ML20210C865

Application for Amend to License DPR-73,revising Tech Specs to Be Consistent W/Current Plant Conditions & Operations & to Provide Transition Through Remainder of Cleanup of Facility.Fee Paid

ML20210C865
Person / Time
Site:	Three Mile Island
Issue date:	04/23/1987
From:	Standerfer F GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20210C858	List: ML20210C856 ML20210C865 ML20210C901
References
NUDOCS 8705060350
Download: ML20210C865 (38)

Text

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METROPOLITAN EDISON COWANY JERSEY CENTRAL POWER AND LIGHT COMPANY' PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR THREE MILE ISLAND NUCLEAR' STATION UNIT.II Operating License No. DPR-73 Docket No. 50-320 Technical Specification Change Request No. 53 This Technical Specification Change Request is submitted in support of.

Licensee's-request to change Operating License No. DPR-73 for Three Mile Island Nuclear Station Unit 2. As a part of this request, proposed replacement pages_for Appendix A are also included.

GPU NUCLEAR By A/ite President /Dirgetor Swornandsubscribedtomethis[2 day of #g/ , 1987.

sd Notary Public

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UNITED : STATES OF' AERICA .

NUCLEAR REGULATORY COM ISSION IN TE MATTER OF DOCKET NO. 50-320 LICENSE NO. DPR-73 GPU NUCLEAR This is to certify that a copy of Technical Specification Change Request-No. 53 to Operating License DPR-73 for Three Mile Island Nuclear Station Unit 2 has been filed with the U.S. Nuclear Regulatory Comission and served to the chief executives of 1) Londonderry Township, Dauphin County, Pennsylvania, 2) Dauphin County, Pennsylvania, and 3) the designated official of the Commonwealth of Pennsylvania by deposit in the United States mail,"

addressed as follows:-

Mr. Jay H. Kopp, Chairman Mr. Fred Rice,-Chairman Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House-Middietown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection PA Dept. of Environmental Resources P.O. Box 2063 Harrisburg, PA 17120 GPU NUCLEAR

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By Vice President /Direotor, TMI-2 Yks Y

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i Three Mile Island Nuclear Station, Unit 2 (TMI-2)- -

Operating License-No. DPR-73 -l Docket:No. 50-320 l

. . Technical Specifications Change Reauest-(TSCR) 1k). 53 The Licensee requests that_the attached Sections l',-2'and 3.and. Bases be.

substituted for existing Sections 1,.2_and 3 and Bases:in-the Technical.

. Specifications. -Due to the extensive' nature of. Technical Specification. Change, Request No. 53, it was decided that the most. efficient manner to accommodate L Lthis: change was to replace the affected sections;in'their entirety._ Revised, copies of_the affected indexes also have been included with this. change and should be replaced as the affected sections are. replaced.

The purpose of the; proposed modifications.is to_ revise the TMI-2' Technical Specifications'to be consistent with current plant-conditions and operations-cand. provide a. logical transition through'the remaindercof the cleanup of' i- TMI-2. The reason and' justification for each change is. described in.the attachment.

! No Stanificant Hazards Consideration ,

i i Page 1 of the attachment presents a No Signficant Hazards. Determination i

pursuant to 10 CFR 50.92.

Amendment Class i

Pursuant to.the requirements of 10 CFR 170, Licensing Fees, an application fee i of $150.00 is enclosed.

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t TECHNICAt$ SPECIFICATION CHANGE' REQUEST 53 This Technical Specification Change Request (TSCR) tis'being submitted to revise the-r LTMI-2 Technical Specificationsito be consistent.with current plant conditions and '

L  : operations ~and;to provide a~ logical-transition through the; remainder of the: cleanup;

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of TMI-2. Transition to the' completion of the cleanup is provided by the adoption'

.of; specific " Facility Modes" and definition of the mode (s) during which each-Technical Specification is, applicable. The' result is a phase-out:of certain; i

.TechnicalESpecifications as the cleanup progresses. -The bases'for the Technical

' Specifications have been reviewed and updated toLbe. consistent with this:and' .

previous'TSCRs.

This Change-Request also assumes NRC' approval of TSCRs'55 and 56.

,- Three " Facility Modes" have been adopted to accommodate the'transttion through~the

-remainder of the cleanup. These. modes are defined as'follows:

.l.- Mode 1_-.The'. current period,-during wh'ich Reactor Vessel.(RV).defueling and other major tasks.are in' process.

2. Mode 2 - The period subsequent,to completion of defueling._of the RV and Reactor Coolant System and prior to completion of the: core debris shipping '

i program.

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3. Mode The period subsequent to shipment of.the last core material off-site.

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. NO SIGNIFICANT HAZARDS DETERMINATION i' 10 CFR paragraph 50.92 provides the criteria which the Comm'ission uses to evaluate a No Significant Hazards _ consideration. 10 CFR 50.92. states that an amendment to a

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facility license involves No Significant Hazards if. operation of.the facility.in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an I accident previously evaluated; or-i 2. Create the possibility of a new or different kind of accident from any accident previously. evaluated; or

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i 3. Involve a significant reduction in a margin of safety.-

1 These proposed changes to the Technical Specifications are'primarily_ administrative in nature and recognize the transition through the remaining milestones'in the 4 post-accident cleanup process. Although-this transition recognizes that opera--

l bility of certain equipment and the monitoring of'certain" parameters are no-longer ~

required subsequent to the achievement of a given milestone,othe resultant overall J protection of the public will have_been enhanced by the combined effects of cleanup i- complimented by retention of selected system operations and plant monitoring requirements.

The transition from Mode 1 to Mode 2.is based on-defueling the RV and Reactor . ,

i Coolant System and precluding criticality. Subsequent to Mode 1,- the systems'and parameters associated with monitoring and protecting the Reactor Core are not required to be maintained. Borated Water Injection capability, Reactor Coolant-t System water control, Reactor Coolant System temperature control and Neutron

? Monitoring are illustrative of systems and monitoring capability which will not be required subsequent to achieving the conditions specified for Mode 2. The recognition that these systems and monitoring capabilities are no longer required does not involve an increase in risk to public health and safety. The former basis

for requiring these capabilities (i.e., the Reactor Core) has been removed.

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During Mode 2,: containment' isolation, containment purge, and sub-positive Reactor I Building (RB) pressure will be maintained to mitigate a possible unplanned release j of radioactivity to the environment. The current Technical Specification to  !

maintain a required boron concentration for Spent Fuel Pool ' A' will be maintain'ed to ensure subcriticality while canisters containing. core material are stored therein. The canisters / debris containers are designed to ensure subcriticality

.during fuel. handling. The previously approved Technical Specification requirements applicable to the primary activity during Mode-2.(i.e.,; hand 11ng and transfer of canisters in the Fuel Handling Building (FHB)) will be maintained. Thus, there is no increased.rlsk to public health and safety resulting from fuel handling activities.

~The transition from Mode 2 to Mode 3 is based primarily on' shipment of all core material off-site. Therefore,_those capabilities relevant'to the safe, temporary .I storage'and handling oficore material on the TMI-2 site are no. longer required.

The deletion of these requirements results in no increased risk to the public since

.the basis no longer exists.

The removal and shipment of the Reactor Core results in a significant decrease in the probability and consequences of accidents previously evaluated. .The fact that the Reactor Core is no longer on the THI-2 site eliminates-possible core-related accidents. Also, accidents indirectly associated with the continuous generation of fission products through operation of the reactor, i.e. Haste Gas Decay Tank failure, have been eliminated. Therefore, the removal-and shipment of the Reactor Core yields.a significant decrease in the probability and number of accidents to be considered; furthermore, elimination of current Technical Specification

-requirements intended to mitigate these accidents does not increase the risk to the public as the basis no longer exists.

The activities related to removal and shipment of the Reactor Core have been analyzed and approved by the NRC. Therefore. the possibility of the evolution of a new or different kind of accident from any accident previously evaluated during defueling and fuel shipment, or subsequent thereto, is non-existant. However, the removal and shipment of the Reactor Core eliminates the possibility of occurence of many accidents previously evaluated. Therefore, present Technical Specification requirements intended to mitigate such accidents are no longer required. Since no new nuclear capabilities have been added, no new or different accidents have been added beyond those previously evaluated.

The removal and shipment of the Reactor Core results in a significant increase in the overall margin of safety for TMI-2. The consequences of a range of events have isen evaluated for THI-2 based on the plant condition subsequent to the removal and

• hipment of the Reactor Core. The consequences of the worst case event for Mode 2, a fuel handling accident, have been previously evaluated and the Technical Specif1-cations required to mitigate such an event remain in place for Mode 2. For Mode 3, the consequences of the worst case event are less than limits that have been recognized as representing no significant threat to public health and safety (e.g.,

less than the guidelines established by 10 CFR 50 Appendix I). Thus, a significant reduction in the scope of the Technical Specifications is appropriate and can be accomplished without an attendent reduction in a margin of safety.

From the above review, it is concluded that the changes proposed in this TSCR do not involve Significant Hazards.

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l GENERAL ADMINISTRATIVE CHANGES

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' DESCRIPTION OF CHANGES-Sections'3.-l.3,c3.1.3.1,3.3.3.3,3.3.3.6,3.4.3,311.3b,':3.6.4,3.6.4.1, 3.6.4.3, 3.7.1, 3.7.2,' 3.7.3, : 3.7.3.1, 3;7.3.2, 3.7.3.3,'. 3.9.14; . Tables 13.3-7,

! 3. 3--10, 3. 3-1,13. 3-3,1 3.~ 3-4, ; 3. 3-8,E 3. 3-9, 3. 3-11, 3. 7-4,~ . and Figure 3.6-1,

have been removed from the Technical Specifications. Numerous

other changes!

" :were made to correct unclear wording'and. typographical errors. Additionally,'

this:TSCR reflects changes proposed tolthe Technical Specif1 cations.vla.the' ~

following-change requests:

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o' - LTSCRL49 submitted' via.GPU Nuclear letter 4410-85-L-0110 dated-

" June 18,3 1985 .

'o TSCR 51 submitted via GPU NuclearLietter-4410-85-L-0135 dated.

July 31, 1985.

o. TSCR'55-submitted via GPU Nuclear letter 4410-87-L-0008 dated.

o January 27,_1987. ..

1 o TSCR 56 submitted via GPU Nuclear, letter 4410-87-L-0029 dated ~

February.,5, 1987.

REASONS AND JUSTIFICATIONS FOR CHANGES I Tables 3'.3-1, 3.3'-3, 3.3-4,_3.3-8, 3.3-9, 3.3-11Eand 3.7-4 were,1dentified as

being " Transferred to the~ Recovery' Operations-Plan." Figure 3.6-1orthe-remaining tables and all the above' referenced sections were ~1dentified as l being " deleted". Since these sections, tables ~and figure no. longer ' serve a

j. useful purpose they have been removed from the Technical Specifications.- The

, typographical-errors and unclear' wordings were corrected to avoid the i perpetuation of errors and potentially ambiguous wording.

2. TECHNICAL SPECIFICATIONS INDEX DESCRIPTION AND REASON FOR CHANGE The Technical Specification Index has been; revised to be consistent with other Technical Specification changes.

JUSTIFICATION FOR CHANGE 1 .

