ML20087D039

Provides Response to NRC Request for Addl Info Re Proposed Decommissioning Plan

ML20087D039
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/09/1992
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-92014, NUDOCS 9201150117
Download: ML20087D039 (57)

Text

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Public Service c e, .< c.--

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• P O BOX 040 DENVER CO 00201 0840 January 9,1992 A, Clegg Crawford Fort St. Vrain gfyg,,

' Unit No.1 P-92014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 ATTN: Dr. Seymour H. Weiss, Director Non-Power Reactor, Decommissioning and

- Environmental Project Directorate Docket No. 50-267

SUBJECT:

PSC. RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION ON Tile FORT ST. VRAIN PROPOSED DECOMMISSIONING PLAN I

REFERENCES:

(See Attached) l

Dear Dr. Weiss:

The purpose of this letter is to respond to the NRC's Request for Additional i Information (RAI),-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. Vrain Nuclear Generating Station and a PSC response to the previous NRC RAI (dated February 8,1991), that were i

submitted to the NRC in References 2 and 3. As committed in Reference 4, this response provides specific PSC responses to NRC Questions No. I1 (Water Cleanup and Clarification System) and No. 38 (Liquid Wastes).

If you have any questions related to the contents of this letter, please contact Mr. M.

H. Holmes at (303) 430-6960.

Very truly yours, 8 (

.A. Clegg Crawford Vice President .r Nuclear Operations ,

b f 9201150117 920109

-PDR ADOCK 03000267 [

.P- PDR t .-

P-92014 January 9,1992 Page 2 ACC:CRB/cb Attachment ec: Regional Administrator, Region IV Mr. J.B. 8aird Senior Resident inspector

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Fort St. Vrain Mr. Robert M. Quillin, Director Radiation Conitol Division Colorado Department of IIcalth 4210 East lith Avenue

. Denver, CO 80220 f.

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P-92014 January 9,1992 Page 3 REFERENCES (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 26,195". (P-91118)

(4) PSC letter, Crawford to Weiss, dated Deceniber 6,1991 (P-91423)

ATI'ACIIMFNI'TO P-92014 PSC -RESPONSE TO NRC QUESTIONS NO.11 & 38 FROM THE NRC RAI DATED AUGUST 30,1991 1

l Attachment to P-92014 January 9,1992 NRC Ouestion 11 (Section 2.3.3.6.2: Water Cleanup and Clarification System)

' Provide analysis ofradiologicalconsequences ofsystem operation. Funher, provide a safety analysis for potential accident scenarios with regard to occupational and public exposure. Include analysis of radioactive material (including tritium) that is released to this system. '

PSC Resoonse:

1. SYSTF31 DESIGN CONSIDERATIONS:

A. Background Information; Decommissioning of Fort St. Vrain requires the dismantlement of the PCRV and PCRV internal components. The sequence of PCRV dismantlement evolutions are described in Sections 2.3.3.7 through 2.3.3.12 of the Fort St. Vrain Proposed Decommissioning Plan [1]. PCRV dismantlement operations will be performed underwater in order to take advantage of the water shielding to maintain radiation exposures ALARA. After flooding the PCRV, dismantlement of the radioactive portions of the PCRV and removal of the PCRV internal components will commence with the PCRV top head removal and then progress downward. This writeup provides a discussion of the expected conditions within the Gooded PCRV, design considerations, and description of components and operations of the PCRV Shield Water System (PCRV Water Cleanup and Clarifiestion System), as well as PSC's responses to the specific NRC concerns identified above.

B. Exoected Conditions Within the Flooded PCILV_;

1. Radionuclides: The radionuclides of concern that will be encountered during dismantlement operations inside the PCRV have been previously identified in the activa; ion analysis provided as Appendix 11 of the PDP. A summary of these radionuclides is provided in PDP Table 3.1-2. A fraction of each of these radionuclides is expected to leach into the water from the graphite when the PCRV is flooded.

The principal radionuclides of concern are Co-60, Fe-55 and Cs-137. Of these, Co-60 will provide the majority of the whole body exposure to occupational workers as Il-1 1

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i Attachment to P-92014 January 9,1992 a result of dismantlement operations. These radionuclides will appear in particulate and lonic form, and the PCRV Shield Water System will be designed to remove the principal radionuclides.

Although not a major contributor to whole body exposures, the other major radionuclide of concern is tritium. _ Since the tritium cannot be removed by processing through filters or demineralizers, it will be processed and released using

-liquid effluent discharge operations in accordance with the revised 10 CFR 20 limits, The maximum tritium concentration shall not exceed the limit specified in the Decommissioning Technical Specifications. [2]

2. Particulates: During PCRV dismantlement evolutions, debris will be generated from handling graphite blocks, concrete cutting operations, insulation, and dross from underwater cutting operations. Relatively large particles of debris (%

to 1 inch in diameter) are expected to be Eencrated from the various cutting methods to be employed during PCRV dismantlement operations, including diamond wire cutting (PCRV top head), oxy acetylene cutting, thermitic rod cutting, and underwater plasma arc cutting. This debris will settle downward in the PCRV water and will be removed by the PCRV Shield Water System. Suitable provisions will be included in the system design to collect this debris and prevent it from damaging system components.

Some graphite dust is expected to become waterborne after the PCRV is flooded. The need to filter this graphite has been incorporated into the design and filter sizing of the PCRV Shield Water System. The possibility of breakdown of the Kaowool insulation (described in Section 2.2.2 and Figure 2.2-8 of the PDP), attached to the PCRV liner immediately outboard of the core barrel, has also been considered.

Ilowever, based on information from the manufacturer, this insulation is not expected to break down when immersed in water and therefore will not be a factor in the design of the system filtration trains.

C. Design ConsideMt2D3 The primary function of the system is to provide water shielding to minimize personnel exposure during dismantlement operations internal to the PCRV. The system will also be designed to provide a means to meet 10 CFR 20 discharge limits for the radionuclides identified above and ensure compliance with 10 CFR 50 11-2

Attachment to P-92014 January 9,1992 Appendix 1 guidance for radioactive liquid waste discharges to unrestricted areas.

Specifically, the system design will provide:

(1). an acceptable method to reduce tritium inventory by liquid efauent dischnge operations.

(2) an acceptable radioactive ligtid waste processing path to reduce the conceatrations of fission and activation prmlucts for discharge to unrestricted areas, as well as co;.<rol radionuclide concentrations in the PCRV water inventory to maintain occupational exposures in the work area ALARA.

In addition to these regulatory criteria, the system will also be designed to meet the following non-regulatory considerations:

(1)' maintain acceptable water clarity to conduct underwater dismantlement operations.

(2) minimize corrosion and biological fouling by suitable chemistry control.

(3) provide a means of initial fill of the PCRV, as well as the ability for mr.keup with demineralized water to compensate for system losses due to effluent discharges.

The recommendations of Regulatory Guide 1,143 " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light Water-Cooled Nuclear Power Plants" [3] will be used in system design, This s- tm will oc designed to maintain occupational radiation exposures within 10 CFr regulatory limits and as low as is reasonably achievable (ALARA). The

-r endations of Regulatory Guide 8 8 "Information Relevant to Ensuring That

. 4tional Radiation Exposures at Nuclear Power Stations Will 13e ALARA" [4]-

s also be incorporated (to the dege: applicable) into system design, D. Dgsjgn.}ixperience from Similar Operations The design of the PCRV Shield Water System incorporates experience gained by the Westinghouse team from the design and maintenance of PWR nuclear power plants, and from the design of similar systems used during the post accident cleanup efforts at Three Mile Island. - The system design has been reviewed with engineers and 11-3

_. . . -- = _ _ . - .

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Attachment to P-92014 January 9,1992 chemists who were involved with the design and operation of the Three Mile Island cleanup system and other similar water systems to insure incorporation of " lessons learned" into the Fort St. Vrain design. Experienced gained in the following areas has been incorporated into the design of the PCRV Shield Water System:

o Syste n turnover rate o- Chrmistry and biofouling controls o Maintenance requirements o- Water clarity to support underwater operations TMI experience was reviewed to determine a suitable system turnover to remove particulate debris from underwater dismantlement activities and to reduce radionuclide concentrations. A system turnover rate of at least 2.5 times per day was determined to be acceptable at TMI and this experience will be factored in the PCRV Shield Water fiystem design. Similarly, consistent with TMI experience and in consideration of the PCRV carbon steelliner, chemistry control will maintain the pil of the shield wate. between 9.0 and 10.0. A hydrogen peroxide system was used at TMI to mimmize biofouling. -Results of this experience indicated that maintaining a concentration of hydrogen peroxide would adequately minimize biofouling in the open air, atmospheric shield water system.

