ML20049J872
| ML20049J872 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/31/1982 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20049J871 | List: |
| References | |
| TAC-47523, NUDOCS 8203290176 | |
| Download: ML20049J872 (47) | |
Text
{{#Wiki_filter:o JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2 DOCKET NO. 50-364 SPENT FUEL POOL MODIFICATION December 1981 8203290176 820319 AMEND 2 3/82 PDR ADOCK 05000364 P PDR
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TABLE OF CONTENTS ) I. Introduction and Conclusions II. Overall Description
-III. Nuclear and Thermal-Hydraulic Considerations
- 1. Neutron Multiplication Factor 1.1 Normal Storage 1.2 Postulated Accidents 1.3 Calculation Methods 1.4 Rack Modification 1.5 Acceptance Criteria for Criticality 5
(1) Neutron Absorber Verification (2) Decay Heat Calculations for the Spent Fuel (3) Thermal-Hydraulic Analyses for Spent Fuel Cooling (4) Potential Fuel and Rack Handling Accidents (5) Technical Specifications IV. Mechanical, Material and Structural Considerations (1) Description of the Spent Fuel Pool and Racks (a) Support of Spent Fuel Racks (b) Fuel Handling (2) Applicable Codes, Standards and Specifications (3) Seismic and Impact Loads (4) Loads, Load Combinations, and Structural Acceptance Criteria (5) Design and Analyses Procedures (6) Structural Acceptance Criteria (7) Materials, Quality Control and Special Construction Techniques (8) Testing and Inservice Surveillance V.l. Cost / Benefit Assessment V.2. Radiological Evaluation V.3. Accident Evaluation Q, Response to NRC Questions AMEND 2 3/812
6 buoyant head greater than the pressure loss, the flow rate 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 flow rate, HPOOL determines the j fuel temperature. l ' HPOOL was used to determine the bulk spent fuel pool water temperature for the normal refueling ddse and the emergency full core offload 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 assumptions and results for these two cases are presented below. A. Normal Refueling:
! Assumptions:
! 1. All storage cells filled.
- 2. 26 years - 1/3 of the core removed each year.
" 27th 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 i Trains Operational Shutdown (108 Btu /hr) Temp. ( F) )i 1 100 19.707 139 ! 2 100 19.707 122 1 150 17.532 135
; 2 150 17.532 120 l
B. Full Core Offload: { Assumptions:
- 1. All storage cells filled i 2. 23 years - 1/3 of core removed each year.
4 24th year - 0.40 of core removed. Emergency full core offload is necessary 180 hours after j normal refueling in 24th year. Full core is III-6 AMEND 2 3/82
, b 1
discharged to the spent fuel pool 10 days
"( after the emergency shutdown.
- 3. 105'F component cooling water temperature. l l
! Results: s s No. of Total Spent Fuel Cooling Trains Heat Load Pool Bulk
. " Operation (105 Btu /hr) Temoerature (*F) 1 30.332 158 2 30.332 _31 The computer code BPOOL is used to analyze the natural 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 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 the racks flows downward to the inlet of the storage racks. The I 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
- 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.
Voiding in the space between fuel assemblies is not possible since these spaces contain poison plates. j III.1.5.(4) Potential Fuel and Rack Handling Accidents The high-density poison racks are of a free-star. ding 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 clean and uncontaminated 13' inch center storage racks. The high-density racks will be installed dry since there is no fuel in the storage pool. The following is a sequence of events for installing the high- ' density poison racks. 4 ( Install and test a temporary crane for handling the Phase I existing racks and the high-density racks. The III-7 AMEND 2 3/82
e 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 a portion of the existing 13-inch center racks, leaving enough racks intact for one emergency full-core offload into the spent fuel pool. Phase III -Remove interferences between new racks and existing floor studs by removing a portion of each stud, where required. 1 Phase IV Install the high-density poison racks into the pool areas left vacant by the removal of the 13-inch center racks. This work may be done gradually over a period of time due to delivery schedules of the high-density racks. Storage capability for one emergency full-core offload will be available at all I times during the spent fuel pool re-rack program. Phase V When one full-core offload storage capacity is I achieved with new racks, remove any remaining 13-inch center racks from the spent fuel pool, g Phase VI Install the balance of high-density racks into the
- spent fuel pool to complete rack installation.
