ML20086M302

Forwards Response to NRC 910830 Request for Addl Info Re Rev to Proposed Decommissioning Plan.Subjs Discussed Include, Dismantling Pcrv Core Components & Final Dismantling

ML20086M302
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/06/1991
From:	Crawford A PUBLIC SERVICE CO. OF COLORADO
To:	Weiss S NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
P-91423, NUDOCS 9112170322
Download: ML20086M302 (46)

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P A S m 840 Demo to 10201+ oB40 December 6,1991 A. Cle0g Crawford Fort St. Vrain

\\he Pindent Unit No.1 P-91423 U.S. Nuclear Regulatory Commission NITN: Document Control Desk Washington, D.C. 20555 A*ITN:

Dr. Seymour H. Weiss, Director Non Power Reactor, Decommissioning and Environmental Project Directorate Docket No. 50-267

SUBJECT:

PSC RFSPONSE TO NRC REQUEST FOR ADDITIONAL, INFORMATION ON TIIE FORT ST. VRAIN PROPOSED DECOMMISSIONING Pi AN

REFERENCES:

(See Attached)

Dear Dr. Weiss:

The purpose of this letter is to respond to the NRC's Request for Additional Information (RAl), forwarded to Public Service Company of Colorado (PSC) in Reference 1. The NRC RAI was developed based on the NRC review of a revision to the Proposed Decommissioning Plan for Fort St. Vrait Nuclear Generating Station

.nd a PSC response to the previous.NRC RAI (dated February 8,1991), that were submitted to the NRC in References 2 and 3. PSC requested a delay in its response to this RAI until the end of November 1991 in Reference 4 The Decommissioning Funding Plan and a revision to the Decommissioning Cost Estimate (originally submitted in Reference 5) were fcrwarded to the NRC in Referenec 6. These documents provide PSC's responses for NRC RAI Questions No.

3 and 4.

The attachment to this letter provides PSC's detailed responses to the following NRC questions:

No.13 Distnantling PCRV Core Components (detailed response for removal of graphite blocks with Hastelloy cans only)

No.16 ~ Final Dismantling (Concrete cutting test results)

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. P-91423 December 6,1991 Page 2 The following is the schedule for PSC responses to the remaining questions forwarded in Reference (1):

December 20,

• No.1I Water Cleanup and Clarineation System 1991

• No. 38 Liquid Wastes January 17,

• No. 9 PCRV Dismanttement Activities 1992

• No.12 PCRV Top llead Concrete and Liner Removal

• No.13 Dismantling PCRV Core Components

• No.14 Core llarrel Removal if you have any questions related to the contents of this letter, please contact hit. hl.

11. Ilolmes at (303) 480-6960.

Very truly yours,

/} /bb,4 iy;.

f A. Clegg Crawford Vice President Nuclear Operations

- ACC:CRil/cb Attachments cc:

Regional Administrator, Region IV hir. J.11.'Ilaird Senic-Resident inspector Fort St. Vrain hir, Robert ht. Quillin, Director Radiation Control Division Colorado Department of Health 4210 East lith Avenue Denver, CO 80220 x

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P 91423 December 6,1991 Page 3 REFERENCl3 (1)

NRC letter, Erickson to Crawford, dated August 30,1991 (G 91178)

(2)

PSC letter, Crawford to Weiss, dated July 1,1991 (P-91217)

(3)

PSC letter, Crawford to Weiss, dated April 26,1991 (P 91118)

(4)

PSC letter, llrcy to Weiss, dated October 25,1991 (P-91371)

(5)

PSC letter, Crawford to Weiss, dated June 6,1991 (P 91198)

(6)

PSC letter, Crawford to Weiss, dated November 15,1991 (P-91392) i w

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n PSC RESPONSE TO TIIE NRC RAI DATED AUGUST 30,1991 9

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HRC_Quntion.Neldection.111LDecomminienina_Cau

'On June 6,1991 PSC stdanitted tyrtGsed c<nt estink1te report uhich is turw tender resiew by the NRC stqf. '

P1C.1012enhc In addition to the Detailed Cost listimate submitted to the NRC on June 6, 1991 (P-91198), PSC provided an update to the original detailed cost estimate with the Decommissioning Funding Plan that was submitted to the NRC on November 15,1991 (P 91392).

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December 6,1991

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NRC Ouestion No. 4 (Section 1.3.2: Funding Plan) i L

'PSC has not responded to questions 4.A and 4.B and indicated a response may be I

provided in the thint quaner this year. 7his remains an open isstor that must be resolved as part ofNRC apprmul of the Decornmissioning Plan. "

t PSC Resacast PSC submitted the Decommissioning Funding Plan to the NRC for review on t

November 15,1991 (P-91392).

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i Attachment to P-91423 December 6,1991 NRC_Qnut_imLNL2ErnimL221J11CluLDismantlemenL&tisticil

• Widle PSC has addres<ed wine of <mr concerns, PSC imut sabmit the final dismant!!ng inethods and a supponing analysis for NRC review. Alternatively, provide description and sqfety analysis for poten:ial options that may be used.

