ML20056B391
| ML20056B391 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/16/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20056B390 | List: |
| References | |
| NUDOCS 9008280219 | |
| Download: ML20056B391 (8) | |
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UNITED STATES '
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NOS. 43 AND 8 TO FACILITY OPERATING LICENSE NOS. NPF-39 AND NPF-85 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-352 AND 50-353 1.0' INTRODUCTION By letter dated June 22, 1990, Philadelphia Electric Company (the licensee) requested an amendment to Facility Operating License Nos. NPF-39 and NPF-85 for the Limerick Generating Station, Units 1 and 2.
These proposed amendments would revise the Surveillance Requirements (SR) of.
the Technical Specifications (TSs) for the refueling platform main and auxiliary hoists to more accurately reflect their actual use.
The following provides a general description of the refueling platform and associated hoists and the proposed changes for the SRs.
The refueling platform is used to transport fuel and reactor core components to and from the refuel floor pools and. cavities, and as a work platform from which underwater activities can be conducted.
The refueling p.latform has three hoists through which many of these activities are-accomplished.
The. main hoist assenibly is suspended from a trolley system on the forward side of the. platform and is used for transporting and orientating fuel assemblies and control rod guides for reactor, storage rack, and shipping cask (fuel assemblies only) placement.
Two auxiliary hoists, each with a 1000 pound operating capacity are provided on either side of the platform.
These hoists are used to perform non-fuel core component activities involving core power monitors, control rods, control rod guide tubes, fuel support casting, neutron source holders, and general servicing aids.
Procedurally, the main hoist is required to be used for the handling of the fuel' assemblies or control rod guide tubes.
During the transfer of fuel assemblies and double control rod guide tubes between the reactor vessel and the spent fuel puol, a potential currently exists for component contact with pool / cavity structures (e.g., portable refueling shield) due to lack of clearance.
This could cause equipment and/or carried component damage.
Therefore, the licensee proposes to change the SRs for the main hoist to allow the normal up stop limit switch to be repositioned no more than 6 inches higher to provide more clearance between a main hoist grapple-carried component-and pool / cavity structures.
This will maintain not less than 8 feet 0 inches of water over the top of active fuel with the pools at normal water level, which will correspond to approximately 6 feet 6 inches of water above the top of the carried fuel assembly.
Also, the licensee 9008280219 900816 DR ADOCK 050 2
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proposed to clarify the main hoist SRs to remove the reference to control rods, since the main hoist is not used for the handling of control rods and. add the phrase "not less than" before the uptravel stop distance.
Also, the proposed TS changes will remove'the requirement for'a fuel loaded auxiliary hoist' interlock by prohibiting the lifting of a fuel assembly with the auxiliary hoist, and also permit less water above the top of a carried component.
Part of the proposed auxiliary haist TS change will clarify the requirements by adding the phrase "not less_than" before the uptravel stop distance.
2.0 EVALUATION Wher. handling irradiated fuel, the radiation dose rates external to the pool surface are highly dependent upon the time interval from reactor shutdown.
TS.Section 3.9.4, " Decay Time," requires the reactor to be subcritical for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to movement of. irradiated fuel in the reactor.
This requirement ensures sufficient time has elapsed to allow for radioactive decay of the short lived fission products.
With the pools at normal water level, the proposed six (6) inch reduction in water shielding (for the main hoist SR) in combination with the handling of a spent fuel assembly 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown, would raise expected radiation dose rate levels at the pool surface from 10.6 millirem / hour to 24 millirer/ hour.
The higher radiation dose rate is still well within the radiation zone designation for the refuel floor pool area (Radiation Zone IV, i.e., <100 millirem / hour, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter reactor shutdown).
Due to the complexity of the activities required to be. accomplished to ready the refuel floor and equipment for core component handling, fuel assembly transfer within six (6) days of reactor shutdown is unlikely.
To. conservatively estimate pool surface radiation' dose rates during fuel handling activities, core off-load and subsequent reload were assumed to occur on the third and 30th-day, respectively, af ter reactor shutdown.
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The estimated pool surface radiation dose rates for a spent fuel assembly l
L having 8 feet 0 inches of water above the top of active fuel three (3) days L
. and 30 days after reactor shutdown are 18.0 millirem / hour (an increase of L
10.3 millirem / hour from the 8 feet 6 inch water shielding condition), and L
4.3 millirem / hour (an increase of 2.4 millirem / hour from the 8 feet 6 inch lL
-water shielded condition).