The revisions to the Technical Specifications Index are administrative changes to make the Index correspond to the revised Technical. Specifications and do ,

not require technical justification.-

. 3. TECHNICAL SPECIFICATION 1.3 " DEFINITION OF RECOVERY MODE" f

DESCRIPTION OF CHANGES The definition of RECOVERY MODE has been deleted and replaced with:

i MODE 1.3' The-FACILITY MODE shall correspond to the plant conditions specified in-Table 1.1.

Table 1.1 has been added to define three facility modes.

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LREASON'FOR~ CHANGE;  ; -

q l Adoption-ofithree Facility Modes ~ whidh directly'correspondLto periods 1' following thel achievement of'the.two-major; milestones remaining.in.the post. .

accident cleanup.of.TMI-2,~ facilitates 1the early review and-timely: transition-

-- .from.one plant status

to another. 'These Facility. Modes allow-the identift ~

cation ~of specific plant conditions Lwhich require various limiting conditions

.- for operation and the associated surveillance r.equirements. 'Once:these-proposed Facility Modes.and their2 applicability.have:been approved,Jthe needl >

> - for:further technical revisions of.the Technical.. Specifications:during the , ,

remainder .of Lthe. Cleanup Period,will be' minimized.' : Alsoi the use ;of: three.

Facility Modes simplifies .the applicability-statement in nthe Technical Spect-:

fications and provides clearer definition.of .the applicabillty of.~each  !

technical; specification. ,

lUSTIFICATION FOR CHANGE -

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[ Since the' period of time bounded;by the.three modes is exactly;the same as( .

-" RECOVERY MODE'," definition of three specific Facility Modes constitutes'an

' administrative change lwith no safety. implications. The Facility Modes are o -based on,the achievement of the two major milestones remaining in the'posty .

accident cleanup'of TMI-2. 1They are. completion of defueling'of the RV_and?the ,

shipment.of the core material off-site. These. Facility Modes. differ from the ~

1 Operational Modes utilized at a normal. power plant in that,once a specific milestone is achieved, the change _from one Facility Mode to another-is an

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4 F evolutionary phase of.the cleanup: process. This evolutionary nature (1.e.,

L progressing from one mode to another) is a significant; element of the basis-of' y this Technical. Specification revision. This evolutionary pr_ocess envisions.

I that systems and equipment which are not. required for a subsequent mode will l be removed from service and deleted from the Technical Specifications: based:on j the applicability statement of a specific Limiting Condition.for Operation.

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While the' definition of the three Facility Modes is:an administrative change, the applicability to specific Limiting Conditions for Operation may have

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!. safety implications. These safety implications, as. applicable, will be~

discussed as part of the justification for the' specific: Limiting. Condition for  !

Operation. "

l The primary nuclear safety requirement;throughout the TMI-2 recovery and .

defueling modes was the prevention oftan inadvertent' criticality. During the Mode 1 operations, prior to and during defueling, . nuclear criticality safety was assured through the introduction of specified concentrations of boric-acid. 6 i soluble neutron poison in the reactor coolant system (RCS) and a combination of limited geometrical arrangements, soluble and solid poisons, and:-

. administrative requirements for other systems and activities. ~

Mode 2 is established when all-defueling canisters have been relocated to the FHB and only a small quantity of residual fuel remains in:the RB. : Mode 3

]- begins after all defueling canisters'have been shipped from.the.TMI-2: site.  ;

! The FHB and RB are, in general, isolated'in terms of neutron' interaction.

E Thus, their systems and activities can be treated-separately. . The criticality- ,

safety criteria currently imposed.for the'FHB will continue to be met during ,

i Mode 2. .During Mode 3, the fuel. remaining in.the FHB willibe significantly-~ l less than a critical mass and inadvertent criticality will~not be a^ concern; Criticality concerns relative to the residual fuel in the RB is the same for-l Modes 2 and 3 and is evaluated below. l i

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Criticality safety is sup' ported by_ removing from each region of the plant-as .

much fuel as reasonably _ achievable, and.by demonstrating that the verified l residual fuel quantitles-and locations constitute a substantially subcritical configuration _under normal and credible abnormal. conditions.

During defueling operations and other recovery activities of Mode 1, a fuel mass of 70 kg was established as the administrative limit for a single volume-

-to. ensure subcriticality. This value represents 75% of-the calculated minimum critical mass of 93 kg. Said calculations have been conservatively based on worst-case assumptions such as a. spherical fuel configuration with_ full -

reflection by unborated water, optimum (i.e., maximum reactivity)~ spacing of__ _

fuel particles and unborated water, 3% enrichment for all fuel, and no assumed

-credit for effects of poisons or structural _ materials.

In Modes 2 and 3, criticality concerns shift from relatively large masses of i -fuel to smaller residual fuel quantitles distributed throughout the_RB. -Since 1.t may not be possible to reduce the total inventory ~to.less than the administrative limit of 70 kg, knowledge of fuel quantitles and distributions must be considered demonstrating the subcriticality of single quantitles of fuel within the system as well as overall system subtriticality based on the total amount and distribution of residual fuel.

- Criticality Indicator Method

-The Criticality Indicator method is useful for this purpose. It is based on

, data in:

e 1. Standard ANS 8.7-1982 (formerly ANSI-16.5), " Guide for Nuclear Criticality Safety in Storage of Fissile Materials,!'

2. J. T. Thomas, E.d., "The Nuclear Safety Guide: TID-7016, Revision 2,"

NUREG/CR-0095 (ORNL/NUREG/CSD-6), U.S. Nuclear Regulatory Commission,-

June 1978.

The methodology is based on the " analogue ~ density" principle _that recognizes that arbitrarily large quantitles of fissile material can be maintained subtritical if separated into smaller volumes,-spaced, and prevented from mixing. It has been determined that a single parameter, the criticality index (CI), can be used to quantify the general interaction properties-(a combina-tion of mass, isotopic. content, moderation, physical form, and volume) engen-dered by a given " container" such that any combination of materials can be stored in any configuration so long as the sum of the CIs does not exceed 100.

The ANS-8.7 Standard imposes the following restrictions on use of_ the method:

1. "The units shall be centered to within 10% of the cell dimension. This restriction may be relaxed to permit freedom of horizontal position provided the unit mass limit is reduced to 60% of the stated value. If this-reduced value exceeds 20% of the unreflected spherical critical mass, the minimum unit surface separation shall be 152 mm (6 in.).

The tabulated values may be applied to other than equilateral units in non-cubic cells provided the unit and cell volumes are maintained and provided the ratio of the dimension of the unit characterizing its shape is approximately equal to the ratio of corresponding dimensions of the cell."

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2. "Any mass limit may be applied to a non-cubic cell equal in volume to that tabulated containing a near equilateral unit provided the largest dimension of the cell does not exceed the smallest by more than a factor of 2.5."

3. The method is intended to apply to compact arrays reflected by close-fitting concrete less than 127 mm (5 in.) thick and to steel containers or shelving less than 12.7 mm (0.5 in.) thick.

Since THI-2 volumes generally will not contain For fuel centered the THI-2 in the spherical reflected designated volume, the 60% reduction is appropriate.

critical mass of 93 kg (noting this is conservative since the unreflected 3 value is larger), masses up to 18.6 kg for containers of at least 15 ft will not exceed the 20% criterion that requires spacing between unit cells. Since The above calculation establishes applicability for the THI-2 material.

CI values are not available for a 3 wt% enriched UO2 , those for 5 wt% are applied as described below.

The method's restriction on shapes is intended to prevent long thin geometries from being used arbitrarily in place of cubes or spheres of the same volume.

For minimum tabulated volumes, the limitation on ratio of dimensions must be applied. The ratio need not be applied to containers that Although are large enough to thick encompass the cube (or sphere) of the referenced volume.

concrete is present at THI-2, it does not fit tightly around the systems structures defined as the storage array. Steel containers are generally thicker than those allowed by the method, but the effect is mitigated in the TMI-2 application by having larger volumes, more spacing, and other substantial conservatisms (as delineated later).

Application The Criticality Indicator method is applied to THI-2 by recognizing that each fuel quantity remaining after defueling will be confined within a particular volume. Since each fuel mass remains only because it could not be removed by reasonably available means, the volume in which it resides constitutes a

" container." The building and equipment components, therefore, can be divided For example, Steam into regions each of which may be assigned a CI-value.

Generator A may be considered to be a single container.

The RV is somewhat different. Taken as a whole, it may contain more than the administrative limit for subcriticality of 70 kg of fuel. The fuel will be distributed among several locations and may be in the form of a tightly adherent film (e.g., the plenum region) in granular form in cracks and crevices (e.g., between the core barrel and core former), or in a congealed mass (e.g., within the core support assembly). Thus, it is considered to be However,

" contained" and CI-values can be assigned on a region-wise basis.

GPU Nuclear will continue defueling until the potential for criticality clearly is no longer existent. Because of this unique condition, however, a detailed analysis of the residual fuel remaining in the RV cannot be performed until defueling is nearly complete and post-defueling conditions can be reasonably inferred.

Figure 1 depicts the principal locations where it is expected that residual fuel may be located. Table 1 gives current preliminary estimates of i

post-defueling masses, locations, and contained volumes external to the RV. i CI-values for 5 wt% enriched fuel (as taken from Reference 2) are listed in Table 2 as a function of fuel mass and volume. These values from Table 2 are used to assign to each region in the reartor building the corresponding 0083P

l Y 'CI-values listed in the third data coluzn of Table 1. The sua for the column, is a CI of 0.27, a value significantly below the 100 allowed by the method.

Thus, it is woncluded that the ex-vessel fuel locations would be substantially subcritical.

The system volumes.shown in Table 1 are very large with respect to the contained mass. All of these latter volumes, in fact, exceed the 15 ft3 maximum used in Table 2-.for assigning CI-values. Each volume could hold 18.2 kg, a value below the "20% of the critical mass" restriction noted above, and be assigned from Table 2 a CI no greater than 0.10 (a value still well below unity, the rough maximum for individual units to which the method may be applied.) Assuming a CI of 0.10 for each region, hypothetical masses are shown in the fourth column of Table 1. The total of the CIs in the fifth column has now increased to 2.0, but is still well below the method's limit of 100 thereby demonstrating substantial subcriticality. This second analysis shows that there is flexibility if one or more of the targets for residual mass (Table 1 first data column) is not attainable.

The conclusion that the system external to the RV is subcritical is more than reinforced by noting the additional conservatism inherent in the following observations.

1. THI-2 fuel has a maximum enrichment of slightly less than 3 wt% while the data in Table 2 assumes the higher value of 5 wt%. For reference, TID-7016, Revision 2, shows that the subcritical mass limits (based on spherical configuration, optimum fuel / moderator arrangement, fuel water reflection by unborated water, and a subcritical Keff-value of 0.95) for 5 wt% and 3 wt% material are approximately 27 kg and 75 kg, respectively.

2. The volumes are often separated by sizable distance (e.g., as shown by Figure 1) as well as by concrete or shielding. The Criticality Indi-cator method allows arbitrary arrangement of the contained volumes including their intimate contact with each other.

3. The THI-2 fuel is likely to be in a dry condition which assures subtri-ticality independent of mass or geometric configuration. The Criti-cality Indicator method was applied assuming optimum moderation with unborated water.

