ML19030A685

05/17/1978 Response to Request for Additional Information Increased Capacity Spent Fuel Racks

ML19030A685
Person / Time
Site:	Salem
Issue date:	05/17/1978
From:	Librizzi F Public Service Electric & Gas Co
To:	Lear G Office of Nuclear Reactor Regulation
References
Download: ML19030A685 (20)

Text

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--1 PS~G Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201/430-7000 May 17, 1978 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. George Lear, Chief Operating Reactors Branch 3 Division of Operating Reactors ...,

I,.+~

Gentlemen:

INCREASED CAPACITY SPENT FUEL RACKS NO. 1 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-272 In response to your letter of March 16, 1978, requesting additional information regarding our application to increase the spent fuel storage capacity at Salem Nuclear Generating Station, we hereby transmit the requested information as Attachment 1 to this letter.

Additional information is provided in Attachment 2 to this letter in response to your requests resulting from a meeting with members of your staff on March 30, 1978. *

• This submittal consists of 40 copies.

Should you have any questions regarding this application, please do not hesitate to contact us.

Very truly yours,

µ~

F. P. Librizzi General Manager -

Electric Production EAL:amg The Energy People 95-2001 (400M) 9-77 I

ATTACHMENT 1 Q. 1 In Section 2.0 of your November 18, 1977 submittal, you concluded that the SFP continuous filtration and demineralization process will maintain a high water quality even after the SFP modification, and therefore "there will be no increase in anticipated radiation levels inside the Fuel Handling Building". Discuss this in more detail including the following:

(a) Identify the expected principal radionuclides and their respective concentrations (uCi/cc) in the SFP water that will provide these "anticipated radiation levels".

(b) Provide the dose equivalent rates (mrem/hr) associ-ated with these radionuclide concentrations above and around the SFP.

(c) Estimate the occupational exposure, based on these anticipated dose equivalent rates, in manual man-rem due to operations associated with fuel handling, etc.,

in the SFP.

A. 1 The expected principal radionuclides in the spent fuel pool coolant are those fission products and daughters that are contained in the spent fuel which will leak or diffuse through the fuel cladding as well as corrosion and activation products contaminating the external fuel cladding. A list of these nuclides is as follows:

CR-51 Y-91M I-130 LA-140 XE-131M MN-54 Y-91 TE-131M CE-141 XE-133M FE-55 ZR-95 TE-131 CE-143 XE-133 FE-59 NB-95 I-131 PR-143 XE-135M C0-59 M0-99 TE-132 CE-144 XE-135 C0-60 TC-99M I-132 PR-144 XE-135 BR-83 RU-103 I-133 NP-239 XE-137 BR-84 RH-103M I-134 KR-83M XE-138 RB-86 RU-106 CS-134 KR-85 H-3 RB-88 TE-125M I-135 KR-85M C-14 SR-89 TE-127M CS-136 KR-87 AR-41 SR-90 TE-127 BA-137M KR-88 SR-91 TE-129 BA-140 KR-89 The concentration of the isotopes in the SFP coolant will not increase significantly as a result of increased fuel storage because (1) dissolved gases will have reach-ed an equilibrium state and (2) other nuclides will be removed by filtration and demineralization. It may be necessary to increase demineralization frequencies and resin changes to maintain the same SFP water quality.

P78 59 01

There are five area monitors located in the Fuel Handling Building. Four of these area monitors have ranges of 0.1 mR/Hr to lOR/Hr. The radiation alarm setpoint for these monitors is 15mR/HR. The high sensitivity and low alarm setting of the monitors, as well as the personnel exposure records, will assure all doses from routine activities in the SFP are kept as low as practicable.

The occupational exposures due to the gaseous release will be nil.

Section 9.11 of the Salem FSAR provides a description of the Fuel Handling Building Ventilation System (FHBVS).

