ML20059L589

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Proposed Tech Specs Changing Design Features Section 6.1 to Permit Removal of CRD & Orifice Assemblies from Core Regions Defueled in Support of Plant Closure Activities
ML20059L589
Person / Time
Site: Fort Saint Vrain Xcel Energy icon.png
Issue date: 09/14/1990
From:
PUBLIC SERVICE CO. OF COLORADO
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ML20059L585 List:
References
NUDOCS 9009270149
Download: ML20059L589 (28)
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,, Fort St. Vrain #1 c Technical Specifications-S ,

Amendment No.-

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Page 6.1- 1 d

l6.11 REACTOR CORE - DESIGN FEATURES Applicability

~ Applies to, the general design features of the reactor core g including fuel, moderator, reflector and reactivity control, i

e Objective To- define the vital' design characteristics of the reactor core to . control changes ~ in the design features of thi fuel, moderator, reflector and reactivity control._

@, i Specification DF 6.1 - Reactor Core. Design Features q.;

'The following1-discussion ' describes the design features which shall be incorporated in the reactor core:

_ Reactor Assembly. l

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Thel reactor core , asists of: (1) removable fuel elements which 'j

_contain the fuel (U & Th), the-moderator.(graphite) and burnable-  ;

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  • poison,-(boron), and, (2) radial _and a'xial -reflectors which-consist of removable reflector -elements and permanent blocks i

l which are made of ' graphite, and in some cases incorporating 1

-boron or structural steel. The reactor core assembly, including q reflector,.has an overall assembly height of about 23.9 feet and j a diameter of about 27.3 feet. The approximate ~ weight of the 1

core assembly is 1,348,000 pounds. The preceding descriptien j

fa includes the core support graphite blocks.

~l .The reactor reactivity ,:on trol. in regions containing fuel _

l consists of pairs of control rods containing boron carbide, '

which are supplemented by burnable poison (boron) in selected fuel elements as esquired. A reserve shutdown sy: tem consisting p# 9009270149 poop 34 ,

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_ ,  : Fort St. Vrain'#1  ;

Technical Specifications '

y Amendment No.- '

Page 6.1- 2 6

l- of hoppers of boronated graphite balls is also provided. t

'A variable orifice' flow-control assembly is' located at the inlet  ;

.l? to'each fueled region to provide adjustment of the coolant flow through the region, j Active Core  :

c I The active core consists of hexagonal graphite fuel elements .

l- stacked in. vertical columns. The fuel elements form the active l core which is essentially a.right circular cylinder. The active l . core is- completely surrounded? by a graphite reflector and f

l. defueling elements. Within the core array, the columns are-l- grouped into-regions containing seven columns each, except for ,

l' outer corner regions which contain five columns each.

l l 'The center column of each of the: fueled regions is a control rod column. -Each control -rod column contains two control rod

.1 channelsnand one reserve " shutdown absorber material channel. l

, Each control: rod channel has a diameter of-.4.0 inches = and the- ~

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F two channels have a centerline pitch spacing of 9.7-inches. The ,

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-reserve absorber material shutdown channel has a diameter of g 3-3/4 inches. The control' rod channels are continuous from the top. face of the top ' reflector and -terminate. ia the bottom reflector at an elevation not greater than 27.0 inches above the

. top face of the core support block. The reserve shutdown absorber channel is continuous from the too face of the-top 4 reflector and terminates in the bottom fuel element at an elevation not greater than 47.5 inches above the top face of the core support block.

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  • 4 U Fort StT Vrain 11
  • ' Technical Specifications

- - ' Amendment No; Page 6.1- Sb 6

'll! second: '(counts seen by- the-_ detectors from- these sourcet)-

.II ' Jper LCO.4.4.1, Table 4.4-4.

Basis for Specification DF 6.1 The above specifications- form the general- design bases and criteria for:the overall design features of the reactor core

which were _used .to evaluate its general performance. Further

~~ details concerning these _ design features are given in

-Section III of the FSAR,-the Safety Analysis Report for Fort St.

