ULNRC-03887, Forwards Response to NRC 980723 RAI Re Callaway Sf Pool Rerack
| ML20237E392 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 08/25/1998 |
| From: | Passwater A UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-MA1113, ULNRC-03887, ULNRC-3887, NUDOCS 9808310252 | |
| Download: ML20237E392 (16) | |
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Antoren Corporation One Ameren Plaza
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1901 Chouteau Avenue PO Box 66149 St. Louis, MO 63166-6149 314.stt.3221 August 25,1998 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station PI-137 ULNRC-03887 Washington, D. C. 20555 TAC NO. Mall 13 Gentlemen:
' MI4 DOCKET NUMBER 50-483 Wggg CALLAWAY PLANT RESPONSE TO NRC RAI LETTER CALLAWAY SPENT FUEL POOL RERACK
Reference:
- 1. ULNRC-3742, dated February 24,1998
- 2. ULNRC-3837, dated May 27,1998
- 3. NRC Request for Additional Information Letter, dated June 4,1998, from K. M. Thomas to G. L. Randolph
- 4. ULNRC-3850, dated June 25,1998
- 5. NRC Request for AdditionalInformation Letter, dated July 23,1998, from C. F. Lyon to G. L. Randolph Reference 1 provided the original submittal of an amendment request to revise the Callaway technical specifications to support modification to increase the spent fuel storage capacity at the Callaway Plant. Reference 2 transmitted revisions to original submittal Chapters 4 and 5 of the Licensing Report to reflect a re-assessment of the proprietary classification ofproprietary versus non-proprietary material.
Reference 3 transmitted requests for additional information. An enclosure to
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Reference 4 provided responses to the requests. Reference 5 transmitted a second set of requests for additional information. The responses to these requests are being
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provided as an enclosure to this letter.
If there are questions on the Enclosure or if additional information is required, please contact us.
Very truly yours, M
Alan C. Passwater jltu' Manager, Corporate Nuclear Services DJW/mlo Enclosure 9808310252 980825 PDR ADOCK 05000483 P
PDRg
I STATE OF MISSOURI )
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CITY OF ST. LOUIS )
l Alan C.
Passwater, of lawful age, being first duly sworn upon oath.says that he is Manager, Corporate Nuclear I
Services for Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his 1
knowledge, information and belief.
'l By he44 Alan C. Passwater Manager, Corporate' Nuclear Services SUBSCRIBED and sworn to before me this 88/'I day of
' 27ae ad 1998.
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PATIWCIA L M W M SS Noterpuus suuteraugu m sr.uxasessfry Mlf00MOSIIONEMIR,4M L.
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cc:
M. H. Fletcher i
Professional Nuclear Consulting, Inc.
19041 Raines Drive Derwood, MD 20855-2432 Regional Administrator U.S. Nuclear. Regulatory Commission Region IV 611 Ryan Plaza Drive Suite 400 Arlington, TX 76011-8064 Senior Resident Inspector Callaway-Resident Office U.S.- Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Kristine M. Thomas (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102-Ron Kucera
-Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102' Denny Buschbaum TU Electric P.O. Box 1002-Glen Rose, TX 76043 Pat Nugent L
Pacific Gas-& Electric Regulatory Services P.O. Box 56 Avila Beach,.CA-93424 i
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u L_____
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l ULNRC-03887 Page 1 of13 Response to Request for Additional Information (RAI) Dated 7/23/98 Question 1:
The May 27,1998, submittal states that administrative controls will be implemented to ensure that dose rates external to the building will meet the limits for the general public. It also states that additional controls will be implemented to ensure that interior building dose rates are maintained within the ranges of the specified zone designations. Describe what additional controls you will take to ensure that the current radiation zone designations are not exceeded. In j
addition to administrative controls, discuss the use of any bridge movement interlocks that may be used to control the placement of fuel assemblies in the fuel racks.