Based on the system design, it is assumed that essentially all radioactivity escaping from the fuel pool as a gas will be collected by the FHBVS allowing the Fuel Handling Building to remain clear of gaseous radio-activity. The primary source of dose received in the Fuel Handling Building would -be direct radiation from the fuel pool and associated cleanup equipment (filters, pumps, and heat exchangers). The FHBVS will effectively control any gaseous releases from the SFP and doses due to airborne activity will be only a small fraction of direct radiation.

Q. 2 Discuss the capability of the spent fuel cooling system to keep the spent fuel pool water temperature. at or below the FSAR design value of 120°F during normal refuelings until the modified pool is filled. If the bulk water temperature may be above the FSAR design value, discuss when this will occur for what period of time. Discuss also, the impact of any higher than design value pool temperatures on the gaseous releases of radioiobines intrigging from the pool.

A. 2 Assuming the instantaneous discharge of one third of a core (one region) annually, the following are the expected pool heat loads and pool discharge temperatures:

P78 59 02

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Pool Heat Pool Heat Load 100 Hrs. Load 150 Hrs.

Region After Reactor Pool Disch. After Reactor Pool Disch.

Discharge Shut.106 B/HR Temp. (°F) Shut. 106 B/HR Temp. (°F)

1. 14.47 125 12.15 120.5

2. 15.63 127.5 13.32 122.9

3. 16.26 128.9 13.95 124.2

4. 16.71 129.8 14.40 125.2

5. 17.09 130.7 14.80 126.0

6. 17.43 131.4 15.12 126.7

7. 17.76 132 15.46 127.1

8. 18.08 132.7 15.67 128.1

9. 18.39 133.4 16.09 128.7

10. 18.69 133.9 16.39 129.3

11. 18.99 134.6 16.69 130.0

12. 19.27 135.2 16.98 130.6

13. 19.55 135.8 17.26 131.1

14. 18.83 136.4 17.53 131.7

15. 20.09 136.9 17.80 132.3

16. 20.35 137.5 18.06 132.8

17. 20.61 138 18.31 133.4

18. 20.86 138.5 18.56 133.9 Assuming an incremental discharge of the fuel assemblies into the SFP 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after reactor shutdown with a time increment between each assembly of one hour, the followi~g are the expected annual heat loads and pool discharge temperatures obtained:

Pool Heat Load 150 Hrs. After Reactor Pool Discharge Region Shut. (106 B/Hr.) Temp. (°F)

1. 11.38 118.7

2. 12.53 121.2

3. 13.15 122.4

4. 13.60 123.4

5. 13.98 124.3

6. 14.33 124.9

7. 14.66 125.6

8. 14.98 126.3

9. 15.29 126.9

10. 15.59 127.5

11. 15.89 128.2

12. 16.18 128.8

13. 16.46 129.4

14. 16.74 130.1

15. 17.01 130.6

16. 17.27 131.1

17. 17.52 131.7

18. 17.77 132.1 P78 59 03

In order to maintain the spent fuel pool bulk water temperature at or below 1200F the following table indicates the number of cooling hours required:

Region Cooling Hours

1. 160

2. 205

3. 235

4. 265

5. 295

6. 325

7. 360

8. 395

9. 435

10. 475

11. 515

12. 560

13. 610

14. 660

15. 715

16. 770

17. 835

18. 900 For a full core discharge, the temperature will not be allowed to exceed 1500F. Prior to a full core discharge, decay heat from the spent fuel assemblies will be dissipated into the RHR system.

