE: -1. - Any fuel essembly ( 24.2%_ U235 everage) mechanically coupled with o l full length CEA may be located :engwhere Lin _ Region 2.
-2. Peripheral cells are those.edjacent to the Spent Fuel Pool well
- or the cask leydown area.
-. , - - . _ - = . - . . - . - - . - - . . . .. . - . - - . .
2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueline Ocerations (Continued)
DaS15 The equipment and general procedures to be utilized during refueling operations are discussed in the USAR. Detailed instructions, the above speci0 cations, and the design of the fuel handling equipment incorporating built-in interlocks and safety features provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety.* Whenever changes are not being made in core geometry one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition. The shutdown cooling pump is used to maintain a uniform boron concentration.
The shutdown margin as indicated will keep the core subcritical even if all CEA's were withdrawn from the core. During refueling operations, the reactor refueling cavity is filled with approximately 250,000 gallons of borated water. The boron concentration of this water (of at least the refueling boron concentration) is sufficient to maintain the reactor suberitical by more than 5%, including allowance for uncertainties, in the cold condition with all rods withdrawn.m Periodic checks of refueling water boron concentration ensures the proper shutdown margin. Communication requirements aliow the control room operator to inform the refueling machine operator of any impending unsafe condition detected from the main control room board indicators during fuel movement.
In addition to the above engineered safety features, interlocks are utilized during refueling operations to ensure safe handling. An excess weight interlock is provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. In addition, interlocks on the auxiliary building crane will prevent the trolley from being moved over storage racks containing irradiated fuel, except as necessary for the handhng of fuel. The restriction of not moving fuel in the reactor for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the power has been removed from the core takes advantage of the decay of the short half-life fission products and allows for any failed fuel to purge itself of fission gases, thus reducing the consequences of fuel handling accident.
The ventilation air for both the containment and the spent fuel poc! area flows through absolute particulate filters and radiation monitors before discharge at the ventilation discharge duct. In the event the stack discharge should indicate a release in excess of the limits in the technical specifications, the containment ventilation flow paths will be closed automatically and the auxiliary building ventilation flow paths will be closed manually. In addition, the exhaust ventilation ductwork from the spent fuel storage area is equipped with a charcoal filter which will be manually put into operatior, whenever irradiated fuel is being handled.*
2-39 Amendment No. 24,U,403, 4+7,433,441,
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2.0 LIMITING CONDITIONS FOR OPERATION 2.8 Refueling Operations (Continued)
The basis for the 100 ppm boron concentration requirement with Boral poisoned storage racks is to maintain the k,y below 0.95 in the event a misplaced unitradiated fuel assembly is located next to a spent fuel assembly. A misplaced unirradiated fuel assembly at 4.2 w/o enrichment condition, in the absence of soluble poison, may result in exceeding the design effective multiplication factor. Soluble boron in the Spent Fuel Pool water, for which credit is permitted under these conditions, would assure that the effective multiplication factor is maintained substantially less than the design condition.
The boron concentration is periodically sampled in accordance with Specification 3.2.
References (1) USAR, Section 9.5 t (2) USAR, Section 9.5.1.2 39a Amendment No.
.- u . _ . ., . -. . . - . _ . - - . . _ _ . . . ~ .
TABLE 3-5 (Continued) USAR Section Test Frequency Reference 10c. (Continued) 4. Automatic and/or manual initiation At least once per plant operating cycle.
of the system shall be denumstrated.
I1. Containment Cooling and 1. Demonstrate damper action. I year,2 years,5 years, and every 5 9.10 lodine Removal Fusible years thereafter. l Linked Dampers 2. Test a spare fusible link. I
- 12. Diesel Generator Under- Calibrate During each refueling outage. 8.4.3 Voltage Relays
- 13. Motor Operated Safety Verify the contactor pickup value at During each refueling outage, Injection Imop Valve $_85% of 460 V.
Motor Starters (HCV-311,314,317,320, 327, 329, 331, 333, 312, 315, 318, 321)
- 14. Pressurizer Heaters Verify contro! circuits operation During each refueling outage.
for post-accident heater use.
- 15. Spent Fuel Pool Racks Test neutron poison samples fo- 1,2,4,7, and to years after dimensional change, hardness change, installation, and every 5 years thereafter.
neutron attenuation change, weight and specific gravity change.
- 16. Reactor Coolant 1. Verify all manual isolation valves During each refueling outagejust Gas Vent System in each vent path are in the open prior to plant start-up.
position.
- 2. Cycle each automatic velve in the During each refueling outage.
vent path through at least one complete cycle of full travel from the control room. Verification of valve cycling may be determined by
, observation of position indicating lights.
- 3. Verify flow through the reactor During each refueling outage.
coolant
- ent system vent paths.