Response 1:
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Administrative controls will be implemented to define acceptable storage locations for spent fuel l
assemblies to ensure that specified radiation zone designations are not exceeded. The administrative controls will be based on calculations and will be documented in the form of plant j
procedures associated with special nuclear material control. The administrative controls will ectablish minimum cooling times for spent fuel assemblies prior to storage in designated loccions. In addition, routine radiological dose surveys in accordance with Health Physics procedures will record dose rates to ensure zone designations are not exceeded. Administrative controls will be adequate to prevent placement of fuel assemblies into unauthorized storage locations, therefore bridge movement interlocks will not be required.
Calculations have been performed to provide one set of administrative controls which may be implemented to maintain zone designations at the same level as the current zone designations (with the exception of the cask loading pit and cask washdown pit areas). The zone designation for the cask washdown pit area (Room 6204 on FSAR Figure 12.3-2) will change from "B" (<2.5 -
mrem /hr) to "E" (>100 mrem /hr) based on the calculated dose rates associated with the most limiting case analyzed. Access controls will be implemented for the new Zone "E" area in accordance with Health Physics procedures. The zone designation above the pool surface in the cask loading pit will be changed from zone "B" to zone "C" which matches the current zone designation above the spent fuel pool given on FS AR Figure 12.3-2.
i Question 2:
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Your submittal states that you will install three spent fuel pool (SFP) racks (with a combined capacity of 279 fuel assemblies) in the cask loading pit in a later campaign. Discuss how the presence of spent fuel in these racks will affect the dose rates in accessible areas adjacent to the cask loading pit both during storage and movement of the spent fuel assemblies into and out of a
the pit.-
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-Response 2:
The increase in dose rates due to fuel stored in the cask loading pit was considered by calculating dose rates at discrete locations around the pit. Initial calculations were based on very conservative fuel parameters (producing source terms beyond the expected limiting values),
which produced dose rates in excess of the current zone designation ranges. Therefore, supplemental calculations were performed to establish stonge limitations which would confirm dose rates within the current zone designation ranges except as established by the response to question 1. The supplemental calculations were also based on conservative fuel parameters, i
since most of the fuel in the cask loading pit is considered to be cooled only 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> subsequent to reactor discharge. This is conservative because the cask loading pit will not contain any freshly discharged fuel. These storage limitations will be implemented through administrative controls based on establishing minimum cooling times for designated storage locations prior to storage. These controls provide a greater distance from the accessible locations to fuel cells which have no cooling time restrictions, and/or a barrier shield using fuel which has had a lengthy cooling time.
The calculated dose rates based on conservative fuel parameters and the worst case configurations allowed by the storage location limitations are as follows:
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1 At the outside of the Fuel Building wall surface at a point directly plant east of the spent fuel l
e location in the cask loading pit, the dose rate is calculated to be 0.08 mR/hr,
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At a point along the wall in the truck bay located directly plant south of the cask loading pit,-
the dose rate :s calculated to be 2.14 mR/hr.
The dose rates listed above fall below the designated limits. At the fuel operating deck (elevation 2047'-6") the approximate 25 feet of water covering the active fuel region, coupled with the adjacent concrete walls, provide more than adequate shielding to maintain dose rates above the water surface of the SFP below the zone C limit of 10 mR/hr. The dose on the fuel operating deck due to fuel stored in the cask loading pit will be less than or equal to the dose due to fuel stored in the SFP, which has similar distance and shielding to the areas accessible to personnel.
l The dose due to fuel in transit remains unchanged from the previous conditions. Fuel movement within the cask loading pit has taken place previously, since the fuel reconstitution station is located within the cask loading pit as discussed in FS AR Section 9.4.1.2. The elevation at _which fuel is raised by the fuel handling machine remains unchanged as does the nominal water depth and fuel parameters. Therefore, the r,hielding provided by the SFP or cask loading pit water is unchanged and the dose rate to the operator and areas surrounding the SFP or cask loading pit will not change.
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Page 3 of13 Question 3:
i Describe any sources of high radiation that may be in the Callaway Plant SFP during diving
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operations to remove the old SFP racks and install the new racks. Discuss what precautions
- (such as use of TV monitoring, tethers, etc.) will be used to ensure that the divers will maintain a safe distance from any high radiation sources in the SFP. Describe how you plan to monitor the doses received by the divers during the reracking operation (e.g. use of dosimetry, alarming dosimeters, remote readout radiation detectors).