Phase VII Remove temporary crane from the spent fuel pool area. i These phases of work will require support work (i.e. leveling of new racks, testing, etc.) to complete the re-racking program. t The outdoor spent fuel cask crane will be used to bring the high-density poison racks from the delivery vehicle into the s spent fuel cask area. The racks will then be moved from the spent fuel cask area, by the temporary crane, into the spent
- l fuel pool. The reverse sequence will be performed to remove the existing 13-inch center storage racks from the spent fuel pool.'
The installation of the high-density noison racks will not increase the potential for a fuel and rack handling accident for the following reasons: o The spent fuel pool is dry and does not contain any ' spent fuel, o The temporary crane, as with the spent fuel bridge crane, can carry loads over the spent fuel cask area III-8 AMEND 2 3/82
and the spent fuel pool only. There is not any 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.
, o 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.l.2.(2).
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. In addition, the racks can follow the cask load path into and out of the cask area. By following this path, the racks will not pass over any safety-related equipment except the spent fuel pool cooling system. Since no fuel is presently stored in the spent fuel pool, the spent fuel pool cooling system is not presently required for a safe shutdown of the plant. III.1.5.(5) Technical Specifications 1 To insure against criticality, the following technical specifications are proposed in figure III-1 on spent fuel storage in the high-density poison racks. f I III.1.5 (5).1 Paragraph 5.6.1.1 cf the proposed revision to the Earley Unit 2 Technical Specifications requires that the i spent fuel storage racks be designed and maintained such that 4 3 the neutron multiplication factor (kegg) in the fuel pool shall be less than or equal to 0.95 when flooded with unborated l water. This represents the most conservative pool condition from a criticality standpoint. III.1.5.(5).2 In addition, paragraph 5.6.1.i of the proposed revision to the Farley Unit 2 Technical Specifications also specifies a naximum enrichment of 4.3 weight percent U-235 l (which equates to 54.25 grams per axial centimeter of the fuel assembly) for fuel loading in the fuel assemblies. This limit i is consistent with the design of the high-density poison racks to preclude criticality in the fuel pool. I i III-9 AMEND 2 3/82
i IV.(2) Applicable Codes, Standards, and Specifications ' The spent fuel storage racks are a welded structure consisting of materials of-U.S. origin. The following chart presents material specifications and alloys used in the rack assembly. , Material Spec. Description Alloy ASTM-A240 Bottom 304SS Grid i l ASTM-A240 Outer Wrapper 304SS ASTM-A666 Inner 304SS Gr. B Tube ASTM-A564 Threaded. Foot Type 630 H-1100 ASTM-SA351 Bottom Grid Foot CF-8C The design of the spent fuel storage rack is as in Section 5 of the AISC Steel Construction Manual with fabrication welding as in ASME Section IX. The racks were also designed and fabricated to meet and utilise 1 the applicable portions of the following regulatory guides, safety review plan sections, published standards, and computer programs.
- 1. United States Nuclear Regulatory Commission (USNRC) 1
- a. Reg. Guide 1.13 Spent Fuel Storage Facility
) Design Basis, Rev. 1, Dec. 1975. 4
- b. Reg. Guide 1.29 Seismic Design Class, Rev. 2, Feb. 1976.
t
- c. Reg. Guide 1.92 Combination of Modes in Seismic Analysis, Rev. 1, Feb.
1976.
- d. Reg. Guide 1.38 "QA Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water j
Cooled Nuclear Power' Plants", Rev. 2, 1977. i
- e. Reg. Guide 1.60 Design Response Spectra for Seismic Deisgn of Nuclear Power Plants, Rev. 1, Dec.
