ML20044A110

Application for Amends to Licenses NPF-39 & NPF-85, Consisting of Tech Spec Change Requests 90-03-0 & 90-04-0 Re Surveillance Requirements for Refueling Platform Main & Auxiliary Hoists

ML20044A110
Person / Time
Site:	Limerick
Issue date:	06/22/1990
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20044A105	List: ML20044A104 ML20044A110 ML20044A113
References
NUDOCS 9006280083
Download: ML20044A110 (12)

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~kp ATTACHMENT 1 LIMERICK GENERATING STATION Unit.1 and Unit 2 Docket Nos. 50-352 50-353 License Nos.'NPF-39 NPF-85

-TECHNICAL. SPECIFICATIONS CHANGE REQUEST Nos. 90-03-0 and 90-04-0

" Proposed-Changes to the Refueling Platform Main and-

' Auxiliary Hoists Surveillance Requirements" Supporting Information for Changes - 11 pages 90062gg${ckhi g[

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Docket Nos. 50-352

, 50-353 License Nos. NPF-39 NPF-85 Philadelphia Electric Company, Licensee under Facility Operating License Nos. HPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2 respectively, hereby requests that the Technical Specifications (TS) contained in Appendix A to the Operating Licenses be amended as proposed herein. The proposed changes to the TS are indicated by a vertical bar in the margins of the pages contained in Attachment 2.

We request the changes proposed herein to be effective by September 1, 1990 to allow their use during the next Unit 1 refueling outage scheduled to begin September 8, 1990.

This Change Request provides a discussion and description of the proposed TS changes, a safety assessment of the proposed TS changes, information supporting a finding'of No Significant Hazards Consideration, and information supporting an Environmeatal Assessment.

Discussion and Description of the Proposed Changes TS changes are proposed to the refueling platform main hoist Surveillance Requirements (SRs) and the auxiliary hoist SRs to more accurately reflect their actual use. The following provides a general description of the refueling platform and associated hoists, and also specific discussions of the existing SRs for both the main and auxiliary hoists and the proposed changes for each.

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 platform has three hoists through which many of these activities are accomplished. The main hoist.-with an 1150 pound operating capacity and dual hoist cables, is deuicated to the operation of the telescoping mast and grapple assembly. This assembly 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 and a single hoist cable, 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 castings, neutron source holders, and general servicing aid placement.

The SRs of TS Section 3.9.6, " Refueling Platform," require main hoist operability to be demonstrated by verifying the following.

1. The overload cutoff shall operrte when hoist load exceeds 1150 pounds.

2. The hoist loaded control rod block interlock shall operate when hoist load exceeds 485 pounds.

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License Nos. NPF-39 NPF-85

3. The redund et loaded interlock shall operate when hoist load exceeds 550 pounds.

4. The uptravel interlock shall operate when the top of active fuel is 8 j feet 6 inches below normal water level.. '

Procedurally, the main hoist is required to be used for the handling of-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 l spent fuel pool, a potential currently exists for component contact with 't

- pool / cavity structures (e.g., portable refueling shield)-due to lack of

- clearance. This could cause equipment and/or carried component damage. '

Therefore, we propose to change the SRs for the main hoist to allow the e normal:up'stop limit switch to be repositioned 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- i inches of. water above.the top of the carried fuel' assembly. Also, we propose to

clarify the main hoist SRs to remove the reference to control rods, since the ,

main hoist is not used for. handling of control rods, and add the phrase "not  !

less than" before the uptravel stop distance.

With respect.to the refueling platform auxiliary hoists Final Safety .

E Analysis Report (FSAR) Section 9.1.4.3,-" Safety Evaluation - Fuel Handling 4 System," and Section 9.1'.4.2.7.1, " Refueling Platform," state that the refueling platform auxiliary hoists-shall provide-at least seven (7) feet of water over  :

the top of any component with the hoist at the normal up position, and that the operational' capacity of each hoist is 1000 pounds.

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l TS Section 3.9.6 SRs require the following to demonstrate auxiliary hoist operability.

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1. The overload cutoff shall operate when hoist load exceeds 1000 i pounds.

2. The control rod block interlock shall operate when hoist load exceeds 400 pounds.

3. The uptravel stop sh611 operate when top of active fuel is 8 feet 6 ,

inches below normal water level. '

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. The following TS changes are proposed.

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' Docket Noso 50-352 f' ,

50-353-License Nos. NPF-39 i NPF 1. Reduce the auxiliary hoist overload cutoff setpoint to 500 1 50 ,

pounds (this proposed change precludes lifting of a fuel assembly withtheauxiliaryhoist).  !

2. . Delete the auxiliary hoist fuel-loaded interlock (i.e., loads in excess of 400 + 50 pounds).

