ML20087C945
| ML20087C945 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/01/1995 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20087C946 | List: |
| References | |
| NUDOCS 9508100094 | |
| Download: ML20087C945 (14) | |
Text
Station support Copartment
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'- A 10 CFR 50.90 l h l PECO ENERGY 75 mi="L,,
965 Chesterbrook eoulevard Wayne, PA 19087-5691 August 1,1995 Docket Nos. 50-352 50-353 Ucense Nos. NPF-39 NPF-85 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Subject:
Umerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 95-12-0 Gentlemen:
PECO Energy Company is submitting Technical Specifications (TS) Change Request No.
95-12-0, in accordance with 10 CFR 50.90, requesting an amendment to the TS (Appendix A) of Operating Ucense Nos. NPF-39 and NPF-85 for Umerick Generating Station (LGS), Units 1 and 2, respectively. This proposed TS change involves revising TS Section 3/4.9.1, " Reactor Mode Switch," in order to provide alternate actions to allow the continuation of core alterations in the event certain Reactor Manual Control System (RMCS) and refueling interlocks are inoperable. Specifically, the current TS stipulate that a control rod can not be withdrawn if the one-rod-out interlock is not operable, and core alterations must be cuspended if specific refueling interlocks are Inoperable. The proposed TS changes w-l! permit the continuation of core alterations with certain interlocks being inoperable, if altemate actions, whict will preserve the intended function of the inoperable interlock, are taken. Information supporting this TS Change Request is contained in Attachment 1 to this letter, and copies of the marked-up TS pages for the LGS, Units 1 and 2, TS are provided in Attachment 2. This TS Change Request is being submitted under affirmation, and the required affidavit is enclosed.
We request that, if approved, the amendments to the LGS, Units 1 and 2 TS be issued prior to January 26,1996, and become effective within 30 days following issuance.
If you have any questions or require additional information, please do not hesitate to contact us.
Very truly yours, G. A. Hunger, Jr.
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Director - Uconsing Attachments 9508100094 950801 Enclosuro PDR ADOCK 05000352 !
P PDR cc: T. T Martin, Administrator, Region I, USNRC (w/ attachments, enclosure)
N. S. Perry, USNRC Senior Resident inspector, LGS (w/ attachments, enclosure) ,
~l mPs^ qt.} '/ R. R. Janati, Director, PA Bureau of Radiological Protection (w/ attachments, enclosure) Aj g
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1 COMMONWEALTH OF PENNSYLVANIA :
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- COUNTY OF CHESTER : j W. H. Smith, Ill, being first duly swom, deposes and says: l That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the foregoing information supporting Technical Specifications Change Request No. C&12 0 for Umerick !
Generating Station, Units 1 and 2. Facility Operating Ucense Nos. NPF-39 and NPF-85, to revise TS Section 3/4.9.1, " Reactor Mode Switch," to provide alternative actions to allow for continuation of core alterations with certain refueling Interiocks inoperable, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
N .ky A Vice President l
Subscribed and sworn to l
before me this /3* day 1
of 1995. l l
b Notary Public Notan.d Seal M.vv Lou Skrocki, Notary Pubic Tredytfnn Twp . Cnester County My Comenssion Expres May 17.1999 Metan.RwvMNaru Asangon ot Notum l
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.; ; ^ . . ' l ATTACHMENT 1 LIMERICK GENERATING STATION
, UNITS 1 AND 2 Docket Nos. 50-352 l 50-353 l License Nos. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 9512-0
' Revise the Technical Specifications to Permit Alternate Actions to Allow for Continuation of Core Alterations
- Supporting information for Changes - 10 pages
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" Attachment 1 Page 1 PECO Energy Company, Licensee under Fac0ty Operating License Nos. NPF-39 and NPF-85 for Limedek Generating Statk>n (LGS), Units 1 and 2, respectively, requests that the Technical Specifications (TS) contained in Appendix A to the Operating Licenses be amended as proposed herein to modify TS Section 3/4.9.1, " Reactor Mode Switch," in order to provide altemate actione to allow the continuation of core alterations in the event certain Reactor Manual Control System (RMCS) and refueling interlocks are Inoperable, while preserving the intended function of the Inoperable interlocks.