Response 3:
Some sediment is expected to have settled at the bottom of the SFP and cask loading pit.
Considering that Callaway Plant has had fuel leakage in the past, it cannot be ruled out that highly radioactive particles, such as fragments of fuel pellets or crud from fuel clad surfaces may be located on surfaces within the SFP or cask loading pit. Other sources of high radiation include the spent fuel assemblies and associated inserts, burnable poison rods, a fuel rod storage rack containing failed fuel pins, and a trash rack containing miscellaneous loose pans. Another source of high radiation on the floor of the cask loading pit is a small piece of torn grid strap from a fuel assembly. Potential sources of high radiation due to contamination include the spent fuel storage racks, the new fuel elevator, fuel handling tools, pool walls, submerged piping or pipe support surfaces. An underwater vacuum cleaner system will be used in conjunction with radiation surveys to ensure appropriate areas are vacuumed as part of dive preparations.
Callaway Plant has purchased special handling baskets for this underwater vacuum system to -
contain small highly radioactive materials such as the piece of grid strap mentioned above.
The Callaway Health Physics Department is aware of the potential for radiation exposure challenges during diving operations in the SFP or the cask loading pit. Callaway Plant diving operation procedures describe the Health Physics coverage associated with each dive. The procedure requires pre-dive surveys with at least two independent survey instruments, with specified agreement criteria between the survey results. The procedure requires the use of electronic dosimeters on the diver. These dosimeters will have remote, above surface, readouts that will be continuously monitored by Health Physics personnel during the dive. The procedure also requires that the diver have available an operable survey instrument so that he can scan j
unknown items. The readout is maintained at the dive control station, and monitored by a Health I
Physics Technician. Constant communications is required between the dive control station and the diver.
The physical methods of controlling the divers spatial location to keep him away from high radiation sources will depend upon the considerations of each dive, and will be determined by the ALARA review in consultation with the dive controller. The primary method of control is to i
- adequately brief the diver on the particularjob, and provide constant visual oversight of the diver l
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from the dive control stat on. The use ofvisual barriers such as air bubbles, ropes, or orange
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safety netting or the use of physical barriers such as tethering the diver umbilical and other control measures will be implemented if the situation is determined to warrant it by Health Physics or the dive controller.
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l Page 4 of 13 The diver will also be in continuous voice communication with the dive controller / Health l
Physics personnel. In the event that an unexpectedly high radiation field / hot particle is encountered by the diver, he will be verbally instructed to take the appropriate corrective action.
Positive control of the diver will be maintained, at all times, through use of an individual on the surface maintaining a diver safety line.
. Radiological surveys are required for the dive area, including entrance and exit travel routes to and from the SFP. These surveys will consist of surveying the general work area and specific
. components with an underwater probe capable of detecting expected radiation levels.-
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Radiological surveys are required after any movement ofirradiated hardware and prior to diving l
after such movement.
l During the planning stage, the sequence of work will be evaluated to ensure safe dive areas are established for any diving activity. Sketches of rack layouts and positions ofirradiated hardware will be reviewed by site personnel, prior to each dive, to ensure sufficient space has been provided between the diver and irradiated hardware. A pre-job brief will be conducted with the diver to discuss radiological conditions and work scope prior to each dive.
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Every dive situation will be evaluated to ensure the safety of the diver.
l Question 4:
Provide the calculated 30 day doses (thyroid and whole body) to the control room operator as a result of a fuel handling accident (occurring both in the fuel handling building and in containment). Describe the calculational method used to arrive at these doses and include all 1
assumptions. Describe any differences that may exist between control room isolation following l
. a fuel handling accident and control room isolation following a LOCA and describe how these l
l-differences would affect operator 30 day accident doses.