The following table indicates the number of hours required prior to discharge of the full core into the spent fuel to assure that the SFP bulk water temperature will not exceed 1500F:

Regions Discharged Regions Discharged During Full Core Case Under Normal Conditions Discharge Cooling Hours

1. 0 1-3 320

2. 1 2-4 355

3. 1-2 3-5 380

4. 1-3 4-6 395

5. 1-4 5-7 410

6. 1-5 6-8 430

7. 1-6 7-9 440

8. 1-7 8-10 450

9. 1-8 9-11 465

10. 1-9 10-12 480

11. 1-10 11-13 495

12. 1-11 12-14 510

13. 1-12 13-15 525 P78 59 04 I

/

Regions Discharged Regions Discharged During Full Core Case Under Normal Conditions Discharge Cooling Hours

14. 1-13 14-16 540

15. 1-14 15-17 555

16. 1-15 16-18 570 The impact of the higher SFP bulk water temperatures on the gaseous release of radioiodines and tritium is discussed below. The sources of tritium in the spent fuel pool are the continuous diffusion of tertiary tritium through the fuel cladding and any fuel cladding defects and, to a lesser degree, from the neutron-boron reaction. The primary mechanism for any additional tritium release would be an increase in evaporation due to higher SFP temperatures. If the design SFP temperature were to be increased to 1500F, the calculated evaporation rate would increase by a factor 2.2. However, the total tritium release resulting from increased evaporation would be somewhat less than this factor, since makeup water would provide dilution.

As is the case for tritium, the expected radioiodine releases would be increased due to an increase in evapor-ation. However, unlike tritium, the partition factor would reduce the amount of iodine which would escape as a gas. Assuming that the SFP bulk temperature reached 1500F, or an increase of 300F over the FSAR design value, the increase in the partition factor would not be significant. Concentrations of radioiodines in the SFP could be reduced by shortening demineralizer resin change intervals in the event of increased liquid concentrations of radioiodines and other radionuclides.

As described in Section 9.11 of the Salem FSAR, the Fuel Handling Buildinq Ventilation System provides means to filter exhausted effluent through HEPA and charcoal filters should significant concentrations of iodines or particu-lates be identified in this exhaust stream. The charcoal filtration would reduce the iodine concentration by an estimated factor of 100. Radioiodine (half life greater than 8 days) releases from Salem Nuclear Generating Station are limited to less than four Curies per year and a design objective of less than one Curie per year by the Environmental Technical Specifications.

Compliance with this limit is assured by a monitoring program which includes continuous sampling for radioiodines. Actual expected manual releases from the Fuel Handling Building are a small fraction of a Curie as the one Curie design objective applies to all release points.

P78 59 05 Q. 3 Discuss the onsite and offsite impact of having a spent fuel pool water level only one foot above the spent fuel assemblies in the pool. Your submittal of November 18, 1977 stated this condition could exist following a spent fuel cask drop accident. Include in your discussion measures which would be taken to insure a greater depth of water over the spent fuel assemblies and how correc-tive actions could be taken to keep personnel exposures within the radiation exposure limits of 10CFR Part 20.

A.3 The statement "In the unlikely event of a cask drop in the transfer pool that results in the loss of water from the pool, the stored spent fuel will still be covered by a depth of water at least one foot above the top of the fuel" did not appear in our submittal of November 18, 1977, but rather was made by the NRC staff as in Section 9.4 of the Staff's Safety Evaluation Report for Salem Units 1 and 2, dated October 11, 1974.

The transfer pool is separated from the spent fuel pool by two gates. These gates will be installed during movement of a fuel cask. No possibility exists of losing the depth of water over the spent fuel assemblies below the nominal depth (EL. 128' - 8"). This action by itself will maintain the personnel exposures within the limits of 10CFR Part 20.

Q. 4 Provide a list of representative loads that might be carried near or over the spent fuel pool. Provide the weight, dimensions, and orientation of each load.

Discuss the load transfer path including whether the load must be carried over the pool, the maximum height at which it could be carried and the expected height during transfer. Provide a description of any written procedures instructing crane operators about loads to be carried near the pool. Provide the number of spent fuel pool assemblies that could be damaged by dropping or tipping into the pool of each representative load carried over or near the pool.