3-20d Amendment No. 44,64,64,M,-W,80,
4.0 DESIGN FEATURES 4.4 Fuel Storage 4,4.1 New Fuel Storagg The new unitradiated fuel bundles will normally be stored in the dry new fuel storage
- rack with an effective multiplication factor of less than 0.9. The open grating floor below the rack and the covers above the racks, along with generous provision for drainage, precludes flooding of the new fuel storage rack.
l New fuel may also be stored in shipping containers or in the spent fuel pool racks which have a maximum effective multiplication factor of 0.95 with Fort Calhoun Type l C fuel and unborated water.
The new fuel storage racks are designed as a Class I structure.
4.4.2 Spent Fuel Storage Irradiated fuel bundles will be stored prior to off-site shipment in the stainless steel lined spent fuel pool. The spent fuel pool is normally filled with borated water with a concentration of at least the refueling boron concentration.
The spent fuel racks are designed as a Class I structure.
Normally the spent fuel pool cooling system will maintain the bulk 14ter temperature of the pool below 12&F. Under otner conditions of fuel discharge s$.he fuel pool water temperature is maintained below 14&F.
The spent fuel racks are designed and will be maintained such that the calculated effective multiplication factor is no greater than 0.95 (including all known
, uncertamties) assuming the pool is flooded with unborated water. The racks are divided into 2 regions. Region 1 and 2 cells are surrounded by Boral. Acceptance criteria for fuel storage in Regions 1 and 2 are delineated in Section 2.8 of Technical Specifications.
4-4 Amendment No. B,43,-75, M3,G3,441,
5.0 ADMINISTRATIVE CONTROIS 5.10.2 The following records shall be retained for the duration of the Facility Operating License:
, a. Records of drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.
, b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup
! histories.
- c. Records of facility radiation and contamination surveys.
- d. Records of radiation exposure for all individuals entering radiation control areas.
- e. Records of gaseous and liquid radioactive material released to the environs.
- f. Records of transient or operational cycles for those facility components designed
- for a limited number of transients or cycles.
- g. Records of training and qualification for current members of the plant staff, h, Records of in-service inspections performed pursuant to these Technical Specifications,
- i. Records of Quality Assurance activities required by the QA Manual.
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
- k. Records of meetings of the Plant Review Committee and the Safety Audit and Review Committee.
- 1. Records of Environmental Qualification of Electric Equipment pursuant to 10 CFR 50.49.
- m. Records of the service lives of all hydraulic and mechanical snubbers which are covered under the provisions of Section 2.18 of the Technical Specifications, including the date at which the service life commences and associated installation and maintenance records.
- n. Records of analyses required by the Radiological Environmental Monitoring l
Program.
l 5.10.3 A complete record of the analysis emplayed in the selection of any fuel assembly to be placed in Region 2 of the spent fuel racks will be retained as long as that assembly remains in Region 2 (reference Technical Specifications 2.8 and 4.4).
5.11 Radiation Protection Program Procedures for personnel radiation protection shall be prepared consistent with the l requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for j all operations involving personnel radiation exposure, i
l l
l 5-19 Order -70/24/80 Amendment No. 59,86,93,99,405,
U.S. Nuclear Regulatory Comission LIC-92-0340A ATTACHMENT B
l ATTACHMENT B Fort Calhoun Station Spent Fuel Pool Reracking DISCUSSION, JUSTIFICATION AND NO SIGNIFICANT HAZARDS CONSIDERATION It is proposed to revise the Fort Calhoun Station (FCS) Unit No. 1 Technical Specifications to allow the fuel rack storage capacity in the spent fuel pool to be increased to 1083 locations. The existing storage racks will be replaced with new high density storage racks. A total of 11 free-standing rack modules in a Discrete Zone, Two Region storage system, will be installed in the Spent Fuel Pool. Two modules, containin are of the design type designated as " Region 1"g a total which of 160unrestricted permits storage cells, storage of fresh fuel up to 4.2 w/o (nominal) U-235 enrichment. The balance of the cells, denoted as " Region 2," will have a prescribed enrichment /burnup restriction.
BACKGROUND Fort Calhoun Station has one spent fuel pool contains spent fuel storage racks consisting o(SFP) which at the present timef a total o The present racks provide adequate capacity for storage of spent fuel while maintaining full core reserve discharge capacity through the end of cycle 15 in 1995. Therefore, to ensure that sufficient spent fuel storage capacity continues to exist at FCS, OPPD has contracted for replacement with high density spent fuel storage racks from Holtec International. The Holtec rack design incorporates Boral as a neutron absorber in the cell walls. The new racks have an ultimate storage capacity of 1083 fuel assemblies, which is expected to extend the full core reserve storage capability through the end of cycle 23 in the year 2007.