The proposed changes to the TS are Indicated on the attached marked-up TS pages contained in Attachment 2 to this letter.
We request that, if approved, the TS changes proposed herein be issued by January 26,1996, and become effective with!n 30 days following issuance.
I i This TS 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 Environmental Assessment.
j Discussion and Descriotion of the Prooosed Chanaes Limerick Generating Station (LGS), Units 1 and 2, Technical Specifications (TS) Section 3/4.9.1, " Reactor -
Mode Switch,' currently stipulates that when the reactor mode switch is locked in the REFUEL position, certain conditions must be satisfied. Specifically, a control rod can not be withdrawn unless the Refuel position one-rod-out interlock is OPERABLE, and CORE ALTERATIONS can not be performed using equipment associated with the Refuel position interlock unless four (4) specific Refuel position intedocks are OPERABLE (i.e., All rods in, Refuel platform position, Refuel platform hoists fuel-loaded, and Sc,'Act platform hoist fuelloaded),
This TS Change Request will revise TS Section 3/4.9.1 to provide attemate actions to allow for the continuation of core alterations in the event certain Reactor Manual Control System (RMCS) and j
refueling Interlocks are Inoperable, while preserving the intended function of the inoperable intedocks.
' The proposed TS changes will provide LGS with the flexibility to minimize delays durin9 refueling outages while maintaining strict adherence to all safety considerations. The specific changes proposed in this TS Change Request are summarized below.
- 1. Provide a standardized format to be more consistent with other TS instrumentation i
sections (e.g., Section 3.3.1, " Reactor Protection System," Section 3.3.6, " Control Rod j Block,* etc.)
l 2. Specify that the refueling platform position applicable to the refueling interlock is the
'over core
- position.
- 3. Delineate that the refuel interlock associated with the service platform is only applicable when the service platform is installed over the vessel.
- 4. Allow all control rods to be verified full-In, then disabled from being withdrawn, as an attemativo to placing the reactor mode switch in the SHUTDOWN position, in the event j the one-rod-out Interlock is inoperable.
- 5. Allow for the selection of multiple attematives in the event the refueling interfocks associated with the refueling platform are inoperable. These alternatNe actions remove one of the necessary components (i.e., a fuel-loaded holst, a refueling platform over the l core, or a control rod withdrawn) required for an inadvertent criticality during refuel operations. In each case, the attemative action will be preceded with a verification of that associated function.
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1 Attachment 1 I Page 2 1
- 6. Allow for the selection of multiple attematives in the event the refueling Interlocks associated with the service platform are inoperable, and the platform is installed over the vessel. These attemative actions remove one of the components (i.e., load on platform i hoist or a control rod withdrawn) required for an inadvertent criticality during refuel I operations. In each case, the alternative action will be preceded with a verification of that assoc!ated function.
Safety Assessment The Reactor Manual Control System (RMCS) provides the operator with a method to make changes in nuclear reactivity so that reactor power level and power distribution can be controlled. The RMCS allows the operator to manipulate control rods. This system and has no design safety functions, and has no function associated with responding to any Design Basis Accident (DBA).
The instrumentation and controls for the RMCS consist of the electrical circuitry, switches, Indicators, and alarm devices provided for operational manipulation of the control rods and surveillance of associated equipment. This system does not include any of the circuitry or devices necessary to automatically or manually scram the reactor. The RMCS circuitry is completely independent of the circuitry controlling reactor scram functions. This separation of scram and normal control rod functions prevents failures in the RMCS circuitry from affecting the scram circuitry.
The RMCS also locludes interlocks which restrict control rod movement (i.e., rod block) under certain conditions as discussed in Section 7.7.1.2, " Reactor Manual Control System - Instrumentation and Controls,' of the LGS Updated Final Safety Analysis Report (UFSAR). The interlock logic for restricting control rod movement is intended to function as a backup to currently established procedural controls. I l
Refueling interlocks, which are discussed in LGS UFSAR Section 7.7.1.15, " Refueling Interiocks - l Instrumentation and Control," are provided to restrict the movement of the control rods and the operation of refueling equipment to prevent an inadvertent criticality during refueling operations. These l Interlocks function as a backup to currently established procedural controls. Criticality is prevented l during fuel insertion, provided that control rods in the vicinity of the vacant fuel space are fully inserted I during fuel insertion. The interlock system functions to prevent inadverent criticality using the following )
mechanisms.
