ML20054J693
| ML20054J693 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/23/1982 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| LIC-82-242, NUDOCS 8206290456 | |
| Download: ML20054J693 (21) | |
Text
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Omaha Public Power District 1623 HARNEY 8 OMAHA, NEBRASMA 68102 e TELEPHONE 536 4000 AREA CODE 402 June 23, 1982 LIC-82-242 f1r. Robert A. Clark, Chief U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Licensing Operating Reactors Branch No. 3 Washington, D.C.
20555
Reference:
Docket No. 50-285
Dear f1r. Clark:
Fort Calhoun Station Spent Fuel Pool Storage Modification The Commission's letter ta the Omaha Public Power District, dated May 20, 1982, requested the District provide additional information regarding the proposed high density spent fuel pool storage rack modi-fication at the Fort Calhoun Station. The District's response is attached.
Si erely,
/
kk
- [W.C. Jones l
r Division flanager Production Operations l
Attachment cc:
LeBoeuf, Lamb, Leiby & MacRae col l
1333 New Hampshire Avenue, N.W.
l Washington, D.C.
20036 l
8206290456 820623 DR ADOCK 05000285 PDR
s Omaha Public Power District's Response to ONRR Letter Dated May 20, 1982 Item 1 Provide the results of an accident analysis which addresses the effects of a vertically dropped fuel bundle which ends up laying horizontally on the i
spent fuel storage racks. The analysis should be similar to the accident analy-sis provided for a fuel bundle being dropped adjacent to the storage rack and should address criticality, reduction in cooling of the assemblies in the rack, and damage to the fuel in the rack and to the dropped fuel assembly.
Response
Various drop accident analyses were performed which address the effects of dropping a fuel assembly into the spent fuel pool (SFP).2 Postulated drop accidents included:
- 1) a straight drop on the top of a rack, 2) a straight drop through an individual cell all the way to the bottom of the rack, 3) an inclined drop on the top of a rack, and 4) one fuel assembly impacting another fuel assembly already in a storage cell. The effects or consequences of a vertically dropped fuel bundle (ascembly) that ends up lating horizontally on the spent fuel pool storage racks are encompassed by postulated accidents #1 and #3 above.
These postulated load drops were reviewed under mechanical, material, structural, criticality and radiological disciplines. The results of these accidents analyses are as follows:
In case 1, there would be structural damage to the top of the cell walls, however, deformation would be limited to 1.64" and the stored fuel assemblics.are nominally 14" down inside the c, ells,.thus the stored fuel assemblies would be", unaffected by this' type' ~ ~f ac'cident'. In case'2, the bottom of'the o
dropped individual fuel cell would fail, however, this kind of an accident is considered highly improbable. This accident would render one storage position unusable.
A direct impact on the fuel pool liner would envelope the radiological consequences of this accident and the consequences have been determined to be within 10CFR 100 limits.2 The radiological consequences due to this postulated load drop would be the same as dropping a fuel assembly directly onto the spent fuel pool floor, which has been previously analyzed as part of the FSAR.
In case 3, an inclined drop of a fuel assembly on top of the rack could cause deformation of a number of lead-in guides.
However, since the fuel assemblies are stored 14" below the top of the lead-in guides, there would be no damage to the stored assemblies.
In case 4, the results have shown that if all the rods were to fail in both assemblies, the total activity released would be less than fuelassemblies,72hoursaftershutdown.gactoftwoofthehighestburnedup the 10CFR 100 limits. This assumes an im 1
Any one of these four postulated accidents would not significantly increase the pool's K factor and there would be no reduction in the ability to cool the spentfu$kgassemblies in the racks.
Item 2 Regarding the use of the Shutdown Cooling system to cool the spent fuel pool, provide, the following:
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4 a.
The licensee states that the shutdown cooling system can be aligned to
_ provide cooling for the spent fuel pool 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after receipt of a high pool temperature alarm. The licensee _did not specify the condi-tion of the reactor (operating mode) during the time the shutdown cooling system is aligned for spent fuel cooling. Verify that the reactor will be in -cold shutdown prior to alignment of the shutdown cooling system for spent fuel pool cooling.
Response
Aligning the shutdown cooling system to provide alternate or supplemental cooling for the SFP is only necessary under full core discharge conditions and then only to maintain the SFP temperature below 140*F, if necessary. Realistically, several days will elapse before commencement of the full core transfer to the SFP, and then the acgnitude of the spent fuel decay heat load would be such the.t the SFP cooling Lystem alone could maintain the SFP temperature below 14007.