1973. IV-3 AMEND 2 3/82
.-r_ , _- , _ _ , . , , , , . - - - . , - - . , - , , . . . , _ . . _ . _ . , , , . _ _ _ . - _ _ , _ , _ , , , _ . , _ . . , _ , _ - . . . , , _ . . . _ _ , . - _,
i
- f. Reg. Guide 1.61 Damping Values for Seismic Design of Nuclear Poser Plants, Oct. 1973.
- g. Reg. Guide 1.31 Control of Ferrite Content in Stainless Steel Weld Metal, Rev. 3, April 1978.
- h. Reg. Guide 1.124 Service Limits and Loading ,
Combinations for Class 1 !
' Linear - Type Component }'
Supports, Rev. 1, Jan. 1978. i
- 1. SRP 3.7 Seismic Design, 1975.
I
- j. SRP 3.8.4 Seismic Category I Structures, ;
1975. i
- k. SRP 9.1.2 dpent Fuel Storage, 1975.
~'
- 1. NRC Guidance on Spent Fuel Fool Modifications, Review and Acceptance of Spent Fuel Storage and Handling Applications (April 14, 1978 revised Jan.
i 18, 1979).
- 2. Industry Codes and Standards
- a. ASME Boiler & Pressure Vessel Code Section IX and Section III, ,
Appendix I, XVII, and Article , NF-4000, 1980 Edition (American Society of Mechanical Engrs.). *
- b. AISC Steel Construction Manual AISC 8th Edition, Dec. 1980 e (American Institute of Steel '
Construction). (
- c. ACI 318-71 Building Code Requirements for l Reinforced Concrete. '
(American Concrete Institute.) ;
- d. ASTM ASTM Standards: A240, A666, A564.
- e. ANSI N45.2 " Quality Assurance Program Requirements for Nuclear Facilities", 1977.
- f. ANSI N45.2.2 " Packaging and Shipping, ;
Receiving Storage and Handling of items for Nuclear Power IV-4 AMEND 2 3/82
Condition 5 - gate drop from 9 in. above the ( rack impactin'g on the top of the rack. B. Load Combinatiens The following load combinations shall be satisifed: Stress Limit Load Combinations Fs
- 1. D+L+T+P Fs
- 2. D+L+T+H Es
- 3. D+L+T+E
- 4. D+L+T+I Condition 1 1.6 Fs (1) .
.' Condition 2 1.6 Fs (1) Condition 3 1.6 Fs (1) 1.6 Fs (1) Condition 4 ( Condition S 1.6 Fs (1)
- 5. D+L+Ti + E2 1.6 Es NOTES (1) Local failure of the fuel support or the top grid However, overall impact interface is allowed.
member stresses shall be limted to 1.60 Fs and resulting rack deformation shall not cause the fuel configuration to reach a keff of 0.95. C. Design Allowable Stresses
= Allowable working stress Fs fs = Calculated stress Fy = Yield stress l !
D. Allowable Stresses (For Stainless) l The allowable stresses are in accordance with ASME Boiler and Pressure Vessel Code Section III Appendix XVII. This is interpreted as being identical to the
( AISC Steel Construction Manual (Section 5). AMEND 2 3/82 IV-8
_ _ - -_ - . ~ 6 , l l V.3.3.(1).(a) The accident aspects of review establish acceptability with 1 respect to subsection V.3.1.(b). Also, the design evaluation of the spent fuel cask crane, as discussed in Farley FSAR subsection 9.1.4.3.2, will not be affected as a result of the re-rack program. V.3.3.(1).(b) See response to V.3.3.(1).(a). V.3.3.(2) 4 See response to V.3.3.(1).(a).
- . V.3.4 The cask drop /tip analysis is provided in subsection 3.1.(a).
V.3.5 The maximum weight of loads which may be transported over the spent fuel is not substantially in excess of a fuel assembly. Farley Unit 2 Technical Specification Section 3.9.7.1 limits 5 the size of the load that can be handled over spent fuel to 3,000 pounds. V.3.6 Since the spent fuel cask will not be handled over or into the vicinity of the spent fuel pool, the proposed modification does not significantly change or impact any previous determinations of Farley Unit 2 Safety Evaluation Reports. Since there will be a negligible change in radiological i
. conditions due to the increased storage capacity of the spent fuel pool, no change is anticipated in the radiation protection l program. In addition, the environmental consequences of a postulated fuel handling accident _in the spent fuel pool, described in FSAR section 15.4.5, remain unchanged. Therefore, there will be no change or impact to any previous determinations of Farley Unit 2 Final Environmental Statements.