- 1) Preventing operation of fuel-loaded refueling equipment over the core whenever any control rod is withdrawn.
- 2) Proventing control rod withdrawal whenever fuel loading equipment is over the core.
- 3) Preventing withdrawal of more than one control rod when the mode switch is in the REFUEL position. ;
The refueling interlock circuitry senses the position of the refueling equipment and control rods.
Depending on the sensed position, Interlocks are actuated to prevent the movement of the refueling l equipment or withdrawal of control rods (i.e., rod block). During refueling operations, no more than one l (1) control rod is permitted to be withdrawn. This function is supported by a redundant logic circuit that i uses the "all-rods-in' signal and a rod selection signal to provant the selection of a second rod for movement with any other control rod not being fully inserted. With the reactor mode switch in the REFUEL position, the interlock circuitry prevents the withdrawal of more than one (1) control rod and the movement of the loaded refueling platform over the core with any control rod withdrawn.
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- Attachment 1 Page 3 The proposed TS changes will not adversely affect the one-rod-out and refueling intedocks. The proposed TS changes will preserve the Intended interlock function, and maintain the same level of protection against refueling errors that could lead to inadvertent criticality, damage to reactor Intemals or fuel assemlies, or excessive personnel radiation exposure.
In addition, the proposed changes associated with TS Change Request 9512-0 were evaluated against those transient events postulated to occur during refueling operations that are described in Chapter 15, l
- . " Accident Analysis," of the LGS UFSAR. An evaluation of each of the applicable transient events i considered in support of this TS Change Request is described below.
Rod Withdrawal Error (RWE) During Refueling Although categorized as an infrequent incident, the RWE is an Anticipated Operational Transient '
(AOT) and was considered in evaluating the proposed TS changes, since the analysis assumes
- that the refueling interlocks act in preventing an inadvertent criticality during refueling operations. J As part of this transient, the potential for an inadvertent criticality is evaluated from the following three (3) different perspectives.
- 1) Fuel insertion with a control rod withdrawn.
- 2) Removal or withdrawal of a second control rod.
- 3) Removal of a control rod without removal of the surrounding fuel.
There is no postulated set of circumstances that can result in a rod withdrawal error whRe in the refueling mode. Protection from fuel insertion with a control rod withdrawn is provided from the l l refueling interlocks which never allow fuel-loaded refueling equipment to operate over the core when a control rod is withdrawn, or a control rod to be withdrawn if fuel-loaded refueling equipment is already over the core. Protection against withdrawing a second control rod is provided from the one-rod-out interlock, which automatically limits rod withdrawal to one (1) .
control rod at a time with the reactor mode switch in REFUEL position. Removal of a control j rod from a cell stil containing fuel assemblies is prevented by the control rod poskion indication i and the physical limitations of the velocity limiter not being able to pass through the fuel support piece (i.e., no upward movement). This ensures that the surrounding fuel bundles are removed prior to removal of a control rod in upward direction. The proposed TS changes w91 not impact .
the removal of control rods with surrounding fuel stil installed.
L in the event the reactor became critical with the reactor head removed (i.e., Operating State B, I and the reactor mode switch assumed to be in STARTUP position), the transient would be i terminated by reactor scram initiated by the Reactor Protection System (RPS). Since the l proposed TS changes do not affect the operation of the RPS, it is only necessary to evaluate criticality prevention from a shutdown refueling condition (i.e., Operating State A).
1 Control Rod MLwporation 1
- The LGS UFSAR asaumes the analysis associated with Control Rod Misoperation to be included l in the RWE transients during refueling, startup, and power operations. Any RWE occurring l during startup or powtr operation is not applicable to refueling operations, and therefore, was l
. not considered in evaluating the proposed TS changes. Since Control Rod Misoperation is
) already considered in the refueling RWE transient, as discussed above, no separate evaluation was necessary in supporting the proposed TS changes.