The CI sums of 0.27 and 2.0 shown in Table 1, therefore, contain additional conservatism above that inherent in the methodology. Thus, by comparing the as defueled CI-values to the maximum allowable CI of 100, the THI-2 ex-vessel fuel locations can be judged to be substantially subtritical subject to verification that the as-defueled residual fuel distribution and the associated system volumes and configurations have a CI-value of less than 100. In the specific case of the THI-2 RV, a separate evaluation will be performed to demonstrate that it is inherently subtritical (i.e., CI-value of less than 100) once the final configuration of the RV can be reasonably inferred.

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TABLE 1 Hypothetical Post-Defueling Conditions Conditions'(2)

Estimated Approximate Fuel Mass Contained Criticality Mass Criticality Reactor Bu11dino Location (kg) Volume (ft3) Index (1) -(ko) Index (2)

Reactor Vessel (3) ---- - - - --

---

------- ---- ^l Upper Plenum Assembly < 11 1400 0.02 18.2 0.1 -1 Horizontal RC Pipe Runs (6)* < 24- 1000(ea) 0.06 18.2(ea) 0.6 RC Pumps (4)* < 12 400 0.04 18.2(ea) 0.4 RB Basement- < 2 248,000 0.01 18.2 0.1 Decay Heat Drop Line < 5 30- 0.01 18.2- 0.1 .

Pressurizer, Spray Line < 6 1500. 0.01 18.2 0.1 and Surge Line-Core Flood Tank 1A < 0.2 410 0.01 18.2 . 0.1 Core Flood Tank 1B < 0.2 410 0.01- 18.2 0.1 RV Drain Tank < 0.1 970 0.01 18.2 0.1 Letdown Line & Letdown Coolers < 4 43 0.01 18.2 0.1-

"A" OTSG Upper Head, OTSG Tubes < 8.2 2337 0.01 18.2 0.1

& "A" OTSG Lower Head /J-Legs "B" OTSG Upper Head, OTSG Tubes < 16 2337 0.07 18.2 0.1

& "B" OTSG Lower Head /J-Legs Criticality Index Total '0.27 2.0

• The number in parentheses indicates the number of the specific category listed, i.e.,

there are six horizontal RC pipe runs. Except.as noted, the values given in the respective columns are the totals for all the respective categories.

(1) Assigned numbers based on comparing mass and contained volume with numbers in Table 2. For multiple units, the CI-value is the sum of individuals CIs assuming the given mass resides separately in each of the units. <

.(2) Based on a maximum mass that could occupy the contained volume while keeping the CI no greater than 0.10. (NOTE: In some cases it may be possible to subdivide these  !

contained volumes depending on the final configuration of the residual fuel.) l (3) A separate evaluationfor the RV will be performed when the post-defueling conditions l for the RV can be reasonable inferred. l 1

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' TABLE 2 VALUE OF CRITICALITY I EEX ASSIGNED TO A V0Lt M IN A rnurorTE REFLECTED STORAGE AREA FOR URANIUM DIDXIDE ENRICHED TO 5 WTY IN 0-235 (The sum of criticality indicators in a storage area shall not exceed 100)

Volume (ft3 ) .67 1.34 2.01 2.68 3.35 4.02 4.69 5.36 6.03 6.70 7.37 8.04 14.74 Tats 1 UO ~

Man fka)2 Equivalent Spherical .54 .68 .78 .86 .93 .99 1.04 1.09- 1.13 1.17 1.21 1.24 1.52 Radius (ft)

Reduction ,

to 60%" Mass Cateanrv 4.5 2.7 A 0.08 0.02 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 5.3 3.2 8 0.14 0.04 0.02 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 6.1 3.7 C 0.21 0.06 0.03 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01-6.9 ' 4.1 D 0.31 0.08 0.04 0.02 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 7.8 4.7 E 0.43 0.12 0.06 0.03 0.02 0.01 0.01 0.01 0.01 0.01 0.01 - 0.01 0.01-8.8 5.3 F 0.59 0.17 0.08 0.04 0.03 0.02 0.02 0.01 0.01 0.01 0.01 0.01 0.01 9.7 5.8 G 0.79 0.23 0.11 0.06 0.04 0.03 0.02 0.02 0.01 0.01 0.01 0.01 0.01-10.7 6.4 H 1.03 0.30 0.14 0.08 0.05 0.04 0.03 0.02. 0.01 0.01 0.01 0.01 0.01 11.7 7.0 I 0.04 0.19 0.11 0.07 0.05 0.04 0.03 0.02 0.02 0.02 0.01 .0.01 12.8 7.7 J 0.05 0.24 0.14 0.09 0.07 0.05 0.04 0.03 0.02 0.02. 0.02 - 0.01-13.8 8.3 K 0.64 0.31 0.18 0.12 0.09 0.06 0.05 0.04 0.03 0.03 0.02 0.01 14.9 8.9 L 0.81 0.39 0.23 0.15 0.11 0.08 0.06 0.05 0.04 0.03 0.03 0.01 16.1- 9.7 M 1.00 0.49 0.29 0.19 0.14 0.10' O.08. 0.06- 0.05 0.04 - 0.04 - 'O.01 17.3- 10.4 N 1.23 0.61 0.37 0.24 0.17- 0.13 0.10 '0.08 0.07 0.05 0.05 0.01-18.5 11.1 0 0.75 0.45 0.30 0.22 0.16 0.13 -0.10 0.08 0.07 - 0.06 - 0.02 -

19.8 11.9 P 0.92 0.56 0.37 0.27 0.20. 0.16 0.13 0.10 0.09: .0.07 0.02-21.1 12.7. Q 1.12 0.68 0.46- 0.33= 0.25 0.19 0.16' O.13 0.11' O.09 0.03 22.5 13.5 R 'O.83. 0.56 0.40 0.31 0.24 0.19 0.16~ 0.13 0.11 0.03-23.9 14.3 S ~1.00 0.68'  : 0.49 - ~0.37 0.29 0.24 0.19 0.16- 0.14 0.04 25.4 15.2 T 1.20 0.82- 0.60 0.46 0.36 0.29- 0.24 - 0.20. 0.17 0.05 27.0 .16.2 U 0199 0.72- 0.55 0.44 -- 0.35 0.29- 0.24 0.21 0.07 17.2 1.19: 0.87 10.67 0.53' O.43 0.35 0.30 0.25 ~ 0.08 28.7 .V-0.64- 0.52- 0.43 '0.26 0.31 0.10-30.3 18.2 -W -1.05 0.81

AA

.0.97 0.77: 0.63' 0.52 0.44 ~0.38 0.12 '

32.1 0.93 0.76 0.63- 0.46, 0.15 :

-33.9 88 1.17 _0.53.

1.12- 0.92 0.76. 0.65 0.56 .0.18 35.8 CC 1.11- _0.92 0.78 0.67. -0.23-

-37.9 - DO ,

1.12 - 0.95- 0.82- 0.28-

~40.0 EE

'1.15 .0.99 0.34s 42.2 FF' 1.21 0.42-44.5 GG -

.0.52 46.9' .HH'

'* Per ANS-8.7-1982 Standard ' reduction to 60% of the mass listed in the first column allows the' fuel .to be distributed arbitrarily within the volume.

50ukCE: ~J. T. Thomas. "The Nuclear Safety Guide: TID-7016, Revision 2." NUREG/CR-0095 (ORNL/NUREG/CSD-6)', U.S. Nuclear Regulatory Commission.' June"

~1978; Tables 4.1 and 4.7. pages 90-91.

FIGURE 1 REACTOR COOLANT SYSTEM COMPONENTS RESSURIZER f li Dp W l{l,],,

l

( '

' 4 j

CORE q; REACTOR

l. (l l FLOOD N '.!l ij1;  ; ! i COOLANT l l t

TAN ." ~'

/ PUMP '

l l ,

[

3

)

g 'i '!!

A

l. ;9 ,'i g

["U. y 6 ::

La i REACTOR COOLANT ll l , t, _

i

%lM,ii; 3'  !-

!' Ti t w I!' DHAIN

[; 2' 1

/ ,

I

\ STEAM GENERATOR "A" LETDOWN LINE REACTOR VESSEL DECAY HEAT DROP LINE STEAM OENERATOR "B"

-4. TECHNICAL SPECIFICATION 1.7 " DEFINITION OF CONTAINMENT INTEGRITY" )

i DESCRIPTION OF CHANGE The sentence " Isolation valves inside the Reactor Building shall be capable of l remote operation from a control station outside of the Reactor Building" has

.been deleted from the definition of Containment Integrity.

REASON FOR CHANGE The quoted sentence was added to Technical Specification 1.7 by the NRC in Amendment of Order' dated October 19, 1984. . The NRC rationale for adding this statement was "The staff has modified a.part of this section to reflect the regulatory relief that was granted GPU by the NRC in an Exemption and Approval of Alternate Design dated July 17, 1984'. The discussion in the staff's letter i states that two manual isolation valves outside-of containment are acceptable in lieu of automatic isolation valves. -It should be note'd that at TMI-2, there are no pressure signals that could automatically initiate valve closure. Instead THI-2 uses remotely operated valves requiring operator action. If a manual valve (instead of a remote / automatic valve) is open, but capable of being closed in all potential accident conditions, then per this definition, containment integrity still exists."

JUSTIFICATION FOR CHANGE Currently, there are manual containment isolation' valves in the RB which are maintained in the "open" position or allowed to be opened in accordance with NRC approved procedures. It might not be possible to close'these valves in all potential accident conditions, e.g., a fire in the containment which requires evacuation of RB personnel. However, in suchLa case, single valve isolation would be maintained utilizing valves outside the RB. This approach complies with the Action Statement of Technical ~ Specification 3.6.1.1,

" Containment Integrity." Thus, deletion of the referenced sentence is solely an administrative change and is not a change' to containment integrity-requirements.

5. TECHNICAL SPECIFICATION 1.21 " DEFINITION OF CONTAINMENT ISOLATION" DESCRIPTION OF CHANGE A definition of Containment ^ Isolation has been added to the Definitions Section.

REASON FOR CHANGE A definition of Containment Isolation 64 bo +jded to support the addition of Technical Specification 3.6.1.2.

JUSTIFICATION FOR CHANGE  !

See the justification for Technical Specifi' cation 3.6.1.2.

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1

6. SECTION 2.'O - SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l 1

DESCRIPTION OF CHANGE.

Section 2.1.3 " Reactor Coolant System Pressure" has been deleted. The state-ment "There are no safety limits which apply.to THI-2" has been added to Section 2.0 " Safety Limits." The bases for the Safety Limits has also been deleted.

REASON FOR CHANGE 1

This change was made to simplify and clarify the Safety Limits section. i Section 2.1.3 " Reactor Coolant System Pressure" was stated as "Not Applicable" -

and no other sections were included in the Safety Limits section. .Therefore, '

the deletion of Section 2.1.3 and the addition of the ' statement "There are no safety limits which apply to TMI-2" clarifles and simpilfles Section 2.0.- The bases have been deleted because there are no Safety Limits.

JUSTIFICATION FOR CHANGE This is an administrative change only and no technical justification'is required.

7. TECHNICAL SPECIFICATION 3.

0.1 DESCRIPTION

OF CHANGE

" RECOVERY" has been changed to " FACILITY".