Inchtde evalua: ions of and methods to minimi:e personnel e.rposure in your safety analysis. "

PSfJicPonic During the weekly telephone conference call of November 13,1991 PSC and the Westinghouse team identified nine dismantienent activities to be performed inside the PCitV for which detailed descriptions and safety analyses would be -

prepared. These descriptions and supporting analyses will be performed in a level of detail similar to that prepared in the enclosed response to NitC Question No.13 for graphite b'ocks containing llastelloy cans.

Detailed evaluations will be prepared for the following activities:

1.

Water Cleanup and Clarification System (llAl Questions 11 and 38) 2.

Itemoval of Top ilead Concrete and Liner (Question No,12) 3.

Itemoval of IIcx illocks with flastelloy Cans (Question No.13) 4 Itemoval of Core iTarrel and Keys (Question No.14) 5.

Itemoval of Hex 111ocks without llastelloy Cans 6.

Itemoval of Side Spacer Blocks with lloron Pins 7.

Itemoval of Large Permanent Side iteflectors 8.

Pemoval of Core Support Floor 9.

Itemoval of Steam Generator Primary Assemblies These activities were selected on the basis of criteria discussed with the NitC:

(1) potential for high radiation levels or Hgh occupational exposures, or (2) unique evolutions with a high degree of difficulty. Per conversations with the NRC staff, both the criteria and the list of proposed evolutions for detailed evaluation appear to be reasonable.

Detailed evaluations for the remaining activities identi0ed in the above list will be prepared in a manner consistent with PSC's response to NitC Question No, 13, and these evaluations will be forwarded to the NitC by January 17, 1992, 9-1 I

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f Attachment to P 91423 December 6,1991 NRC Ouestloa No.11 (Secticall3.6.2: WaltLCkanup and Clarification Syltc]O r

'Provic'e artalysis ofradiologicalconsequences ofsystern operation. Further, provide a sqfety analysis for potential accident scenarios whh regard to occupd!vnal and public exposure. Include analysis of radioacthe snaterial (including Irlthun) that is released to this systern. "

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detailed response to this question will t.e provided by December 20,1991.

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Attachment to P-91423 December 6,1991 NRC Outstion No.12 (Section 2.3.3.7: PCRV Top _ liriiti Conctrlc_aud_Liuri litment

' Provide a sqfety analysis of pn>cedures being developed to minimi:e personnel e.tposure. What matimum radiation levels are expected at uvrker locations during removal ofradioactiw components? Neither the April 26,1991 response nor the Julo i revision to the Decommissioning Plan provide thic infonnation.

• PSC Response:

A detailed response to this question will be provided by January 17, 1992.

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Attachment to P 91423 December 6,1991 NRC Quc10nn_No.13 (Section13.3.E: Dismantling PCRV Certfampenculs)

'PSC'S April 26th response states that three methods are being evaluated for trmoving hlghly radioactive componctusfrom the core. 1he revised Decommissioning Plan cf Jtdy 1,1991 still does not provide sufficient detall on the removal of 272 reficciar blocks containing about 20,000 hastelloy cans, each reading 10,000 R/hr.

What e.rposure lewis undd workers be sub}ceted to daring removal of PCRV radioactive compotu'nts?

A comprehensive sqfety analysis of the removal <f radioactive parts must be submitted as part af the Decommissioning Plan. '

PSC ResPen5C 1.

COMIONENT !)ESCRll'flON A. fi.cattaLuceripilon - Graphile_Dkds t

The-PCRV contains various types of graphite blocks. A listing of the types of graphite blocks that will be removed during decommissioning nctivities include the following:

RADIATION LEVEL GRAPH!TEDIQCK M E

[On cnntatt withnRLhhicMag) 1.

Defueling Blocks

<1 mR/hr 2.

licx Reflector Blocks w/o liastelloy Cans 500 mR/hr 3.

llex Reflector Blocks with liaste!!oy Cans 300 R/hr 1

Laij Permanent Reflectors

< 30 R/hr i.

Side Spacer Blocks w/o Boron Rods

<3 R/hr (incorrectly listed at <3 mR/hr in PDP Table 2.31) 6.

Baron Rods 60 R/hr

-7.

Core Support Blocks 15 mR/hr 8.

W l Clad Reflectors (non-control rod) 300 R/hr 13 1

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Attachment to P-91423 December 6,1991 These top layer blocks will be removed using the Fuel llandling Machine (Film) and current plant methods to transfer the components from the PCitV to a shipping I

container, as discussed in Section 2.3.3.4 of the Decommissioning Plaa and WilS No. 2.3.1.8.2 of the Decommissioning Cost listimate. Use of the FliM will provide the necessary shielding and containment while transferring components from the PCitV to the shipping containers with minimal personnel exposure. The following is a discussion for those blocks that will be removed with manually operated tools.