An evaluation was performed to determine the system filtration and water clarity requ rements necessary to conduct underwater operations. Experience demonstrated the us.cfulness of a surface skimmer system with a capability to perform underwater vacuuming which will be incorporated into system design.

' Maintenance requirements, including approximate numbers of filter and demineralizer

- replacements and associated occupational radiation exposure, were also evaluated in comparison with relevant experience.

II. GENERAL SYSTFAI DESIGN INFORMATION:

The PCRV Shield Water System is shown in Figure 11-1 (Westinghouse Flow L Diagram 201lE26). The system will consist of two parallel trains of equipment, each

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sized for 500 gpm (or 50%) of the total flow. This total design flow rate (1000 gpm) will provide a turnover rate of approximately five PCRV volumes per day. Based on discussions with personnel involved in the TMI cleanup, this rate is considered l

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Attachment to P-92014 January 9 - 1992 adequate. The system will be designed to allow a complete train or individual components to be removed from service for preventive or corrective maintenance.

. Provisions will be' included for the aduidon of another complete train if required.

Tne trains are cross-connected to permit the pumps and filters to be used interchangeably between the two trains.

Maximum Dexibility.will be designed into the system to minimize the impact of individual comp aent failure on system availability, Sufficient valves, bypasses and interconnecting piping will be utilized to allow continued system operation during scheduled maintenance or in the event of a component failure. Remote-indicating radiation detectors will be used to monitor dose rates on components in high radiation areas, such as at strainers, Olters and demineralizers.

A. Filtration Trains:

The purpose of the system Oltration trains will be to maintain PCRV water clarity by removing suspended solids and particulate matter, and to reduce concentratiores of suspended radioactive particulates. In order to maintain optimum water clarity, suction of the PCRV Water Shield System will be taken from the bottom of the PCRV and clarified water will be returned to the top of the vessel. The system will have two filtration trains consisting of the following components:

1. Clarifying Pump Suclinn Strainer: A strainer will be installed in the suction line of each clarifying pump to prevent equipment damage due to large particulate debris. The strainers will be duplex type, and provisions for shielding and radiation monitoring of these filters will be included in the design,

2. Clarifying Pump: Each trcin will have one clarifying pump. The pumps will be horizontal, centrifugal process pumps. Each pump tvill have a capacity of 500 gpm through the associated train of equipment and will return the clarified water to the PCRV.

3. Filter Trains: Each filter train will consist of two filters, with the filters mounted in a series arrangement. Bypasses will be provided to allow each filter to be operated individually or in series with other filters. Five micron filters will be used to remove the graphite particulate expected while Il-5

i Attachment to P-92014 January 9,1992 the PCRV is flooded. The series arrangement for filters will be specified to remove 98% of all particles larger than 5 micron at a maximum differentui pressure of 15 psid. Provisions will be included for shielding and radiation monitoring.

B. Demineraliter Train:

The system wil; also be equipped with a sidestream demineralizer train. The purpose of the demineralizer train will be to reduce concentrations of dissolved radionuclides (speci6cally Co 60, Fe-55 and Cs-137) to levels that will allow discharge to unrestricted areas, as well as reduce concentrations in the PCRV to minimize radiation expore to occupationally exposed personnel. The demineralizer train will consist of the fo!!owing additional components:

1. Demineraliznn Two demineralizers capable of removing the principal radionuclides of concern (Co", Fe", Cs+) will be provided in parallel. The demineralizer train will be sized for a minimum of 10% of the total process flow (100 gpm) and vill remove dissolved radionuclides from the PCRV water inventory. The demineralizer will be enclosed in a shielded housing and will include provisions for remote radiation monitoring. One demineralizar is planned to be in serv!ce at all times.

2. Resin Fines Filter: One cage-type filter will be provided to prevent the loss of resin Gnes from the demineralizer and possible discharge into the PCRV. This filter will be designed for a minimum capacity of 100 gpm and to retain 98% of all particles greater than 5 micron at 15 psid.

3. Ikmiperalizer Pumns: A demineralizer pump will Le provided to take suction fro the demineralizer surge tank and provide the necessary flow through the de.nmeralizers. After ex.tbg the demineralizers and the resin fines filters, the demineralizer booster pump will return the processed water to the top of the PCRV or to the radioactive liquid waste effluent release path when efnuent discharge operations are in progress.

4. Demtr- alizer Surge Tank: A demineralizer surge tank will be provided to isolate the low pressure demineralizer sidestream from the higher pressure clarification filtration trains.

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C. ChemkalAddition Train; I The system will also include a chemical addition train. The purpose of the chemical e dition train will be to minimlic corrosion by suitable chemistry control within the .

PCRV system and to minimize biological foull'*g. The chemical addition train will s consist of the following additional component's t

1. Chemical Addllion Tanks: Two 100 gallon chemical addition tanks will be included, one foi sadium hydroxide and the other for nydrogen peroxide. The tanks will be used to add chemicals to the system for the maintenance of proper chemistry and to control biological ft .!!ing, t

2. ClicmkalAddition Pumpx Two chemical addition pumps will be included. One pump will be provided for injection of sodium hydroxide into the system to maintain chmistry control and pl{ balance, and the other puny will be provided for 10 /tlon of hydrogen Irroxide to control biologicta fouling.

D. & ler System:

Consistent with experience gained conducting underwater operations at TMI. a skimmer subsystem has been included in the system design to maintain adequate surface visibility and to reduce surface contamination. The skimmer subsystem will ,

include a duplex strainer, a skimmer pump, a filter and a floating strainer. By changing the valve positions, this systern can also be aligned to provide underwater vacuuming capability.

E. Sydttn. Controls: ,

The system incorpsrates remote manual controls on a central control panel located in a non congested area on the refueling floor. All major operanons of the system, including flow adjustments and valve positioning, will be performed from the panel.

Pumps will be controlled both locally and remotely from the central control panel.

Alarms for either high differential pressure or high radiation will notify the operator of the need to replace the filters or the demineralizer resins. PCRV water level l indication and PCRV high/ law water level alarms will be included in the design to L facilitate system operation and control, 11-7

Attachment tc P 92014 January 9,1992 P. DrticIaulc11tAhlfeImalletu The strainers, filters, and demineralliers will be shielded to reduce radiation fields in the immediate vicinity of these components during operation. The system will have an isolation valve at the outlet from the PCRV to the suction of the PCRV Shield Water System to allow isolation of the water in the PCRV if a problem should develop in the system. There will also be other valves to allow isolation of portions of the system for maintenance or repair. A connection will also be provided for an additional train, if necessary.

The major components of the system will be prefabricated on skids with drip pans to contain potential leakage, and will be installed in unoccupied areas of the Reactor Building to minimire personnel exposure. Skids that include system filter and strainers will also be shicided. The operating controls and che.nical addition skids will be located at the Refueling Iloor area The skids will be Interconnected with other skids as well as with the suction piping from the bottom of the PCRV and the return piping to the top of the PCRV. The PCRV Shield Water System will be connected to the existing Fort St. Vrain Radioactive Liquid Waste System (System

62) to permit the use of the existing efnuent discharge paths and radioactivity monitoring and controls.

III. SYSTEM OPERATION A. blitial Fill of the PCRV After system installation and check out, the first operation of the system will be during the initial fill of the PCRV. As opposed to normal system operation, the initial fill will be from :he bottom t,f the PCRV via the suction piping. The initial fill will be accomplished prior to the final cutting and removal of the PCRV top head concrete, and prior to gaining access into the PCRV internal cavity. Deminerallred water for the !nitial PCRV fill (estimated to require approximately 325,000 gallons) will be from the existing secondary water treatment system, which is described in a following section.

As the PCRV is being filled, the displaced air and gas will be passed through a portable IIEPA filter system attacheo to a refueling penetration in the PCRV head.