Nevertheless, based upon the above conditions which might necessitate this sup-plemental cooline, the reactor would be in a cold shutdown condition.
Item 2 b.
Until the shutdown cooling system is aligned for spent fuel pool cooling, the pool temperature will increase as the result of inadequate pool cool-ing. No statement was made in the licensee's submittal concerning the high temperature alarm setpoint or the maximum pool temperature until adequate cooling is restored. Our concern is with the effects of the increasing temperatures in the. cleanup system.and the. increase in airborne radio-
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' activity with' respect to offsTte'dosd.""S'picify the high temperature ala'rm
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setpoint. Using a heat load based on NUREG-0800, Standard Review Plan, Section 9.1.3 and Branch Technical Position ASB 9-2 specify the peak pool temperature and provide a discussion on the effects of this temperature on the spent' fuel pool cleanup and the site boundary dosage.
Response
The high temperature alarm setpoint for the spent fuel pool is 150*F.
Following the unlikely loss of the spent fuel pool cooling system, the pool heat up rates would be 6.27*F/hr for normal refueling and 15.40*F/hr for full core discharge conditions.to The decay heat loads utilized for this analysis were based on NUREG-0800 criteria.
Provisions are presently being made to provide a permanent piping tie-in between the shutdown cooling system and the spent fuel pool cooling system to provide supplemental Spent Fuel Pool cooling. ' This permanent system can be aligned within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, instead of the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> required for the temporary hose connections as detailed in the District's March 12, 1982 letter. Conservatively assuming the SFP bulk temperature is 120 F at the initial loss of the SFP cooling system, since the SFP cooling system maintains the pool temperature at or below 120*F during normal refueling conditions, the SFP would reach 132*F af ter 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Similiarly, for full core discharge conditions, commencing at a pool temperature of 140*F,- the SFP temperature would reach 171*F after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. These results are conservatively based on decay heat loads at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown of 11x106 BTU /hr and 27x10 BTU /hr for refueling and full core discharge conditions, 8
respectively.. _ _ _ _ _
The spent fuel pool cleanup system should only be used if the SFP temperature is maintained at or below 150 F, as above this temperature the demineralizer and filter could be damaged. However, the cleanup portion of the spent fuel pool cooling system can be isolated from the remainder of the system to preclude damage to the resins. Assuming the loss of the SFP cooling system, together-with a full core discharge, the SFP cleanup' system would only have to be iso-lated by the-operators for a few hours as the pool temperature will be brought below 150 F several hours after the initiation of the alternate SFP cooling.
The airborne radioactivity,- with respect-to offsite doses, is not expected to.
increase significantly during the two hours that pool cooling is not available.4'8 Item 3 The licensee stated in his submittal-dated March 12, 1982 that the spent fuel pool temperature would be maintained below 120*F.
The licensee did not use NUREG-0800, Standard Review Plan,.Section 9.1.3 and Branch Technical Position ASB 9-2 for calculating the decay heat loads.
Consequently, we are not sure how much conservatism is in the licensees analysis. Therefore provide the following information with the heat exchangers expected fouling factor and pluggage factor-for the life of the plant:
The pool temperature with the maximum normal heat load using the heat load a.
based on NUREG-0800.
Response
The maximum pool temperature utilizing the maximum normal heat-load base.d.on 6
NUREG-0800 criteria (approximately 11 x 10 BTU /hr.) would be 114~.5*F, which is less than the Technical Specification allowable limit of 120 F.
Item 3 b.
The pool temperature with the maximum normal heat load using the heat load:
1.
based on NUREG-0800 and the worst single active failure. Provide a dis-cussion'of the failure and the operation of'the remaining portion of the system.
Response-8 The. maximum SFP temperature,-utilizing the maximum normal heat load (11 x 10 BTU /hr) and assuming the worst single active failure of the SFP cooling system, would be 114.5*F.
This maximum temperature is the same as that detailed in the response above because there is no single active failure of the SFP cooling system that would prohibit the proper cooling of the SFP during normal or refuel-ing conditions. All necessary active components of the SFP cooling system have redundant components that will provide.for proper operation and preclude adverse effects of a single active failure.