REASON FOR CHANGE

" RECOVERY" has been changed to " FACILITY" to be consistent with the change to Technical Specification 1.3; " Definition of RECOVERY MODE."

JUSTIFICATION FOR CHANGE This is an administrative change and no technical. justification is required.

8. TECHNICAL SPECIFICATION 3.1.1.1 "B0 RATED COOLING WATER INJECTION" DESCRIPTION OF CHANGE The applicability of specification 3.1.1.1 has been revised from " Recovery Mode" to " Mode 1."

A " Minimum Temperature" requirement has been added to Action Statement "d".

REASON FOR CHANGE The applicability of 3.1.1.1 has been identified as Mode 1 to clarify under what plant status borated cooling water injection capability is required. The

" Minimum Temperature" requirement was added to Action Statement "d" to be consistent with the Minimum Temperature requirement in LCO 3.1.1.1.c.

l-

! 0083P l

JUSTIFICATION FOR CHANGE The capability for borated cooling water injection has been defined as applic-able only during Mode 1. Once the core has been removed from the RV,.the basis for maintaining this capability.no longer exists. The basis for 3.1.1 states, "The limitation on minimum boron concentration ensures that the core will remain subtritical under.all credible conditions which may exist

.during the long-term cooling mode. The maximum boron concentration'is provided to ensure that precipitation of boron will-not occur in the RCS' and thereby cause possible flow restrictions. The specification requires the OPERABILITY of systems capable of injecting borated cooling water into the RCS within the required boron-concentration limits. The-required' volume of borated water in the BHST provides sufficient water to keep-the core covered in the event of an unisolable leak from the reactor vessel. The specified water volume is sufficient to provide a continuous supply.of water to the vessel during the interim period before the recirculation flowpath from the Reactor Building Sump can be placed in service."

Subsequent to removing the core, the residual fuel in the Re' actor Coolant System will be less than an amount which can support a criticality under any conditions. Since there is less than the amount of fuel required to support a-criticality event and the basis for maintaining a borated cooling water injection capability is to ensure the core will remain subcriti_ cal, there.is.

no requirement to maintain the capability for borated cooling water injection.

Deleting the capability for borated cooling water. injection after the core is removed does not degrade any margin of safety since criticality in the Reactor Coolant System is precluded in both cases.

The addition of the Minimum Temperature requirement to the action statement is an administrative change which corrects an oversight from a prior Technical Specification revision.

9. TECHNICAL SPECIFICATION 3.1.1.2 " BORON CONCENTRATION" (REACTOR COOLANT SYSTEM)

DESCRIPTION OF CHANGE The applicability of specification 3.1.1.2 has been revised from " Recovery Mode" to " Mode 1."

REASON FOR CHANGE The applicability of 3.1.1.2 has been identified as Mode 1 to clarify under what plant conditions boron concentration contr'ol is required in the Reactor Coolant System.

JUSTIFICATION FOR CHANGE The control of boron concentration in the Reactor Coolant System has been defined as applicable only during Mode I because once the RV is defueled and the conditions for Mode 2 are satisfied, the basis for maintaining this control no longer exists. The basis for 3.1.1 states, 0083P

g "The' limitation on minimum boron ~ concentration ensures that the core will remain subcritical under all ~ credible conditions which may exist during the'long-term cooling mode. The maximum boron concentration is provided to ensure that precipitation of boron will not occur in the RCS and thereby cause possible flow restrictions. ~The specification requires the OPERABILITY of systems capable of injecting borated cooling water-into the RCS within the required boron concentration limits. _. The required volume of borated water in the BHST provides sufficient water to keep the core covered in the event of an unisolable leak from the reactor vessel. The specified water volume is sufficient to provide.a continuous supply of water to the vessel during the interim period before the recirculation flowpath from the Reactor Building Sump can be placed in service."

~ Boron concentration is controlled to ensure that in the event borated water is

. required to be injected into the Reactor Coolant system, the boron concentra -

tion will be within the limits' established for' assuring subcriticality.

Subsequent to defueling the Reactor Coolant System, the quantity of residual fuel will be less than an amount which can support criticality under any conditions. Since there.ls less than the amount of fuel required to support a criticality event and the basis for control of boron concentration in the Reactor Coolant System is to ensure the core will remain subcritical, there is no requirement to maintain this control of boron concentration.

Deleting the control of boron concentration in the' Reactor Coolant System once it is defueled does not degrade any margin to safety since criticality is precluded in both cases.

10. TECHNICAL SPECIFICATION 3.1.1.3 " BORON CONCENTRATION" (FUEL TRANSFER CANAL)

DESCRIPTION OF CHANGE The applicability of specification 3.1.1.3 has been revised from " Recovery Mode" to " Mode 1." The specification for boron concentration in the Spent Fuel Pool has been removed from 3.1.1.3 and added as specification 3.1.1.4.

Specification 3.1.1.4 has the same requirements for Spent Fuel Pool boron concentration as was in 3.1.1.3.

REASON FOR CHANGE The applicability of 3.1.1.3 has been identified as Mode 1 to clarify the conditions under which control of boron concentration in the Fuel Transfer Canal is required. The specification for boron concentration in the Spent Fuel Pool has been removed from 3.1.1.3 and added as specification 3.1.1.4 to simplify the control of boron concentration requirements during Mode 2.

JUSTIFICATION FOR CHANGE The control of boron concentration in the Fuel Transfer Canal has been defined as appilcable only during Mode I because once the RV has been defueled and all canisters containing core material have been removed from the RB and the Fuel Transfer Canal, the basis for maintaining this control no longer exists. The basis for 3/4.1.1 states, 0083P i

l

"Minicum boron concentration limits have been provided for the_ Refueling Canal (deep end) and Spent Fuel Storage. Pool "A" to provide assurance

.that any event. involving-these volumes.of water will not result'in a margin of_ safety less than that analyzed for the Reactor Vessel."

The margin of safety analyzed for-the RV is based on.having sufficient fuel-in the RV to establish a concern for possible criticality. Deleting the control of. boron concentration in the Fuel Transfer Canal, once the RV is defueled, does not degrade the margin of safety because there can be no event involving the volume of. water in the. fuel transfer canal which could interact with the RV and result in a criticality concern. Criticality is precluded when-the RV is defueled.

The deletion of the boron concentration requirements for the Spent Fuel Pool from 3.1.1.3 and the addition of Specification 3.1.1.4 transfers the Technical Requirements for Spent Fuel Pool boron concentration control from'3.1.1.3 to-3.1.1.4 and therefore, is an' administrative change and no~ technical justification is required.

11. TECHNICAL SPECIFICATION 3.1.1.4 " BORON CONCENTRATION" (SPENT FUEL POOL "A")

DESCRIPTION OF CHANGE Technical Specification 3.1.1.4 has been.added to define the requirements for Spent Fuel Pool boron concentration control which have been deleted from 4 specification 3.1.1.3. The applicability of specification _3.1.1.4 has been defined as Modes 1 and 2 instead of " Recovery Mode" which was in specification j 3.1.1.3.

REASON FOR CHANGE The applicability of 3.1.1.4 has been defined as Modes 1 and 2 to specify the conditions under which control of boron concentration in the Spent Fuel Pool "A" is required. The boron concentration requirements for the Spent Fuel Pool have been established in a separate specification to simplify boron concentration control requirements in two separate volumes.

JUSTIFICATION FOR CHANGE The-basis for 3/4.1.1 states,

" Minimum boron concentration limits have been provided for the Refueling Canal (deep end) and Spent Fuel Storage Pool "A" to provide assurance that any event involving these volumes of water will not result in a margin of safety less than that analyzed for the Reactor Vessel."

The Canister Handling and Preparation for Shipment Safety Evaluation Report (SER) and the Defueling SER take credit for the maintenance of the specified boron concentration requirement in the refueling canal (deep end) and the Spent Fuel Storage Pool 'A' to assure that accidental rupturing of a canister in.these pools wlli not result in a criticality potential.

The control of boron concentration in Spent Fuel Pool "A" has been defined-as applicable only during Modes 1 and 2 because once the RV has been defueled and all canisters containing core material'have been removed from Spent Fuel Pool "A", the basis for maintaining this control no longer exists.

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i

. Transferring the boron-concentration requirements for the Spent Fuel Pool to a .

new specification 1s-an. administrative change and no technical justification l 1s required.

12. TECHNICAL SPECIFICATION 3.3.1.1 " INTERMEDIATE AND SOURCE RANGE NEUTRON FLUX MONITORS.

DESCRIPTION OF CHANGE The applicability of Specification 3.3.1.1-has been revised from " Recovery.

Mode" to " Mode 1." Additionally, action statements "a", "b",_and "c" have been' revised to delete the Special Report Requirement.

REASON FOR CHANGE The applicability of 3.3.1.1 has been identified as Mode 1 to clarify the conditions under which intermediate and source range monitors are required.

'The deletion of the Special Report requirements from the action statement is consistent with the requirements of 10 CFR 50.73, " Licensee Event Report System."

JUSTIFICATION FOR CHANGE The intermediate and source range neutron flux monitors have been defined as applicable only during Mode 1 because once the RV has been defueled, the basis for maintaining these monitors no longer exists. The basis for 3.3.1 states, "The neutron monitoring instrumentation, which was included in the normal Reactor Protection System Instrumentation, provides information regarding the shutdown status of the core and.it will be used to monitor changes in neutron generation."

Once the RV has been defueled, the shutdown status of the core is assured under-all conditions. Therefore, the neutron monitoring instrumentation can be deleted from the Technical Specifications during modes 2 and 3. In support of this proposed change, GPU Nuclear requests an exemption, pursuant to 10 CFR 70.24(d) from the requirements of 10 CFR 70.24, " Criticality Accident Requirements for the TMI-2 Reactor Building" following completion of Mode 1.

As shown in Table 1 to item 3 of this safety evaluation, the. amount of the residual fuel in the RB, after Mode 1, is expected to exceed the minimum quantitles of special nuclear material for which criticality monitors are required by 10 CFR 70.24. However, pursuant to proposed Technical Specifica-tion Table 1.1, " Facility Modes," one of the criteria for entering Mode 2 is that the possibility of criticality in the RB is. precluded.- Thus, following Mode 1, the requirements of 10 CFR 70.24 should no. longer apply to the THI-2 RB.

. The deletion of'the Special Report Requirements from the action statements is basically administrative in nature. 10 CFR 50.73 requires licensees to submit a Licensee Event Report (LER) when a Technical Specification Action Statement

has not been. satisfied. 'Thus, the requirement to submit a Special Report-is not warranted in this case. Furthermore,.the LER would contain the same information as required by the Special Report.

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13. TECHNICAL-SPECIFICATIONS'3.3.2 and 3.3.2.1 " ENGINEERED SAFETY FEATURE

' ACTUATION SYSTEM INSTRUMENTATION" DESCRIPTION OF CHANGE Technical Specifications 3.3.2 and 3.3.2.1 have been deleted.

REASON AND JUSTIFICATION FOR CHANGE The basis for Technical Specification 3.3.2 states, "Except for automatic starting of the. diesel generators on loss of-off-site power, all automatic features of the ESFAS instrumentation have; been defeated. This action prevents inadvertent actuation of the ESF-systems. 'The diesel. generators will start automatically on loss of off-site power."