Vrain Reload 1 = Test Elements FTE through FTE-8, General _ Atomic Document GLP-5494,-June 30,'1977, and the Safety Analysis Report

. For Re' actor Defueling,-General Atomics Document GA-C19694.-

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<l: Witti the Lreactor permanently shutdown, the Controls Rod Drive and

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ll Orifice Assembly (CRD0A) may be permanently. removed from_ regions

, ll which have.been defueled. Defueling elements contain boronated-h graphite lumped poison rods, no fuel,' no control rod' channels or

' reserve' ; shutdown'-channels. Regions loaded with :defueling-

{ li J l: elements are less' reactive than fueled regions: whose control l _ rods. are fully. inserted. Therefore, the reactivity ~ontrol c and

.l' reserve shutdown functions'normally performed by the CR00As- are g, ~l neither utilized nor required in defueled regions. In addition, y' 17 Ohe flow control function of the orifice assembly is no longer

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required in defueled regions since flow through the remaining l fueled regions will be sufficient following CRDOA. removal from E -l.

all;>defueled regions. The primary closure of the CRD0A is not

, ,l> required to be in place or operable at or below 100 psia per LCO j , l -_ 4.2.7. j g 1 e

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' ATTACHMENT 3 TO P-90273

. NO SIGNIFICANT ,

HAZARDS CONSIDERATION ANALYSIS

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N0'SIGNIFICANT HAZAR0S CCNSIDERATION ANALYSIS-

-I. Evaluation of-Changes

, Changes to -the Design Features (DF) Section 6.1 and the Basis will more accurately reflect the reactor assembly and active core configuration during the defueling process.- As fuel elements are being removed, defueling elements without control rod or reserve shutdown channels -are being added. Therefore, the control rods '

=and reserve shutdown assemblies need not be. replaced in defueled ,

regions. Also, the variable orifice flow control assemblies are not needed in defueled regions and the refueling penetration

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primary closures do, not need to be in place at or below 100 psia r primary system pressure.

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Removal of the control rods from defueled regions is acceptable '

since the rods no longer serve a function. The negative reactivity once provided by control rods in fueled regions is ,

replaced by.defueling elements which contain boronated graphite lumped poison rods in twelve blind holes in each element' .

Moreover, since the defueling elements do not contain control rod ~

channels, the control rods can no longer be -inserted into defueled-regions.

The reserve shutdown assembly of each CRD0A served as a backup to the control rods. Since the defueling elements have no reserve shutdown channels, contain no fuel, and negative reactivity is '

provided by lumped poison rods in the defueling elements, .-the-reserve shutdown hoppers containing boronated-graphite balls are no longer required in defueled regions and-may be removed.

Further- diset;sion and evaluation of defueling element design'is provided in the FSV Defueling SAR submitted to' the NRC in PSC Letter, Crawford to Weiss, dated August 16, 1989 (P-89287) and Amendment No. 74 to the FSV Technical Specifications, NRC Letter,- -,

Erickson to Crawford, dated December 1, 1989'(G-89399).

l Removal ' of the flow control orifice valves from defueled regions was analyzed (GA Document 910144, Attachment 6) with respect to core- flow requirements and decay heat removal. The core flow analysis was performed using the POKE code validated for the FSV reactor core. The analysis determined that given a decay heat value of 0.08 MW, the flow required to remain within the LC0 l 4.1.9 maximum region temperature rise limit of 350 degrees F is '

-0.75 lbm/s. The analysis shows that the core can be operated at '(

this flow rate without reverse flow occurring in any region, even during the worst case scenario when all fuel elements have been removed from a region but defueling elements have not been installed. The calculations were repeated for each region, following the planned.defueling sequence. The analysis is valid for all cases, and is conservative since current decay ht:a t is less than 0.08 MW. 1 t