The new free-stanting high density s ent fuel storage racks will store fuel in two discrete regions of the SFP. Re ion 1 includes two modules with a total of 160 storage cells. Each cell is esigned for storage of fuel assemblies with Uranium-235 initial enrichments up to 4.2 w/o while maintaining the required subtriticality (k.y s 0.95). Region 2 includes 9 modules with a total of 923 storage cells, which are available for storage of spent fuel assemblies with a prescribed burnup restriction. This region is designed to store fuel which has experienced sufficient burnup or fresh fuel which is of a low enough initial enrichment such that storage in Region 1 is not required or allows storage of a fuel assembly not meeting this requirement provided it contains a control element assembly (CEA).
The high density spent fuel storage rack cells are fabricated from 0.075" thick, type 304 stainless steel sheet material. In Region s, panels of the Boral neutron absorber material are interposed between the cell walls and stainless steel retainers, and the cells are separated by a specified water gap. In Region 2, the Boral panels are located between the stainless steel walls without a water gap. The cells are welded together in a specified manner resulting in a free-standing structure which is structurally qualified for all postulated seismic events. The nominal center-to-center spacings of the cells within Region 1 are 9.821" in E-W direction and 10.363" in N-S direction. The nominal pitch in Region 2 is 8.652" in both directions.
1
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4 L
Since a substantial quantity of spent fuel is presently stored in the FCS SFP, i- special administrative controls and/or procedures will be developed to minimize radiation exposure during the installation of the new spent fuel
- racks. The evaluation of postulated accidents with respect to nuclear j criticality and/or radioactivity release has shown acceptable results in that k,4 does not exceed 0.95, including uncertainties, and postulated radiological releases are well within 10 CFR 100 acceptance criteria.
JUSTIFICATION l the j The proposed storage cells tochanges to the Technical 1083, increasing the allowable Specifications entails increasing fuel enrichment to 4.2 w/o U-1 235, and modifying the burnup/ enrichment restrictions imposed on fuel stored j consistent with the design of the replacement racks.
! As the safety analysis in the following section indicates, there is adequate i justification for proceeding with the proposed changes to the FCS storage system, i
- SAFETY ANALYSIS
- 1. Affected Systems - The following systems and subsystems are potentially affected by the proposed modification:
i a. Storace Racks i
j The spent fuel storage array in the pool is directly affected i since the existing racks will be replaced with new free-standing j high density racks.
} b. Soent Fuel-Pool Coolino System a
i Because the stored quantity of fuel in the SFP will be greater than the presently licensed inventory, the decay heat load will be
- greater than the existing licensing basis-value.
! c. Pool Structurg l The gross inertia load on the.' pool structure will be increased due
- - to the increase in fuel inventory in-the pool. The wall
- j. attachments which physically interfere with the racks or their i function will also be trimmed.
i d. HVAC System The rate of water evaporated from' the SFP will. increase due. to
- j. elevated pool water bulk temperature.
4
- e. Purification System
. The radionuclides released to the pool water may increase-due to the increase in the stored fuel inventory. This may affect the
- ability of the purification system to maintain water purity in the 4
SFP.
-2 b
- - _ . . . , _ . , - , _ . , . .-- ~ , _._ e .,__.m-- . .
Safety Functions The safety function or functions of the affected systems, subsystems or components listed are described below,
- a. Storage Racks: The storage racks provide for vertical upright storage of new or spent fuel assemblies in prismatic cell openings. The racks are designed to maintain structural integrity during and after a Maximum Hypothetical Earthquake (MHE) or a Design Earthquake (DE) event.
- b. Cooling System: The SFP cooling system removes decay heat from the spent fuel discharged from the reactor. Thecoolingsystem maintains the pool water bulk temperature below 140 F (which is well below the boiling temperature of water), under normal full core offload condition.
a
- c. Spent Fuel Pool: The SFP provides wet storage for spent fuel which is stored inside the rack. The racks are designed to store the spent fuel in such a manner as to maintain subcriticality during normal and abnormal conditions. The pool floor slab provides the i
requisite support for the storage rack and fuel assembly system during normal and abnormal conditions,
- d. HVAC System: The HVAC system removes the heat generated by the diffusion of water vapor into the pool environment.
- e. Purification System: The Purification System removes particulate and ionized impurities from the SFP to maintain pool water visibility. This system also helps maintain the boron concentration and desired pH balance in the SFP.
Safety Evaluation The safety evaluation is detailed in the attached licensing report,
" Licensing Report for Spent Fuel Storage Capacity Expansion, Fort
. Calhoun Nuclear Station , Fort Calhoun Station Unit No. 1, Docket No.
50-285, Holtec Report HI-92828 dated November,1992 (hereafter referred to as the " Licensing Report").