Using temporary ventilation ducting, the displaced air from the llEPA filter will then 11 8

Attachment to P-92014 January 9,1992 be routed to the installed Radioactive Gas Waste System (System 63) for sampling, and then to the existing Reactor liuilding Ventilation 11xhaust System (System 73)if concentrations permit direct discharge. All gaseous discharges will be in compliance with the Fort St. Vrain Offsite Dose Calculation Manual (ODCM) [5]. The PCRV will be inspected for leaks after the initial fill process is begun. Filling of the PCRV will be stopped at predetermined levels (1/4 core submergence increments) to allow tritium sampling and analysis. The fill operation is expected to take several days.

The Decommissioning Technical Speci0 cations [2] require that the PCRV water be sampled and analyzed daily for tritium concentration during the initial All of the PCRV. Sample frequency may be reduced to weekly after the tritium concentration has decreased to less than 0.1 Ci/ce. Limits have been established in the Decommissioning Technical Specifications [2] to assure that tritium activity concentrations in the PCRV Shield Water System will not exceed those postulated in the decommissioning accident analyses.

B. Normal Systrm_0Pfsttinn Oace the PCRV has been Alled, the PCRV Shield Water System lineup will be restored to take a suction from the bottom of the PCRV, with the return Dow to the top of the PCRV. The system will be opetated to establish and maintain water clarity, water chemistry and to minimize waterborne concentrations of radionuclides.

The demineralizer will be placed iruo service as required. Filters and demineralliers will be monitored for differential pressme and radiation levels to determine when replacement is required, it is expected that approximately 60 days will be available to operate the system to establish water clarity and reduce radionuelide concentrations before the PCRV top head is removed.

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1. System RecirothttjpfLlo the PCRy; The normal operational mode of the system will be with both trains processing PCRV water at a total flow rate of approximately 1000 gpm. Both clarifying pumps will take suction from the bottom of the PCRV (total flow rate - 1000 gpm) and will process the water through two parallel trains of Olters. During system operation, water flows from the Gooded PCRV throt.;h duplex strainers to the suction of the clarifying pumps. From the pumps, the water is processed through the filters before returning to the PCRV. A minimum side stream flow of 10% of the water flow rate will be processed through the I l-9

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Attachment to P-92014 i

, January 9,1992 I

deminerallrcrs for removal of radionuclides, namely Co-60, Fe 55 and Cs-137.

The filter trains will be set up in a series arrangement depending on the  :

turbidity conditions in the PCRV. During initial operation. 50 micron ,

elements will be loaded mto both filter housings in each train. As water clarity improves, filter elements and filter alignment will be changed as needed to support ongoing operating conditions. The clarified water will be returned  !

to the top of the PCRV ;hrough a sparger arrangement which will minimire surface disturbance. ,

2. System Processing Via The Demineralizer:

A minimum sidestream flow of approximately 10% of the total now (100 gpm) will be taken from the return line downstream of the filtration units and processed through demh eralizers. The flow will be adjusted as required to  :

maintain acceptable radiation levels to minimize personnel exposure. Efauent l from the deminerallrcr train can also be routed back to the system return lines for recirculation to the PCRV or, after sampling, routed to tb effluent discharge connections as described in the ~following paragraph. Sultable provisions will also be provided for additional demineralizer capacity as required,-

3. Discharce Via Radioactive Liquid Waste System (System 62):

All liquid waste from the Fort St. Vrain decommissioning will be routed through the existing radioactive liquid waste system (System 62) discharge line that was utilized during normal plant operations. Further details of this system are provided in Figure 2.2-23 of the Proposed Decommissioning Plan [1] and Section 11.1 of the Fort St. Vrain FSAR [6], Discharges w!Il also be performed in compliance with the NPDES permit in effect at that time.

4 Sampling Oncrations:

Initially, releases v'ill be batch mode releases. Prior to liquid efnuent discharge operations, representative samples obtained from the PCRV Shield Water System will be analyzed for principal radionuclides to ensure that the concentrations of radionuclides discharged to the environment do not exceed the values specified in 10 CFR 20. Samples will also be taken to verify maintenance of suitable water chemistry. Sample locations will be included (see Figure 11-1) at the suction line from the PCRV, at the outlet of the filter trains, and at the outlet of the demineralizers.

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P Attachment to P 92014 January 9,1992  !

5. Chemistry Controh in order to minimize corrosion from the carbon steel components within the PCRV, it will be necessary to maintain the pli balance of the PCRV water to a control band of between 9.0 and 10.0 by the addition of suitable caustic.

Additionally, to control biololt eal l fouling, it will be necessary to maintain a

' hydrogen peroxide concentration. A system will be provided for this purpose consisting of two 100-gallon tanks and two chemical addition pumps. The system will initially be used to add chemical during the initial fill of the i PCRV. The system will then be used to batch feed chemicals as required until tritium levels are reduced to a 10:1 whcie continuous efauent discharge operations are acceptable. To support a continuous discharge operation, it will be necessary to continuously feed chemicals.

6. System Interfaern The PCRV Shield Water System will interface with and require support from the existing site systems:

(1) Deminerallred Water System (System 31)

The PCRV Shield Water System will require a supply of demineralized water at a Cow rate of up to 50 gpm at 100 psig, Deminerallred water is required for system makeup,' replacement of water removed by einuent discharge operations, chemical' additions, and to replace evaporative losses. The demineralized water supplied to the PCRV Shield Water System must rnect -

typicalindustry standards for oxygenated, deionized wher. The demineralized water for the initial PCRV 011 and for makeup due to subsequent effluent discharge operations will be from the existing secondary water treatment system.

(2) Radioactive Liquid Waste System (System 62)

Tritium inventories will be initially controlled d subsequently reduced using effluent discharge operations. The discharge from the PCRV Shield Water System vill initially be to the existing plant liquid waste holdup- and monitoring tanks for processing and subsequent discharge. As tritium levels -

are reduced,' the discharge will be directed to the Reactor Building sump.

dischsge line.

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Attachment to P 92014 January 9,1992 (3) lilectrical Power Tie ins to the site supplied power of 480 VAC 3 phase will be required. The skids will be pre whed and transformers are provided to facilitate interconnections.

(4) Compressed Air System A source of dry compressed air at a nominal value of 9') psig is required to support system operations, particularly for dewatering the stralners, the filters and demineralizers.

(5) IIcating, Ventilation and Air Conditioning The PCitV Shield Water System will be located within the Rcactor ilullding to provide the required environmental conditions. No special or additional environmental conditions are required. The system will require a temporary connection to the Reactor Building ventilation exhaust system to accommodate the displaced air during the initial filling of the PCRV.

C. Sy31cm Maintenante Methods for handling the replacement of radioactive strainers, Olter elements and the change out of demineralizer resins will be designed for ease of replacement and will incorporate ALARA concepts, consistent with the recommendations of Regulatory Guide 8.8 [4]. These components will be shielded as necessary to minimire occupational radiation cAposures.

The system will have sufnelent interconnecting piping and isolation valves to allow repair or maintenance on a portion of the system while the remainder of the system continues in operation. In the unlikely event of a leak within the system, the entire system will be isolable from the PCRV, or that portion of the system with the leak will be isolated. The strainers, filters, and demineralizers will be designed to minimize exposure during maintenance. The strainers will be provided with inserts for case of handling during replacement. The filters will be provided with vents and drains, and filter cartridges will be removed into shleided containers to minimize exposure. Strainers, filters and demineralizers will be shielded and provided with radiation monitors within the shielding, 11-12 l

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Attadunent to P 92014 January 9,1992 D. System ikmen! 4 The system will be used to support ongoing decommissioning operations. When the system is no longer required, it will be dismantled and treated as contaminated 110P equipment and piping. Prior to dismantlement, the system will be surveyed and decontaminated or .%poscd of as radioactive waste, as necessary.

IV. RADIOLOGICAL, CONSIDERATIONS l

A. Ecleases from the_Syllcm l Tritium concentrations in the system will be controlled by liquid efnuent discharge 1 operations. (See PSC response to NRC RAI Question No. 38 attached.)

Concentrations of other radionuclides, primarily Co-60, Fe 55 and Cs 137, will be controlled and reduced by reprocessing the PCRV water through the PCRV Shield Water System or the existing Radioactive Liquid Waste System demineralliers.