U Item 3 A discussion of the capability and procedure to remove the spent fuel pool c.
cooling system heat exchanger from service for tube cleaning, tube plugging or retubing. The spent fuel pool cooling system consists of two pumps and i
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a single heat exchanger.
Include in the discussion of the time available to perform these tasks without exceeding any pool' temperature alarm setpoints.
Response
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The District believes the cleaning, plugging or retubing the SFP cooling system heat exchanger will never be necessary since the subject heat exchanger has been designed for a 40 year service life and no problems have been identified to date. Therefore, the District does not presently have a specific procedure for correcting any of these postulated problems.
In the unlikely event that it ever becomes necessary to perform such maintenance, the District would probably com-plete the work during a refueling outage, implement a procedure that would encompass all safety concerns and is approved by the Plant Review Committee, and provide means for alternate SFP cooling to control the pool temperature.
Item 4 Provide the results of an analysis of tl.e effects of dropping light loads onto stored spent fuel. The analysis should verify that the maximum potential kinetic energy contained in all objects of less weight than a spent fuel assembly and its handling tool which will be handled over the spent fuel in the storage racks will not result in fuel damage and a corresponding offsite dose in excess of that determined in the design basis fuel handling accident in the FSAR.
Response
See the District's response to Item 1 above.
Item 5
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a generic letle'F 6as s'ent to S11 licensies which provided On April 14, 1978
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guidance concerning the information to be provided by the. atility when requesting spent fuel pool modifications for the purpose of increasics the number of fuel bundles to be stored in the pool. The licensee submittal did not contain all of the information requested by the generic letter.
Therefo re, provide the following information.
With respect to Section 1.2, verify that no combination of events and/or a.
failures will result in a K f the spent fuel storage arrangement of eff greater than.95.
Response
This was verified in the District's March 12, 1982 letter, page 21, paragraph 3.8 Item 5 b.
Provide a discussion of the onsite tests which will be performed to confirm the presence and retention of the neutron absorber in the racks. The re-sults of the verification tests shall show within a 95% confidence level that there is sufficient amount of absorber to maintain K,ff at or less than.95. _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ - _ _
Response
The prese'nce of the neutron absorber (Boraflex ) is inherent in the design of the racks and will be completely documented by Quality Assurance personnel during construction.
Final verification will be performed by a neutron attenu-ation test on 15% of the cells in each storage rack, selected at random. A neutron source and ja plicable detectors will be used to verify that the two sheets of BoraflexWare in their correct position adjacent to the Aell walls.
Should this test program identify the absence of even one BoraflexW sheet, then all cells will be tested and verified. Verification of the retention of the neutron absorber will be accomplished by the Surveillance Test detailed in the response to Item Sc below.
Item 5 c.
Provide a discussion of the periodic surveillance testing to verify the continued presence of a sufficient amount of neutron absorber in the racks to maintain K below or at.95.
The testing should be performed on a eff statistically acceptable sample size. The frequency of testing should be specified.
Response
A testing procedure similar to the one used for Point Beach Nuclear Plant (see ) will be instituted and testing will be conducted every refueling outage. A sample cell will be tested to demonstrate the presence of a sufficient amount of neutron absorber in the racks to maintain K,gg at or below <.95.
Item 5 d.
Provide a fuel handling accident analysis which includes a quantification of the drop parameters.
Postulated fuel drop accidents must include a straight drop on the top of a fuel bundle in the rack, a straight drop through an individual cell all the way to the bottom of the rack and an inclined drop on the top of the rack. A discussion of the integrity of the racks, fuel pool, and the fuel pool liner as well as the amount of radiation released (compared to the FSAR design basis fuel handling accident) should be included.
Response
See the District's response to Item 1 above.
Item 5 Verify that the seismic excitation along the three orthogonal directions c.
are imposed simultaneously for the design of the new racks.
Response
Seismic excitations along the three orthogonal directions were imposed simulta-neously in the design and analysis of these SFP racks. Please see the District's letter dated March 12, 1982, page 65, paragraph 2 for further discussion.2 3 __
Item 6 Verify thst the seismic loads imposed on the fuel pool liner walls does not
-result in any damage to the liner such as to cause 1) significant releases of radioactivity due to mechanical damage to the fuel, 2) significant loss of water from the pool which could uncover the fuel and Icad to release of radioactivity due to heatup, 3) loss of ability to cool the fuel due to flow blockage caused by a portion or one complete section of the liner plate falling on top of the fuel rack, 4) damage to safety related equipment as the result of pool leakage and 5) uncontrolled release of significant quantities of radioactive fluids to the environs.