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FSV- l Nuclear 1 Engineering:, Division performed _an Engineering R "u

Evaluation (EE-DEC-D033, Rev. B,' Attachment _5) to determine the personnel' radiation- exposure rates at'theLrefueling floor-level "

with the-radiation shielding portion of-the CRD0A removed from a refueling _ penetration' above. a defueled region. The evaluation war based on measured radiation- dose' rates ~ above_-an empty

= refueling penetration, i.e.,_ complete CRD0A removed, and all core regions fueled. Significant' conservatism exists :in thati the  ;

calculations were performed- based. on ' fuel. being_in all core-regions. <

The Engineering- Evaluation concluded that expected radiation  ;

exposure rates at the refueling floor above a defueled region >

with the CRDOA shielding' removed will not prohibit 1 personnel access to the area. In keeping with . good ALARA practices,  !

personnel accessu will be: administratively controlled based on actual radiation measurements taken by Health Physics personnel at the time access is desired.

After CRDOAs are removed from defueled regions, the primary closure assembly, or an alternate sealing device suitable for

  • present reactor _ conditions, will .be' installed in refueling  !

penetrations'above defueled regions. This is in accordance with '

the PCRV closure description provided in TS' Design Features Section 6.2.

,i Also included in this amendment proposal is a correction to the ,

name of the reserve shutdown bal l s .' The balls are made .of "boronated graphite" as described in FSAR Section 3.8.3. This-correction does not involve a technical change.

' II. Conclusion

, Based on the above evaluation,. it is concluded that Fort St.  :

'Vrain activities in accordance with the proposed changes will not 4 involve a significant increase in the' probability lor consequences of-an accident previously evaluated, create the possibi.lity of a new aor different kind:of. accident from any accident previously evaluated, or involve a significant reduction in. a : margin of ,

safety. Therefore, in -accordance with-the criteria in.-10 CFR 50.92, this change-will-not increase any risk-to the health and  !

safety of the public nor does it involve any significant hazards.

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SAFETY ANALYSIS.

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SAFETY ANALYSIS

'1.0; INTRODUCTION Fe rt. St._ Vrain (FSV) was permanently shutdown on August 18,

-1039. By' letter dated December 1, 1989, the .NRC issued Amendment No. 74 to the Facility Operating License, which authorized loading of defueling elements' into the. core 'and;

-permitted PSC to proceed with defueling. By letter dated May.

p 1, 1990, Weiss to Crawford (G-90102), -NRC issued an -order -

confirming PSC's commitment not to take the FSV reactor to ,

criticality and not to operate the reactor at any power . level. .

Currently, PSC has removed fuel from 12 of the 37 ' core tegions,-

and replaced fuel elements in those regions with defueling-elements.

Thel changes to the Design Feature- (DF) Section- of the Technical Specifications (TS) proposed by this submittal address the permanent removal of Control Rod Drive and Orifice Assemblies-(CRD0A) from defueled regions.

In the interest of maintaining qualified personnel to perform operations unique to FSV,_it is PSC's desire'to commence .CRD0A .,

removal from deftaled regions prior to resumption of defueling i

and spent fuel shipping activities. These'defueling activities are tentatively scheduled to begin on or about February 1, 1991. In addition to the potential . economic benefits, CRD0A removal / shipment prior to. February 1, 1991, will enhance 1 personnel morale, attention to detail, and safety awareness,-

With- these-thoughts in mind, PSC is requesting NRC approval of-this TS amendment proposal on or before November 16, 1990.

2.0 BACKGROUND

-The defueling process - is very similar.to refueling the FSV.

reactor. Figure 1 is a map of the Refueling Floor l (RF) indicating ' major components and facilities utilized during the defueling phase. Prior to defueling a reactor core region, the

Auxiliary Transfer Cask (ATC) is used to remove the CRD0A from a top head refueling' penetration (Figure 2) and place it in an equipment storage well (see Figures 3 and 4).