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Attachment 1 Page 4 Misplaced Bundle Accident Although any bundle misplacement would occur wh20 in refueling, all of the consequences postulated for this transient are assumed to occur from power operation, and therefore, this transient does not apply to the proposed TS changes. This event postulates an undetected reduction in thermal margins during subsequent operation following bundle misplacement, and assumes no detection and no protection system response.
Control Rod Drop Accident Although a Control Rod Drop Accident (a DBA) can occur during refueling operations, the consequences of this event are based on the event occurring at rated conditions (i.e., from power operation), since those conditions yield the most conservative (i.e., worst) consequences.
Therefore, a Control Rod Drop Accident occurring during refueling operations would be bounded by the DBA analysis. The analysis assumes that the Average Power Range Monitor (APRM) flow reference high flux scram function responds to terminate this accident prior to exceeding the 280 cal /gm peak fuel enthalpy design limit. During low power operations, a high flux scram having a considerably lower setpoint could be in place from the Intermediate Range Monitor (IRM) or APRM setpoint; however, as previously stated, these lower trips are bounded by the analys!s assumed for the APRM flow reference high flux scram.
Information Sucoortina a Findina of No Sionificant Hazards Consideratip_r1 We have concluded that the proposed changes to the Limerick Generating Stat!on (LGS), Units 1 and 2, Technical Specifications (TS) Section 3/4.9.1, " Reactor Mode Switch," to provide altemate actions to allow the continuation of core alterations in the event certain Reactor Manual Control System (RMCS) :
and refueling interlocks are inoperable do not involve 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.
- 1. The orocosed Technical Soecifications fTS) chanaes do not involve a sianificant )
Increase in the orobability or consecuences of an accident oreviousiv evaluated. I 1
The refueling and one-rod-out Interlocks impose barriers to preclude an inadvertent criticality during refueling operations. Section 7.7.2.15.1 of the LGS Updated Final )
Safety Analysis Report (UFSAR) clearly delineates the functions of the interlocks and the '
criteria used in assessing correct refueling and one-rod-out Interlock operation in the following statement. i I
In all cases, correct operation of the refueling interlock prevents either the '
operation of loaded refueling equipment over the core when any control rod is withdrawn, or the withdrawal of any control rod when fuel-loaded refueling l equipment is operating over the core. In addition, when the reactor mode !
switch is in REFUEL position, only one rod can be withdrawn, and selection of a second control rod initiates a rod block.
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Attachment 1 Page 5 The proposed TS changes provide operational flexibuity while strictly conforming to, and preserving, the intended function of the refueling and one-rod-out interlocks. The proposed TS changes that could affect Interlock capabilities are identified below, along with the appropriate justification to substantiate that the proposed TS changes wHl not result in an increase in the probability or consequences of an accident previously evaluated.
- a. TS Section 3.9.1. ACTION Statement b. The proposed change to this existing TS ACTION will add a verification that all control rods are fully inserted, and then disabled from being withdrawn as a suitable alternative to placing the reactor mode switch in the SHUTDOWN position when the one-rod out interlock is not operable. In addition, the proposed change to this TS section includes a caveat of non-appilcability for those control rods already removed in accordance with requirements stipulated in TS Sections 3.9.10.1 and 3.9.10.2. As indicated in LGS UFSAR which is described in the statement above, it is expected that the refuel and one-rod-out Interiocks will permit the withdrawal of only one (1) control rod at a time with the reactor mode switch in the REFUEL position, and no control rods can be moved when fuel-loaded refueling equipment is operating over the core. By verifying all control rods are Inserted, then disabling withdraw capabilities of all rods, as requested, the most limiting requirements for control rod motion will be met. The potential for having more than one (1) control rod out at a time, or having any control rod not fully inserted while fuel.
loaded refueling equipment is operating over the core, does not exist when applying the alternative. Therefore, the intended functions of the refuel and one-rod-out interlocks are operationally preserved. Since TS Sections 3.9.10.1 and 3.9.10.2 have specific requirements for removing surrounding fuel prior to control rod blade removal, the control rods already removed are no longer required to carry out a safety function in the defueled cell, and as a result would not apply for this specific proposed TS change. From a control rod withdrawal perspective, there is no functional difference between the proposed TS change and the existing, and st0! remaining, TS ACTION of locking the reactor modo switch in SHUTDOWN position.