Response
The SFP liner is designed to support all postulated or effective dead and live loads, including hydrostatic loads, temperature gradients to 212*F and the effects of a design basis earthquake (DBE). Drainage grooves are provided behind the stainless steel liner which permit detection of any liner leakage.
As per Sections 5 and 9.5 of the FSAR, the liner is designed to withstand these seismic loads without failing, and thus, no postulated accident could damage, uncover the fuel or result in radioactive releases in excess of the 10 CFR 100 limits.
The new spent fuel storage racks have been designed not to exceed the spent fuel pool liner plate allowable unit floor loading of 500 psi. Analysis has verified that the unit floor loading imposed by all dead weights, together with the com-bined vertical components of seismic accelerations as determined by the SRSS method, will not exceed the 500 psi limit.2 Thus, the seismic loads imposed on the fuel pool liner walls will not result in significant damage to the liner and there would be no significant releases of radioactivity from the pool liner due to mechanical damage to the fuel; there would be no significant losses of water from the pool which could uncover the fuel, since the liner is designed to support all loads from a maximum credible earthquake; there would be no loss of cooling to the fuel due to flow blockage caused by sections of the liner falling on the racks, since the liner is designed to withstand a DBE; damage to safety related equipment as the result of postulated pool leakage would not occur since no leakage is anticipated; there is no safety related equipment within the postulated leakage area; and uncontrolled releases of significant quantities of radioactive fluids to the environs would not occur, since the liner plate is designed to prevent leakage during the most severe accidents.8 Item 7 Provide the maximum uplift forces imposed by the spent fuel handling crane including the consideration of these forces in the design of the racks and the effects on attachments to the pool liner.
Response
The maximum uplif t forces imposed by the spent fuel handling crane would be approximately 2 kips. The impact of these forces was considered in the design of the fuel racks and on the attachments to the pool liner, and was determined not to be a problem.1,2 l
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l Item 8 In the submittal dated March 12, 1982, the storage of control rods was not addressed. Our concern is if the control rods were stored in the fuel bundle in the pool, then the additional weight might have a significant effect on the seismic analysis.
Verify whether or not more than one control rod at any time will be stored in the spent fuel pool.
If control rods will be stored in the spent fuel pool, verify that the seismic analysis provided in the submittal represents the maximum pool liner loadings on the walls and floor, and the maximum inter-rack reactions with consideration being given to the additional weight of the control rods.
Response
No control rods are presently stored within the Spent Fuel Pool. However, control rods might be stored in the pool during future refueling outages. The SFP racks have been designed assuming control rods would be stored within fuel bundles stored in the spent fuel pool and the mechanical and seismic analyses were based on storing control rods in the racks.2,3 Thus, compliance with the seismic analysis provided in the District's March 12, 1982 letter is still valid and the maximum pool liner loads on the walls, floor ahd on the maximum inter-rack reactions have been previously evaluated and determined to be acceptable.
Item 9 In July 1980 NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," was
, published which requires a_ generic review of the methodology used in routine moving of heavy loads. A heavy 16ad is defined'as any losd which weighs more than a fuel bundte and its handling tools. The re-racking of a spent fuel pool is not considered routine and therefore is not within the scope of NUREG-0612.
Concerning the moving of spent fuel storage racks we have the following questions:
Verify that the special lifting devices for the removal of the existing a.
spent fuel storage racks and the installation of the new spent fuel storage racks meet the criteria of ANSI N14.6-1978 or ANSI B30.9-1971 for non-special lifting devices.
Response
Special lifting devices, if required, will meet the criteria of ANSI N14.6-1978.
Non-special devices used for the removal of the existing spent fuel storage racks and the installation of the new spent fuel storage racks presently meet the criteria of ANSI B30.9 - 1971 as detailed in the District's letter dated June 22, 1981.
Item 9 b.
Verify that procedures are developed, which include the safe load paths, for the removal of existing spent fuel storage racks and modification of new spent fuel storage racks. Verify that all safe load paths for these operations are clearly marked on the floor.
