ML20236V454
| ML20236V454 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 11/23/1987 |
| From: | Doreen Turner GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20236V448 | List: |
| References | |
| 4710-3221-86-01, 4710-3221-86-011-R01, 4710-3221-86-1, 4710-3221-86-11-R1, NUDOCS 8712040315 | |
| Download: ML20236V454 (18) | |
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Nuslear SAFETY ANALYSIS SA f_47_]0-3221-86-011 l Rev. # 1-Page 1 or 14 TITLE SAFETY EVALUATION REPORT FOR -
l CORE SUPPORT ASSEMBLY AND I
LOWER HEAD DEFUELING
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No.
Nuslear 4710 3221-88-011 Title SAFETY EVALUATION REPORT FOR CORE SUPPORT ASSEMBLY Page 2 of 34 AND LOWER HEAD DEFUELING
SUMMARY
OF CHANGE Approval Date Rev.
0 Issued for use. 2/87 1 Revised to permit the use of unborated coolant water 11/87 in the plasma arc torch. The torch coolant inventory is limited so that no more than three (3) gallons are j able to drain into the reactor vessel. Updated the jobhours and person-rem expended for defueling.
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. q 4710-2221-86-011' TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE AND SCOPE 4.0 1.1 Purpose 4.0 i 1.2 Scope 4.0 2.0 MAJOR ACTIVITIES AND EQUIPMENT 5.0 3.0 COMPONENTS AND SYSTEMS AFFECTED 6.0 l
4.0 SAFETY CONCERNS 7.0 !
4.1 General 7.0 4.2 Criticality Control 7.0 4.3 Boron Dilution 8.0 4.4 Hydrogen Evaluation 8.0 4.5 Pyrophoricity 8.0 4.6 Submerged Combustion 8.0 4.7 Fire Protection 8.0 4.8 Decay Heat Removal 8.0 4.9 Instrument Interference 9.9 l
4.10 Release of Radioactivity 9.0 4.11 Reactor Vessel Integrity 9.0 4.12 Heavy Load Drops 10.0 5.0 RADIOLOGICAL CONSIDERATIONS 10.0 i
6.0 IMPACT ON PLANT ACTIVITIES 11.0 )
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7.0 10 CFR 50.59 EVA'..UATION 11.0 l l 8.0 ENVIRONMENTAL ASSESSMENT 13.0
9.0 CONCLUSION
S 13.0 l
10.0 REFERENCES
14.0 APPENDIX A - Evaluation of Loads Handled Over the Reactor Vessel l
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1.0 PURPOSE AND SCOPE 1.1 Purpose The purpose of this. Safety Evaluation Report (SER) is'to demonstrate' that the activities. associated with defueling the upper and icwer coro - fj support assembly (CSA);and the 1ower head (t.H) in:the TMI-2' reactor U vessei can be ar.complished without causing ur; acceptable. risk to?the! .j health snd safety of the public. j 1.2 Scope
'l This evaluation addresses the following activitier.: ')
i o Removai Sf core debris.from upper and lower CSA, including removal i of CSA structural matsrtal when such etructtical futerihi 'and core' lj i
l debris are :ict readii.y separable. !
1 o Removal'of CSA structural Waterial-to gain access to debris i l' deposits'within or below the CEA, R l
3 l o Removal of sections ci the ell!ptical flowilstributor fromt which: )
l core debris ic not readily separable; 1 l
o Removal of sections of elliptical flow distributor toggnin access- q to debris deposits in the LA i 1
o Renoval of core dabris from the LH. I l
o Installation / operation / removal of adcitiona'l equipment in support 1 of the above activities. ~l NOTE: LH oefuelino t.y vacuuming was addressed in Reference l'.
l CSA structural material not plac9d in defueling canisters may be Stored l in the reactor vessel or in other out of vessel temporary containers i which will be addressed in separate documentation. ;
Additional equipment to that discussed in ReferenceL1 required to support these activities consi;ts of:
I o cavitating water jet ;i o plasma arc cutting tool !
o Automatic Cutting Equipment System (ACES) q l o robot manipulators As the CSA/LH defueling operat!ons proceed, tne potential exists that activities or equipment described in this report or Reference 1 will need to be modified or new activities and/or tooling developed. 'Any modifications to existing activities or equipment or the introduction of new activities or equipment will be reviewed and documented in accordance with THI-2 administrative procedures to ensure.tnat no 4.0 0067P/Rev. 1 -i
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potential hazarns or i;afety concerns, not tounded by this SER or '
Peferentf. I are created. N no such hazards or safety concerns are created, CSA/LH defueling may proceed bl. sed en tne new or modified <
activities or equipment without a requirement to revise this SER.