I V-10' AMEND 2 3/82 , I l
- ... - . - , -- = - . - -
RESPONSE TO NRC QUESTIONS ENCLOSURE-1 OF NRC LETTER DATED MARCH 4, 1982 4 NRC Question 220.1: Damping values do not appear to be in accordance with i Regulatory Guide 1.61. Provide justification for the damping values used in the analysis. APC Response: The design response spectra for seismic design Category 1 structures are discussed in the Farley FSAR under design response spectra (section 3.7.1.1) and design response spectra deviation (section 3.7.1.2). The damping values for seismic design for Category 1 structures are discussed in section 3.7.1.3, critical damping values. These values were reviewed and approved by the NRC staff during the operating license review. In addition, the NRC position paper states in paragraph three of Section IV that it is acceptable to use different damping values from Regulatory Guide 1.61 if the response spectrum curves were developed for the particular plant and not from Regulatory Guide 1.60; therefore, the 2 percent and 5 percent values are acceptable. NRC Question 220.2: ASTM A-666 material is not found in the ASME Code. The staff i has previously accepted the use of this material for spent fuel i storage racks under the following conditions: (1) The applicant agrees to qualify the rack material in question to the ASME Code Subsection NE (SA240 material) in all respects and, in addition, to obtain valid test results to justify the higher yield stress allowed by ASTM A-666, Grade B. (2) Allowable weld and base material stresses in all heat-affected-cones are based on the yield strength for SA240 material. (3) Tensile tests indicate that the yield strength of the material used is not greater than 90 ksi. (4) The applicant can provide objective evidence that stress corrosion cracking of both base metal and weldment will not occur. Citing of previous experience would be an acceptable approach. Q-1 AMEND 2 3/82
i j t ' (5) Complete documentation of the applicant's compliance with the above is developed. I The applicant is requested to address compliance with the above requirement or submit an alternate proposal. APC Response: !- (1) The ASTM-A666 material is equivalently qualified to the ASME Code, Subsection NF (SA240 material) with I the exception of the higher yield stress (Grade B) which is documented by certified material test
; reports.
l 3 (2) The allowable stresses in the heat affected =ones of 1 the weld and base material, utilising ASTM-A666 in storage rack construction, are higher than the allowable stresses of ASME-SA240 steel. Neither the i ASME Code or the ASTM Code requires an assumed
- reduction in yield strength in heat affected zones of i~
the weld and base material. Tests nave been
! performed by an independent testing laboratory on samples of welded A666, Grade B material. These i
tests indicate an average yield stress to be 17.9 percent higher than the minimum yield stress ' specified by ASTM for unwelded A666 Grade B material and 76.9 percent higher than the minimum yield stress specified for SA240 material. These welded samples i i failed in the heat affected zone. A copy of the test results is attached. Further,A666 material is ] currently in use in a comparable spent fuel rack j d
?.esign. This design was licensed by the NRC and has
} design stresses in the heat affected zones based upon
! the ASTM Code allowable stresses for the A666 1
-material. The design stresses for the proposed Farley Nuclear. Plant Unit 2 racks are comparable to those currently licensed.