Provide drawings which show the load paths of the new and existing spent fuel racks and all other heavy loads associated with this modification and the spent fuel pool. -
Response
Special Procedures for the removal of the existing racks and installation of the new racks will be written to provide adequate guidance in the handling of the racks.
However, marking of the individual load paths on the floor will not be provided since all areas within the travel of the Auxiliary Building Crane are considered safe, except for the region of the Spent Fuel Pool. Travel is pro-hibited over the SFP, unless specific handling procedures have been approved by the PRC and a crane supervisor is present for bypassing the interlocks.
In the District's letter dated June 22, 1981, in response to NUREG-0612 Section 2.1, we have defined safe load paths (for all areas of crane travel except over the Spent Fuel Pool), by exclusion. Thus, no individual load paths or detailed drawings will be provided.
Item 10 The information provided in the licensee's submittal dated March 12, 1982 did not include a discussion of the capability of the component cooling water system and the raw water system to remove the additional heat from the spent fuel pool.
Based on the heat loads using NUREG-0800 Standard Review Plan, Section 9.1.3 and Branch Technical Position ASB 3-2, provide the results of a revised FSAR analysis which shows the increased heat loads from the spent fuel storage expansion on the component cooling water system and the raw water system.
Include information which shows the design heat load capacities and the imposed heat loads for normal operation with normal refueling and for all design basis accident heat loads. For each system, include an analysis of the capability of the system to remove the new spent fuel cooling heat loads and all normal and accident heat loads while maintaining the original design margins for tube fouling.and plugged.
No single failure should prevent proper spent fuel pool cooling or safe shutdown.
No credit can be taken for any redundant train or component which is not properly qualified for the accident being considered, such as a safe shutdown earthquake, or which requires operator action within 20 minutes (30 minutes if the single operator action is required outside of the control room).
Response
A copy of Table 9.7-2 from Section 9, " Auxiliary Systems" of the updated FSAR is attached and identifies the increased heat loads from the spent fuel storage expansion on the component cooling and raw water systems. Design load capacities, heat loads during normal operation and normal refueling, and all design basis accident heat loads are also detailed in Figure 9.7-2.
The maximum heat load increase from the SFP expansion was calculated to be 2 x 10 BTU /hr for normal refueling conditions; :.hus, the total heat load to be 6
6 removed by the component cooling water system will increase from 28.86 x 10 6
BTU /hr to 30.86 BTU /hr. This additional 2 x 10 BTU /hr does not exceed the Component Cooling Water (CCW) System design basis since it is designed to remove 36.3 x 10 BTU /hr during normal operations (with 3 out of 4 heat exchangers in 6
use). The use of the fourth heat exchanger would increase the heat removal capacity to 48.4 x 10 BTU /hr.
However, each CCW heat exchanger is designed to 6
i remove 136 x 10 BTU /hr during accident conditions, for a total heat removal 6
l capability of 536 x 10 BTU /hr with all four heat exchangers operational.
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The maximum heat load generated in the SFP from a Full Core Discharge was calcu-6 lated to be approximately 27 x 10 BTU /hr. This heat load is 16 x 106 BTU /hr greater than the heat load obtained during normal refueling conditions. Based upon the heat removal capacity identified above for the CCW system, the additional 16 x 106 BTU /hr can sufficiently be removed by the present system. This system's design and analysis has encompassed the design margin for heat exchanger tube fouling and also provides adequate margins for single failures.
The Raw Water System (RWS) also has the capacity to remove the increased heat 4
load generated by the storage of the additional spent fuel. The RWS is designed to remove all of the heat load handled by the Component Cooling Water System, 6
thus it can also remove up to 536 x 10 BTU /hr. Therefore, all of the original e
design margins are fulfilled for both systems, including tube fouling effects.
No single active failure can prevent proper spent fuel pool cooling since all systems have sufficient redundancy to mitigate the failure consequences.
Item 11 Outline by major tasks, the methods to be used in the pool modification. You should specify the man-hours and average dose rate for each task and the expected total man-rem for the entire pool modification.
Response
i Pool modification consists only of replacing, in a predetermined sequence,' the presently installed sp'ent fuel storage racks wi.th the new high density.xacks by~
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s' imply lifting out'the'p're'sent'onds" add l'wering the new bhes in ' lace.