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Response 4:
l Section 15.A.3 of the Callaway Plant FSAR states that only control room radiation doses due to a
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postulated LOCA are presented in FS AR Chapter 15 since a study of the radiological L
consequences in the control room due to various postulated accidents indicate that the LOCA is the limiting ' case. The limiting control room dose from the LOCA is to the thyroid and is listed in FSAR Table 15.6-8 as 25.55 Rem. Additionally, as a part of the supporting information for Amendment 114 to Callaway Plant's Operating License, control room doses resulting from'a Reactor Building fuel handling accident were calculated. The control room thyroid dose calculated for this accident sequence was 8.33 Rem. This value was reviewed and approved by the NRC staffin their Safety Evaluation Report (SER) for Amendment 114. The SER also documented the results ofindependent dose calculations performed by the NRC staff. The results of the independent NRC calculations indicated that the calculated value reported in the
- Callaway Plant's licensing submittal was conservative for the Reactor Building fuel handling accident (FHA).
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The proposed modification to increase the capacity of the SFP racks does not affect any of the assumptions for either the Reactor Building or Fuel Building FHA offsite and Control Room dose calculations. The control room radiological consequences for the Reactor Building fuel handling accident are still bounded by the values reviewed and approved by the NRC staff.
Additionally, we have concluded that control room doses following a Fuel Building FHA are bounded by the Reactor Building FHA radiological consequences. This conclusion is based on a calculation of fuel building FHA Control Room thyroid doses.
The assumptions used in the Fuel Building FHA Control Room dose calculation regarding i
release of radioactivity from the dropped fuel assembly are consistent with the FHA analysis discussion listed in FSAR Sections 15.7.4.5.1.1,15.7.4.5.1.2, FS AR Figure 15A-1, and the Callaway Plant commitment to Regulatory Guide 1.25 " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel
- Handling and Storage Facility for Boiling and Pressurized Water Reactors" as discussed in FS AR Table 15.7-2. The assumptions used regarding the transport of radioactivity inside the Control Building and Control Room are consistent with the information provided in the Control Room dose calculation methodology and parameters discussed in Section 15A.3 and Tables 15A-2,15A-3, and 15A-4 of Callaway Plant's FS AR.
The dose calculations use similar assumptions regarding the timing of control room isolation following LOCA, Fuel Building FHA, and Reactor Building FHA. None of these calculations rely on actuation of Control Room HVAC radiation monitors GK-RE-04/05. Following a LOCA, the Safety Injection Signal initiates a Phase A Containment Isolation Signal. The Phase A Containment Isolation Signal initiates a Control Room Ventilation Isolation Signal (CRVIS).
For the Fuel Building fuel handling accident, a Fuel Building Isolation Signal (FBIS) initiates a l
CRVIS. For the Reactor Building fuel handling accident, Technical Specification 4.9.4.2 i
specifies fuel handling setpoints for GT-RE-22 and GT-RE-33. The high radiation signal from these radiation monitors initiates a CRVIS.
l Question 5:
Verify that all of the assumptions used in the Final Safety Analysis Report (FSAR) fuel handling accident analysis are still applicable. Also verify that the resulting postulated thyroid and whole body doses at the Exclusion Area Boundary and Low Population Zone as a result of a fuel
' handling accident are still valid.
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Response 5:
l' All assumptions used in the FSAR fuel handling accident analysis are still applicable. The resulting postulated thyroid and whole body doses at the exclusion area boundary and low population zone as a result of a fuel handling accident remain valid and are listed in FSAR Table 15.7-8. All assumptions used in the FSAR analysis, such as depth of water over the top of the storage racks, and protection provided to assemblies stored in the racks following the drop of an assembly onto the racks remain valid.
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Question 6:-
Discuss the shipment and disposal of the old spent fuel rack modules.
Response 6:
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The racks will be rinsed with demineralized water while being removed from the pool to remove contaminants to an acceptable level.. Administrative controls will prevent the demineralized water from diluting the pool below acceptable boron concentration limits.' After removal from L
the pool the racks will be bagged, sealed, down-ended, and placed into a special DOT approved -
shipping container. The rack will be braced inside the container, prior to sealing the container, to prevent shifting during transit. The container and enclosed rack will be shipped on a flatbed i
o truck to Manufacturing Science Corporation (MSC) in Oak Ridge, Tennessee for disposal.
- Health Physics personnel will monitor the packaging prior to shipment to assess dose rates and to j
ensure the packaging will prevent dispersal of contaminants. The shipments will be made in accordance with Health Physics and Radwaste departmeht procedures governing shipment of radioactive material / waste and will meet the applicable requirements of 49 CFR and 10 CFR
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' 620.