The increased radiation levels will.be limited I,
to the transfer time between the reactor and the spent fuel pool, which is typically not more than four (4) minutes.
Therefore, with the pools at
. normal water level, the increase in received dose to an individual per trinsfer to the spent fuel pool from the vessel three (3) days after shutdown, and to the vessel from the spent fuel pool 30 days after shutdown, would be approximately 0.69 millirem and 0.16 millirem respectively.
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' A complete core off-load.and reload consists of approximately 1528 fuel assembly transfers, including 300 transfers of non-irradiated fuel.
The total increase in radiation dose received by two individuals on the refueling p
platform'with the pools at normal' water level during a complete core i
H off-load anc ralcad is estimated to be 1.212 rem.
Although this potential 4
increase it, ocaved radiation dose would be notable relative to the past J
refuel ficoi mao
- man-rem total of 32.5 (representing approximately a -
l 3.7 perch t incr %, it will be insignificant rela'.ive to the total
- outage caerem M & 5 (representing approximately a 0.6 percent increase).
The actunt rec +.ivd dose could be less than this value since fuel handling is not anticierAd to occur before the sixth day after shutdown and-approximatoly two thirds of:the fuel will have been irradiated one or two operating cycles.
Both of these factors will reduce the actual radiation levels external to the pool surface and subsequently the accumulated dose.
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' During a refueling outage when fuel assemblies will be shuffled in the' core, approximately one third of the core will be off-loaded.
The total increase in radiation dose received by two individuals on the refueling
- platform with the pools at normal water level during a core shuffle is 1
estimated to be 372.6 millirem.
This increase.is insignificant relative
'f to both the refuel floor outage man-rem total (approximately a 1.2 percent increase), and the total outage man-rem (approximately a 0.2
. percent. increase).
Again, this-estimate is higher than that which would actually be received due to the conservatism of using radiation dose rate
- levels for fuel moves three (3) days after reactor shutdown vice six (6) f days.
The fuel handling accident is discussed in FSAR Section 15.7.4.
The accident is assumed to occur as a result of a failure of the fuel c
assembly lifting mechanism resulting in dropping a raised fuel assemL'y onto other fuel. assemblies.
The accident scenario that produces the largest number of failed fuel rods -is the drop of a fuel assembly and i
grapple mast assembly into the reactor.
The analysis of this scenario-revealed that the calculated exposures for the design basis accident are.
well within the guidelines of 10 CFR 100.
A fuel assembly weighing 700 p :nds was assumed to drop 32 feet and the grapple mast assembly weighing 500 pounds was assumed to drop 47 feet.
The energy available for fuel damage-from these objects was calculated to be 45,900 foot poundt.
Allowing a fuel assembly to be raised to a higher elevation over the reactor will make more energy available for fuel damage than that which is currently available.
The drop distances used in the analysis represent the differences in plant elevation from both the lowest point:
on a carried fuel assembly and the lower surface on the grapple head to the upper channel surface of fuel,in the reactor.
The carried fuel assembly is at a plant elevation where the water above top of active fuel
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't will be changed from 8 feet 6 inches to 8 feet 0 inches (referenced to
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pool normal water level).
The possible drop distances for the fuel and grapple mast' assemblies will increase from not more than 31 feet 5 inches and 46 feet 0 inches to not more than.31 feet 11 inches and 46 feet 6 Linches, n spectively.
The energy which would be available to cau,e fuel i
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. damage assoc ated with these h gher drop d stances would, increase from-its present calculated value of 44,994 foot pounds to 45,594 foot pounds.
1 However, as stated above, the fuel handling accident scenario discussed a
in the FSAR assumed an even greater drop distance and resulting energy availability with which to cause fuel damage.
Therefore, the analysis will continue to. bound any possible main hoist component drop scenario.
The proposed chaliges-to the main hoist TS SRs will not result in any physical _ changes to the refueling platform other than the relocation of the main hoist normal up limit switch.
The limit switch will be relocated on the main hoist grapple mast such that main hoist motion will stop not higher than six (6) inches from its current position. The limit I
switch will be reatte.ched to the mast in a manner similar to that which was originally done. ; No refueling platform control logic circuits will be altered.