The mechanical design of the racks meets all required safety functions stipulated for this equipment in the FCS Updated Safety Analysis Report (USAR). Details of the mechanical configuration are described in Section 3 of the Licensing Report.
The proposed rack arrays have been analyzed to establish their structural integrity under DE and MHE-loadings. Details of the analysis are described in Section C of the licensing report.
The proposed storage expansion will increase the heat load in the pool.
However, analysis has shown that the maximum local water temperature is kept low enough by the existing SFP cooling system such that nucleate boiling or voiding of coolant on the surface of the fuel rod cladding is precluded. The increased pool bulk temperature increases the thermal loading on the reinforced concrete structure and liner of the fuel pool.
The increase occurred for both normal and abnormal conditions.
3
a Reanalysis of the pool structure, however, demonstrated that the 1 integrity of the pool structure and the pool liner is maintained.
. Details of the analysis are presented in Section 8 of the Licensing Report.
Section 9 of the Licensing Report discusses the adequacy of the existing purification system to handle the increased radiological burden.
d The increased evaporation from the pool due to elevated water temperature is safely handled by the HVAC System.
The planned expansion will not increase crud deposition in the SFP since crud deposition occurs during refueling outages and new fuel racks will not affect operation of the clean-up system and/or handling of fuel during refueling outages. The pool clean-up system effectively maintains water clarity and no increase in activity of the clean-up system filters or resins is anticipated.
Safety Margins The NRC process has established that the issue of margin of safety, when applied to a reracking modification, should address the following areas:
- a. Nuclear criticality considerations
- b. Thermal-hydraulic considerations
- c. Mechanical, material and structural considerations
- a. Nuclear criticality considerations The established acceptance criterion for criticality is that thi neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all canditions. This margin of
, safety has been adhered to in the criticality analysis methods for the new rack design.
The methods used in the criticality analysis conforms to the applicable portions of the appropriate NRC guidance and industry codes, standards,
- and specifications. In meeting the acceptarce criteria for criticality in the spent fuel pool, such that k ,, is always less than 0.95 at a 95%/95% probability tolerance level, the proposed amendment does not involve a reduction in the margin of safet)- for nuclear criticality, as defined in the USAR.
- b. Thermal-hydraulic considerations Conservative methods were used to calculate the maximum fuel temperature and the increase in temperature of the water in the SFP. The thermal-hydraulic evaluation used the methods previously employed for evaluations of the present spent fuel racks to demonstrate that the margins of safety for pool temperatures are maintained. The proposed modification will increase the heat load in the SFP, The evaluation shows that the existing SFP cooling system will maintain the bulk pool water temperature at or below 140'F. Ttus, it is demonstrated that the peak value of the pool bulk temperature is considerably lower than the
+
pool bulk boiling temperature at atmospheric conditions. The evaluation also shows that maximum local water tenperatures along the hottest fuel 4
assembly are below the nucleate boiling condition value. Thus, there is
< no reduction in the margin of safety for thermal-hydraulic or spent fuel
, cooling concerns, as defined in the USAR.
- c. Mechanical. material and structural consiaerations The finite element method was used to evaluate the margins of the spent fuel pool concrete structure. The analysis is in accordance with Standard Review Plan (SRP) Section 3.8.4. The evaluation demonstrates that the mar maintained. gin of safety of the fuel pool structural strength is The main safety function of the spent fuel racks is to maintain the s)ent fuel assemblies in a safe configuration through all normal or a) normal loadings. Abnormal loadings which have been considered are the effect of an earthquake and the impact due to the drop of a spent fuel assembly. The mechanical, material, and structural design of the new spent fuel racks is in accordance with applicable portions of "NRC OT
, Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, as modified January 18, 1979; and other applicable NRC guidance and industry codes. The rack materials used are compatible with the SFP and the spent fuel assemblies. The structural considerations of the new racks address margins of safety against tilting and deflection or movement, such that the racks do not impact each other during the postulated seismic events. In addition, the spent fuel assemblies remain intact and no criticality concerns exist. Thus, the margins of safety as defined in the USAR are not reduced by the proposed rerack.
TECHNICAL SPECIFICATION CHANGES In order to implement the proposed changes, it is proposed that the following Technical Specifications be revised:
, Specification 2.8 is being revised to incorporate a change to the fuel enrichment allowed to be stored in the SFP and to place additional restrictions on SFP boron concentration whenever unirradiated fuel is within the SFP.
Specification 3.2, Item 15 on Table 3-5, is being revised to incorporate necessary changes to the test material surveillance program.
Specification 3.2, Item 10 on Table 3-5, is being revised to correct a misspelled word. The word "fuseable" is misspelled and is being corrected to " fusible."