(3) The material certification records document the yield strength of the material to be less than 90,000 psi, j (4) In this particular design the stress corrosion cracking problem is not a concern due to several
, factors. Stress corrosion cracking may occur where extensive intergranular precipitation is found in i combination with tensile stresses. While some residual stresses cannot be avoided in this design, j
the amount of heat input from welding is small and the rate of cooling is rapid. Low heat input, a maximum interpass temperature of 350 F for all welds, and rapid cooling due to the configuration of intergranular precipitation occurs when a high tempe.rature is held for a long period of time; i.e., i n - g_2 AMEND 2 3/82
~
_.M' y_.- ,_ , __m-,--__-,_.7-_.- -- 9-__. ---.-_.,-.-,-e . y ~, ,4 -,4.--. - - _ , - , , , , ,, , y y,=_ =
a typical sensitizing treatment would be 1200 F for 2 hours; which will not be the case. In addition to the above factors all welds will be made with 30SL filler metal and the heaviest weldment (the foot corner casting) will have a 0.03-percent maximum carbon content. These factors also greatly reduce the level of susceptibility to stress corrosion. par has manufactured many spent fuel storage modules and other equipment of welded Type 304 stainless steel for nuclear applications with no indication of stress corrosion cracking problems. (5) Alabama Power Company will document all of the above as part of the normal design process. This documentation will be included in the final design package and will be available onsite for inspection. NRC Question 220.3: Provide details and numerical results of the analysis of loads on the pool liner and the pool structure including the seismic analysis and drop accident. Indicate how the liner will be able to withstand a heavy drop accident. Provide construction drawings of the pool structure and liner. APC Response: The maximum overall floor loading associated with a seismic event (D + L + T1 +E2) is 3731 psf with the allowable being 3800 psf. The maximum local foot load of 164,3000 pounds results in a bearing stress fo 4949 psi with the allowable stress being 5950 psi. The overall pool floor loading associated with a fuel assembly drop (worst case drop accident) onto the storage racks (D + L + T + 1) is less than the loading associated with a seismic event. The drop accident associated with a fuel assembly is discussed in Section IV of the spent fuel pool modification report. The Farley Plant Technical Specifications prohibit the spent fuel pool bridge crane from transporting any loads greater than 3000 pounds over the spent fuel pool. Therefore, no heavy loads will be moved over the pool. The heaviest load that the bridge crane can carry over the pool is a fuel assembly or spent fuel pool gate.. Due to the physical limitations between the racks and wall it is impossible to drop a fuel assembly onto the pool floor. With the new high density storage racks itliner. is improbable that the gate will drop freely onto the pool This improbability is due to the thickness of the gate, the extension of the gate hangers (hooks), the close proximity i Q-3 AMEND 2 3/82 l
l j i i 1 of the new racks to the pool wall and drag afforded by the water in the pool. I i Construction drawings of the spent fuel pool structure and liner are shown in figures Q-1 through Q-13. 3 NRC Question 220.4: l' ! ASTM-A793 and ASTM-A276 material are not found in the ASME
- code. Justify the use of these materials including a
{ discussion of deviations from the material acceptance criteria j of the ASME code. l APC Response: ) j ASTM-A793 has been replaced by ASME-SA351 in Amendment 1 to the Farley Unit 2 Spent Fuel Pool Modifcation Report. ASTM-A276 was included as an optional specification which will i not be used. All material for the racks has been ordered at
! this date and A276 was noc purchased for any material.
- Accordingly, ASTM-A276 will be removed from the submittal.
i NRC Question 220.S: 1
- What is the stress limit for load combination 5, of paragraph IV.(4)B?