The handling of fuel and other core components and the performance of other underwater activities will not be performed differently' from previous refueling activities.
Administrative controls s
Mill not be modified to accommodate these proposed changes.
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Since the. auxiliary hoists are procedurally prohibited from handling fuel, all normal auxiliary hoist activities will involve hoist loads less than 300 pounds.
Therefore, no need exists for the auxiliary hoists to
. have fuel = associated load and interlock capabilities.
The heaviest core component normally handled.by an auxiliary: hoist is the control rod guide tube. -During the handling of a control rod guide tube, the hoist load will be no greater than 292 pounds (i.e., control rod guide tube weight 257 pounds (dry), control rod tube grapple weight 35 pounds (dry)).
Since the auxiliary-hoists are prohibited from handling fuel assemblies (channeled fuel assembly weight.682 pounds (dry)), a 1000 pound capacity is not needed.
Therefore, restricting heist loads to 500 t 50 pounds will have no adverse effect on normel hoist operation.
The 500 t 50
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pound limit is consistent with and will enforce the. administrative controls alrehdy in place to prevent using an auxiliary hoist to move fuel.
3 The 400 i 50 pound auxiliary hoist fuel-loaded signal provides input to the refuel interlock circuitry to indicate an auxiliary hoist on the refueling platform is loaded with a fuel bundle.
Several interlocks are associated with this feature and result in the following:
1.
Prevention of travel of the refueling platform over the reactor while a control rod is withdrawn and any hoist is fuel-loaded.
2.
Prevention of lifting of a fuel assembly fru the reactor with a control rod withdrawn.
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Prevention of withdrawal of a control rod blade with the refueling platform over the reactor and any hoist fuel-loaded.
Since the licensee is proposing to limit the auxiliary hoists' capacity to 500 t 50 pounds, thereby precluding the use of the auxiliary hoists to move fuel, imposition of a 400 1 50 pound fuel-loaded interlock on the auxiliary hoists is unnecessery.
In addition, since the auxiliary hoists are prohibited from handling fuel, specifying the minimum water depth reference requirement to the top of awtive fuel (i.e., 8 feet 6 inches below normal water level) for control rod blade handling is inappropriate.
Minimum water depth requirements for the auxiliary hoists need to be specified such that the reference will be consistent with the use to which the hoist will 'n subjected.
The reference that will be used is not less than 6 feet i inches of water above ey carried component.
This will allow unimposed past o of.
all major cora components through the spent fuel pool to tiie reactor well canal, while Saintaining adequate shielding for the irradiated components being handled.
Currently, during the transfer of core components between the reactor vessel and the spent fuel poc1, the potential exists for ;
r component' contact with pool / cavity structures (e.g., portable refueling shield) due to lack of clearance.
This could cause equipment and/or
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carried component damage.
A control rod blade, one of the larger core components, during transfer from the reactor to the spent fuel, will have approximately six (6) inches of clearance between the bottom of the blade and the floor of the shield bridge in the canal upon implementation of the proposed normal up stop limit.
Permitting non-fuel core components to be raised to a higher plant elevation than previously allowed will increase the radiation levels external to the pool surface.
The control rod blade will create the greatest radiation hazard during handling.
Currently 7 feet 0 inches of water shielding are provided as described in FSAR Section 9.1.4.3.
The calculated average surface radiation dose rates with 7 feet 0 inches and 6 feet 6 inches of water shielding are 10.0 millirem / hour and 27.0 millirem / hour, respectively.
The maximum calculated surface radiation dose rates considering worst case component material compositions would be 20.0 millirem / hour and 54.0 millirem / hour, respectively.
These higher possible radiation dose rates are still well within the radiation zone designation for the refuel floor pool area (Radiation Zone IV, i.e., <100 millirem / hour, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown).
A six (6) inch reduction in water shielding will increase the calculated radiation dose rates by a factor of approximate'v 2.5.
This increase in radiation levels will be limited to the transfer time between the reactor and the spent fuel pool, which is typically not more than five (5) minutes.
Therefore, with the pools at normal water level, the increase in received dose to an individual would average approximately 1.42 millirem per component transferred.
During an outage, vessel to spent fuel pool control rod blade transfers should not normally exceed 30.
The increased radiation dose received by two individuals on the refueling platform with t
4 s the pools at noriaal water level is estimated to average 85.2 millirem.