Specification 4.4 is being revised to incorporate the design
, features of the new SFP racks.
Specification 5.10.3 is being revised to correct references. The reference to Technical Specifications 2.8(12) and 4.8.4 are incorrect, and are being corrected to Specifications 2.8 and 4.4 respectively. It is also proposed to replace the word " bundle" with " assembly" to be consistent throughout the Specifications.
5
- . , e -
4 1
4 l
1 NO SIGNIFICANT HAZARDS EVALUATION >
j' The proposed changes do not involve significant hazards considerations because i
i o)eration of the Fort Calhoun Station Unit No. 1 in accordance with-these c1anges would not:
1.
j (1) Involve a significant' increase in,the probability or consequences of an -
accident previously evaluated. -
i In the course of the analysis, the following potential accident scenarios have been considered:
F a. A spent fuel assembly drop in the Spent fuel Pool (SFP)
! b. Loss of SFP cooling system flow i c. A seismic event
! d. A spent fuel cask drop
! e. A construction accident 1
Probability of an accident
, The increased storage capacity of the SFP has been analyzed for the
! existing fuel handling equipment and procedures, SFP cooling system, and i seismic events. No cask movement is centemplated as part of this-i modi fication. The Auxiliary Buildir.g crane will be used to bring the i racks into the Auxiliary Building. Thus, the proposed modification does j not increase the probability of any of the first four accidents.
i
- With regard to the construction accident, the Technical Specifications i prohibit loads heavier than the weight of a. single spent fuel assembly l plus CEA, and the; tool' for moving that assembly,. from being carried over i fuel stored in the SFP. All work in the SFP area will be controlled and
! performed.in strict accordance with specific written procedures and.
i administrative controls to preclude the movement of a rack directly over-
!- any fuel. Therefore, the probability of occurrence of-a construction l- accident is not significantly increased as a result of the proposed i reracking.
l- In addition, Sections 5.1.1, 5.1.2-and 5.1.6 of-NUREG-0612 entitled ,
! " Control of Heavy Loads.at Nuclear Power Plants", provide guidance for j heavy load handling operations pursuant to a. spent fuel storage rack replacement. Section 5.1.2 provides four alternatives- for assuring the
!: safe handlin of heavy loads during a fuel storage rack replacement.
!. Alternative 1) of Section 5.1.2 provides that the control of heavy
! loads guidel nes can be satisfied by establishing that the potential for.- ,
i a heavy load drop is' extremely small. -
- .i L The Auxiliary Building overhead' electric crane with rated load capacity I of 75 tons will be used for.the reracking operation. The maximum weight r of an individual new rack-or existing rack is less than'10 tons. -The
- i. weight of the lifting fixture is less than 2 tons, which results in a:
minimum factor of safety of 6.25. In addition,-the crane system and the lift ri
' proof. g to be employed in the.rerack operation are single-failure-Therefore, the pote -
- . alternative (1) of Section- 5.1.2 of NUREG-0612 is -satisfied.
L Accordingly .the proposed modification does not involve a significant-L h 6 o
l:
t
' .. --..;.~ .,-.-,_s~ ~~~..--,.-.-.r.. - 1 . i -- ., ~ n? 4, - ,. . , , , n s. .a. . , , n ,,, _ , ,, 4,-, m .,, ,e ,6, i& 4 n
L l
increase in the probabilit i guidelines for defense in y of a load drop accident since NUREG 0612 depth to preve satisfied.
Potential consecuences of an accident l The consequences of a s)ent fuel assembly drop in the SFP were evaluated and it was found that tie criticality acceptance criterion, k,,, s 0.95, is not violated. In addition it was found that there was no significantchangeintheradlolo drop from the previous analyses. gical consequences of a fuel assemblyThe analyse doses are well within 10 CFR 100 guidelines. The results of an ,
analysis show that a dro) ped spent fuel assembly on the racks will not :
distort the racks such t1at they would not perform their safety function. Thus,.the consequences of this type of accident are not significantly changed from the previously evaluated spent fuel assembly drops.
The consequences of a loss of SFP cooling system flow have been evaluated and it was found that sufficient time is available to provide an alternate means for cooling in the event of a complete failure of the ,
cooling system. Thus, the consequences of this type accident are net significantly increased from previously evaluated loss of-cooling system flow accidents.
The consequences of a seismic event have been evaluated. The new rack:,
will be designed and fabricated to meet the requirements of applicable -
portions of the NRC Regulatory Guides and )ublished standards. The new so that the free-standing integrity of theracks racks are anddesigned, the pool structure as are tie is existing maintained racks,during and after a design basis reismic event. Thus, the consequences of a seismic event are not increased from previously evaluated events.