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2.0 MAJOR ACTIVIYIES AND E0M FMENT CSA/LH defueling will be performed in accordance with detailed approved .j procedures. Any of the approved activities performed or tools usad during )
ir.itial and/or core region defuelirig are considered acceptable during CSA/LH j defueling boless specifically precluded. The initial and core region J defueliiig activities and tools are evaluated in Reference 1. Additional ;
operations to be performed during CSA/LH defueling. include:
4 o Core debris and structural inaterial removal from the upper and lower CSA j 1
- o Cutting the ripper and lower CSA within the reactor vessel ;
l i o Core debris removal from the reactor vessel LH Descriptions of tools in addition to those described in Reference 1 to be used j for CSA defueling are provided below. ;
1 Cavitating/ Pulsating Water Jet System / Flushing System
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The cavitating/ pulsating water jet system is provided tc; erode fuel debris from metal surfaces within the reactor vessel and to break up large debris pieces to facilitate removal. The system will flush tightly adherent debris j from vessel structures and will break up the fuel debris into particles ;
amenable to vacuuming. The cystem consists of high pressure discharge pumps '
(approximately 6-15 gpm at 10,000-20,000 psi), cavitating jet nozzles and.
lances, and connecting hoses and piping. The pumps will be mounted on the 347'-6" elevction, be powered by electric motors and take suction from the Defueling Water Cleanup System (DWCS) suction. Any potential siphoning of the reactor vessel inventory as a result of a line break upstream of the pump is limited by the safety systems inherent in the Defueling Water Cleanup System i (DHCS) (Reference 4). Piping downstream of the pump is precluded from siphoning because it is fixed above the Ractor Coolant System (RCS) water level. The cavitating/ pulsating water jet system will be operated using the remote manipulator or other positioner to allow remote manipulation of the device.
Plasma Arc Torch The plasma arc torch is provided to cut electrically conductive materials, such as stainless steel structures, which inhibit access to fuel to be removed. The torch will be operated via the remote manipulator or other positioner to allow remote operation of the torch. The torch is a direct current arc, tungsten electrode, metal burning device. An initial pilot arc will ionize the primary gas, nitrogen, to form a plasma jet. A secondary gas, also nitrogen, is used to aid in flushing away the molten metal from the cut and to provide insulation for the torch head. The total gas flow is approximately 20 scfm. Testing of the plasma arc torch has shown that RCS grade or B-10 enriched borated water cannot be used as tne torch coolant due to the high electrical conductivities of these fluids. Further testing has 5.0 0067P/Rev. 1
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> l determined that the .use of dominera11 zed (i.e., unborated) water results in acceptable torch operation. Consequently, unborated vater will be used as the cooling fluid for the torch. The torch coolant inventory is limited so that j no more than three (3) gallons of unborated water are &51e to dr61n into the 1 Reactor Vessel. Criticality concerns associated with'the use of unborated j coolant water are addressed in Section 4.2 of this SER.
Automated Cutting Equipment System The Autenated Cutting Equ3pment System (ACES) will positico the piasma arc t
torch to cut the ?ower CSA structura! elements to provide access to the fuel )
in the reactor vessel lower head. The equipment that will cperate in the l vessel it; a support frame that provides x-y positioning, a alanictlator arm )
that provides vertical travel, rotation, argular positioning, with & ah lity t:: grip, releas<.> and position the plasma torches. The in-vascel corrponents are pwered by a modified train or' three commercially tva11able plasma power supplies and one ACES power supply, and operated by a control system. The com':nterized centrol system is capable of controlling all five axes of the i in-vsssel equipment and can locate the torch reozzle and nove it over a I pre-determined pcth at contr.olled rates. The very important cuttirig parameter, cerch to work distance, is controlled cor,tinuously ana l l automatically by a serva mo'or t and feed back loop taking its sigr.a1 from ibe l
( tcrch arc voltage. All of the torch operations are pre-programmed after i verification of the progrcm modeled to the in-vessel lower CSA. The controller is located in a Coomand Center outsioe of the containment butjding l and is supported by a ecmputer-assisted-design medel of the lower _CSA. The l operatort are assisted with bcth video monitor and printer output.
Robotic hanipu'ator l Two hydraulic operated manipulator arras willle mountec' on the Manua? Tool Positioner (MTP) or other suitabH masts. One of th? manipulators (Grat'ber) can be used to stabilize the tiTP while the other manipulator (Word is used to help remove debris and structural material after it has been cut. The manipulators will have n separate borated hydraulic power wppl) and will be normally operated from outside tte reactor M lding.
Mechanical Tools i Mechanical Tools will be used to cut structural material (abrasive saw) and i prepare structural material for the plasma arc torch (grinder / milling tools).
Some tools will be powered by a borated hydraSlic power supply, i
3.0 COMPONENTS AND SYSTEM AFFECTED Other components or systems in addition to those described in Reference 1 may be required to conduct the CSA/LH defueling activities. _Where this is the case they will be the subject of separate correspondence.
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4710-3221-86-h11 4,0 SAFETY CCNCERK5 4.1 General An evaluation of the actielties af stciated with CSA/U! defueling ident9)ed the following safety concerns:
o RCS Criticality Contr61 o Boron Otlution o Hydrogen Evoluticm o Pyrophoricity o Submerged Combut:b>n o Fire Protection o Decay Heat Recoval o Instrument Interference o Release of Radioactivity o RV Interjrity l 0 beavy Load Drops l
l Each of these issues are discussed below.