THI-2 TSCR 51 proposed deletion of the TMI-2 emergency diesel generators from the Technical Specifications based on an evaluation which demonstrated that all loads potentially served by the emergency diesel generators either can be interrupted safely for a length of time conservatively assumed necessary to restore off-site power or can be supplied with back-up power from the existing station batteries. Therefore, subsequent.to theudeletion of the Technical Specification requirements for the emergency diesel generators, there is no longer a basis for maintaining the ESFAS Instrumentation operable.

14. TECHNICAL SPECIFICATION 3.3.3.1 " RADIATION MONITORING INSTRUMENTATION" DESCRIPTION OF CHANGE The applicability of Specification 3.3.3.1'has been revised from "as required in Table 4.3-3 of the Recovery Operations Plan" to " Modes 1, 2 and 3 as required in Table 4.3-3 of the Recovery Operations Plan."

REASON FOR CHANGE Modes 1, 2 and 3 have been added to the applicability of 3.3.3.1 to be consis-tent with the identification of modes in Table 4.3-3.

JUSTIFICATION FOR CHANGE The addition of modes 1, 2 and 3 to the appilcability statement is an admini-strative change since there has been no change in the actual applicability of specification 3.3.3.1, therefore no technical justification is required.

15. TECHNICAL SPECIFICATION 3.3.3.4 " METEOROLOGICAL INSTRUMENTATION" DESCRIPTION OF CHANGE The applicability of Specification 3.3.3.4 has been revised from " Recovery.

Mode" to " Modes 1 and 2," and the eight (8) hour time clock in the action statement has been revised to seven (7) days.

0083P

, , , . ;n -. .

n. - - .. .- - - . - -

LREASON FOR CHANGE-

[' iThe:appllbabilityofmeteorological;monitoringrequirementshasbeenlimited.

F :to Modes.lLand 2 because once the last-THI-2 core material has been shipped, l -

-there 1s no longer a need-for this requirement in the Technical'Specifica -

tions. The eight (8) hour. requirement has been changed to seven (7);dayssto' be consistent with the standard B&W Technical Specifications for.this'~

requirementi i: JUSTIFICATION OF CHANGE l The basis for having meteorological Instrumentation states, "The OPERABILITY of the meteorological Instrumentation ensures-that p -sufficient' meteorological data is available for estimating potential.

L radiation doses -to the public as a result of routine or accidental:

Trelease: of radioactive materials to the atmosphere. This: capability is-required to evaluate the_need for: initiating protective measuressto

~

i i- protect the health and safety of the public. -This instrumentation-is' L consistent with the recommendations:of Regulatory-Guide 1.23c 'On-site.

F -Meteorological-Programs,' February 11972."

i . .

Subsequent to shipment off-site of the last core material, there will be:

! various forms of contamination remaining inside :the containment._ 'A spectrum.

I

~

of events:have been analyzed.which involve.the contamination remaining in'the~

containment (reference NUREG/CR-2601, " Technology, Safety and Costs"of Decom -

t missioning_ Reference Light Water Reactors Following Postulated Accidents").

The events analyzed include inappropriately controlled:decontaminetton'activi-ties and the accidental cutting of contaminated piping systems.-.Addl.tlonally, <

j the potential off-site consequences of the_ worst ~ case event during Mode 3, a.

fire in the RB, are bounded by the limits of 10 CFRx50' Appendix-I (see the

, Safety Analysis for Technical Specification 6.8.2.2,1 Item.Noi 42). -Since the-l basis for requiring meteorological instrumentatt.on is "to' evaluate;the:need

for initiating protective measures to protect the health and safety of the-

public" and since the worst case release is bounded by-Appendix I, which -

specifies limits recognized as not being a~ threat to the pubile heal,th and -

i safety, there is no requirement to maintain meteorological instrumentation in a

the Technical Specifications during Mode ~3.

! The proposed seven (7) day timeclock is consistent with'the' requirements of-

! the B&W Standard Technical Specifications and the TMI-2 pre-accident Technical. '

i Specifications, the bases of which state: "The OPERABILITY of the meteoro-

! logical instrumentation ensures that sufficient meteorological data is avall-I able for estimating potential radiation doses to the atmosphere. This capa-i bility is required to evaluate the need for initiating protective measures to ,

i- protect the health and safety of the public (and is consistent'with the

recommendations of Regulatory Guide 1.23, "On-site Meteorological Programs," >

February 1972)." >

The requirement for the eight (8) hour timeclock was incorporated in Technical

- . Specification 3.3.3.4 vla NRC Amendment of Order dated

February 11, 1980. At.

q i- the time of this Order, the RB contained high concentrations of radioactive Krypton-85 and in the event a leak had developed from the RB, it would have-

~

been important to have operable meteorological. Instrumentation to' assess the i consequences of the release. Subsequently, the RB was~ purged, thus the l magnitude of airborne radionuclides in the RB which could be released to the 1

[ environment has been substantially reduced. Therefore, the original:need for 0083P

. - . _ . = - . - . - ~ . - - . - . = , = . . - . w. ......-.-, a -. - ;.,

., - ;_i San eighti(8)Jhour timeclockino' longer extsts and realigning:the~timeclock to be consistent"with;that used in-B&W Standard 1 Technical Specifications isi Jappropriate..

~

[

4. E16; -TECHNICAL-SPECIFICATION 3.3.3.5 " ESSENTIAL-PARAMETERS MONITORING INSTRUMENTATION"-

DESCRIPTION ~0F' CHANGE-cThelapplicab111ty of Technical 4 Specification.3.3.3.'5 has-been revised from:

_ " Recovery Mode" to " Mode 1."~

< REASON FOR CHANGE-iThe: applicability of 3.3.3.5 has been identified as Mode-1.to clarify the ,

plant-status which requires the essential: parameters monitoring ,

-instrumentationLto be operable.

JUSTIFICATION FOR CHANGE ,

.The bases for the' Essential, Parameters Monitoring. Instrumentation states: '

'l "The OPERABILITY of the. Essential: Parameters Moriitoring Instrumentation 1 ensures'that sufficient information is'available on selected. plant ~

-parameters to monitor and assess these< variables."

~

The parameters listed as essential are.'(1) Incore Thermocouples,-(2) RB Nater Level, (3) Borated Water Storage Tank Level and-(4)-Steam Generator Level. '

These parameters are monitored to identify changes- which would : indicate'an -

event requiring action to protect the. health and safety of the.publict Once the RV is defueled, there are no events which could occur for.which these.-

variables would be meaningful Indications of ongoing events or require the-functional capability of the systems monitored.. Therefore,'once the RV'isi defueled, it will no longer be necessary to monitor these parameters. The requirements for RB Pressure, RV Hater Level, Spent ~ Fuel Storage Pool'"A" Hater Level and Fuel Transfer Canal (deep end). Hater Level have been, transferred to separate Technical Specifications. '

17. TECHNICAL SPECIFICATION 3.3.3.7 " CHLORINE DETECTIO'N SYSTEMS" .

DESCRIPTION OF CHANGE The applicability of Specification 3.3.3.7 has been revised from " Recovery' l Mode" to " Mode 1."

REASON FOR CHANGE The applicability of 3.3.3.7 has been identified only as Mode'.1 to specify the plant condition under which chlorine detection capability'isirequired.

JUSTIFICATION FOR CHANGE The bases for the Chlorine Detection Systems states:

"The OPERABILITY of the Chlorine Detection. Systems ensures that an -

accidental chlorine release will be~ detected promptly and the Control. H Room Emergency Ventilation System will automatically isolate the. Control  !

l'; Room and initiate its operation in the' recirculation mode.to provide the'  ;

required protection."

~

-j j' 0083P R I .

r .

Th3 operability of the Chlorine Detection System is required to automatically isolate'the Control Room ventilation system and_ initiate the operation of'the recirculation mode to protect Control Room operators against an accidental-chlorine release. ~0nce the RV is defueled, Technical Specification 6.2.2 no.-

. longer requires licensed operators in the Control Room. Therefore,' concurrent with the change _in plant status, the requirement to maintain operators'in the Control Room is obviated. It follows that-the Chlorine Detection System will' no longer be necessary.

18. TECHNICAL SPECIFICATION 3.3.3.8 " FIRE DETECTION INSTRUMENTATION" DESCRIPTION OF CHANGE The applicability of specification 3.3.3.8 has been revised from " Recovery Mode" to " Modes 1, 2 and 3."

REASON FOR CHANGE The applicability of 3.3.3.8 has been identified as Modes I, 2 and 3 to be consistent with the practice throughout the Technical Specifications of.

utilizing. Facility Mode (s) to identify applicability.

JUSTIFICATION FOR CHANGE Since the period of time. covered by Modes 1, 2 and 3 is exactly the same as that covered by " Recovery Mode," this is an administrative change and no technical justification is required. Fire detection capability will continue to be provided in accordance with the Technical Specifications throughout the Cleanup Period.

19. TECHNICAL SPECIFICATION 3.4.2 " REACTOR VESSEL WATER LEVEL MONITORING" DESCRIPTION OF CHANGE The applicability of Specification 3.4.2 has been revised from " Recovery Mode with the RV Head Remc,ved" to Mode 1." Various editorial changes have also been made to Specification 3.4.2.

REASON FOR CHANGE The applicability of 3.4.2 has been identified as Mode 1 to clarify the conditions under which RV water level monitoring capability is required. The editorial changes were made to clarify the specification and action statements.

JUSTIFICATION FOR CHANGE The bases for RV Water Level Monitoring states:

"The Reactor Vessel Water Level Monitor ensures that indication is available to monitor for changes in Reactor Vessel Water Level. This

. device will provide warning of a leak from the Reactor Coolant System or unexplained increases in Reactor Coolant System inventory which could result in a boron dilution event. . Two independent monitors are required to provide redundancy and to minimize the necessity to discontinue processing because of instrument failures."

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n RV rater level monitoring capability is required during Mode 1 to provide-indication of a leak in the Reactor Coolant Systes or an unexplained increase in the Reactor Coolant System inventory.

Both of these capabilities are. required to provide warning of conditions which might adversely affect the Reactor Core, such as a draindown event or a boron dilution event. Once the RV is defueled it is no longer necessary to maintain water in the RV; consequently, the capability to monitor the water level is no

-longer required.

The editorial changes do not affect the technical requirements of

~

Specification 3.4.2, therefore, no technical justification is required.

-20. TECHNICAL SPECIFICATIONS 3.4.9.1 AND 3.4.9.2 " REACTOR COOLANT SYSTEM" DESCRIPTION OF CHANGE The' applicability statements of specifications 3.4.9.1 and 3.4.9.2 have been revised from " Recovery Mode" to " Mode.l."

REASON FOR CHANGE The applicability statements of 3.4.9.1 and 3.4.9.2 have been revised to clarify the conditions under which control of the Reactor Coolant System pressure and temperature is necessary.

JUSTIFICATION FOR CHANGE Control of the temperature of the Reactor Coolant System is required during Mode I to avoid unacceptable temperature transients which could lead to boron precipitation. Also, monitoring the temperature of the Reactor Coolant System is required to provide indication of possible criticality activity in the core debris. Once the RV is defueled, it is no longer necessary to maintain water in the Reactor Coolant System; consequently, control of Reactor Coolant System temperature is no longer required.