The consequences of a spent fuel cask drop will not be affected by the replacement of the r d s, as no saent fuel cask movement is contemplated as part of this modification. OP)0 does not presently utilize or own a spent fuel sample cask or spent fuel shipping cask. If such casks were to be utilized the single-failure-proof main hook of the Auxiliary Building Crane would be used to move these casks.- In addition, electrical interlocks are installed which hook over the spent fuel pool (NVREG-0612) The .
preclude likelihood travel of of the main occurrence of a cask drop is greatly reduced through-these considerations. Heavy Load paths have been proposed to circumvent carrying a cask over spent fuel. Thus the need for consideration of a radiological release impact of a cask drop accident is obviated. OPfD concludes that an analysis of the rad %1ogical consequences of a cask drop accident is not required.
The consequences of a construction accident have l+en considered. A heavy load will not be carried in the SFP area until all . fuel in the pool has decayed for minimum of two months. This provides. sufficient time for decay of gawous radionuclides in the fuel (gap activity) such that an assumed accidental release of gases .from famage to all stored fuel assemblies would result in a potential off-site dose less than 10%
of 10 CFR 100 ;imits. In 2ddition, load paths were evaluated in accordance with the Criteria contained in NUREG-0612 and found to be 7
_ .._. u .__ _ _ -., . ._ -__ .m, . . _ . . . _ _ . _ _ _ _ . _ _
acceptable. It is concluded, therefore, that the consequences of .
construction accident are not significantly increased from previously evaluated events.
d Therefore it is concluded that the proposed amendment to replace the
, s)entfuelracksintheSFPdoesnotinvolveasignificantincreasein tie probability or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of accident from any accident previously evaluated.
The )roposed modification was evaluated in accordance with the guidance of tie NRC Position Paper entitled, "0T Position for Review and Acceptant.e of Spent Fuel Storage and Handling Applications", appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and
, appropriate industry codes and standards. In addition, several previous NRC Safety Evaluation Reports for rerack applications similar to thir
- proposed modification have been reviewed.
4 No unproven technology will be utilized either in the construction process or in the analytical techniques necessary to justify the planned
- fuel storage expansion. In fact, the basic reracking technology in this instance has been developed and demonstrated in over 80 applications for 4
fuel pool capacity increases previously approved by the NRC.
Based upon the. foregoing, it is concluded that the proposed reracking does not create the possibility of a new or different type accident from any accident previously evaluated.
4 (3) Involve a significant reduction in a margin of safety.
Nuclear criticality considerationi The established acceptance criterion for criticality 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. This margin of safety has been adhered to in the criticality analysis methods for the new rack design.
The methods used in the criticality analysis conformed to the applicable portions of the appropriate NRC guidance and industry codes, standards, and specifications, as listed in the Licensing Report. In meeting the acceptance criteria for criticality in the SFP, such that k,,, is always less than 0.95, including uncertainties at a 95%/95% probability /
confidence level, the proposed amendment does not involve a significant reduction in the margin of safety for nuclear criticality.
Thermal-hydraulic considerations Conservative methods were used to calculate the maximum fuel cladding temperature and the increase in temperature of the water in the SFP.
The thermal-hydraulic evaluation used the methods previously employed for evaluations of the present spent fuel racks to demonstrate that the
' margins of safety for pool temperatures are maintained. The proposed modification will increase the heat load in the SFP. The evaluation shows that the existing SFP system will maintain the bulk pool water 8
- temperature below 1400F. Thus a margin of safety exists such that the
, maximum allowable temperature for bulk boiling is not exceeded for the calculated increase in pool heat load. The evaluation also shows that maximum local water temperature along the hottest fuel assembly is below the nucleate boiling condition value. Thus, there is no significant reduction in the margin of safety for thermal-hydraulic or spent fuel cooling concerns.
Mechanical. material and structural considerations The main safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through all normal or abnormal loadings. Abnormal loadings which have been considered are the effect of an earthquake, the impact due to a spent fuel cask drop, the drop of a spent fuel assembly, or the drop of any
- object used in the rerack modification. The mechanical, material, and structural design of the new spent fuel racks is in accordance with applicable portions of "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications", dated April 14, 1978, as modified January 18, 1979; Standard Review Plan Section 3.8.4; and other applicable NRC guidance and industry codes. The rack materials used are compatible with the SFP and the spent fuel assemblies. The structural considerations of the new racks address margins of safety against tilting and deflection or movement, such that the racks do not impact each other in the cellular region during the postulated seismic events.
In addition, the spent fuel assemblies remain intact and no criticality concerns exist. Thus, the margins of safety are not significantly reduced by the proposed rerack.