3. Change the auxiliary hoist normal up stop from 8 feet 6 inches of  ;

water above the top of active fuel to 6 feet 6 inches of water above -l a carried core component.

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To facilitate limit switch setting, the water shielding requirement will be  !

specified as a minimum value. Therefore, part of the proposed auxiliary hoist-L TS change will clarify the requirements by adding the phrase "not less then" i before the up travel stop distance.

s Safety Assessment  !

-The following safety discussion addresses the proposed SR changes to the main hoist followed by a similar discussion concerning the proposed SR changes to the auxiliary hoists. A general conclusion is then provided.

q p When handling irradiated fuel, the radiation dose rates external to the m pool surface are highly dependent upon the time. interval from reacto* shutdown.

TS Section 3.9.4, " Decay Time," requires the reactor to be subcritical for at a L 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  !

H requirement ensures sufficient time has elapsed to allow for radioactive decay <

Lof the short. lived fission-products. With the pools at normal water. level, the b proposed-six-(6) inch reduction in water shielding-(for the main hoist SR) in, l = 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' ,

i from 10.6 millirem / hour to 24 millirem / 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 /> after 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,

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respectively, after reactor shutdown. The estimated pool surface radiation dose rates.for a spent-fuel assembly having 8 feet 0 inches of water above the top of active fuel three (3) days and 30 days after reactor shutdown are 18.0 millirem / hour (an increase of 10.3 millirem / hour from the 8 feet 6 inch water c shielding condition), and 4.3 millirem / hour (an increase'of 2.4 millirem / hour i from the 8 feet 6 inch water shielded condition). The increased radiation

. levels will be limited-to the transfer time between the reactor and the spent fuel ~ pool. which is typically not more than four (4) minutes. Therefore, with i

.= . Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 i

the pools at normal water level, the increase in received dose to an individual per transfer 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. i 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 platform '

with=the pools at normal water level during a complete core off-load and reload is estimated to be 1.212 rem. Although this potential increase in received radiation dose would be notable relative to the past refuel: floor outage man-rem total of 32.5 (representing approximately a 3.7 percent increase), it will be insignificant relative to the total outage man-rem of 209.5 (representing approximately a 0.6 percent increase). The actual received dose could be less

,- than this value since fuel handling is not anticipated to occur before the sixth l! day after reactor shutdown and approximately 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.

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 estimated to be 372.6 millirem. This increase is insignificant relative 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) days, i

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 assembly lifting mechanism resulting in dropping a raised fuel assembly onto other. fuel -

assemblies. The accident scenario that produces the largest number of failed-fuel rods is the drop of a fuel assembly and grapple mast assembly into the reactor. The analysis of the scenario revealed that the calculated exposures tjf for the design basis accident are well within the guidelines of 10 CFR 100. A fuel . assembly weighing 700 pounds 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-pounds. i

. 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

' Docket Nos. 50-352-50-353 License Nos. NPF-39 NPF-85 the water above top of active fuel will be changed from 8 feet 6 inches to 8 feet 0 inches (referenced to 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 inches, respectively. The energy which would be available to cause fuel damage associated with these higher drop distances would increase i from its present value of 44,994 foot-pounds to 45,594 foot-pounds. However, as stated above, the fuel handling accident scenario discussed 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 changes 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 switch will be reattached 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 t controls will not ue modified to accommodate these proposed changes.

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 rad guide tube. During -t the handling of a control rod guide tube, the hoist load will be no greater thani 292 pounds'-(1.e., control rod guide tube weight 257-pounds (dry), control rod

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guide 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 hoist loads to 500 1 50 pounds will have no adverse effect on normal hoist.

operation. The 500 1 50 pound limit'is consistent with and will enforce the ,

adninistrative controls already in place to prevent using an. auxiliary hoist to move fuel.-

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1he 400 1 50 pound auxiliary hoist fuel-loaded signal provides input to the refuel interlock circuitry to indicate an auxiliary hoist on P efueling platform is loaded with a fuel bundle. Several interlocks ar F sociated with l 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 from the reactor with a control rod withdrawn.  !

,'" Docket Nos. 50-352 '

.' , 50-353 License Nos. NPF-39 HPF-85

3. Prevention of withdrawal of a control rod blade with the refueling platform over the reactor and any hoist fuel-loaded.

-Since we are proposing to limit the auxiliary hoists' capacity to 500 + 50 pounds, thereby precluding the use of the auxiliary hoists to move fuel, imposition of a 400 + 50 pound fuel-loaded interlock on the auxiliary hoists is unnecessary. In addTtion, since the auxiliary hoists are prohibited from handling fuel, specifying the minimum water depth reference-requirement to the top of active 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 be subjected. The reference that will be '

used is not less than 6 feet 6 inches of water above any carried component.