B.

IkutomLGaphite RdaiEilh.Huldley_ Cans - Delaikilk.w&lis The table in Section A above indicates that the hex reflector blocks with llastelloy cans have one of the highest radiation levels of those irradiated components to be removed from the PCitV with manually operated tools. The liastelloy can hex reDector blocks are shown pictorially in Figure 131.

Each block has 0.531-inch diameter holes to accommodate liastelloy cans. There are 270 hex reDector blocks that contain 72 IIastelloy cans and 4 hex reDectors that contain only 2 liastelloy cans

- each. The llastelloy cans are 0.51 inches in diameter, approximately 8 inches long, and contain boronated graphite. The llastelloy cans are shown in Figure 13-2. The location of flastelloy can hex reflector blocks in the PCllV is shown in Figure 13 3.

- Table 2.31 of the PDP indicates that the expected radiation level for the llastelloy cans is 10,000 It/hr on contact. This value represents an infinite plane source of pure flastelloy with no shielding from the surrounding graphite. Ilowever, since the llastelloy cans will not be removed from the graphite blocks and grouped together, the infinite planc source will not exist and the radiation Geld of 10,000 It/hr will not be present. Therefore, calculations were performed to determine the radiation levels on contact with the surface of the hex renector blocks with llastelloy cans by modeling the block itself as an infinite plane source. These calculations provide an i

estimated contact radiation level of 300 It/hr on the surface of the hex rencetor blocks with llastelloy cans.

As stated previously in PSC's April 26th response, the llastelloy cans in the hex reflector blocks are not expected to fall out of the block. Therefore, removal of the

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liastelloy cans using -a dumping or tipping operation will not be attempted.

itemoving the Ilastelloy cans from the graphhe blocks will require the use of some mechanical method (broaching, cutting, pressing, and/or crushing). Afler considering the methods for removing the }{astelloy cans and comparing them to the alternative 13 2 e-sgs,.

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Attachment to P 91423 December 6,1991 of leaving the llastelloy cans in the bhicks, it was decided that the llastelloy cans

- 1 would be left in the blocks. leaving the llastelloy cans in the bk>cks eliminates the need for special equipment to remove the liasulloy cans, simplines the process of

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Dl'.SCR I PTION O F Til E G R A PillTE llLO C K it D IOV A l. PRO C ED UR E i

A.

firRctalAIIangement of the Work Atr.tG2r_QntphittllkskjkRWXal The arrangement of the work area that will tyrically be used during removal of all types of graphite blocks is shown in Figure 13 4. To gain access to the core for y

removal of the core components, the top head will have been cut and removed to form a large hexagonal opening (See PDP Section 2.3.3.7; further detailed di:cussion of the top head removal will be provided in response to RAI Question No.12). The final breach into the PCRV will be neomplished by cutting and removing the remaining concrete and liner in a ciret.lar shape corresponding to the PCRV inside diameter. Prior to the final breach of the PCRV liner, the PCRV will have been filled with water to a level approximately 4 feet above the graphite blocks, but 1

below the top of the PCRV liner. Sultable controls will be implemented to prevent water flom splashing or the water level from approaching the top of the exposed PCRV liner. These controls are necessary to prevent water Dom entering the region between the PCRV liner and the concrete and allow potential contamination of the concrete. The Work Platform will have been installed on the ledge at the bottom of the hex opening into the PCRV.

The Work Platform will be installed with the capability of rotating to provide access to_ all areas of the core.- It will-have three access openings to allow insertion and removal of tools and components, and to pcrmit up to three operations to proceed in parallel._ A floor will be' installed between the platform and the walls of the PCRV at the level of the Work Platform. There will be three jib cranes installed on the refueling door level to service the access openings in the platform. The Reactor lluilding crane will also be avadable to service the platform and the remainder of the-refueling Door area. A ventilation system will be installed to provide control for airborne contamination and tritium, Air will be drawn from the refueling Door to the Work Platform, down:through the access _ openings in the platform, and then exhausted to the Radioactive Gas Waste System (System 63) for sampling and the

- Reactor 13uilding Ventilation (exhaust) System _(System 73) for discharge. The 13-3 w -^

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2 Attachment to P-91423 December 6,1991 airnow from uncontaminated areas to contaminated areas througl. the work platform will minimite personnel exposure to airborne and surface contamination.

The area on the Work Platform will be quite large, approximately 43 feet across the corners of the hexagonal opening. This will provide the capability to move personnel on the Work Platform to a considerable distance away from an operation when a signincant radiation field is enco,mtered.

IL llMicl10LCalics.EtQtdorJ]l01.llemoval Scqqcots The sequence of operations for removal of tl e llastelloy can hex reflector block can be broken into eight distinct steps:

1.

Removal 5.

Unloading 2.

Staging 6.

Dewatering

3. l.cading 7.

Drying 4.

Transfer 8.