4.2 RCS Criticality Control The evaluations provided by Rderences 1,11, and 16 galerally bouna this .
coricern during CSA/tH defueling. The turch cooling system has, by design, a maximum unborated coolut invcatory of les? than tour (4) gallons. However, l the maximun amount of n borated water that could drain into the Reactor Vessel from the ccMant rystem when it is in its operating position is no more than three (3) Gallons. As this quantity of unborated water exceeds the previous . ,
l li1 nit of two (2) gallem et tabliMed in Raference 16, a criticality analysis . '
ws performad to demonstrate that the use of unborated coolant for-the plasma l are torch wocid not pose a criticality safety concern..-Reference 17 provides
[ the t'a11s, assumptions. and bound ng fuel models used in the' plasma < arc torch criticality analysis. Bhsed on iMe results of the analysis, it is' concluded thdt the plasma arc torch, with a maximum drainable coolant system inventory of three (3) gallons of unborated water, can be used to dismantle the LCSA, includina t're elliptical flow distributor' head, without developing a l criticality safety concern within the Reactor Vessel.
l The above- conclusion is cased on the following operational limitations.
o The plasma arc torch will only be used to cut the LCSA.
l o The maximum amount of unborated water that can drain from the torch coolant system when the torch is in its operating position is three (3) gallons, o Flushing of the plasma arc torch coolant-system with the torch within the vessel can only occur if there are no known leaks in the' coolant system, the torch is in-its hcme position, there is at least one-foot separation between the torch tip and significant debris quantities, and the gas purge is operating. Otherwise, the torch must be removed from vessel prior to connection of the flushing tie-in.
o The maximum inventory of unborated water permitted in the flush system storage tank is 15 gallons.
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1 4710-322 M 6-011' i o Following the. loss of. coolant inventory', thi torch w st it removed and repaired.before refilling the torch cooling systti o If in-vessel flushing of the: torch is being performed;.no load hsndling l operations (heavyorlight).arepermittedinorabovethereactorvess ;
4.3 Boron Dilution j Boron dilution concerns during CSA/LH defueling are bounacd hy the eval Mtionst provided by References -l 'and 12. To proclude the possibility of a_ hydraulic . -1 l fluidileak leading to a possible critichi configuration of'.fue and modereton, ;
all hydraulic fluid used with CSA/LH defueling tools will be torated to a4 '
least 4350 ppm natural boric acid. !
4.4 Hydrogen Evol'Jtion p Smallquantitiesof.hydrogengasgeneration(?essthan1/10SCFM);h111be'c .
L by-product of the. plasma arc cutting tool operation underatar. This hydrogen '!
will be diluteJ by.the off-gas treatment cystem, as i'egdired,.and.thosi.a- "
combustible concentration will not occur within the reactor builoing. 0ther a hydrogen related safety issues are bounded.by'the evaluations provided'1n '
Reference 1.
4.5 Pyrophoricity Pyrophoricity concerns during CSA/LH defueling are bound 6d-by' evaluations provided in References 1 and 14.
4.6 Submerged Combustion The use of underwater burning devices (e.g. . plasma arc torch) crestss a heat- i source not previously considered. This additional heat source is not.expectad to create a combustion concern since the plasma arc torch will be operated. '
i underwater. Additionally, testing of thermic torch and plasma arc burning-devices on alumina filled zirconium tubes underwater did not produce an.y-sustained ignition (Reference 5 and 7). It is. considered reasonable not to postulate a combustion reaction of exposed fuel debris dueLtoLoperation of-the !
plasma are torch.
1' 4.7 Fire Protection The evaluation provided by Reference 1 bounds this concern during CSA/LH defueling.
4.8 Decay Heat Removal Decay heat removal concerns during CSA/LH defueling are generally' bounded by the evaluation provided in Reference 1. .'The maximum power requirements.for i the plasma arc torch are 1000 amps at 200 volts DC. Operation of the torch underwater will provide a significant heat source; however, continuous operation is not probable due to the need to. reposition the torch. Even if the torch were to operate continuously for one hour, it would raise the RCS temperature only approximately two.(2) degrees. The RCS temperature will be monitored to preclude an uncontrolled water temperature increase.