A. 4 The loads to be carried consist of the fuel assembly and the fuel handling tools. The following list specifies the length, weight, and orientation of each tool:

Equipment Length Weight Orientation Fuel Assembly 14 I 1615 lbs. Vertical Spent Fuel Assembly 32' 350 lbs. Vertical Handling Tool Burnable Poison 32' 650 lbs. Vertical Rod Removal Tool P78 59 06 The equipment listed above will be limited in vertical movement by the fuel handling crane limit switches.

The fuel assembly will be limited to a height of 15" above the spent fuel racks. All other equipment will be maintained, when not lifting a load, approximately 15' above the spent fuel racks. Dropping of one of the tools could damage the funnel portion of approximately 36 cells if the tools rotate while falling and drop perpendicular to the top of the cells.

The Technical Specifications for Salem dictate that loads in excess of 2500 lbs. shall be prohibited from travel over the spent fuel assemblies in the SFP. The fuel handling crane is equipped with overload limit switches which will render the crane inoperable if a load in ex-cess of 2500 lbs. is attempted.

Q. 5 Has any water containing radioactivity been introduced into the pool? If so:

(a) provide the estim~ted volume, if any, of contaminated material (e.g. spent fuel racks, seismic restraints) expected to be removed from the spent fuel pools and shipped from the plant to a licensed burial site because of the modification.

(b) discuss also the man-rem exposure to perform the proposed pool modification.

A. 5 The spent fuel pool has never been contaminated with water containing radioactivity. Also, the proposed modification is to be performed in a dry pool.

Q. 6 Discuss the existing or proposed instrumentation to indicate pool water temperature and level. Include its capability to alarm.

A. 6 Normal water level in the spent fuel pool is EL. 128'- 8" A level switch actuates on high level at EL. 129'- 2" and on low level at 128'- 2". The switch contacts which open on high and low level are used as input. to the control room overhead annunciator. The switch contact also actuates a local alarm near the spent fuel pool.

A local temperature switch and indicator alarms at 1250F + 20F. The alarm is sent to the control room overhead annunciator. The power source for this alarm is from a vital 118 V A-C instrument bus.

Q. 7 Provide the overall dimensions of the storage rack modules including the Bumper Plates. Also provide the nominal distance between the bottom of the modules bases and the pool floor.

P78 59 07 The lOxlO module is nominally 15' 3-3/4" high, 9'1-1/2" wide and 9'1-1/2" long. The Bumper Plates extend approximately 5-1/2" from the module. The modular base is approximately l'- 3 1/4" from the pool floor.

Q. 8 Provide the dimensions of the fuel pool.

A. 8 The spent fuel pool is 37'long and 28.5' wide.

Q. 9 State the nominal depth of water that will be maintained in the spent fuel pool. What is the minimum depth of water in the spent fuel pool, (i.e. when it is at the level of siphon breakers)?

A. 9 The nominal depth of water in the spent fuel pool is 39' 2" the lowest possible level will be 14' - 6".

This low level will occur when the gates separating the spent fuel pool from the transfer pool are removed.

Q. 10 Identify the redundant pool makeup water sources available along with flowrates and duration of these sources.

A. 10 Salem FSAR Questions 9.10 and 9.48 provide a description of the makeup water supply for the spent fuel pool.

Q. 11 State whether it is possible to place a fuel assembly either between the storage rack modules or between the outer periphery of the storage rack modules and the fuel pool walls. If so, what is the closest possible distance it can be placed to a fuel assembly which is in the storage racks? Also, what is the maximum possible neutron multiplication factor in the pool assuming that the outside assembly is in the most reactiv position and assuming that the storage racks are filled with the most reactive fuel assemblies with no boron in the water.

A. 11 Is is not possible to place a fuel assembly either between a storage rack module or between the outer peri-phery of the storage rack modules and the spent fuel pool walls.

Q. 12 Provide a more complete description and written docu-mentation on the experiment listed as Case 4 of Table 3.1

7. This documentation should include the limits of possible experimental errors. Also discuss the geometry and the assumptions that were made when calculating this experiment with the NITAWL-XSDRNPM-KENO-IV Programs.