- b. TS Section 3.9.1. ACTION Statement c. This existing TS ACTION requires that core alterations be suspended in the event that a refueling interlock is not operable. The proposed TS change to this TS ACTION ieaves this requirement in place, but makes this ACTION specifically applicable to the refueling platform, and adds three (3)m additional ACTION alternatives. The wording for changes to this TS section are such that implementation of any one of the three (3)m altematives can be substituted for suspending core alterations. The proposed wording for these three (3) new attematives and justification is provkfed below.
- 1) Verify all control rods are fully inserted and disable withdraw capabilities of all control rods *".
Since this attemative ensures all control rods are, and will remain fully inserted, all required conditions of the associated refueling and one-rod-out interlocks are mot. The refueling interlock is satisfied since a fuel.
loaded refueling platform operating over the core would be assured that all control rods are fully inserted and prevented from being withdrawn.
The one-rod-out interlock is satisfied since control rod withdrawal is
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- l disabled for all control rods, which is an even more conservative !
' requirement than the one-rod-out interlock itself. Whle operating in this - )
configuration, there will be no associated travel or hoist restrictions for .!
the refueling platform over the core, which is normal for the current l refuel interlock design. The potential for having any control rc 1 not fully inserted whle a fuel 4oeded refueling platform is operating over the core, ,
does not exist when applying this proposed alternative. Therefore, the l intended function of the refueling platform refuel interlocks are l operationally preserved with the implementat!on of this proposed J attemative, and there wtl be no increase in the probablity of occurrence .
of an accident. This proposed altemative also maintains an exclusion (via a reference to the *** footnote) for control rods removed in accordance TS Sections 3.9.10.1 and 3.9.10.2. This exclusion does not apply to inadvertent criticality concems, as previously discussed in item 1.a above.
- 2) Verify Refuel Platform is not over core (limit switches not reached) and disable Refuel Platform travel over core.
As previously stated above, LGS UFSAR Section 7.7.2.15.1 stipulates -
, that the refueling platform position interlocks initiate a control rod block j whenever a fuel-loaded refueling platform is over the core, and stop a j fuel-loaded refueling platform from moving over the core if a control rod i 13 already withdrawn. This specific proposed TS change satisfies both these requirements by precluding the possibuity of the platform from I being over the core. If a control rod is being withdrawn, the platform will not be over the core, and the withdrawal wul be in accordance with the current design. If a control rod is already withdrawn, disabling ,
platform travel over the core,' before reaching the over-core limit !
switches, la performing the same function as the existing refueling platform reverse and forwani motion blocks. Therefore, the potential for i I
having any control rod not fully inserted whle a fuelloaded refueling platform is operating cwor the core, does not exist when applying this proposed altemative. The intended refueling interiock functions are operationally preserved with the implementation of this proposed attemative. ;
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- 3) Verify that no Refuel Platform hoist is loaded and disable all Refuel
,. Platform husts froir, picking up (grappilng) a loed.
As previously stated above, UFSAR Section 7.7.2.15.1 stipulates that blocking control rod withdrawal with a refueling platform over the core, and restricting refueling platform travel from going over the core with a ;
control rod already withdrawn, are based on the refueling platform hoist I being fuel-loasted. An unloaded platform without grappling capabHitles poses no threbt to erroneous fuel bundle or control rod removal, and . ,
elim! nates the pon,ntial for having any cowol rM not fully inserted whue l a fuel-loaded refueSng platform is operating o,u the core. Therefore, l Impenenting this proposed attematNe opentkwlally preserves the
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inten:!ad interlock functions.
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3 :m.4 y Attachment 1 1 Page 7;
- c. TS Section 3.9.1. ACTION Swamerti. The proposed TS change adds thism TS ACTION section to specify the refueling interlock requirements for the service -
platform, since the applicablity of ACTION Statement c above is being revised i to specifically address refueling interlocks associated wit 1 the refueling platform.