The RV is currently at atmosphere pressure with the head and plenum removed.

Repressurization during Mode 1 requires NRC approval to assure all concerns during a pressurized condition with fuel in the RV have been reviewed. Once the RV has been defueled, there is no longer a' basis for NRC review and approval.

21. TECHNICAL SPECIFICATION 3.5.1 " CONTROL ROOM" DESCRIPTION OF CHANGE The applicability of specification 3.5.1 has been revised from "During Core Alterations" to " Mode 1 during Core Alterations."

REASON FOR CHANGE The applicability of 3.5.1 has been revised to identify the Facility Mode during which 3.5.1 applies and to be consistent with the practice throughout the Technical Specifications of identifying the applicable mode (s) for each specification.

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JUSTIFICATION FOR CHANGE Since this specification' applies only during " Core Alterations" which occur only during Mode 1, this change is only an administrative change and no technical justification is required.

22. TECHNICAL SPECIFICATION 3.6.1.1 " CONTAINMENT INTEGRITY" AND 3.6.1.3

" CONTAINMENT AIRLOCKS" DESCRIPTION OF CHANGE The appIlcability of Specifications 3.6.1.1 and 3.6.1.3 have been revised from

" Recovery Mode" to " Mode 1."

REASON FOR CHANGE The appilcability of 3.6.1.1 and 3.6.1.3 have been revised to clarlfy which plant conditions require containment integrity as defined by 3.6.1.1 and 3.1.6.3. This change also recognizes that containment isolation provisions have been added for Modes 2 and 3 in Specification 3.6.1.2 and 3.6.1.6.

JUSTIFICATION FOR CHANGE Containment Integrity is generally understood as requiring double isolation of all containment penetrations and the containment airlocks. Double isolation is necessary to protect against the single failure of an active component to seal a penetration, thereby allowing an uncontrolled release of radioactive materials to the environment. Once the RV has been defueled, the potential for the creation of a condition which could result in an uncontrolled release of significant quantitles of radioactive materials to the environment has been essentially eliminated and the containment's primary function is that of a barrier to the release of residual contamination which remains inside the containment. Specifications 3.6.1.2 and 3.6.1.6 provide for containment isolation during Modes 2 and 3. The single isolation of containment penetrations and airlocks provided for in 3.6.1.2 and 3.6.1.6 is adequate to provide the passive contamination barrier function required for Modes 2 and 3 as opposed to the pressure boundary and active single failure protection provided by the double isolation requirements in 3.6.1.1 and 3.6.1.3.

Therefore, the combination of Specifications 3.6.1.1, 3.6.1.2, 3.6.1.3, and 3.6.1.6 provide adequate protection of the public for all Facility Modes.

23. TECHNICAL SPECIFICATION 3.6.1.2 " CONTAINMENT ISOLATION" AND 3.6.1.6

" CONTAINMENT AIRLOCKS" DESCRIPTION OF CHANGE Specifications 3.6.1.2 and 3.6.1.6 and the associated surveillance requirements have been added to the Technical Specifications.

REASON FOR CHANGE Containment isolation requirements have been added for Modes 2 and 3 to provide appropriate provisions for maintaining the containment as a contamination barrier during these two facility modes.

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' JUSTIFICATION FOR CHANGE Specifications 3.6.1.1 and 3.6.1.3 have been revised to make the double isolation requirements of Specifications 3.6.1.1 and 3.6.1.3 applicable only

'during Mode 1. ' Specifications.3.6.1.2 and 3.6.1.6 were added to define appropriate containment isolation provisions for Modes 2 and 3. The single isolation provisions provided in 3.6.1.2 and 3.6.1.6 are adequate for Modes 2.

and 3 because of the change in the functional requirement of the containment once the RV is_defueled. During Mode 1, the containment functions as a pressure boundary and a contamination barrier. However, once the RV is defueled, the potential _ for a significant. pressurization event and component failure is eliminated and only the contamination barrier function is required to be maintained. Therefore, the isolation provisions and associated survell-lance intervals in 3.6.1.2, 4.6.1.2, 3.6.1.6, and 4.6.1.6 (Reference R0PCR 38) are adequate to assure this function is maintained.

'24. TECHNICAL SPECIFICATION 3.6.1.4 " INTERNAL PRESSURE" DESCRIPTION OF CHANGE The applicability of Specification 3.6.1.4 has been revised from " Recovery Mode" to " Modes 1, 2, and 3."

REASON FOR CHANGE The applicability of 3.6.1.4 has been changed from " Recovery Mode" to Modes 1, -;

2, and 3" to be consistent with the remainder of the Technical Specifications in the use of Facility Modes to identify applicability.

JUSTIFICATION FOR CHANGE Since the period of time represented by " Recovery Mode" is the same as that represented by " Modes 1, 2, and 3," the changing of the applicability to be consistent with the other Technical Specifications is an administrative change ,

and does not require technical justification. Internal pressure in the Containment will continue to be monitored and controlled in accordance with the Technical Specifications throughout the Cleanup Period.

25. TECHNICAL SPECIFICATION 3.6.1.5 " AIR TEMPERATURE" DESCRIPTION OF CHANGE The applicability of Specification 3.6.1.5 has been revised from " Recovery Mode" to " Mode 1."

REASON FOR CHANGE The appIlcability of 3.6.1.5 has been revised to clarify the plant condition under which the control of the Containment air temperature is required.

JUSTIFICATION FOR CHANGE The bases for Technical Specification 3.6.1.5 states:

"The average air temperature of the Containment atmosphere is currently being maintained between 50'F and 130*F. This condition will maximize 0083P l

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~ the service-life cf the instrumentation and equipment-installed in the

~

Containment and ensure that Reactor Coolant Systea temperature does not drop below 50*F via LOSS-TO-AMBIENT. This temperature will ensure that boron will remain in solution. Continued OPERABILITY of these items is required to continue monitoring and mitigating the March 28, 1979 incident."

The Containment average air temperature is required during Mode I to be-maintained between 50*F and 130*F to protect instrumentation inside the-Containment and to be consistent with maintaining the-Reactor Coolant System

. temperature above 50*F to prevent boron precipitation. Subsequent to RV defueling and draining of the' Reactor Coolant System.. boron precipitation will no longer be a concern and the 50*F limit will no longer be necessary. Also, once the RV is defueled, it will no longer be necessary to " continue mont-toring... the March 28,'1979 incident" and any instrumentation which is to be maintained inside the Containment will not require a Technical Specification on maximum Containment air temperature to establish acceptable environmental conditions. Although the Containment air temperature may be maintained in approximately the same range during Modes 2 and 3, it will primarily be for-the comfort of personnel, and such control is more properly left to station procedures.

26. TECHNICAL SPECIFICATION 3.6.3.1 " CONTAINMENT PURGE EXHAUST SYSTEM" DESCRIPTION OF CHANGE The applicability of Specification 3.6.3.1 has been revised from " Recovery Mode" to " Modes 1,2 and 3." The seven (7) day timeclock in the action statement has been changed to " prior to resuming purge operations."

REASON FOR CHANGE The applicability of 3.6.3.1 has been changed from "During Purge Operations" to '? Modes 1, 2, and 3" to be consistent with the remainder of the Technical Spectf1 cations in the use of Facility Modes to identify applicability. The seven (7) day timeclock is an unnecessary burden since activities requiring operation of the Containment Purge Exhaust System cannot be performed until at least one (1) train is restored to operable status.

JUSTIFICATION OF CHANGE Since the period of time represented by " Recovery Mode" is the same as that represented by " Modes I, 2, and 3," the changing of the applicability to be consistent with the other Technical Specifications is an administrative change and does not require technical justification. Containment Purge Exhaust System will continue to be maintained in accordance with the Technical Spect-fications throughout the Cleanup Period. The change to the action statement timeclock is also administrative in nature since those activities requiring operation of the Containment Purge Exhaust System cannot be performed until at least one (1) train is restored to operable status.

27. TECHNICAL SPECIFICATION 3.7.6.1 " FLOOD PROTECTION" DESCRIPTION OF CHANGE The applicability of Specification 3.7.6.1 has been revised from "at all times" to " Modes 1, 2 and 3."

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REASON FOR CHANGE The applicability of 3.7.6.1 has been revised to be consistent with the identification of modes in the applicability statement.

JUSTIFICATION FOR CHANGE Since " Modes 1, 2 and 3" and "at all times" are the same time period, revising 3.7.6.1 applicability from "at all times" to " Modes 1, 2 and 3" is in admini-strative change only and no technical justification is required. Flood protection will continue to be maintained in accordance with the Technical Specifications "at all times."

28. TECHNICAL SPECIFICATION 3.7.7.1 " CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM" DESCRIPTION OF CHANGE The applicability of Specification 3.7.7.1 has been revised from " Recovery Mode" to " Mode 1."

REASON FOR CHANGE The applicability of 3.7.7.1 has been revised to clarify the plant conditions which require the operability of the Control Room Emergency Air Cleanup System.

JUSTIFICATION FOR CHANGE l

The Control Room Emergency Air Cleanup System is required to be maintained operable to protect the Control Room operators in the event of a unit accident and to maintain Control Room habitability in the event of chemical release events. In accordance with Technical Specification 6.2.2, the Control Room is required to be manned while there is fuel in the RV. Once the RV is defueled, there will be no requirement to man the Control Room. Therefore, there-is no need to provide the protection for the Control Room operators as described in 3.7.7.1.

29. TECHNICAL SPECIFICATION 3.7.9 " SEALED SOURCES" i

DESCRIPTION OF CHANGE The applicability of Specification 3.7.9 has been revised from " Recovery Mode" to " Modes 1, 2 and 3."

REASON FOR CHANGE The applicability of 3.7.9 has been changed from " Recovery Mode" to " Modes 1, 2 and 3" to be consistent with the remainder of the Technical Specifications in the use of Facility Modes to identify applicability, l JUSTIFICATION FOR CHANGE Since the period of time represented by " Recovery Mode" is the same as that represented by " Modes 1, 2 and 3," the changing of the applicability to be consistent with the other Technical Specifications is an administrative change and does not require technical justification. Sealed sources will continue to be controlled and tested in accordance with the Technical Specifications during the Cleanup Period.

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30. TECHNICAL-SPECIFICATION 3.7.10.1 " FIRE SUPPRESSION WATER SYSTEM" DESCRIPTION OF CHANGE The appilcability of specification 3.7.10.1 has been revised from " Recovery Mode" to " Modes I, 2, and 3."

REASON FOR CHANGE The applicability of specification 3.7.10.1 has been changed from " Recovery Mode" to " Modes 1, 2 and 3" to be consistent with the remainder of the Tech-nical Specifications in the use of Facility Mode (s) to identify applicability.

JUSTIFICATION FOR CHANGE Since the period of time represented by " Recovery Mode" is the same as th'at represented by " Modes 1, 2 and 3," the changing of the applicability to be consistent with the other Technical Specifications is an administrative change and does not require technical justification. The Fire Suppression Water System will continue to be maintained in accordance with Technical Specifications during the Cleanup Period.