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4.9 Instrument Interference Iss6es regarding M strumentfinterforcsce cause:1 by'the ute of t e plasma arc ,
torch are bounded by the ovaluaticq provided in Reference 7.
j 4.10 Release ef' Radioactivity The central 3or*e of the plason u. reaches temperatures of 20,000'F to- ,
50,000*F knd .ls rempletely . ionized. However, this hign energ) it quickly I dissipated and primarily heats the conductive metal'. It is expected that' fuel
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on the eetal 59tfarfs will also bt. heated to the ligold'or capor'statei . Most i fuel so heated will immediately.ox!dize, traasfer its teat to the curroending i wrter, resolidify and sink. soluble !sotopes tracoed in the fuel math x may j baccee c'icsolved in the water. This poulble incuase in the concentration of-radioactivity is- not ~ expected to bo . prohibitive or exceed that 'observet in 'the. l) cere drilling prograa. Safety concerns associated witr the' release of .
radioactivity from the reactor vessel to the environment ure bounded by the ]
evh?uctions in Reference 1. 'l 2.11 Raattor VAssel Integrity Damage to tha reattor vessel due to the operation of berning devices inside the vessel ha: Ocen confidered. Initially, the operation'of such dovices;is '
chysically limited to inside the confines of the core support structure and ;
the ellipttcal flow distributor where the torch 1:; more than one-foct away j from the reactor vessel wall. Cutting overations wi.ll begin.do the top of the '
CSA and wl!1 sequentially cut through the lower grid, lower grid flor distributor, lower grid forging, in-core instrument stopc! t plate to the elliptical flow distributor. Since- torch accus to the ellipti:a1 flor distributor is physicaliy precluded by the CSA structere until the upper.
layers are remaved, the elliptical flow' distributor (whibh is more than one foot from the reactor vessel wall) will' te cut with the plasma arc torcb only i after considerable esperience is , gained by its use elsewhere in the reactor vessel. Therefore, th arc or fine of such ;buraing devices; operating j
underwater, will always be operated at least 6 foot frcca the reactor vessel wall. Propagation of an arc through one-foot'of water is not possible, thus, damage.to the reactor vn sel wall due to the operation of-burning devices is precluded even when cutting the elliptical ' low distributor. :
Additionally, the use.of other tools that could potentially impart excessive loads to the incore instrument tube nozzles .or damage the reactor vessel wall (e.g., abrasive / water jet cutting system) will be limit-ar) to use within the confines of the. core-support structure and the elliptical flow distributor until most of the fuel within the lower CSA has been. removed after which procedural limitations will be applied. . Mechanical cutting devices, such as the abrasive saw, grinding wheel, cavitating water jet and impact hammer are not of sufficient size or power to damage the reactor vessel wall and, therefore, do not create a safety issue.
During the removal of fuel debris from the lower head, care will be exercised to prevent excessive loads on exposed incore nozzles. If, during the process -
of removal of fuel in the vicinity of an incore nozzle,' observations; indicate that the nozzle has suffered damage due to excessive temperatures, work will be halted and the situation evaluated further.
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1 Other reactor vessel integrity safety concerns (e.g., Assessment of. potential damaae to incore nozzles from pulling on incore instrument strings) are .
boundsd by the evaluations.provided in Reference 6.
4.12 Heasy Load. Drops After A portior, of the Icwer flow distributor has been rooved, the'incore l instiument nozzles and the reactor vessel lower head will be. exposed to~the !
potential of impact by dropped.lcads. Prior to that time,. the CSA structure- )
will preclude the dropping of heavy loads on the incor6 nozzles, i Appendix A describes analyses which de.nonstrate that load drop characteristics. .
csso;15ted with LH defueling do not significantlysc a promise tha integrity of- h '
the reactor vessel after the elliptical-flow distributor is remoued or l sectioned. In addition, the consequences'of the totaltfailure of en incore l nozzie have been previously evaluated. It was concludeo that GPU Nucleat has .
l the capability to promptly detcct a totally failed nozzle penetration and car I l
mEintain the RCS level at or above the reactor vessel nozzles (References'6- !
and 16),
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l The potehtial for a load drop. accident into.the reactor vessel is minimized by.
f careful sontrol of load handling activities'and the.u:e of. load handling j l equipment WM ch has bean conservatively designed and tested. Load handling '
l activities are performed ir, accordance with approved procedures:for such activities including 4000-PLN-3Di.02, "THI-2 Lifting and Handling Program;" '
Each specific Joad handling activity is controlled by a Unit Work. Ins'.ruction or procedure. Load handling activities will be performed by personnel who have been trained and qualified for these activities.
5.0 RADIOLOGICAL CONSIDERATIONS Based on a comparison of activities associated with Reference 1 to those associated with CSA/LH defueling, it is concluded that the radiological considerations associated with CSA/LH defueling are bounded _by Section.5 of Reference 1. An update of the jobhours and person-rem expanded to date for ,
all defueling activities is provided in Table 5.1. The overall estimated J occupational exposure to complete reactor vessel defueling remains at I approxinntely 1400 person-rem.
TABLE 5.1 Job-hours and Person-rem Expended Through October 1987 Activity Jobhours Person-rem Preparations, installations 3,930 100 !