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Spent fuel assemblies must be moved during this total operation in order to sequentially empty the row of racks to be removed and store the fuel in the rows of new racks after installation. Although the total inventory of stored spent fuel assemblies will have to be moved, only approximately 44 of these assemblies will be fuel that was discharged since the last reactor outage. The total accumulated dose during these fuel handling operations is expected to be of the l
same order as that experienced during a normal refueling. Based on the 1980 t
refueling outage, the total dose for operations in the fuel pool area was.32 l
man-rem, and therefore, the dose due to fuel handling'during rack installation would be expected to be less than 1 man-rem.
Any significant increase in personnel dosage due to this modification beyond that incurred during regular -efueling outages would be from the rack deconta-mination process. A conservative estimate is that any increase will be consider-ably less than that experienced dering the changeout of the SFP filter. Using one-third of the dose estimate for ft? ter changeout of.6 man-rem for 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> elapsed time, results in a dose rate of.615 _au-rem /rwar-hour. Assuming 6 man-hours to decontaminate one rack, the total dose for 21 rada would be 1.89 man-rem.
Adding this estimate to the dose due to fuel handling operations (less than 1 i
man-rem) results in a total dose for the entire pool modification of less than 3 i
man-rem.
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Item 12 Outline the actions that will be taken to assure that occupational doses during each task of the pool modification will be ALARA.
Response
1.
A health physicist will be present at all times during the re-racking to monitor for excessive airborne or high radiation by utilizing portable or hand-held radiation monitoring instruments.
2.
Area Radiation Monitors will be used to alarm on a high radiation signal.
3.
Personnel shall be required to wear appropriate clothing as determined by the health physicist to preclude contamination.
4.
As the racks are pulled out of the water, they will be washed.
5.
All rack decontamination areas will be enclosed by a cover-all to reduce airborne contamination.
6.
The District is considering the possibility of also deconning the racks in the pool by using chelating agents, along with agents to dissolve surface films and remove surface crud. None of the chemicals which will be added to the pool will initiate stress corrosion cracking or change the pH signifi-cantly.
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- 7., All personnel wiil be requ'iYed ta undergo 2 days of ihdiation training. _.
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Elem 13 Providt an estimate of the increase in annual man-rems from more frequent chang-ing of the demineralizer resin and filter cartridge estimates.
Response
l The installation of the new fuel racks will have a negligible impact on the annual man-rem accumulated, since the filter and demineralizers are changed only at refueling outages. This change out results in a total exposure of approxi-mately 0.6 man rems every 1-1 years.
Please see the District's letter dated March 12, 1982, page 68, Section 8.1 for further discussion.
Item 14 Identify the monitoring system that will be used, and its location in the spent fuel pool area, that would warn personnel whenever there is an inadvertent increase in radiation levels that would trigger the alarm set point.
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Response
The spent" fuel pool area is monitored by an area monitoring system. Actual dose rates can be read locally and in the control room. The monitor will alert personnel in the SFP area when dose rates exceed the setpoints. This monitor, RM0-85, is located outside the SFP wall as shown on Figure 11.2-3, attached.
Health Physicists carrying portable monitoring instruments will also provide a further means to identify increases in radiation levels. Additionally, a radia-tion monitor is located on the spent fuel handling machine to indicate dose rates in this area.
All effluents from this area are discharged through the auxiliary building stack, via the auxiliary building IIVAC system and are monitored through gaseous effluent and particulate monitors RM0 60/61/62.
In the event that these gaseous effluent monitors indicate activity levels are in excess of Technical Specifica-tion limits, the auxiliary building flow paths are manually closed. The exhaust ventilation ductwork from the spent fuel storage area is equipped with iodine absorbers which are manually brought into operation whenever irradiated fuel is being handled.6 Item 15 Describe the methods used to preclude spent fuel pool water from overflowing onto the spent fuel pool area floors.
Response
' A' level' indicator'is p'rovided'with'in the# iool"tB monitor spent fuel po"ol water
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level and alarms both locally and in the control room upon reaching a high level setpoint. Additionally, the SFP contains a curb or lip at the top of SFP which will preclude minor overflowing or spilling onto the floor.
Item 16 Specify the present dose rate in occupied areas outside the spent fuel pool concrete shield wall and provide an estimate of the potential increase of this dose rate if the space between the spent fuel and inside concrete shield wall is reduced due to the modification.
Response
The space between the spent fuel pool and inside concrete shield wall will not be reduced due to this modification.