P78 59 08 A. 12 Benchmark Documentation The reported values in lines 4 and 5 for keff calculated by CCELL-KENO-III are in error. These should be 1.038 in line 4 and 1.037 in line 5. Figure 1 (attached) describes the geometrical makeup for the critical experiment listed as line 4 of Table 3.1-7.

The comparison of the KENO-III and KENO-IV analysis in Table 3.1-7 with ORNL critical experiments repre-sents the most current and applicable benchmark data available for poisioned fuel storage rack applications today. To date, these*ORNL critical experiments are unpublished.

Limits on Experimental Errors Fuel enrichment equals 4.95 + .05 w/o U-235 The + .05 w/o uncertainty in-enrichment would have an effect of :=: + .002 tJ, k.

Lattice Reactivity The greatest reactivity difference measured from the critical condition for any of the five lattices was

+ 12.1 cents. Hence the measured keff value for each of the lattices has an uncertainty of ~ O.OOlk.

Model Assumptions The KENO-IV 123 (group) reactivity calculations incor-porate the same general assumptions made for the FSR calculations; for the benchmark calculations:

1. The NITAWL code was used to obtain U-238 cross section data adjusted to account for resonance self shielding by the Nordheim integral method. (123 group XSDRN X-S data was used).

2. XSDRNPM was then used to spatially cell-weight the cross sections from NITAWL for input into KENO-IV.

This allows one to (in the Salem case) represent the fuel region in KENO as a homogeneous mixture. Thus, for the 3 benchmark calculations using KENO-IV, all regions were represented discreetly except the fuel region which was homogenized as described above.

P78 59 09 Q. 13 What is the maximum temperature and the maximum temper-ature gradient due to gamma heating in the concrete pool wall when the maximum power assemblies (i.e., 55.4 KW/assembly) are placed in the storage locations which are closest to the pool walls?

A. 13 Calculations show that the maximum temperature expected in the concrete pool wall is less than 5%

greater than the temperature of the SFP bulk water temperature.

Q. 14 Provide a description of the onsite test you intend to perform to verify within 95% confidence limits, that a sufficient number of Boral Plates in the installed racks will contain the required boron content to maintain the keff ::. 0.95.

A. 14 Stringent in-process inspection and process controls are imposed during manufacturing of the Baral to assure that the Boral plates contain at least 0.020 gms B-10/cm2. Additional measures assure that the plates are properly inserted in each completed storage cell.

In addition to these controls, KENO-III analyses were performed with one Boral poison plate missing in a 5X5 storage cell array with a keff of 0.928

+ 0.004. Consequently, the design assures that the keff criteria of 0.95 or less is met for all conditions previously delineated in the license amendment submittal as well as for the unlikely event of one Boral plate missing.

Neutron transmission tests will be performed on the completed rack modules prior to placing any fuel in the racks.

As prior tests and inspections have verified the uniformity and minimum boron-10 content of the Boral plates, the purpose of these additional attenuation measurements is to further confirm the results of the process controls previously discussed (i.e. to provide further assurance that only Boral plates are contained in the completed storage cells).

The tests will be performed by inserting a shielded moderated neutron source into one storage cell and a neutron detector into an adjacent storage cell. Prior to the actual tests, calibration tests will be performed which will provide the necessary counting statistics for demonstrating whether any Boral plates are missing or do not contain the boron poison material.

P78 59 10 The tests will be performed on the foll~wing basis:

1. A 100% storage cell inspection of the first rack module.

2. A 10% random sampling of the storage cells in the remaining rack modules.

3. Should any one measurement demonstrate that a Boral plate is missing or does not contain the boron poison material, a 100% inspection program will be conducted on all rack modules.

A control procedure will be prepared and utilized during the measurements and a permanent record will be generated and retained for each measurement.