31. TECHNICAL SPECIFICATION 3.7.10.2 " DELUGE / SPRINKLER SYSTEMS" DESCRIPTION OF CHANGE The applicability of Specification 3.7.10.2 has been revised from " Recovery Mode" to " Modes 1, 2 and 3." Additionally, an asterisk (*) has been added to Specifications d, e, f, h, 1, and j to indicate that their respective systems are required to be operable only when charcoal filters are installed.

REASON FOR CHANGE The applicability of 3.7.10.2 has been changed from " Recovery Mode" to " Modes 1, 2 and 3" to be consistent with the remainder of the Technical Specification in the use of Facility Mode (s) to identify applicability.

JUSTIFICATION FOR CHANGE Since the period of time represented by " Recovery Mode" is the same as that represented by " Modes 1, 2 and 3," the changing of the applicability to be consistent with the remainder of the Technical Specification is an admini-strative change and does not require technical justification. The Deluge and Sprinkler Systems will continue to be maintained in accordance with the Technical specifications during the Cleanup Period. However,'an Asterisk (*)

has been added to those locations where charcoal filters may be installed to indicate that their respective deluge and/or sprinkler system is required to be operable only when charcoal filters are installed since the potential for a fire hazard does not exist when charcoal is not installed at these locations.

32. TECHNICAL SPECIFICATION 3.7.10.3 "HALON SYSTEM" DESCRIPTION OF CHANGE The applicability of Specification 3.7.10.3 has been revised from " Recovery Mode" to " Mode 1."

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REASON FOR CHANGE

~

The applicability of 3.7.10.3 has been revised to clarify the plant conditions.

which require the operability of the Halon System.

' JUSTIFICATION FOR CHANGE The purpose of the Halon System is to protect circuits and equipment in specific areas of the plant from the propagation of a fire'such that functions required for safe shutdown-and core protection are not disabled. Once the core material has been removed from the RV there will be no applicable circuits and equipment because there will be no core to protect or shut down.

With the RV defueled, the functions provided by the'Halon System are no longer required, therefore, they can be deleted during Modes 2 and 3.

33. TECHNICAL SPECIFICATION 3.7.10.4 " FIRE HOSE STATIONS" DESCRIPTION OF CHANGE The applicability of Specification 3.7.10.4 has been revised from 'l Recovery' Mode" to " Modes 1, 2, and 3."

REASON FOR CHANGE The appitcability of 3.7.10.4 has been changed to be consistent with,the practice throughout the Technical Specifications of identifying the applicable mode (s) for each specification.

JUSTIFICATION FOR CHANGE Since the period of time represented by " Recovery Mode" is the same as that' represented by " Modes 1, 2 and 3," the changing of the appilcability to be consistent with the remainder of the Technical Specifications is an admint-strative change and does not require technical justification. The Fire Hose Stations will continue to be maintained in accordance with the Technical Specifications during the Cleanup Period.

34. TECHNICAL SPECIFICATION 3.7.11 " PENETRATION FIRE BARRIERS" DESCRIPTION OF CHANGE The applicability of Specification 3.7.11 has been revised from " Recovery Mode" to " Modes 1, 2 and 3."

REASON FOR CHANGE The applicability of 3.7.11 has been changed to be consistent with.-the practice throughout the Technical Specifications of identifying the applicable mode (s) for each specification.

JUSTIFICATION FOR CHANGE Since the period of time represented by " Recovery Mode" is the same as that represented by " Modes 1, 2 and 3," the changing of the applicability to be consistent with the remainder of the Technical Specifications is an admint-strative change and does not require technical justification. The Penetration 0083P

Fire Barriers will continue to be maintained in accordance with the Technical Specifications during the Cleanup Period.

35. TECHNICAL SPECIFICATIONS 3.8.1.1 "A.C. SOURCES." 3.8.2.1.1 AND 3.8.2.1.2 "A.C.

DISTRIBUTION" AND 3.8.2.2.1 "D.C. DISTRIBUTION" DESCRIPTION OF CHANGE The applicability of Specifications 3.8.1.1, 3.8.2.1.1, 3.8.2.1.2 and 3.8.2.2.1 have been revised from " Recovery Mode" to " Mode 1."

REASON FOR CHANGE The applicability of Specifications 3.8.1.1, 3.8.2.1.1, 3.8.2.1.2 and 3.8.2.2.1 have been revised to clarify the plant conditions which require the operability of A.C. Power and D.C. Power.

JUSTIFICATION FOR CHANGE The basis of the above referenced specifications is "The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during opera-tion ensure that sufficient power will be available to supply the safety related equipment required to maintain the unit in stable conditions during recovery from the March 28, 1979, accident." Following Mode 1, no safety related equipment will be required to maintain the unit in a safe and stable condition as the RV will be defueled. Thus, the necessity to maintain Technical Specifications which require the operability of Electric Power Systems will have been eliminated.

36. TECHNICAL SPECIFICATIONS 3.9.1 " SPENT FUEL POOL 'A' WATER LEVEL MONITORING" and 3.9.2 " SPENT FUEL POOL 'A' HATER LEVEL" DESCRIPTION OF CHANGE

" Modes 1 and 2" have been added to the applicability statements of Specifications 3.9.1 and 3.9.2.

REASON FOR CHANGE

" Modes 1 and 2" have been added to the applicability statements of 3.9.1 and 3.9.2 to be consistent with the practice throughout the Technical Specifications of identifying the applicable mode (s) for each specification.

JUSTIFICATION FOR CHANGE Since these specifications apply only when canisters containing core material are in Spent Fuel Storage Pool "A" and Modes 1 and 2 are the only Facility Modes during which canisters containing core material will be on-site, this change is only an administrative change and no technical justification is required.

37. TECHNICAL SPECIFICATIONS 3.9.3 " FUEL TRANSFER CANAL (DEEP END) HATER LEVEL MONITORING" AND 3.9.4 " FUEL TRANSFER CANAL (DEEP END) HATER LEVEL" DESCRIPTION OF CHANGE The applicability statements for Specifications 3.9.3 and 3.9.4 have been revised to read " Mode 1."

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REASON FOR CHANGE The applicability statements for 3.9.3 and 3.9.4 have.been revised to clarify

, the plant conditions which require water level control in-the Fuel Transfer-Canal and to be consistent with the practice throughout the Technical Specifications of identifying the applicable mode (s) for each specification.

JUSTIFICATION FOR CHANGE The bases for Fuel Transfer Canal Water level states:

"The water level in the Fuel Transfer Canal (deep end) has been esta-blished to limit the dose rate, due to the storage of the plenum assembly and Canisters, to acceptable levels."

The basis for water level control in.the Fuel Transfer Canal is radiation protection for personnel when there are fuel canisters and/or the plenum assembly stored in the Fuel Transfer Canal. During Modes 2 and 3 there will be no canisters stored in the Fuel Transfer Canal. Therefore, water level control will not be required for purposes of personnel radiation protection from canisters. During Modes 2 and 3 the plenum may be stored in the Fuel Transfer Canal; however, there will be a significant reduction of activity in building. Therefore, with fewer people in the RB requiring radiation protection and no canisters stored in the Fuel Transfer Canal, radiation.

protection is more properly addressed through plant procedures and 10 CFR 20 compliance rather than the Technical Specifications.

38. TECHNICAL SPECIFICATION 3.9.12.1 AND 3.9.12.2 " FUEL HANDLING BUILDING /

AUXILIARY BUILDING AIR CLEANUP EXHAUST SYSTEM" DESCRIPTION OF CHANGE The applicability of Specifications 3.9.12.1 and 3.9.12.2 has been revised from " Recovery Mode" to " Modes 1, 2 and 3."

REASON FOR CHANGE The applicability of 3.9.12.1 and 3.9.12.2 has been revised to be consistent with the practice throughout the Technical Specifications of identifying the applicable mode (s) for each specification.

JUSTIFICATION FOR CHANGE The changing of the applicability to be consistent with the remainder of the Technical Specifications is an administrative change and does not require technical justification. The FHB/ Aux 111ary Building Air Cleanup Exhaust Systems will continue to be maintained in accordance with the Technical Specifications during the Cleanup Period.

39. TECHNICAL SPECIFICATION 3.9.13 " ACCIDENT GENERATED WATER DESCRIPTION OF CHANGE The applicability of Specification 3.9.13 has been revised from " Recovery Mode" to Modes 1, 2, and 3,

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REASON FOR CHANGE-The applicability of.3.9.13 has been revised'to be consistent with the practice throughout the Technical Specifications of identifying the applicable mode (s) for each specification.

JUSTIFICATION FOR CHANGE The changing of the appilcability, to be consistent with the remainder of the Technical Specifications, is an administrative change and does not require technical justification. Thus, this. specification will require NRC approval of_ procedures for discharge of ACCIDENT GENERATED HATER throughout-the Cleanup-Period.

40. TECHNICAL SPECIFICATION 3.10.1 " CRANE OPERATIONS CONTAINMENT BUILDING" DESCRIPTION OF CHANGE j The applicability of Specification 3.10.1 has been revised from " Recovery Mode" to " Mode 1."

REASON FOR CHANGE The applicability of 3.10.1 has been revised to clarify the plant condition which requires the control of Heavy Loads in the Containment Building.

JUSTIFICATION FOR CHANGE The basis for this specification states, "A load drop into the Reactor Vessel may cause reconfigurations of the core debris and/or structural damage which could hinder recovery efforts. A load drop in the Incore Instrument Seal Table and/or guide tubes may result in an unisolable leak from the Reactor Vessel. The restriction on movement of loads in excess of the nominal weight of a fuel or control rod assembly and associated handling tool over these areas is to mitigate the potential consequences stated above in the event this load is dropped."

Following Mode 1, the RV will be in a defueled, subtritical condition; there-fore the bases for controlling Heavy Loads inside the containment will have been eliminated.

41. TECHNICAL SPECIFICATION 3.10.2 " CRANE OPERATIONS FUEL HANDLING BUILDING" OESCRIPTION OF CHANGE The appilcability of Specification 3.10.2 has been revised from " Recovery Mode" to " Modes 1 and 2."

REASON FOR CHANGE The applicability of 3.10.2 has been revised to clarify the plant conditions which require the control of Heavy Loads in the FHB.

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., +. .

-= -- . , - .- - . - . ,. . -.- ---

JUSTIFICATION FOR CHANGE m, '

Heavy Loads-are controlled inside the FH8 to reduce the probability of a. load drop causing damage:to canisters containing core material. Subsequent-to, . '

- Modes'.1 and 2 all core debris will have-been shipped off-site,-thus,-the: basis?

for controlling Heavy-Loads inside the FHB is. eliminated.

42. TECHNICAL SPECIFICATION 6.2.2 "TMI-2 ORGANIZATION"-

"This requirement applies during Mode l'only" has been added to paragraphs '

'6.2.2.a 6.2.2.b., and 6.2.2.c. Additionally, paragraph 6.2.2.d has been ._.

. revised to clarify that the restriction on the minimum shift crew only-applies. l during Mode 1..