In fact, this distance will be increased at all locations and accordingly, the dose rate outside the concrete wall is not expected to increase beyond present levels.1 i
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REFERENCES 1.
Des ~ign Criteria, Minimum Pitch Poison Design Spent Fuel Storage Racks, for Fort Calhoun Station Unit One, prepared by Wachter.
2.
Mechanical Analysis, Minimum Pitch Poison Design Spent Fuel Storage Racks, for Fort Calhoun Station Unit One, prepared by Wachter.
3.
Seismic Analysis, Minimum Pitch Poison Design Spent Fuel Storage Racks, for Fort Calhoun Station Unit One, prepared by Wachter.
4.
Thermal and Hydraulic Analysis, Minimum Pitch Poison Design Spent Fuel Storage Racks, for Fort Calhoun Station Unit One, prepared by Wachter.
5.
WAI Surveillance Testing Program, Minimum Pitch Poison Design Spent Fuel Storage Racks, for Fort Calhoun Station Unit One, prepared by Wachter.
6.
F.S.A.R. Fort Calhoun Station Unit One Sections 5, 9.5, 9.7, 9.8, and 11.2 7.
Radiological Analysis, Minimum Pitch Poison Design Spent Fuel Storage Racks, for Fort Calhoun Station Unit One, prepared by Wachter.
8.
Criticality Analysis, Minimum Pitch Poison Design Spent Fuel Storage Racks, for Fort Calhoun Station Unit One, prepared by Pickard, Lowe & Garrick.
9.
OPPD Response to Sections 2.1 and 2.2 - 2.4 Enclosure 3 NUREG-0612
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10.' " Spent Fuel Poo'l'Hea't Up Rates";- CAlculit' ions perfo'rmed by 'OPPD','
une 10,1982.
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11.
OPPD Response to Sections 2.2 - 2.4 NUREG-0612, " Control of Heavy Loads".
e I
i TABLE 9.7-2 COMPONENT COOLING WATER SYSTEM HEAT LOADS (Btu /hr x 106) 3.6 Hrs.
27.6 Hrs.
Post-DBA Post-DBA Normal Operation after Shutdown after Shutdown Cond. A*
Cond. B**
Shutdown Cooling Heat Exchangers 122.00 37.00 117.50 Letdown Heat Exchanger (balanced letdown) 1.54 Storage Pool Heat Exchanger 11.00***'
S.I. Tanks Leakage Coolers 2.40 Containment Air Cooling Coils 4.920 :j 420.00 R. C. Pump Seal and Lube Oil Cooling 3.15 i
3.15 3.15 RWDS Process Coolers 7.50 Primary Plant Sample Coolers
.45
.45
.45 Charging Pump Coolers
.30
'l
.30
.30
.30
.30 CEDM Seal Area Coolers
.30
.30
.30 S.I. and Spray Pump Coolers l
.15
.15 1.05 1.20 Control Room Air Conditioning
.30
.30
.30
.30
.30 30.86***'
126.65 41.65 421.65 119.30
- Post-DBA Cond. A - Only containment air recirculation, cooling and iodine removal system coils in operation.
During the ipitial phase of a DBA, containment spray will initiate; however, component cooling water to the shutdown cooling heat i
exchanges is only supplied upon receipt of a Recirculation Actuation Signal (RAS) which occurs later in the accident, when air cooling loads are significantly reduced.
- Post-DBA Cond. B - Only containment spray system, in recirculation mode, in operation.
- Changed due to increased Spent Fuel Storage
ATTACHMENT 1 L
WACIITER ASSOCIATES, INC.
54@0 W8LLIAM FLYNN HjCHWAY h j CIDSONIA, PENN F'l.VANI A 15044 (4-4 443 7590
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SURVEILLANCE TEST PROGRAM
~
FOR BORAFLEX I NEUTRON. ABSORBER USED IN SPENT FUEL RACKS AT
~ POINT BEACH NUCLEAR PLANT Prepared For WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT WEP-T-152 Apr3.1 20,
. 1979 Rev. 1, 7/12/79
- 4. #
Rev. 2, 8/7/79 Prepared By /- [,,d. [q, f{,
W./J.' Wachter
.. i 7/12/79 9
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0 SURVEILLANCE TEST' PROGRAM FOR BORAFLEX I NEUTRON ABSORBER USED IN SPENT FOEL RACKS AT.