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Question 7:'
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- Discuss how the storage of the additional spent fuel assemblies in the Callaway Plan' SFP will
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L affect the releases of radioactive gases (specifically Kr-85, I-131 and tritium) from the SFP.
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, Response 7:
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Release of radioactive gases by Callaway Plant will remain a small fraction of the limits of 10.
CFR 20.1301 and the' design objectives of Appendix I to 10 CFR 50 following the j
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implementation of the proposed modification to increase the capacity of the Callaway Plant SFP.
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This conclusion is based on the following supporting statements:
The halflives of short lived nuclides such as I-131 are short m comparison to fuel cycle i
length, therefore, short-lived nuclides are present only in freshly offloaded fuel. The quantity.of freshly offloaded fuel placed into the SFP each refueling outage is not affected by' the number of spent fuel assemblies being stored in the SFP. Therefore, the
L Inventories oflong-lived fission products (e.g.' Kr-85, I-129, and ternary tritium) in spent E fuel assemblies will decrease slowly within individual fuel assemblies over years in storage. Therefore, an increase in the number of spent fuel assemblies stored in the SFP p
would increase the total SFP inventory of these radionuclides. However, these L
radionuclides are not released in significant amounts from the stored fuel to the SFP water, even for failed fuel since the fuel pellet temperature of stored fuel is not high enough to create sufficient gas pressure in the gap to overcome the static pressure of the
. SFP water, q
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The radioactivity in the SFP water is independent of the number of assemblies in the
? pool. This activity originates in the reactor coolant system (RCS) and is introduced into the SFP when the SFP and RCS are connected during refueling outages.
The increased number of spent fuel assemblies in storage will raise the heat load on the SFP and could result in an increase in the evaporation rate. Other than a small amount of tritiated water released by evaporation, SFP radionuclides are non-volatile and consequently are not released from the pool water. The increased evaporation rate of tritiated water would result in an increase in gaseous tritium released in Callaway Plant's -
effluents. However, the discharge of gaseous radioactive effluents will continue to be a J
. small fraction of the limits of 10 CFR 20.1301 and the design objectives of Appendix I to 10 CFR 50.-
Question 8:
Discuss how the storage of the additional spent fuel assemblies will affect the releases of a
ndioactive liquids from the plant.
Response 8:
j The number of spent fuel assemblies in storage does not affect the release of radioactive liquids l
from the plant. The contribution of radioactive materials in the SFP water from the stored assemblies is insignificant relative to other sources of activity, such as time reactor coolant j
system. The volume of SFP water processed for discharge is independent of the number of fuel L
assemblies stored in the SFP.
Question 9:
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Discuss your plans to use a vacuum to remove any crud or other debris from the floor of the SFP before and during the SFP reracking project. Also describe any radiation surveys that will be
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performed (from the pool rim or by divers in the pool) to map dose rates in the SFP.
Response 9:
During reracking, a Tri-Nuc underwater filtration unit with associated filters will be used to vacuum the SFP. A backup system will also be available. Vacuuming will consist ofvacuuming crud after a series of"old" racks have been removed. Vacuuming will occur in areas in which l
the diver may have to enter, provided the dose rates in the area warrant vacuuming.
l Performance of radiation surveys is discussed in the response to question 3.
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. Question 10:
Is full core off-load the general practice for planned refueling outages?
Response 10:
Yes. -A full core off-load is the general practice for planned refueling outages.
Question 11:
If full core off-load is the general practice, then the corresponding heat load is the one considered for normal operation. Table 5.8.1 states that for the postulated full come discharge, the maximum allowable heat load is 63.41 MBtu/hr, which corresponds to a maximum bulk SFP temperature of q
170 F. This limit exceeds the American Concrete Institute (ACI) Standard 379, which states in q
part,."for normal operation or any other long term period, the (SFP) temperatures shall not exceed 150 degrees." If the planned refueling is generally a full core off-load, provide an evaluation on how long the SFP temperature would be above 150 F, and the justification for why the 170 F limit is acceptable to meet the intent of ACI-379.