Q. 15 In Section 3.7.1 of your February 14, 1978 submittal you state that the racks are designed to Section III, Subsection NF. Do the criteria for materials, fabrication and installation, and examination conform to articles NF-2000, NF-4000, and NF-5000 respectively?

If not provide the standards used.

A. 15 Specifications utilized for materials, fabrication and installation, and examination are in general conformance with the requirements of the ASME Code Section III, Subsection NF, Articles NF-2000, NF-4000, and NF-5000.

Code stamping of the fuel storage racks is not required.

Q. 16 With regard to the seismic analysis, provide information on how the floor level response spectrum was obtained for the storage pool floor. Does this spectrum conform to that approved for the FSAR? Is the seismic excitation along to three orthogonal directions combin~d in the analysis?

A. 16 The seismic response spectra for the spent fuel storage pool floor was generated from the horizontal and vertical time-history accelerations calculated at the level of the pool floor. The total acceleration history evaluated at that floor level was extracted from the seismic analysis of the Fuel Handling Building and applied to a single degree of freedom oscillator. By varying the frequency and damping of the oscillator, the full range of the response was generated for each given earthquake motion. For the fuel storage rack analysis, the result-ing response spectra were broadened at the peaks by 15%

in accordance with the provisions of Regulatory Guide 1.122. The response spectra used in the analyses con-form to those approved in the FSAR.

P78 59 11 Seismic excitations along the three orthogonal directions are combined by the SRSS method in accordance with the requirements of the Standard Review Plan, Section 3.7.2, and Regulatory Guide 1.92.

Q. 17 Provide the following information on the damping values used in the seismic analysis.

a. Identify the damping values used.

b. Are these values consistent with those approved in the FSAR?

c. Are these values in accordance with criteria in Regulatory Guide 1.61?

d. Is additional damping credit taken due to submergence in water? If so, detailed test data and analytical results are requested.

A. 17 a. One percent (1%) damping was used for OBE and three percent (3%) damping was used for SSE.

b. The damping values used are consistent with those approved in the FSAR.

c. Regulatory Guide 1.61 allows up to 2% damping for the OBE and 4% damping for the SSE for welded steel structures. The damping values used are below these levels, which is more conservative.

d. No credit was taken for additional damping due to submergence in water.

Q. 18 Provide an analysis to show that the safety factors to protect against the overturning of racks under all possible service conditions are in accordance with Section 3.8.5.II.5 of the Standard Review Plan.

Explain in more detail how transmission of vertical shear forces prevents overturning as stated in Page Sa of your February submittal.

A. 18 Calculations have been performed which show a factor of safety against overturning of 1.7 for dead+ OBE loads and 1.45 for dead + SSE loads.

As shown in Figure 1.2-3 of the February submittal, the interties are attached to the modules by bolts.

These bolts are torqued to a specified minimum tension to clamp the interties to the peripheral beams and prevent rotation of the interties relative to the peripheral beam. As the modules attempt to over-turn, the intertie acts as a cantilever with its free end fixed against rotation. The shear force on the module due to the interties creates a moment which resists overturning of the module.

P78 59 12

Q. 19 Provide the increase in floor loadings in the spent fuel pool due to the proposed rack design and its effect, if any, on the overall structural stability and seismic response of the auxiliary building. Does the pool structure still meet the allowable limits imposed on the design by the FSAR?

A. 19 Pool floor loadings due to the fuel storage racks and fuel are as follows:

Total dry weight of rack module with fuel 196,100 lbs.

Total buoyant weight of rack module with fuel 175,600 lbs.

Floor loading due to buoyant weight of rack module with fuel 2,110 lbs/ft.2 Maximum load on any 6" diameter module screw foot 53,100 lbs.

The increase in floor loading due to the proposed spent fuel storage racks is well under 1% of the total mass lumped at that level in the analytical model. The walls were investigated for the seismic effect of the heavier racks and stored fuel. It is concluded that the new high density racks have no appreciable effect on the structural stability and seismic response of the Fuel Handling Building. The pool structure meets all allowable limits imposed on the design in the FSAR.