APC Response: This has been added in Amendment 1 to the Farley Unit 2 Spent Fuel Pool Modification Report and is 1.6Fs-
)
NRC Question 220.6: , Provide a description of all items which may be moved over the spent fuel assemblies. State which of these heavy objects is < the critical one during operation and which is critical during ; installation. , APC Reponse: During installation of the new spent - fuel storage racks no i spent fuel assemblies will be located in the spent fuel pool. The heaviest loads which can be carried over the spent fuel racks during operation are the fuel assembly and the spent fuel pool gate. The analysis of these drop accidents onto the 1 storage racks is discussed in Section IV of the spent fuel pool modification report. Floor loadings associated with these drop accidents are discussed in response to question 220.3. a e { g_4 AMEND 2 3/82
NRC Question 220.7: Provide the numerical results of the structural analysis of the racks for all pertinent loading conditions. Provide structural drawing' of the racks. The following table summarizes the maximum stresses of the rack components for each pertinent load combination. The following calculated seismic stresses are based on the cumulative conservations associated with bounding the OBE and DBE response spectra and using the 2 percent damping values. Figures II-4 and II-5 of Section II of Alabama Power Company's licensing submittal contain adequate structural drawings for the assembly of the spent fuel modules. 7 x 8 RACK MAXIMUM STRESS INTERACTIONS (ACTUAL / ALLOWABLE (a)
.135 .120 Load Comb. Cans Cans Grid Cruciform (b) Foot D+L+T+P -
.27 .38 .25 .32 D+L+T+Hx -
.23 .41 .26 .32 D + L + T + Hz -
.25 .40 .26 .31 (c) D+L+T+ It -
.92 .58 .50 .47 (c) D+L+T+ I: .54 .44 .46 .45 .35 (c) D+L+T+ I3 -
.27 .63 .62 .39 D+L+T+E .71 .90 .51 .28 .38 D+L+T1 + E2 .51 .65 .37 .20 .27 6 x 7 RACK MAXIMUM STRESS INTERACTIONS (ACTUAL / ALLOWABLE) (d)
.135 .120 Load Comb. Cans Cans Grid Cruciform (b) Foot D+L+T+E .66 .88 .44 .28 .21 D+L+T1 + E2 .47 .63 .32 .20 .15 NOTES:
(a.) 7 x 7 rack bounded by the 7 x 8 rack. (b.) Fuel support cruciform. (c.) I 4 & Is are bounded by these drop conditions. (d.) Static load cases are bounded by the 7 x 8 racks. NRC Question 220.8: The description of the plates which are to be attached to the pool floor in order to provide a smooth surface is not understood. How can such devices eliminate the described obstructions? Provide sketches to illustrate how the plates are to be installed and how they will interact with the racks. AMEND 2 3/82 Q-5
._. - _. . .- . . . . _ - _ = - . - _ ~ - _ - _ - _ . . _ _ _ _ - _ - - - _ - - _ . -
APC Response: 1 i The only floor obstructions that require bridging are the two weld seams. The remaining weld seams and all embedment bolts q are far enough away from the rack feet to eliminate interference. The detailed arrangement, shown in figure Q-14, reduces loads on floor weld seams for following reasons: (1) Coefficient of friction between the machined foot and 1/2-in. plate is less than the coefficient of -I friction between the 1/2-in. plate and liner plate. Therefore, the rack foot should slide on the 1/2-in.
- plate rather than the 1/2-in. plate sliding on the
- liner plate.
l 1 (2) If the 1/2-in. plate were to slide with the foot it ' would taken an incremental amount of force to exiet between the 1/2-in plate and liner plate to allow the foot to slide on the 1/2-in. plate. This incremental
- force would be insignificant and not affect the weld seam.
1 ) NRC Question 220.9: i 3 a Discuss the method used to account for the effect of sloshing Water on the fuel pool walls. I j APC Response: i i The effects of sloshing on the spent fuel pool wall were predicted based on the method of analysis presented in Chapter I i 6 AEC Publication TID 7024, " Nuclear Reactors and Earthquakes." The analysis for SSE conditions indicates a maximum water surface displacement of 11 in. and maximum. 1 convective pressure of 40 psf due to sloshing. Normal water surface level in the pool is approximately 16 in. below the top of the pool walls. Since the minimum thickness of the concrete walls for the pool is 42 in., the convective pressure on the pool walls due to sloshing will be negligible. The rack analysis accounts for the pool water by utilizing the fluid coupling element of "ANSYS" which is based on the paper by R. J. Frits, "The Effects of Liquids on the Dynamic Motions of Immersed Solids." NRC Question 220.10: Provide and justify the time history and floor response spectra I used in the analysis of the fuel rack assembly, and of the spent fuel pool as well. Describe the methods by which seismic i Q-6 AMEND 2 3/82