Packaging Although only the removal of the llastelloy can hex reDector blocks is discussed in the following paragraphs, the remaining graphite blocks will be removed in a similar manner. Details of the eight steps are provided below.

(1) Removal: The blocks will be lifted from their posi ion in the PCRV core t

area, and placed in an intermediate staging area that is below the surface of the water (see Figures 13-5 and 13-6) using manually operated long handled tools (Lilt's) attached to an overhead crane that is operated by personnel on the Work Platform. The workers will be working through one of the three access openings in the rotary Work Platform installed over the flood:d PCRV.

The tool for handling the llastelloy can hex renector block will be an expanding collet type similar to that presently used in the I uel llandling Machine. The end of the tool will be inserted into the reverse counterbored hole in the top of the block with the end of the tool retracted. The end if the tool will then be expanded in the l' rt;er diameter in the lower portion of the a

hole and the block will t.e lifted utilizing an overhead crane, J

(2) hgist; After removal from the PCRV core area and.hile still submerged, the blocks will be lifted and placed in an intermediate stand 13-4 i

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attached to the work platform (See Figures 13-5 and 13 6), that will temporarily store the block safely below the surface of the water.

l (3) L~m A shielding bell with integral grappling tool will then be lowered into position on guide pins as shown in Figure 13 6. A grappling tool j

will be lowered from the inside of the shielding bell and grapple and lift the block into the shielding bell, The shielding bell guide pins and the storage stand will provide the necessary alignment for grappling The actual raising-of the block will be accomplished in a few minutes.

After the block has been loaded into the bell and the shield!ng bell lifted to just above the floor of the rotary Werk Platform, a catch pan with absorbent material (see Figure 13 7) will be installed under the shielding bell to contain possible drippings of contaminated water during transport to the dryer / shipping

.j liner. The catch pan will be strong enough to retain the block in the shielding bell in the unlikely event that the grappling mechanism should fail. The catch pan will also provide limited shielding for the bottom of the shield bell.

j llowever, this shielding will not be sufficient to fully shield and protect the workers on the platform from the indirect scattering that will occur out the bottom of the shielding bell, Therefore, during hiading operr.tions radiation levels in the immediate vicinity of the shielding bell will be closely monitored and personnel access to the affected area will be limited by administrative j.

_ procedural controls.

-l The expected dose rates on the Work Platform, both with and without the shielding bell are shown in Figure 13 8. As shown, the radiation field from a single unshielded block containing 72 Hastelloy cans is estimated to be 16 R/hr at one meter. This conservative estimate represents a point source with

.j 12 Curies of Co-60 and neglects shielding of the llastelloy can afforded by the-graphite or distribution of the source in the graphite. Shielding the graphite block with the equivalent of four inches oflead reduces the radiation levels to 0,05 R/hr at three feet, as shawn in Figure 13-8; (4) Itanskn As the shielding bell is moved (using the Reactor fluilding crane or the jib cranes) from the PCRV to the dewatering device and to the dryer, nonessential personnel will be required to stay clear the area to create a clear path for movement of the load.

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i Attachment to P-91423 December 6,1991 (5) UnloadiDgi Unloading of the shielding bell into the dewatei:ng device will be accomplished by removing the catch pan and lowering the bell oc th, opening in the dewatering device. An alignment fixture will be used a:

necessary to lower the block into the dewatering device.

- (6) Ikraltriog; The graphite blocks contain blind holes that will have collected a small amount of water while submerged in the PCRV, and therefore it is necessary to plaec the blocks in a dewatering device to drain the water from these blind holes. A schematic of the dewatering device is shown in Figure 13-9. After 15

• blocks are unloaded from the shielding bell into the dewatering device, the 11ocks will be rotated (tipped) approximately 90 degrees. The water that is drained from the graphite blocks will drain from the dewatering device back into the PCRY. The blocks will then be moved from the dewat-ing device into the shielding bell, the catch pan installed, and then moved to the block dryer.

(7) DIylog Dryer units will be set up on the refueling floor. A ': hematic of the block dryer arrangement and its loading and unloading positions are shown in Figure 1310. The shielding bell will be positioned on alignment pins over the dryer, the catch pan removed, and the block lowered into the dryer. Pneumatic push rods will progressively push the blocks through the dryer. Airyill be drawn into the dryer through spring loaded louvers near the

_ block exit end and exhausted near the block entry end by use of temporary ducting to the Radioactive Gas Waste System for sampling and to the Reactor Building Ventilation System for discharge, it is not planned to heat the air for drying.

(8) Eadarine: The flastelloy can hex reflector bhxks will be re loaded into the shielding bell from the dryer and transferred to the packaging area. The llistelloy can hex reflector block will be placed into a shielded shipping

- container (see Figure 1311), using an alignment fixture as necessary to ensure plae, ment for efficient use of available space. - After the shipping container is filled, the top will be installed.

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RADIOLOGICAL CONSIDERATIONS A.