There are three methods for releasing water from the system:

(1) From the discharge of the PCRV Shield Water System deminerallrcrs (primary release path);

(2) From the PCRV Shield Water System return line downstream of the -

filter trains; and (3) Directly from the PCRV to the Radioactive Liquid Waste System (with no processing) via a cross-connection from the PCRV suction line; All three paths win be capable of discharge to: (1) the existing radioactive liquid waste tanks; (2) the Reactor Building sump discharge line; or (3) a PCRV Shield Water System waste holdup tank (see sheet 2 of Figure Il 1).

Initially, the estimated tritium concentration in the PCRV will limit allowable discharges to less than 10 gpm. During this period, system output will be discharged to the existing radioactive liquid waste tanks. Samples will be taken and analyzed for tritium as well as other principal radionuclides.11ased on sample results and the limits established in the Fort St. Vrain ODCM [5], an allowahic release rate will be determined using the cooling tower blowdown now to dilute the discharge. Once concentrr.tions of tritium and other radionuclides have stabilized and are reduced to acceptable levels to allow direct discharge from the PCRV the second discharge flow Il-13

i Attachment to P-92014 January 9,1992 ,

path (to the Reactor Building sump discharge line) will be used. The third discharge path provides a path for releasing collected drains from the PCRV Shield Water System after sampling and verl0 cation of the allowable discharge rate.

B. C_ontrol of Tritium  :

Tritium levels in the PCRV will be controlled by efnuent dischar. e operations, it is estimated that 500 Curles of tritium will be released from the graphite blocks to the shield water system in the first month of operation. (See PSC's response to NRC RAI Question No. 38 for additional analysis of the amount of tritium projected to be released to the PCRV Shield Water System.) As conditions permit, the PCRV Shield Water System will discharge to the outlet piping of the Reactor Building sump pumps during the discharge operations. The controls and administrative procedures that governed liquid ef0uent releases from the radioactive liquid waste system during pknt operation will also govern the releases from the PCRV during decommissioning.

The revised 10 CFR 20 MPC limit for tritium discharge to the environment is 1 E(-3) pCi/ml. Sampling and analysis prior to discharge or during releases will ensure that 4 this limit is met. The PCRV water whl be mixed with cooling tower blowdown water for discharge. The discharge flow rate will be established in accordance with the Fort St. Vrain ODCM [5] prior to radioactive liquid release to assure the limits of 10 CFR 20 and 10 CFR 50 Appendix I are not exceeded,- At a dilution of 2000 gpm, up to 10.9 Curies can be discharged per day. As the total curie content in the PCRV decreases, the allowable bleed rate is expected to increase.

C. Non-Occ.upallenal Radiation Exposure The doses from the projected liquid tritium discharge have been calculated based on the guidelines established in Regulatory Guide 1.109 [7), and are presented in Section 4.2.4.1 and 4.2.4.2 of the " Supplement to Applicant's Environmental Report - Post Operating License Stage" [8).

D. Occupational Exposure The occupational radiation exposure that will result from the normal operation and maintenance of the PCRV Shield Water System has been calculated to be a total of 2.5 Person-Rem. Table Il-1 (attached)identines assumptions regarding maintenance requirements for system compr nents during the dismantlement period. (See also Il-14

. , , . . . . , . ~, ,-

l Attachment to P-92014 January 9,1992 Section 3.2 of the PDP Detailed Decommissioning Cost Estimate [9].) The whole l,ody exposure is primarily due to Co 60 exposure, since the exposure due to tritium will be a small fraction of the total external exposure. This total represents the radiation exposure associated with system maintenance, Diter replacements, deminerallrcr resin change out, and final removal of the system. The filters and I deminerallrcrs will be shielded during operation and will be designed for case of opening, removal and replacement to minimize occupational exposure. (The projected occupational exposure to personnel from accident scenarios due to tritium and Co-60 are provided in PSC's response to NRC RAI Question No. 38.) The details of the occupational radiation exposure estimate for Alter replacement are showa in Table 11-1.

V. SYSTD1 SAFETY ANALYSIS The maximum credible accident involving the PCRV Shield Water System would be the rupture of the system, resulting in the release of the entire liquid contents of the Dooded PCRV. The accident scenario has been postulated and analyzed in Section 3.4.7 of the PDP [1]. This accident scenario conservatively assumed that the theoretical maximum amount of tritium (1 ES Curies) is transferred to the PCRV shield water from the graphite blocks. Furthermore, it is assumed that the entire PCRV inventory of tritiated water spills into the Reactor Building sump / keyway and Goods the basement Door to a height of two feet. In analyzing this accident, atmospheric dispersion factors were calculated using the guidelines provided in Regulatory Guide 1.145, "Atmospherie Dispersion Models for Potential Accident Consequences Assessments at Nuclear Power Plants" [10]. Furthermore, it was conservallvely assumed that all releases to the environment were ground level releases. Dose conversion factors were taken from NUREG-0172 [11]. The dose to an individual standing at a point on the EPZ 100 meters from the Reactor Building was calculated to be 34.8 mrem whole body and lung dose for a two hour period.

This accident scenario was also analyzed in Section 5.0 of the Supplement to the Applicant's Environmental Report Post Operating License Stage [8], using the AIRDOS-EPA computer model to assess the radiological impact on the general public within 30 miles of Fort St. Vrain. For this analysis, the terrestrial food chain model presented in Regulatory Guide 1.109 [7] was used. The maximum individual dose was calculated to be 4.6 E( 6) mrem red bone marrow dose and 4.6 E(-6) mrem to

. the lung. These analyses determined that the radiation exposure to the general public as a result of an accident involving the PCRV Shield Water System is very low. In L 11-15

l Attachment to P-92014 January 9,1992 all cases, the radiological consequences from the postulated accident scenario are well within the 25 Rent whole body dose and 300 Rem to any specific organ guidelines established in 10 CFR 100. Moreover, the radiological consequences are a small 1 fraction of the one Rem whole body dose and five Rem to any specific organ j guidelines cited in the EPA Protective Action Guidelines [12]. j Limits have been estab!!shed in the Decommissioning Technical Specifications [2] to assure that tritium activity concentrations in the PCRV Shield Water System will not exceed those postulated in the decommissioning accident analyses.

VI. SAFETY ANAIJSIS CONCLUSIONS The proposed PCRV Shield Water System will be designed for operation in a safe, analyzed condition and the system will not pose an undue risk to the health and safety of the general public and occupationally exposed personnel. Based on the design, operation, radiological consid: rations, and bounding accident conditions for the system, the following conclusioi 're determined to be applicable:

A. The design of the PCRV Shield Water System will meet the anticipated processing requirements for the decommissioning of Fort St. Vrain, including periods when major processing equipment may be down for maintenance. Adequate processing, recirculation, and temporary holdup capabilities will also be available during periods of liquid waste generatlor.. In accordance with the guidelines of Regulatory Guide 1.143, system design will also provide suitable provisions to (1) control leakage and facilitate system operation and maintenance, and (2) prevent and collect spills from storage tanks.

13. Administrative controls will be incorporated into the system operation so that the total quantity of all radioactive material released during decommissioning activities to unrestricted areas will not exceed limits established in the revised 10 CFR 20, and therefore will not exceed limits for non-occupationally exposed personnel.

C. Administrative controls will be incorporated into the system operation so that the total radiation exposure will not result in estimated annual dose or dose commitment to occupationally exposed workers associated with 11-16

l l

1 l

Attachment to P 92014 January 9,1992 normal system operation and maintenance in excess of allowable 10 CFR 20 occupational exposure limits.

D. Administrative controls will be incorporated into the system operation so  ;

that the total concentration of radioactive materials in liquid efnuents j released to unrestricted areas during liquid discharge operations will not exceed 10 CFR 20 limits.

l l

11-17 L

l l

L 1

- _ _ , . . , . - ~ - . . _ . _ _ _ . _ _ _ . _ _ _ . _ _ . . _ _ . _ . . _ . . . _ _ . _ _ . _ . .

Attachment to P 92014 Jauuary 9,1992 REFERENCES

1. PSC letter, Crawford to Weiss, dated November 5,1990, " Proposed Decommissioning Plan fe :he Fort St. Vrain Nuclear Generating Station", (P-90318) (Revised July 1991, P-91217),

2. PSC letter, Crawford to Weiss, dated August 30,1991, " Decommissioning Technical Specifications' (P-91278).

3. USNRC Regulatory Guide 1.143, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light Water-Cooled Nuclear Power Plants", Rev.1, October 1979.

4. USNRC Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will lic .

ALARA', Rev. 3, June 1978.