Response 11:
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(As stated in response to question 10, full core ofiload is the normal practice during refuehng 1
outages.. The 63.4 i MBtu/hr decay heat load limit is not considered to be a long term normal g
f operation heat. louliThe heat load of 63.41 MBtu/hr corresponds to a steady state bulk pool
' temperature of 170 F when the heat balance is calculated assuming one train of cooling in ~
operation, and the design basis maximum coolant water heat exchanger inlet temperature of 105 j
F. Given the initial conditions of 170 F bulk pool water temperature and a decay heat load of q
63.41 MBtu/hr, it has been determined that bulk pool boiling will not occur following a
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postulated 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> duration loss of forced cooling.
Using a conservatively calculated decay heat load, it has been determined that the maximum time period that the pool temperature would be above 150 F is less than 9 days per fuel cycle.
The 170 F bulk pool temperature is considered acceptable for the following reasons:
a) It is common practice for the bulk pool temperature at many plants to exceed the l
recommended 150 F limit during short durations. In fact, the current licensing basis j
allows the bulk pool temperature to approach approximately 160 F, as stated in FSAR Section 9.1.3.3.
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b) As noted by the. question, the 170 F temperature represents a peak, off normal,.
- temporary condition. During the time of this bulk pool temperature peak, the concrete
._ temperature will not be homogeneously elevated to 170 F because ofits thermal inertia.
The concrete elevated temperature would be based on a thermal gradient occurring
- between the bulk pool temperature and the temperature at the outer surface of the
7 Page 9 of13 concrete pool enclosure. This gradient passes through the water / liner interface film resistance, the liner plate, any small air pockets between the liner plate and the concrete surface, and finally the massive concrete cross-section itself. Therefore, only a portion of any particular concrete section is expected to approach the temperature of 170 F.
c) The evaluation of the concrete pool structure considers the thermal gradient across sections, in accordance with the methods recommended by ACI-349, by considering the 170 F temperature in the pool as a normal condition. In fact, the pool structure evaluation also considers the more extreme accident case of pool boiling with an even greater temperature condition.
d) It is assumed that the "(ACI) Standard 379" referenced in the question represents a typographical error and is intended to refer to ACI 349, " Code Requirements for Nuclear l-Safety Related Concrete Structures (ACI 349).". Appendix A, Section A.4.1 states, "The
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following temperature limitations are for normal operation or any other long term period.
The temperatures shall not exceed 150 F except for local areas, such as around penetrations, which are allowed to have increased temperatures not to exceed 200 F."
This short section of the Standard does not include any detailed definitions for " normal operation" or "long term period". However, it is implied that normal operations are l
considered to occur over long term periods. The ACIimplied intention of normal -
operations appears to differ significantly from the conditions required to produce the temporary 170 F bulk pool temperature, therefore the 170 F bulk pool temperature results in concrete temperatures exceeding 150 F for an acceptable period of time (i.e.,9 days) and in only a portion of any concrete cross-section (i.e., local areas), as allo-d by the
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ACI Code.
Question 12:
Is there a procedure for performing the outage specific evaluation of heat load? What value is
- used for the maximum pool heat load?
Response 12:
L Procedure EDP-ZZ-00014, " Reload Design Control and Coordination," requires that a fuel cycle L
specific decay heat load calculation be performed. This calculation addresses SFP decay heat i
loads. A calculation was performed for Callaway Plant's current Fuel Cycle 10 using both the current limiting decay heat rate and the proposed new limiting decay heat rate of 63.41 MBtu/hr.
. Question 13:
Discuss sources and capacities of make-up water and the methods / systems (indicating seismic design category) used to provide make-up water, as well as the time needed to set up a path for make-up water to the SFP.
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Page 10 of13 Responst 13:
Technical Specification Surveillance Requirement 4.9.11 requires the water level in the fuel l
storage pool be determined to be at least its minimum required depth corresponding to 23 feet i
above the top ofirradiated fuel assemblies at least once per 7 days when irradiated fuel assemblics are in the fuel storage pool. Surveillance procedure OSP-ZZ-00001, " Control Room l
Shift and Daily Log Readings and Channel Checks" requires logging the level on daily logs. A.
l SFP low level alarm is received prior to a reduction in water level below the minimum Technical l
Specification required depth. The margin between the SFP low level alarm and the Technical Specification minimum water level will be maintained equal to or greater than the current margin of 23 inches.