Oscupational Radhlion Expolyics The occupational radiation exposures that will result from the removal of the llastelloy can hex renector blocks have been estimated and are presented in Figure 3.5-2 of the Cost Estimate (WBS No. 2.3.3.7). The occupational radiation exposure estimate for the removal of one llastelloy can hex rencetor block is 0.104 per. son-Rem and is shown in Table 131. Multiplying by 272 to account for all llastelloy can hex reflector blocks yields the 28.36 person rem shown in Table 3.5 2 of the Cost Estimate (attached to this response). Procedures will be implemented for the removal activities described above to maintain occupational radiation exposures within regulatory limits and as low as is reasonably achievable (ALARA), consistent with 10 CFR 20 and Regulatory Guides 8.8 and 8,10.

B.

Ofsilc.f,3posuIps From Accidgnis Section 3.4.5 of the PDP analyzed a postulated heavy load drop accident. As described in Section 3.4.5 of the PDP, it was assumed that a container with a single unsectioned large side reflector block (with a total releasable activity of 1477 Curies) falls 100 feet to the level of the truck loading bay. The cantainer spills its entire i

contents on the truck bay floor. The whole body and lung doses to an individual standing at a point on the EPZ 100 meters from the Reactor Building were calculated to be 4.66 mrem and 133 mrem, respectively.

In analyzing this accident, atmospherie dispersion factors were calculated using the guidelines provided in Regulatory Guide 1.145, "Atmospnerie Dispersion Models for Potential Accident Consequences Assessments at Nuclear Power Plants". NUREG-0172 dose conversion factors were also used in the calculations. This analysis determined that the radiation exposure to the general public as a result of a heavy load drop is very low. The radiological consequences from the postulated accident scenario are well within the 25 Rem whole body dose and 300 Rem to any specinc organ guidelines established in 10 CFR 100. The radiological consequences are also a small fraction of the one Rem whole body dose and Ove Rem to any specific organ guidelines cited in the EPA Protective Action Guidelines.

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GRAPHITE I)10CK TYPE CduiCMllCm 1.

Hex Reflector blocks without 4.43 Hastelloy Cans (bottom reflector graphite blocks) 2.

Large Permanent Side Reflectors 1477 3.

Side snacer Blocks without Boronated Pins 16.1 4.

Boronated Pins 0.12 5.

Hastelloy Cans 0.19 Therefore, the maximum Curie content of a hexagonal block with Hastelloy cans will be:

(72 Hastelloy cans @ 0.19 Curies /can) + (4.43 Curies) = 18.1 Curies Although there are 13.7 Curies of activated material in the Hastelicy cans, it is contained in the cans Ar.d is e readily disp;rsible. Therefore, the only radioactivity that is available for ri 2f a a hexagonal reflector block with Hastelloy cans

- during an accident is Fo 4.<-

?.mes in the hexagonal reflector block. This small Curie content is considerably less than and is bounded by the 1477 Curies that was

. assumed to be involved in the heavy load drop accident described in Section 3.4.5 of the PDP. Furthermore, the current radioactive waste packagirg plan calls for loading 44 hexagonal reflector blocks into a container that will be transported via a shielded cask. As such, the dispersible Curie contents of the shipping container (44-

' blocks x 4.43 Cuties = 194.9 Curies) will also be bounded by the 1477 Curies a',sumed in the PDP Section 3.4.5 accident scenario.

l In addition, the Decommissioning Technical Specifications establish requirements for the integrity of the Reactor Building and operation of the Reactor Building ventilation i

exhau:. system ti.S will ensure that the offsite doses under normal and abnormal conditions during graphite handling activities are well below 10 CFR 100 guidelines.

13-8

Attachment to P-91423 December 6,1991 Therefore, it can be concluded that the radiological consequences from any postulated handling accident involving the hexagonal reflector blocks with llastelloy cans would also be a small fraction of the 10 CFR 100 limits and would be bounded by the consequences predicted for the heavy load drop accident of PDP Section 3.4.5.

C.

Radinactive hte Generated A tabular list ng of the estimated radioactive waste disposal volume for each WilS i

was providw n Figure 3.2-1 of the Decommissioning Cost Estimate, provided to the NRC on lane 6,1991 (P 91148). The estimated dispasal volume for the hex reilector blocks with Hastelloy cans is 816 cubic feet and the burial classincation is

" Class C".

The radioactive waste will be classified in accordance with an established waste classification compliance program, and will be handled in accordance with the radioactive material controls discussed in Section 3.2.6 of the PDP. The packaging

elected for disposal of the hexagonal reflector blocks will comply with the equirements specified by 49 CFR,10 CFR 71, and the disposal site criteria. The radioactive waste processing, packaging, and shipping activities will be per fo med in accordance wit' wriiten procedures as discussed in Section ~).3.3 of the PDP.

IV.