This will allow unimposed passage of all major' core components through the spent fuel pool to the reactor well canal, while maintaining adequate shielding for the: irradiated components being handled. Currently, during the transfer of core components between the reactor vessel and the spent fuel pool, the potential

-exists for component contact with pool / cavity structures (e.g., portable .'

refueling shield) due to lack of clearance. This could cause equipment and/or y carried component damage. A control rod blade, one of the larger core l- 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.

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Permitting non-fuel core components to be raised to a higher plant E elevation than previously allowed will increase the radiation levels external to L the pool surface. The control rod blade will create the greatest radiation L 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 L radiation dose rates with 7 feet 0 inches and 6 feet 6-inches of water shielding l- 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, Lrespectively. These higher possible radiation dose rates are still well within

the radiation zone designation for the refuel floor pool area (Radiation Zone l

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 i

. reduction in water shielding will increase the calculated radiation dose rates

'by a factor of approximately 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 the pools at normal water level is estimated to average 85.2 millirem. This potential increase-in received radiation 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 the past total outage

' Docket Nos.- 50-352 l

.. 50-353 License Nos. NPF-39 NPF-85 i

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 predicated since the control rod blade design providing the above estimates was an advanced type (General Electric DURALIFE 215) not currently in use at LGS. The advanced type 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, u

The fuel handling accident discussed in FSAR Section 15.7.4 and summarized above is-also pertinent to the safety discussion concerning the auxiliary

-boists.

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 provided 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 + 50 pounds, half of the currently allowed limit, the energy which would be available to cause fuel damage would decrease L from an approximate value of 45,000 foot-pounds to 22,790 foot-pounds.

l 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 is discussed in FSAR Section 15.4.1.1. The transient considered was an inadvertent criticality

-due to the complete withdrawal or removal of the highest worth control. rod during refueling. However, the core is designed to remain'subtritical and meet shutdown requirements with the highest worth rod withdrawn. During refueling operations, system interlocks are provided to assure that inadvertent i 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.

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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

'inse'rted 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 without prior removal of the four adjacent fuel assemblies. The withdrawal of the highest worth control rod during Docket Nos. 50-352

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License Nos. NPF-39 t NPF-85 refueling will not result in criticality and additional reactivity insertion is  :

precluded by interlocks and physical design.

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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 the hoists will be maintained in the  !

same manner. No refueling platform control logic circuits will be altered. The i handling of core components and performance of other underwater activities will not be performed differenly from previous refueling activities. Administrative controls will not be modified to accommodate these changes.

In conclusion, although the proposed changes to the SRs for the main and i auxiliary hoists will result in minimal increases in occupational radiological exposures, the applicable design analyses remain bounding and all regulatory 1 requirements will continue to be met so that the-proposed changes will not adversely affect safety.

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b Information Supporting a Finding of No Significant Hazards Consideration We have concluded that the proposed changes to the LGS Units 1 and 2 TS, which revise the SRs for the refueling platform main and auxiliary hoists, do not constitute a significant hazards consideration, in support of this -!

-determination,.an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below, .

f 1): The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The main hoist normal up stop limit switch is proposed to 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. This position is well within the hoist design operating-range. .No other physical changes will be performed on the. main hoist or the refueling platform as a result i of this TS change. The main hoist will operate in the 6 xact same. manner 4 after'the switch relocation as it did. prior to the proposed change.

Therefore, handling of fuel will not be done any differently than that '

which is done presently and that which was considered in the fuel handling accident. The main hoist will not be any more likely to drop a carried g fuel assembly.or the mast assembly after the switch relocation.

The refueling platform main hoist will be allowed to raise a load not more than six (6) inches higher than that which is presently permitted creating a-greater possible drop distance for a fuel assembly. This situation will create fuel and grapple mast drop distances of not more than 31 feet 11 '

inches and 46 feet 6 inches, respectively. However, the fuel handling .* '

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9- " Docket Nos. 50-352 5  :- ., 353 License Hos. NPF-39 NPF-85

-accident. assumes drop distances of 32 feet 0 inches and 47 feet 0 inches, respectively. Therefore, the analysis presented in FSAR Section 15.7.4 will= continue to be conservative and bound the condition that would exist after the main hoist normal up stop limit switch is raised.

Restricting refueling platform auxiliary hoist loads to less than that'of a fuel assembly will h c e no effect on the likelihood of the fuel handling accident. The-fuel-handling accident scenario which produced the greatest

-fuel damage considered the drop of a fuel assembly and grapple mast assembly from the refueling platform while over the reactor. The fuel assembly is the heaviest component transferred.by the refueling platform-and the. main hoist grapple mast assembly is a significant portion of the total drop mass considered in the accident scenario. By comparison, the auxiliary hoists'are procedurally prohibited from handling fuel and they have no mast assembly. -Their potential drop mass and any resulting energy would be much less than that for the main hoist. Since the auxiliary-hoists play no role in the fuel handling accident analysis, a change to them will not affect the likelihood of the analyzed accident.