The proposed TS changes for new this TS section retah the existing .
requirement to suspend core alterations if the service platform associated refueling interlock is not operable, unless the service platform is not installed over vessel. The specific proposed TS changes add two (2) new additional ACTION altamatives. The proposed wording for these two (2)m ACTION -
statements are such that implementation of any one of the two (2) m altematives can be substituted for suspending core alterations.' Not odorcing operabulty requirements on the service platform refueling interlocks whe. The -
service platform is not over the vessel does not pose an inadvertent criticality concem since there is no associated hoist to manipulate fuel bundles or control rods. These two (2)m altematives are not applicable unless the service platform is installed over the vessel, and are described below.
- 1) Verify all control rods are fully inserted and disable withdraw 1 capabilities of all control rods ***.
This altemative ensures that all contt01 rods are, and wNi remain, fully inserted which meets the required cor,#!ent for proper refueling and one-rod-out interlock operation. The refueling interlock is satisfied since a fuel-loaded service platform hoist operating cver-core is assured that all control rods are fully inserted and prevented from being withdrawn.
The one-rod-out interlock is satisfied since all comrol rods are disabled, an even more conservative requiremert than the one-rod-out interlock itself. While operating in this configuration, there will be no associated hoist restrictions for the service platform, which is normal for the current refuel interlock design. The potential for having any control rod not fully insertad while a service platform hoist is fuel-loaded over the core, does not ex'st when utNizing this proposed attemative. Therefore, the intended function of the service p;eiven refuel interlocks are operationally preserved with the Nplementation of this proposed attemative. This proposed attemative also maintains an exclusion (via a -
eference to the proposed *** footnote) for control rods removed in accordance with the requirements of TS Sections 3.9.10.1 and 3.9.10.2.
This exclusion is not applicable to inadvertent criticality concems as discussed in item 1.a above. Y, l
- 2) Verify Service Platform hoist is not loaded and disable Service Platform hoist from picking up (grappling) a load.
As previously described above, UFSAR Section 7.7.2.15.1 stipulates that blocking control rod withdrawal with the service platform over the core is based on the service platform hoist being fuel-loaded. An unloaded holst without grappling capablities poses no threat to erroneous fuel'
' bundle or control rod removal and eliminates the potential for having any control rod not fully inserted whNe a fuel-loaded service platform is operating over the core. Therefore, implementing this proposed altemative operationally preservr.s the intended refueling interlocle 5 functions.
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Attachment 1 Page 8 As discussed in the LGS UFSAR, the use of the refueling and one-rod-out interlocks are evaluated from a prevention, not a mitigation, perspective. A Rod Withdrawal Error (RWE) transient event during refueling is concemed w'th an inadvertent criticality, and assumes the reactor vessel head is off, and the plant is shutdown (i.e, Operating State A). As described in the LGS UFSAR under Nuclear Safety Operational Malysis (NSOA)
Event 16, it is assumed that the Reactor Protection System (RPS) terminates the event should the reactor actually reach Operating State B (i.e., head off and not shut down),
which is conditional on the reactor mode switch being in the STARTUP podion. The proposed TS changes only pertain to the refueling and one-rod-out interlocks. S!nc6 these Interlocks act only in a preventive mode, the consequences of an inadvertent criticality accident during refueling remain unchanged.
Since the proposed TS changes are lim!ted to the one-rod-out and refueling interlocks, they do not affect the reliability of the associated equipment. The proposed TS changes specify attemative actions that can be taken in the event that an interiock is inoperable.
These alternative actions serve to ensure the faHed interlock function la pressrved, and do not affect the probability of malfunction of the interlocks.
The one-rod out and refueling interlocks, as evaluated in the LGS UFSAR, are designed to preclude an inadvertent criticality during rAeling operations by piacing strict controls on fuel bundle and control rod manipulations, using the following mathods.
- a. Preventing operation of a fuel-loaded refueling platform or service platform hoist whlie over the core if a control rod is already withdrawn.
- b. Preventing a fuelloaded refueling platform from traveling over the core if a control rod is already withdrawn.
- c. Preventing any control rod from being withdrawn if a fuel-loaded refueling platform or servfee platform is already operating over the core.
- d. Preventing the withdrawal of more than one control rod at a time with the reactor mode switch in the REFUEL position.