SAFETY ANALYSIS CONCLUSIONS Du6g the hexagonal renector block removal activities, the radiological hazards will be a r.itored and evaluated on a routine basis. All work activities associated with wval of the graphite blocks will incorporate effective radiological controls to r

n exposures ALARA. All workers will be provided instructions in radiation orcWtion concep's commensurate with the radiological hazards that they will encountez during the removal of the graphite blocks, including instructions to the workers concerninF actions required during unusual conditions.

It is estimated that the removal of the Hastelloy can hex reDector blocks will result in a total occupational radiological exposure of 28.36 person Rem. Furthermore, postulated accidents involving the llastelloy can hex reflector blocks will result in offsite radiological consequences that are a small fraction of the guidelines established in both 10CFR100 and EPA Protective Action Guidelines, it is therefore concluded that the activities associated with the removal and disposal of the Hastelloy can hex 13-9

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Attachment to P-91423 December 6,1991 teflector blocks will not pose an undue risk to the health and safety of the general public nor to the decommissioning workers,

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t Attachment to P 91423 December 6,1991 TAllLE 13-1 OCCUPATIONAL ltADIATION EXPOSUllE ESTlhlATE FOR IIASTELLOY CA.N IIEX REFLECTOlt IlLOCK REhlOVAL NO.OF FIELD PERSON FERSON Qlt!,'ATION EQEhCES nL&B

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2.3.1 PCRV INfTI A1. PREPARATION / DIS ASSEMB LY l-I 2.3.1J Modify Main Crem 682 341 0.1 0.03 1

2.3.1.7 Desamance PCRV Teodose 25.590 12.795 0.1 1.28 i

2.3.1.8 Roanove Core Dersents with Fuel Headig Machine 2.3.1.8.1 PCRV Regiou C_cesarelat Devices 2.208 1.104 1.7 1.88 2.3.1.8.2_ Receove Metal Clad & Control Rod Reflector B!ccks 12.267 6.134 0.4 2.45 2.3.1.9-He Purir.ation Compoeem Wells 2.013 1.007 1.0 1.00 2.3.1-SUBTOTAL 21.381 6.64 2.3.1 HP& QA COVER 4GE(f f %)

0.352 0.73

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2.3.1-l7DTAL 23.733 7.37 2.3.2 SillELDED ACCESS TO PCRV 2.3.2.3 hel FCRV Conting Tubes & Teodos Conduits

,4.110 2.055 1.0 2.06 2.3.2.4 Center Access Penetration 1.190 595 J1 0.65 2.3.2.5 PCRV Shielded Weser Sywue 4.980 2.490 1.0 2.49 2.3.26 Am%rw Caesar.1-Contrrd 3.633 1.816 0.3 0.54 2.3.2.7 Cut Core Top Head 21,660 _10.830 1.1 11.91 2.3.2.8 Flood NRV 180 90 0.6 0.05 2.3.2.9 - $ PCRV Cavity Shieldal Work Platform 1.325 663 1.0 0.66 2.3.2 ^

SUBTOTAL -

15.539 18.36 2.3.2 HP A G4 COVETt4SE(f f %)

2.039 1.02 2.3.2 TOTALE 20.578

- 20.38' 2.3.3 DISMANTLE PCRV CORE 2.3.3.3 Defeating samenes 16.683 8.342 1.7 14.18

~

2 3.3.4 and Parm Hex Reflector Blocks 16.683 8.341 3.7 30.86 2.3.3.5 tarse side Refiscaer Blocks -

27.782 13.891 3.6 50.00 2.3.3.6

- Boramend Spacer Elorecean 16.683

- 8.342 -

1.9 15.85 2.3.3.7 Hassalloy Cam Hen Reflecto_r Riocks 8.341 4.170 6.8 28.36 2.3.3.8 Core suppert alocks and Posts 2.780 1.390 1.8 2.50 2.3.3 SUSTD7AL 2.3.3 RP A G4 COVERAGE (71 %)

44.476 141.75 1

4.892 15.59 2.3.3 TOTAL -

49.368 157.34 i-l'

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Attachment to P-91423 December 6,1991 NRC Ouellion.NL.LLISection 2.3.3.9: core Battrilemoral)

_1he Jtdy 1,1991 trvision selects a thennal cutting rnethodfor core barrel retnaval.

Provide procedurrs and related sqfety analysis for sninhnitation of occupational e,rposure so personnel. "

t ESC Resnonse:

A detailed response to this question will be provided by January 17, 1992.

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'We smderstand that you plan to conduct testing that demonstrates the suita'rtity of sectioning the concrete with a diamond wire through the tendon tubes.

Please provide your test results. '

PSC RIiSPONSit 1.

INTRODUCTION in the initial submittal of the Proposed Decommissioning Plan (PDP), PSC madt the following statement:

"The inner row of vertical PCRV tendon tubes are suitably positioned for removing the beltline activated concrete, llowever, in the event that t tese tendon tubes prove to be unsuitable for the initiation of diamond wire < uts, new vertical holes will be cored drilled."