Operations 34,899 343 Maintenance / Support 20,570 311 Decontamination and Removal
- O O TOTALS 59,399 754 ;
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- No activity associated with final decontamination and' removal of.defueling d equipment has been performed as of January 1,.1987,lthus no jobhoursiand l person-rem are given. Note, decontamination. maintenance in the!rea'ctor 1 building is;not considered part of this activity.
q 6.0 IMPACT ON PLANT ACTIVITIES The major potential impact of CSA/LH'defueling on plant activities is the effect of fuel movement in Unit'2:on operations in Unit 1.. Based on the il evaluation provided in Reference 1 ilnd the similarity of the activities '
considered in Reference 1 to those itctivities within the scope of.this SER. It h is concluded that the CSA/LH defuel! lng operations in Unit.2 will.not affect-personnel in Unit 1. j 7.0 10 CFR 50.59 EVALUATION ~ ,
1 10 CFR 50, Paragraph 50.59, permits; the holder of an operating.11 cense to.make:
changes to the facility or perform h test or experiment, provided.the: change,. l l test, or experiment is determined not to be an unreviewed safety ques tion and i I does not involve a modification of the plant technical specifications.
10 CFR 50, Paragraph 50.59, states a. proposed change involves an unreviewed .;
safety question if: ~
- a. The probability of occurrence or the consequence of an accident or- ..
I malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
- b. The possibility for an accident or malfunction of a'different type than. I any evaluated previously in the safety analysis report may be created; or ;
- c. The margin of safety, as definea in the basis for any technical specification, is reduced.
J Although there are notable differences between the proposed defueling activities for THI-2 and routine activities described.in the FSAR, the consequences of postulated accidents are not different and as demonstrated in Reference 1, are sufficiently similar to be compared. Reference'l compared l
two (2) potential events during defueling, a canister drop accident and a Krypton 85 release, to two (2) events described in the FSAR, a-fuel handling accident and a waste gas decay tank failure. The comparison demonstrated that on a worst case basis, the consequences of the FSAR events bound the consequences of any defueling-related event.
A variety of postulated events were analyzed in this SER for CSA/LH defueling. The analysis of these events provided in Section 4 results in the conclusion that the postulated events are-bounded by previoui; evaluations and/or do not result in an unanalyzed condition.
To determine if CSA/LH defueling activities involve an unreviewed safety question, the following questions must be evaluated.
Has the probability of o'currence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report been increased?
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'4710-3221-86'-011 1 A Yariety of events were analyzed.in Reference 1. It was demonsti. e d that these events were bounded by comparable events analyzed in.the F M . It was !
shown that the potential consequences from these events were.substantially j less than the potential ccesequences of. comparable events analyzed in the . i FSAR. Section 4 of this SER demonstrates that the consequences of: potential j
i svents during CSA/LH defueling are bounded by previous evaluations, 1
The proposed' activities for CSA/LH defueling will' eventually create a hole l'n j
the lower CSA, exposing a large area of the lower RV to direct impact from a. j heavy load. Appendix A provides a summary analysis of the potential damage to l the reactor vessel LH incurred by.a heavy load drop. . This analysis concludes '
that the potential load drop of either a loaded canister in sleeve or the Manual Tool Positioner and Manipulator (MTP/M) directly on an exposed incore instrument nozzle may result in leakage of the.RCS. However, the.MTP/M,'as discussed in Appendix A, will only be in position to cause. damage when it is 1 being taken in or out of the reactor. vessel. The majority of the time, the- i
.MTP/M will be in position within the reactor vessel where the drop height is 1 minimal. The canister sleeve handling tool and the Canister Positioning !
l System (CPS) both have locking cevices to pievent dropping of.a loaded l l cac. ster and sleeve. The it,cking device oh the canister sleeve handling tool j is verified to be engaged prior to lifting the canister and sleeve. The j locking device on the CPS is verified to be.engag?d after t% canister sleeve is positioned on the CPS. In addition, previous evaluations have shownuthat 4 sufficient leak detection and mitigation equipment it, Available and operable 'l should a load drop event, which damages the. reactor vessel LH,' occur. j The design features and administrative controls as; described in Reference 9- )
ensure that the probability of a load drop is minim 17ed. CPL' Nuclear will l also take special precautions for the handling of the MTP/M during j installation / removal from the reactor vessel.
By considering postulated events and reviewing various safety mechanisms. I l 1.e., fire protection and decay heat removal, it has been demonstrated that i CSA/LH defueling activities will not adversely affect equipment classified as !
important to safety (ITS). Consequently, it is concluded that the probability ;
l of a malfunction of ITS equipment or the consequences 'of a malfunction of ITS i equipment has not been increased. I Therefore, it is concluded that the proposed activities associated with CSA/LH ,
defueling do not increase the probability of occurrence or the consequences of any remaining accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
Has the possibility for an accident or malfunction of a different type, than ;
a_ny,n evaluated previously in the safety analysis report been created? '
i The variety of postulated events analyzed in Reference 1 considered'a spectrum '
of event types which potentially could occur as a result of the defueling '
process. A comparison of those events with comparable events in the.FSAR demonstrated that the event typus postulated for the defueling process are similar and bounded by the FSAR. In addition, no new event type was identified which was different than those previously analyzed in the FSAR.