Q. 20 In Section 3.4 of your February 14, 1978 submittal you state that materials are compatible with pool water that contains a nominal concentration of 2,000 ppm boron. Describe your surveillance program to maintain this leve Also, discuss the quality of the pool water in terms of pH value and available chlorides and fluorides.

A. 20 No surveillance is required to assure that the pool water contains a nominal concentration of 2,000 ppm boron.

The design of the proposed spent fuel storage racks is such that the keff will be below 0.95 even with unborated water.

The pH value of the SFP water, which depend on the con-centration of boric acid, will range from 4 to 4.7.

Chlorides and fluorides in the water will be a maximum of 0.15 ppm.

P78 59 13 Q. 21 On Page 37b of your February 14, 1978 submittal, you state that comparison tests were used to develop load deflection curves. Briefly describe these tests and the results. Provide a comparison of dynamic results to static results. Describe how dynamic stress-strain curves were developed for tensile and shear modes that occur in impact loading conditions.

A.21 Compression tests were performed on each of two test specimens. Each of the specimens was made of a 2' long section of the Boral poison spent fuel cell together with the flared insertion guide attached at the top. A steel bearing plate was attached to the bottom for support. The load was applied by thick square steel plate mounted on the load head of a 120 kip static test machine. The plate, representing the bottom of a fuel assembly, contacted the top of the insertion guide in a manner which represented the alignment and orientation of the fuel assembly with respect to the fuel cell at the assumed time of impact. An increasing load was slowly applied and the load-deflection recorded until the insertion guide was significantly deformed. In each of the two tests, the structure deformed elastically up to the onset of localized inelastic buckling. In the first case where the four corners of the fuel assembly contact the mid length of the four sides of the cell, initial inelastic deformation occurred at a load of 4,500 lbs. The maximum load of 25,000 lbs. occurred at a deflecton of 2.5 inches. In the second case, where the fuel assembly is centered on a group of four fuel cells, initial inelastic deformation occurred at a load of 5,000 lbs for the one cell tested. The maximum load of 10,000 lbs. occurred at a deflection of 0.8 inches.

In both cases, local crushing of the cell was limited to the upper 7 inches of the insertion guide. Energy absorption at maximum load was 3,300 ft.-lbs. for the first case and 428 ft.-lbs. for the second case. The impact in the second case involved four cells, giving an energy absorption of 1,712 ft.-lbs. Energy continued to be absorbed with deflections beyond the point of peak load.

The mechanical strength and auxiliary properties of type 304 stainless steel have been shown to be relative!

insensitive to strain rate over a wide range. It is therefore concluded, that the static load tests describe above, adequately represent the response to fuel cell impact at the low velocities involved (approximately 7ft.

per second)

• P78 59 14 Q. 22 On Page 35 of your February 14, 1978 submittal, you state that the mass of water enclosed in the storage rack is lumped together with the masses of the fuel assembly and rack structure. Is the water surrounding the modules included? Provide details on the method used to calculate virtual masses (e.g. potential theory). What virtual mass do you use for the horizontal excitations and for the vertical excitations.

A. 22 Only the mass of water entrained in the module is included in the model. The maximum gap between the modules and any of the pool walls is 7 inches. The additional mass due to this water is considered insigni-ficant. Virtual mass was computed on a per storage position basis as follows: Horizontal virtual mass was taken as a mass of water in a 10- 1/2" square, the height of the fuel storage cell plus the mass of water entrained in the fuel storage cell gap less the mass of water displaced b~ fuel and the structure, and was found to be 1.18 lb-sec / in. or 456 lbm. Due to the 6" diameter flow holes provided in the bottom of each storage cell for coolant flow, vertical motion of the water was not considered to be restrained, and no vertical virtual mass was added.