' At the time 'the initial PDP was written there was a question whether the dian:ond wire cut could be started from a tendon tube because of the sharp transition that was made from the vertical tendon tube into a 1/2" kerf. To further understand this concern, refer to Figure 16-1 which shows the construction of the diamond cu-ting wire. It was considered possible that a spring or diamond embedded bead Iouk get caught on the sharp, steel-edged corner, preventing an effective sawing actio.1 to round the corner.

It is standard practice with diamond wire cutting to manually feed the wire until the corners of the rraterial being cut are rounded and the wire feeds smoothly without

" humps end jerks". Once the rounding is completed, the diamond vcire is then po ver fed. This process has proven successful on numerous applications with square corners on concrete.

II.

PCRV MOCK-UP TESTING To resolve these concerns, a' mock-up was constructed to perform test cuts to -

demonstrate the suitability of sectioning the concrete with a diamond wire through ;he tendon tubes.

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? Attachment te P-91423

December 6,1991:

A.

Mock-urLTmLQbjettlyn The objectives of the mock up tests were as follows:

-1.

Test the suitability of initiating a cut through _the tendon tubes.

2.

Test methods of cutting the liner plate, insulation and cover plates.

3.

Demonstrate the diamond wire method to the Decommissioning Project Team.

B.

, Description of thgfCRV Mgh-ity

. A test block was constructed to duplicate a section of the verucal wall of the FSV PCRV. Steel reinforcement, consisting of number 8,9 ano 18S bars, was included in locations in accordance with the PCRV construction drawings. Iforizontal a'id vertical post tensioning tendon tubes were also included in the test block in order to provide a cutting sequence similar to actual site work.

A thermal barrier and liner plate were placed on the front side of the mock up. The thermal barrier was layered with a 1/4" steel cover plate,21/2" Kaowool insulation, another 1/4" steel cover plate and another 21/2" Kaowool insulatiou covering a 3/4" steel liner plate, On the concrete side of the 3/4" liner plate,1" square tubes and concrete embedments were welded to the liner plate per the PCliV construction drawings. Figure 16 2 shows the mock-up prior to filling with concrete aggregate, including reinforcements, mt:d lugs, tendon conduits, cooling tubes, liner plate, insulation and cover plates.

The conente was mixed following the " Job-Mixed Concrete for PCRV", General

- Atomic Co. Specification 11-R-17, Appendix A, that was used for the original PCRV construction and concrete strength was approximately 6000 psi. The cerse and fine aggregate consisted of crushed porphyritic andesite shipped from Colorado to a local

' Cincinnati concrete batch plant. Aggregate grading was in general conformance to

" Aggregate for PCRV Concrete", General Atomic Co. Specification Il-R 13, gradation B,C, and E. Figure 16 3 shows the completed mock-vb prior to test cuts.

i 142

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Attachment to P 91423-December 6,1991 C.

Descrintion of Diamond Wire Cutting System Figure 16-4 shows the dhmond wire system used during the mock up test and similar to that which will be used to section the PCRV concrete. The diam'ond wire system consists of a diamond matrix wire inade to length for each individual cut and a hydrrulic drive system._ Diamond wire is routed to envelope the cut area. The wire

- is then guided back to a drive wheel located on the drive system. The wheel rotnes and pulls the wire through the cut area. Figure 16-1 shows a detail drawing of a segment of the diamond wire.

D.

Dc;cription of thc.fullkTcM_ano the Cut Sequsacs Initial activity on the mock-up was the thermal cutting of various combinations of the

- insulation cover plate, Insulation and liner plate. The methods tested included Oxy-Acetylene, Plasma and Oxy-Lance, with the Oxy Lant.e being the most successful.

Diamond wire test cuts were ;>crformed on the following configurations:

(1)

Concrete and tendon tubes (2)

Concrete, tendon tubes ~, and liner plate (3)

Concrete, liaer plate, insulation and covef plate

' Figure 16-5, Proposed Cut Locations, depicts the planned cuts during the mock-up test. Not all cuts originally planned were performed because satisfactory results precluded the r.ced for further development. Numerous cuts were made, however,

- the test cut <. of interert are Cut Numbers 3,7, and 9. The following is a description of these cats:

Cut

-Sinther.

Drictintion -

3'

-(See Figure 16-6) This cut simulates a vertical radial cut of the PCRV belt line region. The wire was routed down the vertical tendon tube, through a 1/2" horizontal kerf, and then up the face of the PCRV liner.

This routing required the wire cut to start on two square corners, one with a tendon tube, and the other with the liner plate.

Tne 3/8" diamond wire successfully cut through the tendon tube, steel reinforcement plate, cooling tube and 3/4" liner plate. The insulation L

16 3 m

Attachment to P-91423 December 6,' 1991 and insulation cover plates were removed prior to this cut by thermal cutting methods.