responses due to three components of earthquake loading were combined. APC Responce: The response spectrums used for the seismic rack anaysis were conservatively developed from an enveloped response spectrum of both the OBE (2 percent) and DBE (5 percent) curves. The compucer program "SIMOKE" was then used to generate the time histories for each of the three envelope curves (N-S, E-W, Vert.). The spent fuel pool floor response spectra are provided in Farley FSAR sections 3.5, 3.7.1.1, and 3.7.1.2 which were reviewed and approved by the NRC staff as part of the operating license review. There are no rooms below the spent fuel pool. The pool floor rests on concrete columns and earth as shown in figures IV-4 and IV-7. The pool is a part of the auxiliary building. Floor response spectra for auxiliary building elevation 121 ft-0 in. are shown in table Q-1. The computer models used in the time history analysis for the new spent fuel storage racks are three-dimensional. This model was excited simultaneously with the three orthogonal earthquakes. The resulting stresses are actual stresses with no method of combining required. The seismic analysis was performed on the spent fuel pool structure for each horizontal direction and vertical direction. The loads from the analysis for the vertical direction were combined with loads from each horizontal direction separately; i.e., loading 1 in N-S and vertical, loading 2 is E-W and vertical. This procedure is in accordance with FSAR Section 3.7.2.1 which was reviewed and approved by the NRC staff as part of the operating icense review. Q-7 AMEND 2 3/82
T_ABLE Q-1 PERCENTAGE OF CRITICAL DAMPING FACTORS 1/2 Safe Fafe-Shutdown Shutdown Earthquake Earthquake (E') 0.05 Ground 0.10 a Gro'and Type of Structure Acceleration Acceleration
. Vital piping 0.50 1.00 Welded steel plate 1.00 2.00 assemblies Welded steel frame 2.00 5.00 structures Bolted and riveted 3.00 5.00 steel Reinforced concrete 2.00 5.00 structures and equipment supports Prestressed concrete 2.00 5.00 structures Soil damping 4.00 7.00 l
t l l Q-8 AMEND 2 3/82
l 1 a'
- l
! l s NRC Question 220.11: l Provide a detailed discussion of the methods of analysis used to calculate stresses due a fuel handling uplift accident and the results of the analysis. J 1 APC Response: i A detailed finite, element model of the rack was developed. The I deadweight and live load of the fuel were acplied plus the uplift load at the center of the rack. The resulting stresses 1 are discussed in response to question 220.7. 4 NRC Question 220.12: I Provide a detailed discussion of the methods of analysis used to calculate stresses due to thermal loads and the results of the analysis.
- APC Response
The racks are freestanding with no fixed end thermal stresses resulting. The thermal concerns will be included by decreasing the material level. yield strengths for the applicable temperature The results of rack stresses consider these thermal reductions in yields in response to Question 220.7. NRC Question 220.13: - - - If venting of the " containment pocket" for the poison material is not provided, explain the method used to mitigate the structural effects of gas buildup. { APC Response: i The poison material is vented as discussed in Section II of the Alabama Power Company Spent Fuel Pool Modification report. i NRC Question 220.14: I Indicate whether fabrication and quality control of spent fuel pool racks are in conformance with subparagraph NF of ASME , code, if not, identify and justify the deviations. APC Response: l In as much as this is not a code stamp project the i certification requirements of NF4000 do not apply. This is also true for the design requirements of NF3000 and the marking requirements of NF8000. With these exceptions, GCA/ par's i I Q-9 AMEND 2 3/82
\
quality and f abrication programs meet ASME Section III Subsection NF. NRC Question 220.15: Indicate if this proposed modification conforms with the NRC position on fuel pool modifications entitled "OT position for review and acceptance of spent fuel storage and handling applications" issued on April 14, 1978, later amended on January 18, 1979, if not identify and justify the deviations. APC Response: (1) Use of ASTM material (A666) instead of ASME. See response to question 220.2 for further detail. (2) Use of utility supplied response spectra with 2-percent (OBE) and 5-percent (DBE) damping.. See response to question 220.1 for further detail. The proposed modification meets the intent of the NRC position paper, in regards to cections III and IV, except for the minor deviations noted above. Q-10 AMEND 2 3/82 ) J
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