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_c The ATC -(FSAR 9.2;2) is an'. integral.part of the' fuel handling system-in that it is- capable of performing all preparatory functions > including . handling activated or -contaminated components in the Prestressed Concrete -Reactor Vessel- (PCRV)-

top' head and other facilities,secessible from the-RF. -The ATC is capable of safely _ removing (or replacing) CRD0As from the PCRV Equipment Storage Wells (FSAR 9.2.3), or Hot' Service Facility-(HSF). The- ATC operates withethe same Reactor Isolationi Valve- (RIV)' as' the Fuel Handling Machine and also

-with a shielding adapter used on the equipment storage wells.

(Figure -4).- The ATC^ is. transported between the reactor and-

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other RF facilities by +he Reactor Building Overhead Crane.

( Figure . 5) .

The ATC consists of a hoisting mechanism mounted in the top of aLcylindrical, gas tight' shielding unit with a gate and centering device bolted to the bottom face of_the structure. A lift extension is permanently bolted to the top of the ATC and mates with. a' specially designed crane lift extension attached to the Reactor Building Overhead Crane.

Adequate Jcontainment - of radioactive gases -and particulate matter is provided at all times by the ATC and its gate, 0-isolation _ val ve s ', shielded adapters, -wells, etc. The decay

-heat emanating from objects stored within the cask is sufficiently low so as'not_to warrant a heat removal system.

Radioactive. gases _ and particulate matter contained within the ATC are: discharged .through an internal filter to the radioactive gas waste system via thel-fuel har.dling purge system. In the-event decontamination of the ATC and grappler is.' required, these. operations will be performed in existing

-facilities on the Refueling Floor.

= Adequate shielding is provided to protect operating personnel at all times.

The ATC _is interlocked _ to prevent inadvertent discharge of its contents. As a result, the utilization of the ATC for these operations represents no undue hazard to the health and safety of the public or to the plant personnel.

The RIV (Figure 6) provides a sealed and shielded cover for open PCRV penetrations or other RF facilities when the ATC is taken away. The RIV is a horizontal, gate-type, sealing valve, consisting >of a nodular cast iron housing in which a cast block can be. moved by means of an electric-motor-actuated acme screw.

Gas tightness between the valve housing and reactor, and between the housing and gate is accomplished with inflatable seals. Sealing between the valve housing and the ATC is accomplished with a Gask-O-Seal. For accommodation of the ATC-on the RIVs, 15 clamp blocks are screwed on the circumference of the valve.

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T-The' . only dynamic parts of the RIV penetrating the- PCh/ primary -  !

T l boundary are-sealed actuating- shafts- for the.~ valve': gates.  :

There'are static penetrations for purging, instrumentation,'and controls, all-of which are provided with conventional seals.

-Differential pressures across seals and valves are extremely low. During fuel-handling and maintenance in.the reactor, LC0

'4.7.1 requires the PCRV to be depressurized to atmospheric t pressure or slightly below.

The failure and/or misoperation of a valve leading tu a stop at .

an unwanted position,-or the inadvertent closing of the valve on a CRD0A. or the' ATC mast could damage the. valve gate but-would not lead to a break in the primary coolant, barrier formed' -

-by the inflatable seals associated with the RIV.

In general, location of' the seals is such that they are not .

directly exposed to the effects of failures of the . handling 4 mechanisms. Since the reactor vessel is maintained at i atmospheric pressure or slightly below during defueling, ,

, failure of1 a- primary seal at any point in the containment  !

boundary formed by the ATC would result in a slow inleakage of t air into the reactor, rather than a release of primary coolant.