Section 4 of this SER demonstrates that the potential events postulated for CSA/LH defueling are bounded by previous evaluations and do not create the possibility of occurrence of an accident or malfunction of a different type '
than evaluated previously in the safety analysis report.
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E Hasthemargin'ofsafety.asdefinedinthebasis'forany'tacEnibal specification been reduced? 3 93q Technical Specification safety margins at TMI-2 are concerned with critica'lity contrci and prevention of further core.. damage.due to overheating. Tecnnical Specification safety margins will be maintained throughout the CSA/LH j defueling-process. Suberiticality is ensured by establishing the'RCS boron t '
concentrationatgreaterthan4350ppmor'equivalentandensurinythatthis concentration is maintained by monitoring the boron concentration and 4 ,
inventory levals and by isolating potential _deboration' pathways.l Systems will ?
remain in place to add borated cooling water.to the. core.in'the event of.an j unisolable leak from the reactor vessel to prevent overheating. and cpotet tialt criticality. _ The introduction of unborated water _ from the: torch coolink 1 system will not present a criticality concern.
{
No Technical Specification changes are required.to conduct the~attivities '
bounded by this SER. .'
In conclusion, the CSA/LH defueling activities do not: j o Ir. crease the probability of occurrence or the'c' consequences of an accident or malfunction of equipment important to safety previously l evaluated in the safety analysis report, or o Create the possibility for an accident or malfunction of a diffe' rent type than any evaluated previously in the safety analysis repert, or.
s o reducethemarginofsafetyasdefinedinthebasisforany[echnical l specification. -l Therefore, the CSA/LH defueling activities do not constitute an unreviewed safety question.
8.0 ENVIRONMENTAL ASSESSMENT Based on Section 8.0 of Reference 1 and noting the similarities between the
~
i activities considered in Reference 1 to those activities within the scope of s this SER, it can be concluded that the proposed CSA/LH defueling activities s can be performed with no significant environmental impact. >\
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9.0 CONCLUSION
S j Activities associated with CSA/LH defueling have been described and /
evaluated. The evaluations have shown that.the radioactivity releases to the-environment that will result from the planned activities will not exceed-allowable limits. It has been demonstrated that the consequences of postulated accidents with respect to potential core-disturbances will nol '
compromise plant safety. The evaluations have also shown that the tasksiandL tooling employed follow the continued commitment to maintain radiation / '
T a i
exposure levels ALARA. Therefore, it is concluded that CSA/LH dafueling, e s ,
1 activities can be performed without presenting undue risk to the health and y safety of the public.
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'30.0' REFERENCES. 1 j( I l
> 1. Safety Evaluation Report for Defueling the TMI-2 Reactor.-Vessel, - '
Revision 10, 15737-G07-108, May 1986, ai
/ 2. Technical Evaliation Rep @ t.fdr Defueling Canisters, Revision 2,.
15737-2-Q03-114,\ January 1956.
l 3. Safety Evaluatirp Repor'c for. Canister Handling l and Preparation for r
[\
Shipment, Revis on 3, 15737-2-G07-111,oJune 1986.
- 4. Technical Eraluation Report for Defueling Water Cleanup System,: ~
j Revision 8, 15737-2-G03-106,. December 1985. l1
- 5. EG&G Plasma-Arc Test Report.:LCSD-34, April.30,'1986.
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- 6. GPU Nuclear letter to W.D.' Travers,.USNRCi 4410-86-L-0162 dated- ,
September 1$. 1986,
Subject:
Core Bore Operations, and Attachments. j tr u
- 7. GPU Nuclear letter to W.O.: Travers,'USNRC, 4410-86-L-0143 dated' i August 27, 1986,
Subject:
'Use of Plasma Arc Torch.
- 8. Tech cal Specification Change No. 46'
.s . . ;
) 9.
Safeth Evaluation Rep' ort for Heavy Load Handling ovf the TMI-2 Reactor ;
p ,, Vessel, Revision 0, 15737--2-G07-110.-April 18, 198 'p t - ]
)
- 10. GPU Nuclear letter to W. D. Travers .USNRC, 4410-86-L-0160 dated -
, September 9, 1986
Subject:
End Fitting Storage. I n, y s
,' 11. Criticality Regrt for the Reactor Coolant System., Revision 0, d 15737-2-N09 001, October 1984. )
s i b- Haq+sdifiysis: Potential for Boron Dilution of Reactor Coolant
, 'N' 12.
}/ 'T Systes, Kevision 2.
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'Y 13. TMT-2 p actor Building Purge - Kr-85 Venting, GEND-013. 1 i ( :
-I ; N. GPU TPD/(41-127, Revision 0, "hhnical Plan for Pyrophoricity"., '
i' Decemben 1984. -
a-3 5 t 15. Eubnded{ ore Stratification Sample Achisition Activity, GPU Nuclear
/# ,etter44?ba6-L-0122datedJuly11,1905
> 3 w ,n e s .
?h
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- 16.