The LGS UFSAR indicates that a single component failure does not cause an interlock failure and that a single interlock failure does not cause an accident. The proposed TS changes provide attemative actions that can be taken in the event of an associated componont or interlock malfunction. Implementing the proposed TS changes will continue to ensure that the Intended Interlock functions are maintained and operationally preserved, as described in the LGS UFSAR.
Therefore, the proposed TS changes do not involve an increase in the probability or consequences of an accident previously evaluated.
- 2. The crocosed TS chances do not create the nossib!!ity of a new or different kind of accident from any accident oreviously evaluated.
The proposed TS changes only pertain to the refueling and one-rod-out interlocks. The I
refueling and one-rod-out interlocks impose barriers to preclude an inadvertent criticality l during refueling operations. The proposed TS changes provide operational flexibalty,
) while strictly conforming to, and preserving, the intended function of the refueling and j
one-rod-out interlocks. There is no other potential faBure mode for these interlocks than has already been evaluated and described in the LGS UFSAR. Implementation of these
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Page 9 proposed changes will maintain and operationally preserve the intended interlock functions. Therefore, the malfunction of any associated component or interlock will not adversely impact the plant and any other equipment important to safety, directly or indimetly.
Therefore, the proposed TS changes do not create the possibalty of a new or different kind of accident from any previously evaluated.
- 3. The orooosed TS channes do not involve a slanificant reduction in a marain r# safety.
The proposed TS changes only affect the TS associated with the one rod-out and refueling interlocks. The associated TS Bases Section 3/4.9, " Refueling Operations,"
states that the one-rod-out and refueling interlocks maintain conditions during refueling actMtles that reinforce refueling procedures and reduce the rMential for the probability of occurrence of each of the following conditions:
- a. Inadvertent criticality,
- b. Damage to reactor intemals or fuel assemblies, and
- c. Exposure of personnel to excessive radioactivity.
The proposed TS changes do not adversely affect the one-rod-out or refueling interlocks. The associated interlocks will continue to perform the refueling functions ,
required to maintain the'present margin of safety. The proposed TS changes only contain attemative actions that can be taken in the event an interlock is Inoperable.'
l These proposed attemative actions ensure that the intent of the interlocks is preserved, and that there is no reduction in the ability of the interlocks to maintain adequate refueling conditions.
The proposed TS changes will preserve the intended interlock functn., .s, and maintain ,
the existing level of protection against refueling errors that could lead to an inadvertent criticality, damage to reactor intamals or fuel assemblies, or excessive personnel radiation exposure. The one-rod-out and refueling inteecks will continue to function with their present degree of reliabRity. The proposed TS changes will continue to maintain strict controls on fuel bundle and control rod manipulations to avoid inadvettent criticality. The proposed TS changes provide the same level of assurance regaring the' l manipulction of control rods during refueling operations as that currently described in the LGS UFSAR, and as discussed below,
- s. Preventing operation of a fuel-loaded refueling platform or service platform hoist whlie over the core if a control rod is already withdrawn.
- b. Preventing a fuel-loaded refueling platform from traveling over the core if ;
a control rod is already withdrawn. -
- c. Preventing any control rod from being withdrawn if a fuel-loaded i refueling platform or service platform is already operating over the core.
- d. Preventing the withdrawal of more than one control rod at a time with the reactor mode switch in the REFUEL position.
Therefore, the proposed TS changes do not involve a reduction in a margin of safety.
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, , L. * * ,' j Attachment 1 i Page 10 - l l
Information Sucoortina an Environmental Assessment An Environmental Assessment is not required for the changes proposed by this Change Request because the requested changes to the LGS, Units 1 and 2, TS conform to the criteria for " actions eligible for categorical exclusion,' as specified in 10 CFR 51.22(c)(9). The requested 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 1 types or significant increase in the amounts of any effluent that may be released offsite. In addition, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Conclusion
- The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed ,
change to the LGS, Units 1 and 2 TS and have concluded that they do not involve an unreviewed safety l quer. son, and will not endanger the health and safety of tae putdic.
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a ATTACHMENT 2 UMERICK GENERATING STATION l UNITS 1 AND 2 Docket Nos. 50 352 60 353 ,
Ucense Nos. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 95-12-0 UST OF AFFECTED PAGES (MARKED-UP PAGES) jlnit 1 Unit 2
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