Ten equipment storage wells are provided for the storage _of new L , or contaminated CRDOAs, as' well as other components. 'The wells  ;

are gas tight, shielded and provided with a'normally sealed ,

drain at the bottom to facilitate decontamination activities,

'The equipment- storage wells provide a passive function only, with no heat loads or_ ventilation requirements. -Further, the t amount of radioactive material ' contained in-the control and

  • orificing assembly represents no hazard even _if an accidental release'of_ contaminated gas were to occur. Safety of personnel-in the Reactor Building is adequately prov;ded by shielding, sealed storage,'and purging of the contaminated gas by the fuel .

handling purge system. 3 The HSF (Figure 7) is provided to allow the' decontamination,

-service and repair of contaminated and activated equipment such as the CRD0As. The ATC is used to deposit a CRDOA into the HSF r so that select parts can be removed from the CRDCn and prepared t 3f for off-site shipment. Service operations can be performed j# either indirectly with manipulator equipment.or manually after

-decontamination. Provisions are made for partial shielding of portions of an assembly while work is proceeding on an' exposed [

portion. The shielding design of the facility conforms to

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applicable regulations and is adequate for 1 activated / contaminated CRD0As.

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-The facility ! consists?-of two areas, one for relatively short' items-such as the helium-circulator, andE the other for long

-items such ast the CRDOA. Both areas are serviced byLtwo-

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- remotely' operated manipulators fer disassembly and handling of components. The longer area is serviced by a movable platform.

which straddles the CRDOA- supported at- preselected: verticali locations by means of retractable support pins. At the bottom-of the service platform lower pit,-provisions have been made, for a water pool which acts as shielding for possible storage-of. highly contaminated items if this should-ever beLneeded.

Ventilation- is provided by-a fresh air inlet located below the .

service floor elevation and an outlet exhaust blower and filter assembly located near the top-of the cell with ducting to the Reactor Building. exhaust. The filter assembly is removable by either manipulator. -

The- ' structural- design of the facility. is such that not significant release of radioactive material will-occur.

The design and operation of the HSF is such that radioactive-materials 'are adequately contained under all_ normal; -

and abnormal conditions to protect the health and safety.of the

-public. Adequate shielding, ventilation and: administrative controls are:provided to protect-operating personnel.

Adequate shielding is provided for all personnel, machinery and facilities as discussed above. Shielding is designed - to meet-the FSV Shielding Design Criteria as detailed in.FSAR Section 11.2.2.

CRDOA. Di sa s sei..bly Currently, CRD0As from defueled regions are replaced into

. refueling penetrations or into an equipment storage well. Each JCRDOA 'is comprised of many individual components and subassembliesLas is shown in Figurer 8. The complete assembly may be separated at the point indh , ted on- Figu:e 8, with the lower portion permanently removed from the PCRV, and prepared for disposal as radioactive waste.

The control rod drive and orifice control mechanism, known as the 600 Assembly- (Figure 9) and shown in FSAR Figure,3.8-9, or' an alternate- sealing device suitable for present. reactor conditions, wil.1 be installed in the refueling penetration 1 above a defueled region. The CRD0A's position relative to the reactor : ore, PCRV- concrete, primary seal and secondary seal is shown in Figures 2 and 10. .

Each of the components which are an integral part of a CRP0A are discussed below, and can be seen in Figure 8. Reactivity control provided by the CRD0As in fueled regions is discussed in detail in FSAR Section 3.5.3.

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~ (( o Control Rods-The- control-- rods, also callAd a' orber strings,'are shown in

' Figure 11. There'are two' rods associated-with each CRDOA.

. The": rod . pair is attached , to the rod drive mechanism via cables and can be- raised or lowered -simultaneously, as "

desired.for reactivity control.in a fueled region. Prior to removal by the ATC from a fueled region, . the rod ' pai r - i s fully retracted into the _ control. rod guide tubes and mecha'nically locked into place. The ATC is then used to remove' the CRD0A.from the refueling penetration.for-storage in an equipment storage well or disassembly. Following replacement' of '"el elements in a region with defueling' elements, the cons.ol rods no longer serve a . function. If

+ .the CRD0A is placed.back in the defueled region, the control rod pair. remains fully retracted and locked in place- since defueling elements ~contain no control rod channels, q j

Orifice Valve l As shown inJ FigureL8, the variable-orifice flow-control assembly is attached to the CRD0A, lower end. Its position in. the reactor in relation to a region is shown in Figure

12. The orifice valve provides adjustment of helium coolant flow through the region. The need~to retain the orifice valve:in-a defueled region is analyzed in Attachment 6, GA Document = No, 910144,. dated June 21, 1990.