Fort St. Vrain Offsite Dose Calculation Manual, dated August 29,1990.

5.

6, Fort St. Vrain Final Safety Analysis Report, Rev. 9, July 1991.

7. USNRC Regulatory Guide 1,109, " Calculation of Annual Doses to Man From Routine Releases of Reactor - Effluents for the Purpose of Evaluating Compliance with 10 CFR 50 Appendix 1", Rev.1, October 1977.

8. PSC letter, Crawford to Weiss, dated July 10,1991; " Supplement to Applicant's Environmental Report - Post Operating License Stage, for Proposed Decommissioning of the Fort St. Vrain Nuclear Generating Station" (P-91219).

9. PSC Ictter, Crawford to Weiss, dated June 6,1991t " Fort St. Vrain Decommissioning Cost Estimate - PROPRIETARY" (P-91198).

10. USNRC Regulatory Guide 1.145, "Atmospherie Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants", Rev.

1, February 1983, 11-18

Attachment to P-92014 January 9,1992

11. NUREG-0172, " Age Specific Radiation Dose Commitment Factors for a One Year Chronic Intake", October 1977.

12. EPA

• Manual of Protective Action Guides and Protective Actions for Nuclear Incidents", EPA 520/1-75-001 A, U.S. Environmental Protection Agency, January 1990.

1 l, 11-19

=

l i

Attachment to P-92014 January 9,1992 TAlh..

• FII!rER REPLACEMENT OCCUPATIONAL RADIATION EXPOSURE ESTIMATE ,

No. of Field Person t Operation Ectken (mtlht) Ilmc Rein

1. Open Shield 2 36 2 min 0.0012

2. Open Filter floushig 1 150 2 min 0.006 and Attach Crane 1 36 2 min

3. Lift Filter to 2 36 5 min 0.006 Shielded Cask 4 Close Shielded Cask 2 36 1 min 0.0012 TOTAL 0,0144 TOTAL OCCUPATIONAL RADIATION EXPOSURE FOR TIIE PCRV SillELD WATER SYSTEM (MAINTENANCE ITEMS) -

Maintenance _QDcInlion EcneD3rin

1. Filter Replaecments: 1.5 Person Rem o Approx. 0,015 Person Rem per replacement

• 100 replacements

2. Demineralizer Resin Change-out: 0.6 Person Rem e 0.050 Person Rem per change out e 12 change outs

! 3. Other Maintenance and Removal: 03 - Person-Rem TOTAL MAINTENANCE ORE BUDGET 2.5 Person-Rem 11-20 L

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Attachment to P-92014 January 9,1992 NRC Ouestion No. 38 (Section 3.3.2.2: Ilquid Wages)

A. PSC estimates that about 500 curies of tritium uvuld be teleased to the PCRV shield waterfrom the 100,000 curies of tritium in the graphite (one half of one percent). PSC's April 26th response andJuly 1st revision to the Decommissioning Plan states that the estimated tritium releasefrom graphhe blocks to PCRVshield unter is ' based on data. " 7he source qf this data must be provided and its accuracy and applicabilityjustified. Inchtde in these evaluations the structure of graphite blocks, e.g. unciad and material composition.

B. PSC's July 10,1991 Supplement to Environmental Report, page .1-12 states that feed and bleed operations uvuld be used to dilute 535 curies of tritium to

• one-half of10 CFR Part 20 limits using 2000 gallons qf smter per minutefor about one month. This is based on the data evaluation above. Any change in this evaluation must be reflected in this release plan. Provide verification that the planned releases are consistent with AURA principles and Envinmmental Protection Regulations related to 10 CFR Part 51. Discuss other potential release options considered.

C. Also evaluate the potential contamination of large whomes of concrete with tritham from nuter leaks in PCRV penetrations or liner, from drying of wet graphite blocks,from unter spills during cask and radioactive meterial handling andfrom evaporation ofunterfrom open PCRVpool surface. As discussed with your stqf', tdtlated unter of hydration in the concrete of a reactor svom at a 5 MW, heavy unter, research reactor prevented its releasefor unrestricted use qfler extensiv: decontamination eforts (NiiREGICR-3336 ' Summary Report Ames Laboratory Research Reactor *).

PSC Resimn$ti i

1.

SUMMARY

OF TRITIUM ANALYSIS During the Fort St. Vrain decommissioning project, the PCRV cavity will be flooded

with water to provide shielding and contamination control. Flooding the PCRV . vill

! result in the release of radionuclides (that exist in the PCRV as a result of activation and plateout)into the water. One of the radionuclides of primary concern is tritium, since a fraction of the tritium inventory is expected to leach out of the graphite blocks l

38 1 1

r Attachment to P 92014 January 9,1992 into the water and the tritium cannot be removed by conventional processing mean.,

employed by the PCRV Shield Water System. The amount of tritium to be handled by the PCRV Shield Water System and potential exposure to personnel depends on both the total amount of tritium present in the graphite and other components inside the PCRV, and the fraction that is released to the water, liowever, since measured data on the actual tritium concentrations that are in the Fort St. Vrain PCRV graphite components and the rate at which the tritium icaches into the water from the Fort St.

Vrain graphite do not exist, the amount of tritium that enters the PCRV water has been estimated, based on (1) a conservative calculation of the total amount of tritium produced during power operation (i.e. 100,000 curies) and (2) actual measurements of tritium leach rates from British Magnox reactor graphite, it is estimated that approximately 500 Curies (or 0.5% of the total tritium inventory) will enter the water. The PCRV Shield Water System is being designed to process this tritium inventory for discharge using the existing liquid effluent discharge path.

An assesstnent has also been made of the impact if the maximum theoretical amount of tritium (100,000 curies) is released into the PCRV shield water. Included are impacts on air handling, tritiated water disposal, contamination, and personnel protection, it was found that these impacts, although significant, can be managed without undue safety hazards and within reasonable costs. Allowing for this extreme case, decommissioning can proceed and will be accomplished within the decommissioning cost estimate previously submitted to the NRC, as will be discussed in this response, in addition, with considerations for the worst credible accident and this extreme case, decommissioning will also be accomplished without undue risk to the safety of the public,

11. ANALYSIS OF TRITIUM GENERATION IN FORT ST. VRAIN A. Sources of Tritium From Iltatter_Qneration Chapter 11 of the Fort St. Vrain Final Safety Analysis Report (FSAR) [1] provides a detailed discussion of tritium generation in the Fort St. Vrain reactor. As stated in the FSAR, the main sources of tritium are:

(1) ternary fission in the fuel particles (2) n,p reaction with lie-3 in the coolant (3) n,or reaction by Li-6 impurities in the graphite and o.her components j 38 2

---

w g- +--p - +g--- + , _----- - , - 7p, _ ,&-7 g- + .-

Attachment to P-92014 January 9,1992 The tritium produced in the fuel by ternary Ossion totals about 10.000 Curles and is expected to be contained by the fuel particle coatings. This tritium will be removed when the fuel is removed prior to decommissioning. The tritium produced from lle-3 reactions was minimired by using helium low in lle-3 and totaled about 3000 Curies.

This tritium was either removed by the gas purl 0 cation syste 3 or was absorbed in the core. Most of this absorption would have occurred in fuel graphite blxks near the coolant channels, and the majority of the tritium generated by this reaction will also be removed when the reactor is defueled prior to decommissioning.

Table 3.1.2 of the Fort St. Vrain Proposed Decommissioning Plan (PDP) [2]

identines the amounts of tritium determined to be present in the PCRV as a result of the activation analysis. Ilowever, the removable graphite redector blocks located at the top, bottom and sides of the core were not included in the results of the activation analysis since it was assumed at that time that these ,emovable reDectors would be removed. Ilowever, current plans are for these removable reneetor blocks to remain in the PCRV after the fuel elements are removed and replaced with defueling elements. Therefore, it is possible that these removable reneetor blocks could be in the PCRV when it is Gooded, and the total tritium inventory should be revised to account for the contents of these blocks.

Table 38-1 provides a t imary of the key parameters of these hexagonal removable renector blocks. These blocks are comprised of 11-327 and 11-451 graphites.