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It ATTACHMENT 2 Question 1 In Section 3.7.4 of your February 14, 1978 submittal, the maxi-mum impact force for the wall restraints is stated to be 21,800 lbs., calculated by non-linear analysis. Provide the correspond-ing force calculated by linear elastic analysis.

Answer 1 The total horizontal seismic reaction at the base of a row of four racks, calculated by the linear elastic analysis, was 218,000 lbs. The net load on the wall restraints would be 7,000 lbs., maximum, with a balance being resisted by the fric-tion between the rack feet and the pool floor. As stated in the February 14, 1978 submittal, the wall restraints were de-signed for the higher 21,800 lbs. impact load.

Question 2 What was the upper limit of friction coefficient assumed to be for the design of the rack feet.

Answer 2 The rack feet were designed based on the assumption that they were fixed against sliding on the pool floor, i.e., that the friction coefficient could be infinitely high. The feet were, consequently, conservatively designed to resist all of the horizontal seismic forces, without assistance from the wall restraint members.

Question 3 What provisions have been made to prevent blockage from the fuel storage cell coolant flow inlet, and what would be the consequences if such blockage should occur?

Answer 3 The spent fuel is cooled by convective flow of coolant up through the fuel bundle from a 6" diameter hole in the bottom plate of the rack. The water returns to the bottom of the pool by traveling down outside the racks between the outside rack sur-face and the pool wall. The maximum clearance between the wall and the rack is 7", and the space at the bottom of the rack is 6". Thus, it is extremely unlikely that any suffi-ciently sized object dropped or inserted in the pool could be carried to a position where it could fully block the flow of water into a storage cell.

In the unlikely event that flow to a cell did block the fuel storage cell coolant flow inlet, an analysis was performed to evaluate the possible consequences. It was assumed that flow is fully blocked from the cell containing a freshly dis-charged fuel assembly generating 55 kW of decay heat. Using

an approach presented by Lahey and Moody,l a conservative countercurrent two-phase, flow analysis was performed. It was found that the evaporation (steaming) rate from the fuel bundle was about a factor of 5 lower than that which would limit reflooding through the location of minimum flow area at the upper tie plate.

The average heat flux resulting from the blockage of the cool-ant inlet would be less than 600 BTU per hour per square foot.

This value is substantially below critical heat flux even with a void fraction as high as 99%, for which the critical heat flux would be about 4000 BTU per hour per square foot at the top of the fuel bundle. With a heat flux less than critical, the maximum clad temperature would be effectively at the separation temperature, or 2400F.

Question 4 Examine the consequences of the gate which separates the spent fuel pool from the transfer pool, falling on the spent fuel racks.

Answer 4 It is not possible for a gate to fall on the spent fuel racks.

The gates are restrained by a safety cable which prevents the gates from falling on the spent fuel racks. The gate, which weighs approximately 3,370 lbs., is lifted not by the fuel handling crane, but rather by a separate hoisting system trav-elling overhead directly above the area of the gates in the spent fuel pool and the Decontamination Pit.

Question 5 How would flow be stopped if liner perforation occurred?

Answer 5 Flow from the leak protection system would be stopped, should liner perforation occur by installation of a plug on the leak detection system tell-tale drain from which leakage would ori-ginate.

Question 6 How long would a leak go undetected, should perforation of the liner occur?

1 R. T. Lahey, Jr. and F. J. Moody, "The Thermal-Hydraulics of a Boiling water Reactor," American Nuclear Society (1977)

Answer 6 Should perforation of the liner occur, depending on the size of the perforation and the location of the perforation, the rate of water leakage can be determined. Assuming that un-restricted flow occurs out of one leak detection system line located at the bottom elevation of the pool, the leak would go undetected for approximately two minutes. The leak detection system is designed so that any leak from the spent fuel pool will be collected in a sump. A sump pump installed in the sump will start upon water reaching a preset level. Actua-tion of the sump pump will automatically actuate an alarm in the control room.

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