Additionally, the proposed amentiment of f " Amendments That Are Considered Not most closely likely to resembles Involve Significantexample Hazar (X)ds Considerations" as rovided in the final NRC adoption of 10 CFR 50.92, 51 FR
. This example indicates that an amendment is not likely 7751 to inv(March olve a 6, 1986cant signif hazards consideration as follows:
(X) An expansion of the storage capacity of a spent fuel pool when all of the following are satisfied:
- 1. The storage expansion method consists of either replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies or placing additional racks of the original design on the pool floor if space permits.
The FCS spent fuel pool rerack involves the replacement of the present capacity racks with a design which, by requiring burned fuel or low enrichment fresh fuel be stored in Region 2, allows closer spacing of the stored spent fuel cells. Region 1 is designed for allowing safe storage of fuel enriched to 4.2 w/o U-235.
9
I i )
- 2. The storage expansion method does not involve rod
- consolidation or double tiering.
. The FCS racks are not double tiered and all racks will sit on the spent fuel pool floor. Additionally, the amendment application does not involve consolidation of spent fuel.
- 3. The k,,, of the pool is maintained less than or equal to 0.95.
The design of the spent fuel racks contains a neutron absorber, Boral, to ensure that the k.,, remains less than
! 0.95 under all conditions (with unborated water in the pool). Additionally, the water in the spent fuel pool is a maintained at refueling boron concentration, providing i further assurance that k,,, remains less than 0.95. The i analysis demonstrates that only 80 ppm boron is required to l meet the reactivity requirement for the accident condition.
Therefore the minimum boron concentration requirement for theFortdalhounpoolcanbereducedto100ppmwithalarge
! margin of safety.
! 4. No new technology or unproven technology is utilized in l either the construction process or the analytical techniques necessary to justify the expansion.
The rack design has been licensed at least 15 times. The technology for the construction processes and analytical i
techniques remain substantially the same as those in more than 15 other storage rack projects. Thus, no new or unproven technology is utilized in the construction or analysis of the proposed FCS spent fuel racks.
This submittal meets example (X) as presented in the supplementary
} information accompanying publication of the Final Rule as an example of situations which are considered not to involve significant hazards 4
considerations. Finally, it is recognized that the only external effect
- of the proposed change is the increase of less than 0.01% in the total quantity of heat rejected to the anvironment.
Therefore based on the above considerations, it is OPPD's position that this proposed amendment does not involve significant hazards considerations as defined by 10 CFR 50.92 and the proposed changes will not result in a condition which significantly alters the impact of the Station on the i environment. Thus the proposed changes meet the eli categorical exclusion set forth in 10 CFR 51.22(e)(9)gibility criteria forand pursuant to 10 C 51.22(b) no environmental assessment need be prepared.
4 10
U.S. Nuclear Regulatory Comission LIC-92-0340A ATTACHMENT C
A CONPARISON OF THE OPPD LICENSING SU8MITTAL TO THE NRC'S OT POSITION FOR REVIEW AND ACCEPTANCE OF SPENT FUEL POOL STORAGE AND HANDLING APPLICATIONS (APRIL 14, 1978 REVISION JANUARY 18,1979)
, OPPD Licensing i Set'- Submittal Ref.
- U tsition Description Section 11 Overall Description 2.0 111 Nuclear and Thermal Hydraulic 4.0, 5.0 Considerations 111.1 Neutron Multiplication Considerations 4.1
!!!.1.1 Normal Storage 4.1 1.1.a 4.2.1, 4.3.1 !
1.1.b 4.1 1.1.c 4.1 1.1.d 4.5, 4.6 1.1.e 4.1, 4-2.1 .
1.2 Postulated Accidents 4.8
- 1. 2. l;1 ' (a) 4.8.3, 7.2 1.2.1 1 (b) 4.8.5 1.2. J2 9.1.2 1.2. f3 4.8.4 1.2.d4 4.8.1, Table 4.8.1 1.3 Calculation Methods 4.4,4.5,4,6 1.4 Rack Modification 1.4. p Table 4.3.1,4.5,4.6 1.4. I 4.5,4,5.2.4 1.4. h 4.5 1.4. . il 4.3.2, 4.3.3, 4.5 1.4.d. f2 a 4.5.2.5, 4.6.3.5 1.4.d.il2 b 4.5.2.3, 4.6.3.3 1.4.d.(2 c - -
- 4.5.2.1, 4.6.3.1 1.5 Acceptance Criteria for Criticality 4.0 1.5.(1) Neutron Absorber Verification 10.0 1.5.(2) Decay Heat Calculation for the 5.3 S)ent Fuel 1.5.(3) Tiermal Hydraulic Analysis of Spent 5.2, 5.4-5.8 Fuel Cooling 1.5.(4) Potential Fuel and Rack Handling 2.5, 7.2.1, 7.2.2 Accidents 1.5.(5) Technical Specification T.S. 2.8, 4.4.2 1
1 1
.- ~ , , .