Althogh more than one third of these removable reDector blocks were removed from the PCRV during the three core refuelings and replaced with unirradiated blocks, it is conservatively assumed that all blocks were irradiated over the full 890 EFPD reactor operating lifetime and were exposed to the same relatively high integrated thermal neutron Dux as the large side reDectors. Based on these assumptions (and a Lithium impurity of 0.1 ppm), a tritium inventory of 3,500 Curies is estimated to be contained in the removable hexagonal reDector block graphite.

38-3

i Attachment to P 92014 January 9,1992 Therefore, the maximum tritium inventory in the graphite that could exist in the PCRV when it it flooded is: i Large permanent side renectors 82,588 Ci (84.5 %) l Boronated side spacer blocks 11,532 (11.8%)

removable hexagonal renector 3,500 ( 3.6%)

b*ocks Core support blocks and bottom 48 ( < 0,1 %)

reflectors with hastelloy cans TOTAL 97,638 Ci For the purposes of estimbting the amount of tritium in the graphite, a tritium inventory of 100,000 Cuhs is assumed. Per o rms of the fixed price contract, WestinEh ouse is obligated to remove and dispose of the entire tritium inventory of up to 100,000 Curies.

B. Dncription of Fort St. Vrain Graphites Retalnine Tritium Figures 38-1 and 38-2 provide plan and elevation views of the core area that identify the key components and type of graphite used for each. From the standpoint of tritium generation, the graphite in Fort St. Vrain is primarily of two types. The primary graphite of interest is the liLM graphite, which was used for the large side reflector blocks and the side spacer blocks. IILM graphite is less pure than the graphite used in fuel elements, with a maximum specincation for lithium of 2 ppm.

The second graphite of interest is the relatively high purity graphite (either H-327 or 11-451) that was used for the fuel elements and removable hexagonal reDector blocks.

This high purity graphite is specified to have less than 0.1 ppm of lithium. The amount of tritium contained in these blocks is expected to be relatively low in comparison to the tritium activity in the HLM graphite blocks.

Tab!c 38-1 provides a summary comparison of the major properties of the craphites found in the core area. As seen from the table and the above description of sources -

of tritium, over 96.4% of the tritium p;oduced from the Li 6 reaction is produced in -

the HLM graphite. In particular, due to the neutron Dux depression caused by the boron pins in the side spacer blocks and lithium's thermal neutron capture cross-38-4

l 1

i' Attachment to P 92014 January 9,1992  :

secth%, most of this tritium will have been generated in the large permanent side refic ; tor blocks. These blocks are solid (with no coolant holes) and have a relatively low surface to volume ratio, which is expected to result in lower release rates, as discussed later. Figure 38-2 also identl0cs the outer extreme limits of hexa;onal reflector blocks that can be removed by the Fuel llandling Machine (Film) prior to removal of the PCRV top head. This figure clearly shows that removal of any of the ItLM graphite blocks to obtain tritium data is not possible with the Film due to the size (weighing approximately 1500 lbs each) and location of these blocks. Additional information can be obtained by reviewing Figures 2.2 5 and 2.2-10 of the PDP. [2]

Conservative calculations of the total amount of tritium that could potentially be in the graphite blocks have been made.13ased on the maximum specification value for lithium impurity in the llLM graphite, the thcoretical maximum amount of tritium that can be in the total system is approximately 100,000 Curles. This estimate was derived in the activation anal) sis [2] using the calculated neutron flux in the graphite ,

and the actual reactor power history. The calculated Aux was derived from one-dimensional neutron transport calculations in the axial up, axial down, and radial directions. These calculations produce the maximum Oux in each direction and therefore are expected to overestimate the activation, especially in corner regions.

An additional conservative assumption is that all of the tritium produced in the graphite has been retained, i.e., none has diffused out during reactor operation or subsequent shutdown.

The activation analysis also e,timated the curie content of other radionuclides pnxluced in the graphite. Of these, Co-60 will be the major contributor to worker dase (a total of 3,000 to 10,000 Curies of Co 60 may also be retained in the graphite). The Co-60, like the tritium, is also expected to slowly enter the water

= system, but the cobalt is capable of being removed by the PCRV Shield Water System demineralizers. Another radionuclide of concern is Ibropium (Eu), however, previous measurements of Fort St. Vrain graphite [2] Indicate that Europium will be only a small fraction of the Co-60 activity and therefore is not a major concern during an early dismantlement process.

l 38-5 L ,

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Attachment to F 92014 January 9,1992 111. . ANALYSIS OF TRITIUM Hl:1. EASE INTO Tile PCRV SillELD WATER SYSTEM A. Dnttiption of British Graphile Tnthw 7

Data on tritium leaching from graphitc obtained by the British [3] is considered to be directly relevant to determine the fraction of the tritium inventory like!y to be leached from the Fort St. Vrain graphite after the PCRV is flooded. These llritish measurements were made in support of decommissioning of the Magnox and AGR plants, and form the basis for disposal planning for irradiated graphlh for the European Community. The graphite in the Ilritish tests is typical of that used in the  ;

M5gnox. and AGR reactors. Key parameters for the liritish graphite test samples are also included in Table 381 for comparison with data for Fort St. Vrain graphites. -

The liritish leach rate measurements were carried out follov ing !AllA guidelines [4]

with a slight modification to expose a larger graphite surface area to the leachant.

The 13ritish data were taken for several cases; those most appropriate for the Fort St.

Vrain case are measurements of leaching in simulated ground water and in demineralized water. Two graphite samples were tested in each test. The measurements were made by placing a graphite sample into a relatively small amount ,

of water in order to produce tritium concentrations that could be easily and accurately measured, llowever, since the number of hydrogen atoms in the water dwarfs the number of tritium atoms in the graphite samples, any effect of the amount of water or its circulation should be negligible.

For the two samples of Ilritish graphite that were tested in demineralized water, the leach rate of the tritium was measured to decrease with time starting at about 0.1%

per day and declining to below 0.0001% per day after several months, as shown in Figure 38 4 (Figure 6.6 from [3]). Table 6.6b [3] identifies the cumulative fraction of tritium leached after 100 days in deminerallred water to be 0.52%. The fraction leached versus time in deminerallied water is shown in Figure 38 3 (Figure 6.2 taken from [3]).

11. Besults of Graohite Testing Performed.hylhe French Measurements of tritium leaching from irradiated graphite in distilled water were performed by the French [5]. These measuren :nts were carried out on two types of 38-6

Attachment to P 92014 January 9,1992 graphite with different porosities obtained from gas cooled reactors. leaching of the unimpregnated graphite for greater than 90 days resulted in fractional tritium amounts in the water of 0.004% to 0.3%, which are less than the results observed in the liritish testing. In addition, the French results support an interpretation that reducing graphite porosity (i.e., higher density) will retard the tritium leach rate.

C4 Dc.lenDjaallerLufjhtfort St. Vrain leach. Rate As noted previously, the leach rate of the tritium in the liritish test was rneasured to decrease with time starting at about 0.1% per day and declining to below 0.0001 % -

per day after several months, as shown in Figure 38-4 (Figure 6.6 from [3}).

Applying these values to Fort St. Vrain, a curve of tritium release rate versus time was prepared with an initial tritium release rate of 0.5% of the tritium inventery in the graphite released in about the first month after flooding the PCitV. Use of this release rate results in a release of 500 Curies from the graphite and absorbed by the water, based on an assumed initial tritium inventory of 100,000 Curles in the core graphites. Thereafter, the tritium release rate from the graphite is assumed to continue to decrease, falling to a release rate of less than 0.0001% per day within several months, consistent with the results of the liritish test.

The tritium in the PCRV Shield Water System will be gradually removed by efnuent discharge operations. Based on allowable discharge limits and the tritium release curve constructed from the Ilritish data, an estimate was made c' Qe amount of tritium in the PCRV water system. This estimate is shown in 4 F are 38 5. This curve is icvised from that previously provided in F,gure 3.31 of the PDP [2] and Fyure 4.'2-1 ed the Supplement to the Environmental Report - Post Operating Singe  ;

[6), and assumes that discharges will be in accordance with the revised 10 CFR 20 limits as discussed below. This curve is based on the tritium leaching and discharge rates discussed in the following section on tritium release and monitoring.

D. Comparison of lidtlih Test Results vsJort St. Viain Assump.llen$

' The impurities in the British graphite are lower than those in the Fort St. Vrain large

side reflector graphite. Therefore, typleal tritium levels in the liritish graphite are l expected to be more. than 10 times less than expected in the large side renector

- blocks, even though the irradiation exposure of the llritish test samples was higher.