7 (See Figure 16-6) This cut simulates a vertical back cut of the PCRV beltline region. The 3/8" diamond wire was routed down a vertical tendon tube, through 1/2" horizontal kerf, and up a vertical tendon tube.

The start for this cut included two right angle steel tendon tube corners.

This cut successfully demonstrated that a vertical cut could be made between vertical tendon tubes.

9 (See Figute 16-7) This cut simulates a vertical radial cut of the PCRV beltline region. The cut was initiated from kerf to kerf rather than from a tendon tube. Also, the insulation and cover plate were left in place.

This cut demonstrated that a vertical radial cut could be made without removing the insulation and insulation cover plates.

111.

CONCRETE CUTTING TEST CONCLUSIONS A.

Demonstration of Test Objectives The tests successfully demonstrated the following te.st objectives:

(1)

Ability to start a diamond wire cut from the tendon tubes thereby demonstrating the suitability of sectioning the concrete with a diamond wire through the tendon tubes.

-(2)-

Ability of. the diamond wire.to cut through the liner plate, insulation and insulation c.over plate and connrmed the use of this method for those operations ' previously identified in the Fort St. Vrain Proposed Decommisrioning Plan. Use of this cutting technique was demonstrated to be an implovemen ovci thermal cutting techniques in that it minimizes the spread of contamination and requirements for permnnel protection.

(3)

Succestful demonstration for Fort St. Vrain Decommissioning

' Project personnel to further their understanding of the diamond wire cutting

-process.

c 16-4 1

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EAttachment to P-91423 December 6,1991 B.

Other lessons Leargd

_(1). Contamination Control: As witnessed during the demonstration, the control of the cutting slurry will be the basis for the radiological controls associated with diamond wire cutting. The demonstration was also beneficial

~ for the future planning and design of engineered contamination control containments. For the FSV Decommissioning, approximately 5000 f!' of the 2

total 16000 ft of cut surface will be through activated concrete. The slurry from these cuts will contain activated concrete particles which will become mobile via the water used as a epolant or lubricant for the cu@, process.

The coolant water is independent of the PCRV Shield Water System. Concrete cutting of the PCRV sidewalls will be performed after the PCRV has been drained, and the cutting operations will not be performed underwater.

Airborne contamination can be introduced as a liquid mist as the wire exits the cut and at lo.;ations where the wet wire turns around a sheave or as airborne sohd if the slurry is allowed to dry.

Containment of the slurry will be by means of plugging openings or embedded pipe within the concrete, constructing a water collection system to collect (l'c cutting slurry, decant the slurry and recycling.the water. Airborne and loose

-surface contamination control will be achieved by containing the wire path and drive unit in a containment tent served by llEPA ' ventilation.-

(2). Occunational Exposure Controh Personnel exposure to radiation will be minimized by controlling access to areas with elevated dose rates, by limiting the stay time in these areas, and by the use of shielding. The equipment and controls for diamond wire cutting allow them to be positioned in areas that will-maintain worker dose ALARA, The process requires only a small percentage -

of the time near the source. = Specifically, this tinie is associaied with the g

installation of turning sheaves and threading the wire for the belt line cuts.

The source for this activity is the activated concrete, liner plate, insulation and insulation cover plates.

16-5

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• y Attachment to P 91423 December 6,1991 NRC Ouestion No. 38 (Section 3.3.2.3: Liquid Wastes)

• PSC esthnates that about 500 cutles of tritium uvuld be released to the PCRV shleid unterfrom the 100,000 curies of tritium in the graphite (one half of one percent).

PSC's April 26th response and July 1st revision to the Decornmissioning Plan state that the estimated trkh,m release from graphite blocks to PCRV shleid nuter is

' based on data.

• The source of this data must be ptvsided and Hs accuracy and applicabilityjustified. Include in these evaluations the structure ofgraphite blocks, e.g., unciad, and material composition.

I PSC's July 10,1991 Supplement to Environmental Repon, page 412 states thatfeed t

and bleed operations uvuld be used to dilute 535 curies of trHium to one half qf10 CFR Part 20 thnits using 2000 gallons of unter per minutefor about one month.

Jhls is baaed on the data esuluation above. Any change in this evaluation must be rc:llected in the release plan. Provide verification that the planned releases are consistent with AIARA principles and Environmental Protection Regulations related to 10 CFR Pan 51. Discuss other potential release options considercd Also evahtate the potential contamination oflarge svhanes ofconcrete with tritium from unter leaks in PCRVpenetrations or liner,from drying of urt graphite blocks, from water spills during cask and radioactive material handling andfrom evapora?lan qf unterfrom open PCRVpool surface. As discussed with your stqlf, trillated water of hydration in the concrete of a reactor room at a 5 MW, heasy unter, rescatch reactor prevented hs release for unrestricted use qfter extensive decomamination ejbns (NUREGICR 3336

• Summary Report Ames Loboratory Research Reactor *).

• L'SC.KUDM A detailed response to this qmtion will be provided by December 20,1991.

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