- Reporty.' Limits 6f Foreign Materia 3 s LAlliwed t ri dhe THI-2' Reactor ;
Coolarit Sptem During,0duNing Aci)vU.in, '
Revision 1, 15737 N09-002, !
'(>3eptember 665. I F ), 3 i ,
/ w s a 1,1
' CriticalitySafats)AssessmentforUsing'tiePlasmaArcTorch'toCutthe.
LQA,'15737-2-N-0l 4, November 1987.
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/ t A L' \, f 5 s -
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= AP.P..f.NDIX A EV/J.UATI,0NS OF LOAD D40PS OVER TOE REACCOR VESSEL l During core support assemoly and lower head defueling, The lower core support asiambly (CSA) will have pieces cut from it and remcy/ad to gain access to core debels, hentually, a hole will be created through the lower CSA, exposing a large area of the lower reactor vessel head to direct impM t from heavy loads. Analyses have been performed to better dt:termine the potentiay damage which could be i incurred by the incore nozzles due to dropped loads.; 10 provide the analyses )
eported herein, simple calculations. were employed in order to ascertain if further j more conplex analyses were warranted. l l The followini objects were considered as potential accident loads:
l TABLE A Maximum ych_ievable Droo Heights for Q isideced Objects l \
1 DROP DISTANCE DROP DISTANCE OBJECT IN AIR ** IN HATER * ,
A. Light Outy Pole 52'-0" 36'-7" B. EndEffegtortundlingTool 56'-0" 36'-7" C. Loaded Defueling Canister 5'-6" 36'-7" D. Loaded Defueling Canister in Sleeve N/A 24'-0" E. Manual Tool Positioner w/ Manipulator 21'-7" 36'-7" l
- Distancf to bottom, inside surface of Reactor Vessel Lower Head
- Drops are sequential - ffrst air then water The analyses, in order to maintain a simplistic apprcach, made the following major assumptions:
- 1. Upon impact, all kinetic energy of the falling object is transmitted to the f instrumentation nozzle and results in strain. It:< assumption is conservative '
since some of the energy would also be converted t.o strain in the dropped cbject and the lower reactor vessel head. {
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- 2. lhe compressive stress-strain curvesfor a short column of inconel is identical I to the tensilt 5, tress-strain curve. This assumptier; is conservative since j ductile metals will fa51 in tension before they do in compression without buckling.
- 3. The static stress-strain curve for inconel is appropriate for dynamic l loadings. This assumption may be s1fsitly ut ccitservative as some metals )
l exhibit higher strength but lower dnility vith increasing load application ;
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- 4. The strain is uniform over the entire nozzle. This is not a conservative assumption as-the real possibility of the nozzle bending is neglected (see page 3 for bending considerations). .Use of this assumption gives.an upper bound on the permissible drop heights.
- 5. Virgin material properties were used for the nozzle and weld materials.. This assumption;is slightly non-conservative since the nozzle material properties at 1200*F have not been determined.
The objects under consideration when dropped through water will be subject to drag which could vary significantly, depending on the orientatio'n of the falling' object to the direction of movement; An examination of the potential coefficients'of drag for various sharp edged bodies indicates drag-coefficients varying from 0.5 to !
1.5. This indicates that the drag coefficient will have'a significant effect on j the calculated impact velocity for a water drop height of.30 feet or'more. In lieu- '
of actually calculating drag coefficients for all dropped objects, a' range of drag coefficient from 0.5 to 1.5 was used.
Assuming that the impact load is entirely in the axial direction and along the centerline of the nozzle, an upper bound on the permissible drop heights can be established.
It is conservative to assume that all the kinetic' energy of the impacting object I must be absorbed in the nozzle. Since the nozzle's stress-strain curve is known '
the limiting impact velocity can be determined. Knowing the impact velocity allows the determination of the drop heights by iteration. q The following drop heights were calculated.
TABLE B Allowable Drop Heights Cross Maximum Air Drop Water Drop Object Weight lbs.
Sectional Strike area-in.2 Velocity-in/sec height-ft 0.5 height-ft j 1.5 0.5 1.5- .i A 150 2.8 2120 >52.0 >52.0 36.6 36.6 B 500 9.6 1160 >56.0 >56.0 36.6 36.6' C 3350 154 449 --
> 5.5 34.1 36.6 :
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D 5100 254 364 -- --
19.6 >24 !
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E 4500 128 388 -- --
21.7 25.4 A comparison of the calculated drop heights versus the. criteria previously given in Table A shows that even for the very low drag coefficient (0.5) items A & B (the Light Duty Pole and the End Effector Handling Tool) satisfy the given criteria.
The loaded defueling canister with the minimum drag coefficient misses the water drop height criteria by about two feet (34.1' vs. )
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i 36.6') and the loaded defueling canister with sleeve misses by about four, feet '
(19.6' vs. 24'). Note that with the maximum drag coefficient of.1.5 both criteria are met. The manual tool positioner does not attain the maximum drop height criteria by several fh t with either drag coefficient.