1 Procedure OTN-EC-00001 provides two sources for make-up to the SFP depending on the SFP boron concentration. The normal source of makeup water is from the Reactor Makeup Water tank via the Reactor Makeup Water pumps. The second source of makeup water (as specified in this procedure) to the SFP is from the Refueling Water Storage Tank via the SFP cleanup pumps.
If a SFP low level alarm is received, procedure OTA-RL-RK076D, "SFP LEV HILO" provides for actions to restore level. If SFP level is stable, procedure OTN-EC-00001 is performed to restore level. If SFP level is decreasing, off-normal procedure OTO-EC-00001, " Loss of SFP / Refuel Pool Level" identifies four sources of makeup water for restoring level.
The table below provides information on the various sources of make-up water to the SFP.
WATER CAPACITY PUMPS SEISMIC Time to begin SOURCE (gpm)
(pump / piping) makeup (min. approx.)
Reactor Makeup 20 RMW pump No 30 Water (RMW) Tank Refueling Water 300 SFP cleanup No 30 Storage Tank pumps (RWST)
Recycle Holdup 100 Recycle No 90 Tank (RHUT)
Evaporator l
feed pump I
Essential Service 25 ESW pump Yes 30 i
Water (ESW)
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System (two redundant trains) l l
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Page 11 of13
. Question 14:
The February 24,1998, submittal stated that additional racks will be placed within the cask loading pit during a later campaign, When do you anticipate installing these cask loading pit racks? Would newly off-loaded fuel be placed in the cask loading pit? If not, what controls would be in place to ensure that newly off-loaded fuel would not be placed in the cask loading
- pit? How do you ensure that sufficient cooling is provided to the fuel located in the cask loading pit? -
Response 14:
i Based on current projections, the reracked SFP will provide sufficient capacity to allow a full core discharge until Refuel 24 in the fall of 2020 without the need for fuel storage in the cask loading pit. Callaway Plant does not anticipate the installation of racks in the cask loading pit until after the year 2015. Racks may never be procured to install in the cask loading pit if permanent storage options become available. Callaway Plant will use administrative controls to
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ensure that newly ofIloaded fuel is not stored in the cask loading pit. The administrative controls 1
will be documented in the form of plant procedures associated with special nuclear material control.
If spent fuel is stored in the cask loading pit it will be adequately cooled by the passive, buoyancy driven exchange of water with the SFP. The sufficiency of cooling provided to the cask loading pit is demonstrated by the three-dimensional computational fluids model, which i
- includes the cask loading pit and the interconnecting slot. The cask pit is assumed to contain the j
average background decay heat generation Computational fluid dynamics (CFD) calculations 1
indicate that the flow rate through the interconnecting slot is sufficient to completely replace all the water in the cask loading pit in less than one hour. The temperature contours through the l
interconnecting slot show no substantial difference between the pool bulk temperature and the -
cask loading pit temperatures.
Question 15:
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.With regard to the FLUENT 3D model, what are the model assumptions? Was a sensitivity
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8 study done with regard to the different physics and correlations options? Has the model been proven to be grid independent?
Response 15:
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The following assumptions (12 total) for the CFD modeling have been paraphrased from the appropriate local temperature analysis calculation package.
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The decay heat load is conservatively selected to bound earlier calculated decay heat load j
limits. This CFD evaluation is therefore performed for a higher heat load (65 MBtu/hr j
vs. 63.4 MBtu/hr) than is actually permitted.
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Page 12 of13 l
2.
The decay heat load contribution from the hottest 1/3 core in the SFP is conservatively based on a bounding fuel rod heat flux value. This causes the decay heat from the hottest i
fuel assemblies to bound any actual condition.
3 3.
The analysis uses the geometry of a Westinghouse 17x17 Standard fuel assembly. Of the fuel assembly types stored in the SFP, this fuel assembly has the lowest cell free flow l
area. This conservatively maximizes the hydraulic resistance and resulting local j
temperatures, j
4.