E The tritium may, however, be assumed to be distributed in a similar fashion in the 38 7 L

1 l~

l. ._,,,- .. . - - , - -. -

r Attachment to F-92014 January 9,1992 -

two cases since its formation methal (i.e., n,a reaction with Li-6) is the same.  ;

I Several other factors are expected to affect the leach rates as determined by the ,

liritish testing. These factors have been evaluated in comparing the liritish test  !

samples and the Fort St. Vrain llLht graphite, which is the principal graphite of concern since it contains over 96.4% of the tritium retained in the core. These factors include: (1) surface-to volume ratlo; (?) graphite density; (3) reactor power history; and (4) type of primary coolant.

(1) Smface to-Volume Ratio: One of the most important differences between the graphite in the liritish tests and the Port St. Vrain llLh1 graphitc in the -

PCitV is the signincantly smaller surface-to volume ratio of the Port St. Vrain l llLht graphite. This ratio is significant since the actual diffusion of tritium from ,

the graphite depends on the amount of surface area exposed to the water.

Greater sJrface area will allow greater water ingress !nto the graphite matrix and reduce the tritium migration distance before it will dissociate into the water. A higher surface area will therefore result in a higher leach rate and greater ,

fractional release, liowever, the relationship between surface to-volume ratio and tritium leach rate is not linear since the water ingress into the graphite is a complicated function.

The surfm-to-volume ratios for the large side reflectors and the boronated side

. space, Nocks are approximately 1/20th and one half, respectively, of the surface-tomohre oth of the 11ritish test samples. Therefore, tritium leach rates for thtst ? M paphites should be less than those determined in the llritish testing.

Conwgnty, use of the 11ritish leach rate would be expected to bound the tritium release rate from the larger Fort St. Vrain graphite blocks, both in terms of rate of release and total release fraction.

(2) Graphite Density l Appreciably lower density implies that the graphite would have a higher porosity i within its graphite matrix, liiglier porosity, as indicated in French data, allows more sites for the tritium to collect in interstitial holes with the graphite matrix.

This, in turn, implies weaker bonding for this tritium, and more rapid release, liigher porosity will also allow water to more easily penetrate the graphite matrix, which will also allow more rapio release of the tritium (i.e., higher leach rate and greater fractional release).

38-8 l

I l

l l

Attachment to P.92014 January 9,1992 Table 381 compares the Fort St. Vrain llLM graphite with the llritish Magnox test samples, and unitradiated densities of both are nearly equal, while the irradiated density of the liritish graphite is less than the llLM grapniie.

Therefore, based on density properties of irradiated graphite, the teach rai nr the llLM graphite should be less than that determined by the liritish testing.

(3) Ructor Power Uhwy Reactor power operation will subject the graphite matr8x to damaging irradiation l cffects. One possible result is increased radiatlos damage to the interstitial l

. graphite inatrix, which can increase the porositv of this matrix. 1 Comparing the llLM graphite with the liritish test samples, both were exposed ,

to a thermal neutron flux of approximately 3 Hl3 neutrons /(ernNc), llowever. l due to the low power history of Fort St. Vrain, the llLM graphite was subjected to only 890 EFPD, whereas the liritish test sampics were exposed to over 3500  ;

EFPD. Therefore, the potential damage to the liritish test surnples would be expected to be greater than that experienced by the llLM graphite, resulting in more irradiation damage to the tiritish graphite, increased porosity, and a higher leach rate than the llLM graphite under thek circumstances.

l (4) Type of Primary CN] ant The effects of primary coolant were also; valuated, liritish Magnox reactors use '

carbon dioxide as the primary coolant, whereas helium was used as the primary coolant in the Fort St, Vrain reactor. Carbon dioxide, when used as a evolant at elevated temperatures within a graphite-moderated reactor, will react and remove carbon from the graphite moderator and produce carbon monoxide. Over s time, this process will decrease the graphite's density and increase its porosity. +

As noted above, this increased porosity will increase both the tritium leach rate and the fractional release of the total tritium inventory, llelium, whan used as the primary coolant, will not react with the graphite and therefore v dl not affect the density or porosity of the graphite. (As noted previously, lie-3 will react to ercate tritiuni, some of which will be retained within the core graphites; however, as discu , sed, it is expected that nearly all of this tritium is retained in the fuel element gr sphite that will be removed prior to beginning decommissioning.) - As noted wit!' reactor power history, the potential damage to the liritish test samples would i e expceted to be greater than that 38 9

i l

t Attachment to P 92014  !

January 9,1992 {

experienced by the llLM graphite, resulting in increased poroshy and a higher  ;

leach rate than the llLM graphite under these circumstances. l 5

($) Qupnarlson of Key Paramgica For each of the above parameters, the properties of the Fort St. Vrain-IILM graphite are expected to result in a more conservative behavior than the liritish  ;

Magnox test samples with respect to tritium leach rate and the fractional release  ;

of tritium, llLM surface to volume ratios are significantlylower, indicating that l water irgress will not occur a rapidly and tritium migration to the graphite surface wllt take significantly longer. Irradiated densities of the 11LM is greater l

than the llritish graphite samples, indicating lower porosity and a lower leach rate in the llLM graphite due .to density. liffects of both reactor power history and j primary coolant favor the 11LM graphite, since the effect on increased porosity ,

should be greater in the liritish samples than in the 11LM. o Therefore, the leach rate for the IILM graphite is not expected to be greater than that determined for the liritish Magnox graphite samples, and use of the leach rate determined by the liritish test in demineralized water should represent a conservative upper bound on the leach rate that should be experienced when th ,

PCRV is flooded and the llLM graphite is immersed in water.

IV. TRITIUM PROCESSING, RELEASE AND DISPOSAL OPTIONS A. Tritium Release Alternatives Since tritium cannot be removed from the water by processing, it must either' be- ]

diluted to releasable levels or disposed of as radioactive waste. The release option 1 was chosen for the low tritium conecatrations expected in the PCRV Shield Water  !

System. The following tritium release options were evaluated for the Fort St. Vrain decommissioning program.

(1) Construction of a large shallow water impoundment for solar evaporation of tritium. i

^

(2) Installation of a series of mechanical evaporators for forced evaporation of tritium.

(3) Use of a discharge system (PCRV Shield Water System) to discharge tritium through the existing Fort St. Vrain liquid effluent release pathway, g

y .38 10 ,

l'

~ . - . , u. a = -. - -..-..--.-.-..__.___=.--..__...-.-a

L Attachment to P-92014 January 9,1992 Solar evaporation was evaluated and deemed inappropriate at this time for the following reasons:

1 The evaporation rate c.u not be controlled.

2. Rainfal! or snow couk. hd6 tae amount of tritiated liquid tc be evaporated.

3. Required security and access fencing.

4. Difficulty in preventing migrating birds and small animals from entering the area.

5. Potential occupational radiation exposure to decommissioning personnel.

6. The impoundment presents an addnional accident source term, specifically in the event of rupture of the impour.dment.

7. Costs associated with lining the impoundment.

8. Disposal of sludge and ultirnate decontamination of the impoundment area.

Mechanical evaporators were also evaluated and deemed inappropriate at this time for the following reasons:

1. Large throughput mechanicalnaporate", are expensive and require constant attention.

2. L,arge number of evaporators are required for the uccessary throughput and operational flexibility in the event of evaporator breakdown.

3. Concern over input stream to evaporators (i.e., oils and detergents).

4. Potential occupational radiation exposure to decommissioning personnel.

5. Even,ual decontamination and disposal of evaporators is requirco.

The PCRV Sh.Jd Water System, discharging to the existing Fort St. Vrain plant liquid effluent stream, was selected as the best possible rebase path for tritium in the PJ RV shield water. In addition, occupational and public doses from effluent discharge operations are the lowest of all the alternatives evaluated. This effluent discharge path was fully analyzed in the Fort St. Vrain Safety Analysis Report [1]

and the original Fapplement to the Environmental Report [7], and the impacts during normal plant .gerations were determined to be acceptable to the NRC, Moreover, the PSC environmental monitoring program, which complies with the Regulatory Guide 1.21 [8], has confirmed no significant impacts to the environment due to discharges of tritiated water over the last fifteen years of operation.