POINT BEACH NUCLEAR PLANT
~
1.0 PURPOSE This test program is desi'gned to provide an on-going
~
proof test of the adequacy of the Boraflex I material in the spent fuel pool environment.
. 2.O FACILITY AND EQUIPMENT DESCRIPTION
.s.
A test sample will" consist of a 1-7/8" square by 0.1" piece.of Boraflex I clad with. 02" thick 304 stainless steel.
The
~
top and bottom edges of the sample are' vented.
Thirty of these samples are mounted on the back of a channel, see Figure 1.
The channel is attached inside a standard I
' poison assemb1y
,,.see;WAI. drawing.128-16.,
~
The surveillance sample assembly is, placed in the pos-
.ition agreed upon by WEP and WAI.
It will be examined l
and installe'd at Point Beach; see WAI drawing 128-20 for location.
3.0 TEST DESCRIPTION
~
The surveillance sam'ple train, Figure 1, will be install'ed in the predetermined poi' son. assembly.
During the first
+
refueling after the installation of the new high density racks, two of the hottest (maximum burnup) fuel assemblies removed from the reactor will be placed on each side of the surveillance sample train.
At each r.efueling, the fuel assemblies.on each side of the surveillance specimens n
- - - - - - - ~ - - - - - - -
will be replaced by two new hot fuel assemblies from the new batch of spent fuel.
The fuel in this position will always be fresh assemblies while all,the other poison assemblies will probably be exposed to freshly irradiated fuel only once every 15 years.
Thus the.
test sampics will receive the most irradiation exposure rads gamma,'while the average region of up to 2 x 1011 in the. spent fuel pool will receive about 2 x 10 rads.
1' At the end of the first five years, the surveillance
' samples will have received the normal irradiation ex-posure expected for the poison material.
This test program will test the Boraflex I. installed in the north pool and the Boraflex I installed in the south pool one year later.
~
4.0 SM1PLC REMOVAL The poison assembly (TP-1) containing the sample t. rain is taken out of the pool and the sample train removed.
Three adjacent. samples are cut fron the bottom end of the' train for testing.
The train is reinstalled and the poison assembly reinstalled.
5.0 SAMPLE EXM1INATION Representative samples of the poison material used in the Point Beach sp'ent fuel storage racks'have been tested at'the University,of Michigan; all the critical propcrties have been determined and are listed in Brand Industrial Services Inc. Report 748-10-2.
The surveillance samples will be tested sufficiently to verify that they are conforming to the performance of..
_~~~~"~~~:=~.
3O'%W WW'
tha raaterial in the University o'f Michigan Test Program.
5.1 BASELINE MEAsunEMENTS r
The string of 30 samples will be examined prior to in-stallation iri the test assembly.
The following data will be recorded for each sample (see Figure 2).
1.
Phys.i, cal Dimension - 9 thickness measurements
,3 length measurements
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3 width measurements 2.
Hardness in Shore A Durometer at s,ame point as the
^
thickness me,asurements o
3., Weight 4.
Neutron attenuation measurement
- In addition, nine thickness measuremhnts Will be taken of each sample after it is encased in stainless steel.
The control material (archive material) will be stored
. at the Point Beach Nu'elear Plantl I
.~
^.-
J 5.2 MEASUREMENTS AT END OF SAMPLE LIFE 8
Each sample will be examined at end of life.
The following data will be recorded:
l '.
Physical D'imension per ASTM-D-1042 or equivalent Before cladding'is removed - 9 thickness measurements..
After stripping cladding
- 9 thicknes's measurements
- 3 length measurements
- 3 width measurements
/
'2.-
Weight 3.
Hardness in Shore A Durometer per ASTM-D-2240 or equivale~t ',
n 4.
Neutron attenuation measurement *'
- Phoenix. Memorial Laboratory, University of Michigan -
liev. 2
5 N
P-1 Poison Assembly
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- 110 x 2" SQ Boraflex
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FIGURE 1 4
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DATE
....... SUDJECT_
SHECT NO.-
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_ DATE. _
-_ ___._._ _ PrIOJ. N O. -
WEP-T-152
- l. Sample.No j i
. 2. Date _-
3/16" 13/4" 3/4" 3/16"
- 3. Measurement By:
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- 4. Amount of Irradi;
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