I OPPD Licensing Section Submittal Ref.
OT Position Description Section IV Mechanical Material ani Structure 3.2, 3.3 Considerations (1) Description of Spent Fuel Pool and 7.1, 7.2, 8.1, 8.2 Racks
& Support of Spent fuel Racks 3.2, 6.2 b Fuel Handling 7.1, 7.2 Applicable Codes, Standards and 3.4 Specifications (3) Seismic and Impact loads 6.3 (4) Loads, Loads Combination and 6.5, 7.4 Structural Acceptance Criteria Design and Analysis Procedures 6.4, 6.6 Structural Acceptance Criteria 6.0, 8.5 Materials, Quality Control and 2.4, 2.5, 3.0, 10.0 Special Construction Techniques (8) Testing and Inservice Inspection 10.0 V Cost / Benefit Assessment 11.0 V. 1 Environmental Assessment 1.1 Special Needs for Increased Storage Capacity 1.0 1.1. ) Reprocessing 11.1 1.1. p Discharge Schedule 1.0, Table 1.1.1 1.1. g fuel Assemblies Stored 4.7, Table 1.1.lB 1.1. I Control Rod Assemblies 46 CEA's 1.1. [ Storage Capacity 1.0 1.1. J Full Core Discharge Capacity 1.0 1.2 Construction Costs 11.2 1.3 Alternatives to increased Storage 11.0 Capacity 1.3.(a) Reprocessing Facilities 11.0 1.3.(b) Independent Spent Fuel 11.0 Storage Facility 1.3 (c) Away from Reactor 11.1 1.3.(d) Replacement Power it is more cost effective to rerack i the pool than to shutdown Fort Calhoun Station.
2 i
l
}
i l
j OPPD Licensing i Section Submittal Ref.
! OT Position Description Section i
I 1.4 Resources The expansion of the i spent fuel pool j capacity is expected
, to require less than 4
.0001% of total
+
world output of stainless steel and
- Boron Carbide.
' Experience has shown j that the production 4
of Boron Carbide is
! highly variable and i depends upon need i and can easily be
, expanded to i accommodate
, worldwide needs.
I J 1.5 Maximum Water Temperature 5.8, 11.3
- of Spent fuel Pool a
V. 2 Radiological Evaluation i 2.1 Radioactive Wastes 9.2
! Annual Report
' Radionuclides Table 9.4.1 2.3.(a)
!- 2.3.(b) Dose Rates 9.5, OPPD Radiation 4 -
Protection Program l 2.3.(c) Airborne Concentrations 9.4, and Measured in 4 accordance with OPPD
! Radiation Monitoring ,
j Program.
1 2.3.(d) Increased in Airborne Releases 9.4 To be Monitored
} -in Accordance with
- OPPD Radiation ,
Monitoring Program.
2.3. Rad Waste 9.4 2.3. Dose Rates - Increase 9.4 i 2.3. Emission Directly from fuel 9.5 2.4 Rack Decontamination _ 9.2, excluding wts.
- V. 3- Accident Evaluation: **9.1.2, 2.5 4
3.1.(a) Cask Drop **9.1.2-3 l
t i
! OPPD Licensing Section Submittal Ref.
i OT Position Description Section i
j 3.1.(b) Overhead Cranes, R.G. 1.104 Overhead system meets NUREG-0612 1 3.2 Cask Movement 9.1.2 ,
! a) Accident Aspects of Review 9.1.2 3.3.l()[b) 3.3.11 3.3.(2 )i, Accident Aspects of Review Accident Aspects of Review 9.1.2
} 9.1.2 3.4 Cask Drop Analysis **9.1.2 .
! 3.5 Maximum Weight of Loads 7.1, 7.2 ;
i 3.6 Radiological Consequences of i Cask Drop **9.1.2 ,
j
- Table 1 '
j Existing Fort Calhoun Station $FP Racks to be Removed Module Size Capacity Quantity Est, wt. (1bs): Total. Cavities l
i >
! 8x9 72 1 15,800 72 i
!. 7x9 63 7 13,825 441
. 6x9 54 4 11,850 216 Totals 12 159,975 729
- The licensee does not presently utilize or own a spent fuel sample cask or spent fuel shipping cask. If such casks were to be utilized, the single failure-proof main hook of the Auxiliary Building Crane would be used to-
. transport these casks. This characteristic, in conjunction with the presence i of electrical interlocks which preclude travel of the main hook and auxiliary i hook over the spent ~ fuel pool, would greatly reduce the likelihood of i occurrence of a cask drop, obviating the need for consideration of 1 radiolo ical release impact of such an accident. Therefore, we conclude that i' an anal sis of the radiological consequences of a cask drop accident is.not require .
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