All storage locations are assumed to be 50% blocked at the top of the racks, thereby increasing the hydraulic resistance of the rack cells. This will conservatively bound any realistic cell blockage scenario.
5.
All pedestal cells, which have no baseplate holes and must therefore rely on the cell wall side holes, are assumed to be located together in the SFP, This highly conservative assumption creates a large flow restriction that affects a relatively large region of the.
SFP, substantially worsening the local flow and temperature fields in that region.
6.
The calculated hydraulic resistance parameters are conservatively worsened to bound any
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small deviations in fuel assembly geometry and rack construction tolerances. The two -
calculated parameters, permeability and inertial resistance factor, are worsened by 15%
and 50%, respectively.
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7.
The assemblies from the hottest 1/3 core are assumed to be located in the region occupied entirely by pedestal cells. This conservatively locates the majority of the decay heat load
)
L in the region with the highest hydraulic resistance. This is an extremely conservative assumption as the inertial resistance in this region is nearly twenty-times greater than that for non-pedestal cells.
I' 8.
The minimum bottom plenum height in the SFP is used for both the SFP and the cask loading pit, even though the cask loading pit bottom plenum is actually much larger.
This conservatively maximizes the hydraulic resistance and resulting local temperatures.
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9.
' Along each wall of the SFP and the cask loading pit, the minimum gap is used for all locations along that wall. Localized larger rack-to-wall gaps (i.e. near transfer canal l
gates) are neglected. This conservatively minimizes the downcomer area, thereby maximizing the hydraulic resistance and resulting local temperatures.
l 10.
The downcomer spaces between racks are neglected. This conservatively minimizes the I
downcomer area, thereby maximizing the hydraulic resistance and resulting local temperatures.
g 11.
Bounding minimum pool dimensions in the east-west and north-south directions are used instead of the nominal pool dimensions. This conservatively minimizes the downcomer-area, thereby maximizing the hydraulic resistance and resulting local temperatures.
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12.
The total peaking factor is applied to the heat flux in the fuel cladding temperature
. calculation. This conservatively increases the decay heat generation rates, thereby l
maximizing the resulting fuct clad temperatures.
With all CFD codes, the ability to select appropriate mathematical models to correctly evaluate physical phenomena is based largely on knowledge gained through firsthand experience in modeling the phenomena. The FLUENT code is no exception. Once an acceptable combination of mathematical models, boundary conditions and discretizing schemes are discovered, however, similar situations can be evaluated using the same combinations.
Holtec's thermal analysts have developed a substantial amount of expertise with the use of the FLUENT code to evaluate SFP thermal-hydraulic phenomena, particularly buoyancy-driven flows in porous media. The mathematical modeling options selected by the analysts reflect the experience gained through performing many analyses.
The computational burden associated with modeling of this class of problems is substantially increased by the range oflength scales involved.- While a typical pool has dimensions on the -
order of 40 feet, rack-to-wall gaps can range from a few inches to less than an inch. Because buoyancy forces dominate the flow field in a SFP, it is usually difficult to subdivide a large complex problem into several smaller, simpler ones. To ease the computational requirements to a
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reasonable level, deviation from discretization schemes typically recommended by CFD code l
user's manuals becomes necessary.
For the recently docketed (50-382) and reviewed Waterford Unit 3 license amendment request, l
an exhaustive independent review of the CFD modeling of SFPs was performed. This study, i
submitted to the NRC and approved by Amendment No.144 for Waterford Unit 3, performed a
)
sensitivity analysis of a Holtec CFD model which included storage of freshly discharged fuel in the Waterford cask loading pit. This study demonstrated that substantial computational grid.
refinement and the use of an alternate turbulence model Renormalization Group (RNG) had an insignificant impact on the peak local temperature result thus indicating grid independence. The
' independent review also recommended that a three-dimensional model would be more appropriate for evaluating fuel storage in the cask loading pit. The Callaway Plant CFD model is 3
three dimensional.
1 Tbn combination of the many conservative assumptions incorporated into the FLUENT analysis, l
the prior sensitivity studies and the experience of Holtec's analysts in using the code for SFP analysis provides significant confidence in the results of the analysis.
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