Removal'of the auxiliary hoist fuel loaded interlock feature through load switch recalibration will not adversely impact the Refueling Interlocks

.(RIs). The Ris serve to restrict the movement of the control rods and the operation'of the refueling platform to reinforce operational procedures that prevent the reactor from becoming critical during refueling operations. A hoist with a 500 + 50 pound limit will not be able to handle -

(i.e., lift or transport) a fuel assembly-(682' pounds). Fuel can only be handled with the-main hoist and its associated grapple mast assembly which

=will not be'affected by this change. Therefore, the absence of a fuel-loaded. interlock on the auxiliary hoists will not increase-the likelihood-of an inadvertent criticality due to fuel handling errors.

The proposed change will allow the refueling platform auxiliary hoists to raise,their loads not more than six (6) inches' higher than presently permitted, creating a greater possible drop distance for any carried

' Lcomponent. However, since the proposed change restricts the normal ~ load to 500 pounds, half of the current value, the' energy available with which to cause fuel damage would be less than the current value. Therefore, the total energy available from an auxiliary hoist dropped component will continue'to be less than,that produced from the main hoist drop scenario

, discussed in the FSAR analysis of a fuel handling accident.

Finally, the auxiliary hoists are procedurally prohibited from fuel handling activities and the proposed changes to the auxiliary hoists SRs will reinforce these restrictions. Therefore, an inadvertent criticality event during-refueling as discussed in the FSAR will not be affected.

In summary, the proposed changes to the main and auxiliary hoist SRs do not involve an incr ase in the probability or consequences of any accident

'previously evaluated.

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2)- The proposed changes do not create the possibility of a new or different t 4 kind of accident from any accident previously evaluated.

The main hoist' normal up stop limit switch in its proposed location will ,

A function in the exact same manner as it does now (i.e., the strike plate on '

the 10 inch mast section will operate' the limit switch lever. arm). No j v

other-physical changes are to be performed on the refueling platform as-a a

. result of this proposed change. The refueling platform control logic

• d circuits will not be altered. The handling of fuel and the performance of

[ underwater activities will not be done any differently from those presently o performed.

The proposed TS changes to the SRs for the auxiliary hoists will not modify  :

or result in any hardware changes to the refueling platform or its  !

L Lauxiliary hoists. No refueling platform control logic circuits will'be ,

L altered.- The handling of core components and the performance of underwater i activities will not be performed differently from those presently l performed. '

Therefore, the proposed changes to the SRs of the main and auxiliary hoists y .do not create the possibility of a new or different kind of accident from J l any previously evaluated.

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l (3) The proposed changes do not involve a significant reduction in a= margin of' safety.

N The TS' Bases for the refueling platform state in part that the operability requirements will ensure that-a refueling platform hoist has sufficient '

. load capacity for handling fuel assemblies and control rods. The load'-

capacity of-the main hoist is not affected by-the proposed changes. The.

L potential drop distances and resulting energies for a fuel assembly and the grapple mast' assembly during the postulated fuel handling accident would '

. remain bounded by the analysis presented in the-FSAR. The potential for radioactive releases'following.a fuel handling accident would remain 'i bounded by the analysis presented in the FSAR.

Physically, the capacity of each auxiliary hoist is not being altered since I l no material changes are being undertaken and the hoists will be maintained i L in the same manner. However.-to reinforce, administrative controls  !

H presently in place to ensure fuel assemblies are handled in a safe and 1 controlled manner utilizing only the main hoist grapple mast assembly, the auxiliary hoist load switches will be reset so that a fuel assembly can not' L be handled. The potential energy.available from a~ dropped auxiliary hoist {

L' carried component would be less than that which is currently available and f therefore is bounded by the current FSAR analysis. The potential for i radioactive releases following a component drop is also bounded by the current FSAR analysis.

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Docket Nos. 50-352

. 50-353-License Nos. NPF-39 NPF .

Therefore, the proposed changes to the SRs of the main and auxiliary hoists do not involve a reduction in a margin of safety.

Information Supporting an Environmental Assessment

.An environmental-assessment is not required for the changes proposed by this Change Request because the requested changes conform to the criteria for

" actions eligible for categorical exclusion" as specified in 10 CFR 51.22(c)(9).

The proposed changes will have no impact on the environment. The proposed changes do not involve a significant hazards consideration as discussed in the

_ preceding section. The proposed changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed changes _do not involve a

~significant increase in individual or cumulative occupational radiation exposure as previously discussed in the Safety Assessment.

Conclusion The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the TS and have concluded that they do not

-involve an unreviewed safety question, or a significant hazards consideration, and will not endanger the health and safety of the public.

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