This potential increase in received rtdiation dose will be insignificant relative to the past refuel floor outage man-rem total of 32.5 (representing approximately a 0.26 percent increase), and past total outage man-rem of 209.5 (representing approximately a 0.04 percent increase). The actual radiation levels and total received dose are expected to be less than those predicted since the control rod blade design providing the above estimates was an advanced typa (General Electric DURALIFE 215) not currently in use at LGS. The advanced typ? of control rod blade has a longer in-vessel life than those currently in use, and therefore would become more activated than the control rod blades currently in use.
The fuel handling accident discussed in FSAR Section 15.7.4 and summe.rized above is also pertinent to the safety discussion concerning the auxiliary hoists.
Allowing a core component to be raised to a higher plant elevation over the reactor vessel will increase the potential energy available for fuel damage provived the drop weight is maintained the same. The greatest possible distance through which an object (assumed to be at least one (1) foot long) could drop would increase from not more than 45 feet 1 inch to not more than 45 feet 7 inches, a 1.1 percent increase in the drop distance. However since auxiliary hoist load will be restricted to 500 1 50 pounds, half o,f the currently allowed 'iimit, the energy which would be available to cause fuel damage would decrease from an approximate value of 45,000 foot-pounds to 22,790 foot-pounds. Therefore, all auxiliary hoist component drop scenarios possible will continue to be bounded by the current analyses.
A control rod removal error during refueling activities it discussed in FSAR Section 15.4.1.1.
The transient considered was an ir. advertent critically due to the complete withdrawal or removel of the highest worth control rod during refueling.
However, the core is designed to remain suberitical and meet shutdown requirements with the highest worth rod withdrawn. During refueling operations, system interlocks are provided to assure that inadvertent criticality does not occur because two control rods have been removed or withdrawn together.
Refueling interlocks are provided to accomplish the following.
1.
Prevent refueling platform travel over the reactor core if a control rod is withdrawn and fuel is on the hoist.
2.
Prevent control rod motion if the refueling platform is over the reactor core and fuel is on the hoist.
These interlocks back up requirements that all control rods be fully inserted when fuel is being loaded into the core. Another interlock that
is provided involves the reactor mode switch.
With the mode switch in the " Refuel" position, only one control rod can be withdrawn at a time.
Finally, the design of the control rod blade does not physically permit its removal from the reactor since the fuel support piece and control blade are designed so that the blade can not be removed from the reactor i
without prior removal of the four adjacent fuel assemblies.
The withdrawal of the highest worth control rod during refueling will not result in criticality and additional reactivity insertion is precluded by interlocks and physical design.
The proposed changes to the TS SRs on the auxiliary hoists will not result in any physical changes to the refueling platform or its auxiliary hoists.
This change will not alter the physical load capacity of the auxiliary hoists since no material changes are being performed and 9 hoists will be maintained in the same manner.
No refueling platfr control logic circuits will be altered.
The handling of core com ents i
and performance of other underwater activities will not be performed differently from previous refueling activities.
Administrative controls will not be modified to accomodate these changes.
We have reviewed the licensee's analyses and agree with their evaluations.
We conclude that the proposed changes to the SRs fer the main and auxiliary hoists will result in minimal increases in occupational radiological exposures, the applicable design analyses remain bounding and all regulatory requirements will continue to be met so that the proposed changes will not adversely affect safety.
The proposed changes to the SRs of the TSs are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
These amendments involve a change to a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements.
l The staff has determined that these amendments involve no significant increase in the amounts, and no significant chance in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding.
Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement nor I
environmental assessment need be prepared in connection with the issuance of these amendments.
4.0 CONCLUSION
The Commission made a proposed determination that these amendments involve no significant hazards consideration which was published in the Federal Register (55 FR 28479) on July 11, 1990 and consulted with the Commonwealth of Pennsylvania.
No public comments were received and the Commonwealth of Pennsylvania did not have any comments.
o' i The staff has concluded, based on the considerations discussed above, that:
(1) there is reesonable assurance that the health and safet public will not be endangered by operation in the proposed manner,y of the
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- 2) sgch activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the connon defense and the security nor to the health and. safety of the public.
1 Dated:
August 16. 1990 l
l Principal Contributors:
SDembek RClark i
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