The conclusion drawn;by this analysis is that the variable orifice assembly j can be- removed from a defueled region--with no deleterious {

effect on flow through fueled regions.  !

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'A' radiation shield is contained in the housing just below- j the-control. rod drive and orifice contr.ol mechanisms (Figure j.

8). This shield assembly extends downward =to the' lower end '

of the refueling penetration. Starting at the top of the

-. shield directly under the-drive mechanisms is approximately

8. inches of lead, below which is approximately 14.5 .. inches of boronated graphite. These two thicknesses of material <

comprise the-. shield plug which sits on approximately -45  :

inches of granular graphite.

i Public< Service Company of Colorado Ench.::ri'g Evaluation,

- EE-DEC-0033, Rev. B'( Attachment 5), de. ermined the r adiatha i exposure; rates at the Refueling Floor above a defueled region wi.th the CRDOA radiation shield removed. The ,

conclusion of the evaluation is that normal ALARA practices, 2 1.e. , Health Physics administrative controls, are adequate to ensure personnel safety while working on the Refueling >

Floor or above a defueled region.

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Reserve' Shutdown System F 'Alreservesbutdownsys'temindependentofthenormal~ control'.

rod aystem:is provided.. The system -consists of a hopper

- (container), _ rupture - disc,- and guide tube ~ (Figure 8).

Neutron _ absorbing . material, in the form of .,boronated graphite balls,--is stored in a hopper within each CRDOA from which it can be released, if required, by the operator and.

allowed to- fall into channels in' the fuel: elements. A cylindrical channel with. closed bottom:is provided in each fueled region connected by guide tubes to the un brsides of the storage hoppers. The absorber material. is introduced

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F in;o the core channels through the guide tubes from the E hoppers .( Figure 12). Refer to FSAR Section 3.8.3 for further detail.

The hoppers ~'are -

integral to each'CRDOA located within the refueling penetrations. The absorber material is released.

from each hopper _by a rupture disc actuated by gas press ~ure which is introduced from external helium sources.

If the CRD0A is placed back i_nto a defueled region, the external helium sources for rupture: disc actuation are not reconnected. This will prevent inadvertent pressurization of.the' hopper and subsequent insertion of' the reserve shutdown material for which there is no corresponding-channel.-

CRD0A Primary Seal The. refueling penetrations in- the top head of the PCRV 1 provide access for removal and replacement of fuel, reflector, ,and: defueling elements. These penetrations

currently'contain CRDOAs which can be removed for carrying out defueling operations through the penetrations. Figure 10 shows one of *.hese penetrations with .the-' CRD0A in-position.

The CRDOA ' housing has a flange approximately. 45 inches below

, its top face. This_ fits'into a counterbore provided in -the

, refueling penetrat_1on, approximately 46' inches deep. The CRD0A'is bolted into the penetration at this point. The

, housing forms the primary closure for the penetration, and is held in position by a flange bolted to_a. support ledge in the base of- the penetration liner. This. support ledge is

~1ocated approximately 48 inches below the- top of the penetration liner. The primary closure seal is an elastomer 0 ring face seal. The secondary closure member is a flat, hexagonal plate assambly bolted to the top flange of the penetration iU.ar anci sealed by an elastomer 0-ring face seal, Reference FSAR Section 5.C.2.5.1 for additional details.

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3.0 ANALYSIS w .