A more realistic evaluation of the criteria for the dropped fuel canister indicates that the loaded canister when in a "droppable" position is a) within the Canister Positioning System (CPS) sleeve or b) within the port of the shielded work platform or c) over the port in the shielded work platform. For each of the positions from' which it might drop, it would strike the CPS first thereby decreasing its velocity. Further, the criteria of all the impact energy being transmitted to the incore nozzle is highly conservative' relative to the fuel canister; a. vessel with a 1/4" thick shell. In all likelihood dropping the. fuel canister on end.onto the ,
incore nozzle will result in significant bending and possibly puncture of the I bottom head of the defueling canister and little or no deflection of the incore nozzle. Consequently, only the loaded canister in sleeve and the Manual Tool {
Positioner and Manipulator do not satisfy the drop cr.iteria. The canister sleeve handling tool ann the CPS both have locking-devices.to prevent dropping of a loaded-canister and sleeve. The locking device on the canister sleeve handling tool is verified to be engaged prior to lifting the canister and sleeve. The locking device on the CPS is verified to be engaged after the canister sleeve is positioned on the CPS. Additionally, the dropping of a loaded canister and sleeve can only occur during a transfer of the sleeve from a loading position to the top position J on the CPS. Consequently, the loaded canister and sleeve have a very low probability of dropping.
The Manual Tool Positioner will be in a position where it is more than 22 feet above the reactor vessel lower head less than one percent of the time it is in the l
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reactor vessel. Obviously, most of the time this tool is in the vessel after holes l have been cut through the elliptical flow distributor is when it is being used to !
perform work on either the lower CSA or the RV bottom head. The drop height from '
these positions is minimal. Further, when the tool post is fully retracted it is at elevation 313'-6" or approximately 22 feet above the lower head. 'In this position the tool is supported on its rails and not on the lifting rig. 1 Consequently, the tool has a very low probability of dropping.
All of the above analyses considered that the dropped tool struck the exposed l incore nozzle on centerline. A realistic condition exists whereby the impacting object strikes the nozzle off-center creating both an axial load and a bending '
moment. An impact load on the nozzle taper would produce a lateral load and an additional moment would be created.
The magnitudes of the lateral load and bending moment are difficult to establish.
However, by again using the energy approach and simple inelastic equations for the deflection of an end loaded cantilever beam, the maximum energy absorbed can be compared with that for the axial load only condition.
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' Analysis has determined that the nozzle is capable of absorbing as a side load only about 6% of that which it can absorb as an axial load. If a substantial part of the postulated impact energy is applied horizontally the nozzle is likely to fall.
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parallel to the.inside surface of the reactor vessel lower head. Therefore, nozzle failure due to off-center loading ~could fail the nozzle but not cause significant leakage since the in-vessel segment of the 3/4" schedule'160 inconel pipe and its ';
weld would likely remain.
The potential of punching a hole through the lower head is greatest for.an axial i impact load on the incore instrument nozzle. As a worst case it was assumed that the ultimate axial load on the nozzle had to be taken in direct shear of the lower reactor vessel head shell. The stresses determined are well below the ultimate l- strength of the vessel wall. An undamaged nozzle, therefore, cannot be pushed through the vessel wall.
Of the potential failure mechanisms, it is concluded that t.he worst anticipated incore nozzle failure mechanism is shearing off the nozzle at the inside surface of the reactor vessel lower head.
As previously noted, the 3/4" schedule 160 portion of the instrument tube which penetrates the vessel wall is welded directly to the vessel wall. The 2" 0.D. -j incore instrument nozzle is welded separately to the vessel wall and the 3/4" ~
pipe. Failure of the nozzle is unlikely to fail the 3/4" pipe to vessel' weld.which provides the penetration seal. For conservatism, however, it is assumed that this weld fails as a result of the postulated load drop accident.
Failure of the tube-to-vessel-wall weld will not result in the tubes being forced out of the lower head by the head of water'in the vessel. The. tubes consist of schedule 80 stainless steel pipe and are supported at'the floor below the vessel.
The maximum clearance, taking into account manufacturing tolerance, between the.00 y of the tube and the ID of the bore in the vessel wall is 0.010 inches. There is insufficient flexibility in the tubes to allow them to drop the 5 1/2 inches required to fall free of the bottom of the vessel head, j t
Incore tube failure outside of the vessel is not considered credible. Consequently the only credible leakage path from the vessel following a heavy load drop is' .
through the annulus around the tube penetrations through the vessel wall. This leakage has previously been calculated to be approximately 0.40 gpm per no zle penetration.
This analysis indicates that a potential load drop directly onto an uposed incore instrument nozzle may result in the leakage of reactor' coolant water through the nozzle - vessel hole annulus. Previous submittals have shown that sufficient leak detection and mitigation equipment is available and operable to combat leakages due to the discharge of an entire incore nozzle (125 gpm) from the reactor vessel.
i Consequently, the capability exists to promptly detect the existence of any failed .
incore instrument nozzle and to maintain the reactor coolant system water level at l or above the reactor vessel nozzles, i l
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