~ REASONS FOR CHANGE s

"This requirement applies during' Mode.1 only" has been added to paragraphs l- ~ 6.2.2.a,.6.2.2.b, and.6.2.2.c to clarify the applicability of the respective. .

requirements. The restriction on the minimum shift crew in paragraph 6.2.2fd'

! has been revised to be consistent with-the revision to paragraph 6.2.2.a.

j JUSTIFICATION FOR CHANGE 10 CFR 50.54 paragraphs (m)(2)(11) and.(m)(2)(111)Lestablish' requirements for-j licensed operators for " fueled" nuclear power plants. Subsequent to defueling q j- the RV, TMI-2 will no longer be " fueled" and the requirements for licensed-F operators will no longer apply. The conditions for the existence of Modes 2-

!. and 3 require that the RV be defueled, consequently-the requirements of 10 CFR  :

i 50.54 (m)(2)(11) and (m)(2)(111) will not apply during Modes 2 and 3. Since t

! the requirements of 10 CFR 50.54 (m)(2)(11) and'(m)(2)(111) wlII not apply-  ;

during Modes 2 and 3, it follows that the requirements of paragraphs 6.2.2.a ,

and 6.2.2.b of the Technical Specifications also will only apply during Mode

1. . Therefore, the addition of "This requirement applies during Mode 1.

, only" is a clarification of the intent of these paragraphs'and not a technical j change in applicability.  ;

Paragraph 6.2.2.c stated that "An individual qualifted in radiation protection ,

procedures shall be on-site when fuel is in the reactor." The conditions for the existence of Modes 2 and 3 require that the RV be defueled. Therefore, l the addition of "This requirement applies during Mode 1 only"its a clarifica-

tion of the intent of this paragraph and not a technical change in the! -

[ applicability.

Paragraph 6.2.2.d stated, in part, "The Site Fire Brigade shal1 not include 3 members of the minimum shift crew necessary for safe shutdown of the Unit and~

L any personnel required for other essential function during a fire emergency."

This paragraph has been revised to clarify that the restriction on the minimum shift crew applies only during Mode 1 which is consistent with the justification for 6.2.2.a.

43. TECHNICAL SPECIFICATION 6.8.2.2. " PROCEDURES" i

i, DESCRIPTION OF CHANGE l'

i The phase "During Mode 1" has been added to indicate that the criterion of l- Specification 6.8.2.2 does not apply following the completion-of Mode 1.

b

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REASON FOR CHANGE The criterion of Specification 6.8.2.2 is based on the limits established in 10 CFR 50 Appendix I. Following completion of Mode 1, the most significant events that will occur will be the handling and transfer of canisters containing core material in the the FHB. Based on the experience achieved in performing these activities and the controls established in current NRC-approved procedures, GPU Nuclear believes that the potential for the occurrence of a canister handling accident resulting in exposures exceeding 10 CFR 50 Appendix I limits is highly unilkely. Thus, the criterion of Specification 6.8.2.2 will no longer be applicable after Mode 1.

JUSTIFICATION FOR CHANGE TSCR 55 proposed the following modification to the Specification 6.8.2.2:

" Procedure of 6.8.1.a and changes thereto which alter the distribution or processing of a quantity of radioactive material the release of which could cause the magnitude of radiological releases to exceed 10 CFR 50 Appendix I limits shall be subject to approval by the NRC prior to implementation."

To justify deletion of the above specification following Mode 1, the following evaluation demonstrates that the potential for activities to exceed 10 CFR 50 Appendix I limits, during Modes 2 and 3, will either be precluded or minimized to a degree such that the health and safety of the public will not be jeopardized.

The inventory of radionuclides remaining on-site during Modes 2 and 3 will be significantly reduced from that existing prior to the accident. This results from natural decay, removal of the fuel (which represents the largest concen-tration of radionuclides) and processing and shipping radioactive waste. The remaining radioactivity can be characterized as residual contamination located primarily in closed piping systems or surface films closely adherent to equipment or structural surfaces. Exceptions are the reactor building basement (282' level) and the fuel canisters stored in the FHB.

The radioactivity in the RB basement is dominated by the block wall enclosing the stairwell and elevator. Radionuclides (primarily cesium and strontium) have been absorbed into the concrete block structure during the period when the wall was (partially) submerged in the highly contaminated water collected in the RB basement during and following the accident. Since the radioactive material is embedded in the concrete, it is not readily available as a near term source for airborne release. Over longer periods of time, however, mechanisms related to diffusion and leaching by cyclic changes in moisture content may transport a fraction of the radionuclides in the block wall to the surface where it can become available for suspension.

Even though this fraction is expected to be small, the inventory of the block wall (i.e., an estimated 17,000 Cl of Cs and 8000 Cl of Sr) could make any suspension of radionuclides reaching the surface a significant airborne source term. For the purpose of this analysis, a fraction (i.e., conservatively estimated to be 31.) of the radionuclide inventory in the block wall, and to a lesser degree from other surfaces containing some residual contamination, is assumed to reach the surface and contribute to the inventory of "suspendable" contamination. This results in a total of about 1000 Cl of "suspendable" 0083P

m - . . . - _. . .

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conta

Inatten in the vicinity'of the block wall 'which, fer the purpose'cf this

~

analysis, is considered the largest single. potential. source fcr airborne

contamination.

~ ~

Analysis of core samples of-the block wall and concrete in the R8 basement

Indicate cesium to strontium ratios'of near 1:1,1or lower, from comparisons-of 6 and y emissions,'as opposed to the ratio ~of approximately 2:1, noted ..

4

.above, for a sample analyzed by destructive. assay. The conservative 1:1 Cs'to i

Sr ratio-(e.g., 500 Cl of.Cs and 500 Citof Sr) was used for.the assumed

. . postulated: accident airborne , concentrations following-defueling.

' .Another Important ftctor in the consideration of residual contamination =1s the content of transurantc elements. Although-the. quantity and configuration.of l fuel remaining after completion of defueling is insufficient to be of concern 1 with respect to criticality it is necessary to examine the potential'contri-

! bution it could make-to radiological source terms.

The core inventory of significant transuranic elements (i.e., greater than 4 0.1% of the core transuranic inventory, on a curie basis) remaining after

( eight (8) years of decay is shown in Table 3. On the basis of the. samples  !

! analyzed to date, as.well as the analyses of the course of the accident the 3 transuranic elements can be assumed to be associated with residual fuel. . For

an eight (8) year decay time, Table 3 indicates.a plutonium / americium mixture

of 1.23 E+5 Ci for the total.of 98,000 kg of fuel., or 1.25 C1 per kg of fuel.

Most of the residual fuel remaining following'defueling will,be fixed in the-form of very thin surface flims on reactor components or deposits in inacces-

sible locations of the reactor coolant system. Therefore, they will not 1- contribute significantly to the airborne' source term. During the accident, a i small quantity of fuel fines were carried to the RB basement by reactor j coolant escaping through the struck-open relief valve. About'1.7 to 3.2 kg of j these fuel fines are estimated to be mixed with other solid materials in the i sediment in the RB basement based on information available'in mid-1986. Much

1. of this sediment will be removed prior to the.end of defueling. However, j because of the difficult operating conditions in the RB basement, it is i assumed that a fraction of the fuel fines in the sediment will remain after i removal of the bulk of the sediment. As a reference point for the calculation j of potential off-site consequences, it is assumed that no more than about 2 kg.

j of fuel will remain as a suspendable source following defueling.

[ ' Additionally, various other radionuclides will be'present in the residual '

contamination; however, they are not specifically evaluated here since.the most significant radionuclide contribution to the off-site dose impact results' j} from the transuranics listed in Table 3 and Sr-90, and will be bounded by

j. consideration of these other radionuclides.

1

! For the postulated bounding conditions (i.e., a fire affecting essentially all i of the suspendable inventory in the vicinity of the block wall (e.g., 1000 Cl) j and the 2 kg of " loose" fuel) an experimentally determined suspension fraction

of 5 x 10-4 was used <

Reference:

J. Mishima and L. C. Schwendiman, t " Fractional Airborne Release of Urantum (Representing Plutonium) During the Burning of Contaminated Hastes," BNHL-1730, Battelle Pacific Northwest Laboratories, 198h. This results in a fission sgurce term of 5 x 10-3

curies and a transuranic source term of 1.2 x 10-3 curies assuming a 99%

filter efficiency.

l '

4

)

I

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.i.

TABLE 3 CORE INVENTORY OF SIGNIFICANT TRANSURAMIC ISOTOPES REMAINING AFTER 8 YEARS DECAY I ISOTOPE C1 INVENTORY C1 FRACTION Pu-238 7.73 x 102 6.30 x 10-3 Pu-239 8.98 x 103 7.32 x 10-2 Pu-240 2.38 x 103 1.94 x 10-2 Pu-241 1.09 x 105 8.87 x 10-I Am-241 1.71 x 103 1.40 x 10-2 TOTAL 1.23 x 105 1,00 1 - Significant means greater than 0.1% of the core transuranic inventory, on a curie basis.

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The cff-site dose from the accidental release cf these fission products and transuranics were evaluated using an emergency dose calculation routine similar to that used to calculate off-site doses for emergency planning purposes. The meteorological conditions for this case were chosen to produce a X/Q of 6.lE-4 which is normally used as the worst case accident X/Q for this type of evaluation. The accident calculations include only the inhalation dose pathway since it is expected that the other pathways (e.g., the ingestion pathways) can be restricted by protective actions under accident conditions.

For this case, the maximum dose to an Individual from fission products and transuranics is calculated to be 3.9 mrem to the bone which is well below 10 CFR 50 Appendix I limits.

Additionally, during Mode 2, there will be canisters containing fuel material in Spent Fuel Pool 'A'. These canisters have been demonstrated to be criti-cality safe and any potential accident scenarios associated with these canisters are addressed in NRC-approved safety evaluations. The only accident scenarios which has the potential to exceed 10 CFR 50 Appendix I limits is a canister drop resulting in a rupture of the canister and spilling of its contents (

Reference:

" Safety Evaluation Report for Defueling of the THI-2 Reactor Vessel," 4350-3261-85-1, Revision 10, and " Safety Evaluation Justifying the Non-Selsmic Design of THI-2 Post-Accident Systems,"

4430-7322-85-1, Revision 0).

However, the expertence gained to data with canister handling coupled with the safety factors of the canister handling equipment and the controls established in current NRC-approved procedures provide a high degree of confidence that a canister drop accident is highly unlikely to occur. Similarly as discussed in the TER for Defueling Canisters, 15737-2-G03-114, Revision 2, leakage of canister contents is not expected for the drop heights tested based on tests performed by Babcock and Hlicox. Therefore, even in the event of a canister drop, an off-site release which could exceed Appendix I guidelines is unlikely.

Thus, based on the rationale provided above, the need to have NRC approve any procedures at THI-2 following the completion of Mode 1, with the single exception of those procedures regarding disposal of Accident Generated Water, is no longer extant and this requirement can be deleted.

Further, based on the proposed revision to Specification 6.8.2.2 in TSCR 55, the following Technical Specifications and Recovery Operations Plan Sections require procedures approved pursuant to Specification 6.8.2.2:

o Section 3.1.1.1, "Boration Control" o Sections 1.7, 3.6.1.1, and 4.6.1.1, " Containment Integrity" o Sections 3.6.1.3, 3/4.6.1.3, and 4.6.1.3, " Containment Airlocks (Mode 1)"

Since the above referenced specifications are app 11 cable only during Mode 1, the proposed change to Specification 6.8.2.2 does not affect these specifications.

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