3.1' Permanent Removal of Control Rod Drive and Orifice Assemblies L , from Defueled Regions l

i The CRDOAs p ovide the following functions:

, 1. The contru rod pairs and drives provide rr. activity control j: in fueleo regions as is described in FSAR Section 313.

[ 2. The variable-orifice flow-control assembly provides adjustment of helium coolant flow through each of the fueled regions (FSAR Section 3.2.2.7).

3. A radiation shield is contained within the CRDOA just below L the control and orifice drive mechanisms (FSAR Section 3.8.1.1.1).
4. A reserve shutdown (RSD) system, functionally independent of the normal cuntrol rods, is provided. Boronated graphite balls are contained in a hopper which is an in'.egral part of the CRD0A. Should the normal control rods fail to insert for any reason, the RSD material can be released into the -

RSD channel in fueled regions.

5. The CRD0A housing fo>ms the primary closure for the 37 refueling penetrations. The closure system is described in ,

FSAR Section 5.8.2.5.1.

Amendment No. 74 to the FSV Design Featuus Technical Specification approved the use of defueling elements containing boronated graphite lumped poison rods for reactivity control, '

no fuel, no control rod channels, and no reserve shutdown channels. .Since reactivity control is not a concern in defueled regions, CRDOA Functions #1 and #4 above are no lonu ?r "

required in defueled regions.  ;

Removal of the flow control orifice valves from defueled regions has been analyzed and determined acceptable since flow

-through the remaining fueled regions will be sufficient for decay heat removal. The attached analysis (GA Document 910144) determined that, given the decay heat value of 0.08 MW, the flow required to remain within the 350 degree F maximum region temperature rise 1 sit of LCO 4.1.9 is 0.75 lbm/s. The primary coolant circuit can be operated at this flow rate without reverse flow occurring in any region, even during the worst case' scenario when all fuel elements have been removed from a region but defueling elements have not been installed. The core flow analysis was performed for each region using the POKE code validated for the FSV reactor core, and, since the current decay heat value is less than 0.08 MW, the analysis is conservative. Therefore, function #2 is no longer required for defueled regions.

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within each CRDOA is not required for defueled regions in the permanently shutdown FSV reactor (Function #3). The EE concluded that expected radiation exposure rates- at the refueling floor above a defueled region with the CRDOA shielding removed will not prohibit personnel access to the area. Normal ALARA practices will be adequate based on actual radiation measurements taken by Health Physics personnel at the time access is desired.

Since the FSV reactor has been permanently shutdown and depressurized, the primary penetration closure (Function #5) is no lorger required to be in place. FSV TS LCO 4.2.7 requires all primary and secondary penetration closures be in place and operabic anytime ths PCRV is pressurized to greater than 100 psia. The primsry closure portion of a CRDOA can be detached from the remainder of the assembly. This portion of the CR00A, or an alternate seal ng device suitable for present reactor conditions, will be installed in refueling penetrations above defueled regions.

Initially, PSC will retain 37 CRDOA 600 Assemblies on-site, although their disposal may be pursued at some future date.

The 600 Assembly includes the primary closure for each refueling penetration. This practice is censistent with past outage practices, and it will ensure PSC's capability of establishing rated PCRV conditions, with the exception of the degraded steam generator main steam ringhtaders which were previously de-rated because of material deficiencies. PSC will request NRC aciroval prior to disposal of PCRV penetration closures.

4.0 CONCLUSION

The proposed amendment to the Design Features Section of the Technical Specifications would allow removal from the PCRV of CRD0As in defueled regions that have no required or useful function c, A q any planned or postulated defueling or shutdown cond i' '.o'i .

Approval of this proposeC amendnient will not be detrimental to the common defense and security and will not create any undue risk to the haalth and safety of the public.

PSC has concluded, bared on the preceding analysis, that adequate structural Integrity, cooling, shielding, and shutdown margin will be maintained in the FSV reactor core during defueling, with removal af control rods, orifice valve, r&diation shielas, tnd reserve si.otijown assemblies from defueled regions.

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