38-11

Attachment to P_-92014 January 9,- 1992 The current tritium discharge pathway is modeled by the Fort St. Vrain Offsite Dose Calculation Manual (ODCM) from this release pathway. This pathway currently has adequate measuring and monitoring capabilities for the anticipated discharge level (both water quantity and curie content). This pathway provides adequate water to dilute the anticipated quantity of tritium to below the revised 10 CFR 20 MPC limits.

In summary, the discharge of tritium to the existing Fort St. Vrain liquid efauent stream provides the most advantageous method for tritium release during decommissioning. In addition, it is an accepted and demonstrated safe method that

. will minimize bcth occupational and public doses.

B. Solidification of Highly Tritiated Water In the unlikely event that the amount of tritium entering the water greatly exceeds the expected levels, and the effluent discharge release method cannot be used, attemate disposal methods are available. In this case, after trithim pickup by the water is complete and suitable containers are in place, a feasible contingency plan is to remove the water from the PCRV in its entirety and solidify it for disposal. For cost estimating purposes, an acceptable solidification process would be to use Aquaset, which has a solidification efficiency of 45 gallons in a 55 gallon drum. The disposal effort requires about I hour per drum.

Appropriate radiological controls would be implemented during the solidification of the tritiated water to maintain external and internal radiation exposures ALARA. The

, occupational dose resulting from this operation would be negligible and any incident involving a small spill of tritiated water during solidification operations would be bounded by the scenario described in Accident Scenario 3 "Small Spill" (described

~ in paragraph V.D. of this response.)

Major costs associated with this method include the Aquaset ($250 per drum) and-additional waste disposal costs of $300 - $1100 pe: drum, based on projected disposal costs between $O/ft'- $140/ft'. The total disposal cost for the water in the PCRV

. could range from $5 million - $10.0 million (including labor, materials, transportation, and burial costs), Because of the increased costs and inability to continuously improve water quality, the solidification method wou, not be used unless the tritium level greatly exceeds expected levels. Solidification of highly tritiated water is discussed here to demonstrate that suitable technology is currently -

38-12

l Attachment to P 92014 January 9,1992 available, and to establish a bounding cost for disposal should the preferred method not be suitable.

C. Irillum Release and Monitoring The PCRV will be filled with approximately 325,000 gallons of water. Filling of the PCRV will be stopped at predetermined levels (1/4 core submergence increments) to allow tritium sr.mpling and analysis. No discharge will be made until the trend of tritium concentration is determined. The initial concentration of tritium in the PCRV (approximately 5 days after fill) is estimated to be less than 0.40 pCi/ml, based on 500 Ci of tritium diluted in 325,000 gallons of water.

The Decommissioning Technical Specifications require that the PCRV water be sampled and analyzed daily for tritium concentrations during the initial 011 of the PCRV, Sample frequency may be reduced tn weekly after the tritium concentration l i as decreas.ed to less than 0.1 pCi/cc. Limits have been established in the Decommissioning Technical Specifications to assure that tritium activity concentrations in the PCRV Shield Water System will not exceed those postulated in the decommissioning accident analyses.

Once the trend of tritium concenn.! ion in the PCRV is established, discharge will begin. Water from the PCRV.will be processed through the PCRV Shield Water System (See RAI #11 for details) and a side stream will be transferred to a liquid waste holdup tank in the existing plant Radioactive Liquid Waste System (System 62).

The tank will be sampled for tritium and other principal radionuclides. 13ased on sample results and the limits prescribed in the Fort St. Yrain Offsite Dose Calculation Manual (ODCM) [9], an allowable release rate will be determined. This method of redundant monitoring will ensure that the desired discharge concentration (less than MPC) is not exceeded.

After sampling, the liquid in the liquid waste holdup tank will be initially discharged l - at a rate from 1.4 - 10 gpm and diluted by the cooling tower blowdown flow prior l - to release to the surrounding surface v, ter. The minimum cooling tower blowdown

. flow of 1100 gpm defined in the ODCM will ensure a dilution factor of more than

-100. Figure 38 5 shows the projected decrease in PCRV tritium concentration assuming a discharge of up to 10.9 Curies per day (2000 gpm cooling tower blowdown discharge) until tritium concentrations drop below levels where this ratt 38-13

i

- Attachment to P-92014 -

January 9,1992' can be maintained without requiring dilution to meet the revised 10 CFR 20 MPC  !

limits. Tritium concentration will continue to be reduced, and after approximately  :

3 months of effluent discharge operations, the PCRV water tritium concentratins will be low cnough to allow direct discharge to the environment (i.e., less than 10

- CFR 20 MPC limits). Discharge at a slower rate (1100_ gpm, resulting in a release rate 'of 6 Curies /dayLor less) would extend the time to reach the 10 CFR 20 '

concentration limits.

LWater consumption requirements to support the dilution of the vessel water may range from the minimum flow mte of 1100 gallons per minute to a Dow rate of more than = 2000 gallons . per minute. Existing-site-capacity can accommodate these requirements.and'no additional water sources are required. FSAR Section 2.5.1 -

details the make up water supply which includes water diverted from the South Platte

~ River and St. Vrain Creek.

In sumniary, the estimated 500 Cories of tritium released from the graphite blocks into the PCRV shield _ water will be sampled, analyzed and discharged through the existing' Fort- St.' Vrain liquid effluent release pathway. The established liquid

~

~e ffluent pathway will be used to dilute the effluent to less than the revised 10 CFR-20 limits. The liquid effluent pathway uses cooling tower blowdown (nominally 1100

- to 2000 gpm) to dilute releases to ensure that the releases are consistent with 10 CFR 50*Appndix'I and 10 CFR 51; The controls and administrative procedures that

. governed releases of radioactive liquid wastes during plant operation will'also govern "the releases from the PCRV-during decommissioning. Sampling and analy_ sis prior-

to discharge or during releases will ensure that the limits of 10 CFR 20 and 10 CFR c 50 Appendix I are not exceeded. .

V. RADIOLOGICAL CONSIDERATIONS A~ . Non Occupational Radiation Excontrs The effluent discharge operation will be capable of discharging up to 6 Curies per

1. ' day using an 1100 gpm blowdown rate,'or up to 10.9 Curies per day using a 2000-gpm blowdown rate. Using'the existing Fort St, Vrain ODCM [9], the projected annual dose to a member of the public is estimated to be less than 1.0 mrem. This assumes that all-500 Curies are discharged in a manner consistent with the revised 10 CFR-20 limit in approximately 3 months, and a dose rate conversion factor of

38-14 1

' Attachment to P-92014 January 9,1992 0.226 mrem /hr/pCi/mlis used. The resulting dose to any member of the public due to the proposed dischargt, is a fraction of the guideline provided in 10 CFR 50 Appendix I (i.e., less than 3 mrem total body or 10 mrem to any organ).

B. Qctupational Radiation Ex31ure From Normal OperedinDS Some limited occupational exposure to tritium is expected to occur while performing decommissioning work activities located over the Gooded PCRV. The resulting personnel dose from tritium is expected to be a small fraction of any doses resulting from external radiation.

The work activities most likely to involve exposure to tritium will be those performed on or around the rotary work platform over the PCRV. (See the previous PSC response to NRC RA1 Question No.13 [10] for conceptual drawing.) The work platform will be equipped with an air handling system to exhaust the air from the space between the work platform and the open PCRV pool surface, and discharge it to the Radioactive Gas Waste System (System 63) for sampling and to the Reactor Building Ventilation System (System 73) for exhaust. 'ihis air Dow will virtually eliminate any exposure to workers on the work platform due to tritium evaporation from the open PCRV pool surface in addition, workers handling wet items will be

_provided protective clothing to minimize direct contact with tritiated water.

A tritium bioassay-program will be impicmented as part of the Decommissioning Radiation Protection Program and tritium air samplirig equipment will be available to ensure the timely detection and assessment of individuals likely to be exposed to tritium. Personnel exposure to tritium during decommissioning is expected to steadily -

decline as the PCRV water tritium concentration decreases due to the ef0uent discharge- operation. In fact, the tritium concentration in the PCRV water is projected to be reduced to 0.01 Ci/cc after about 2 months of discharge assuming a 2000 gpm blowdown rate (See Figure 38-5). According to Regulatory Guide 8.32,

" Criteria.for Establishing a Tritium Bioassay Program" [11], operations involving tritium concentrations below 0.01 pCi/cc are sufficiently low that a bioassay program is not specincally required. Additionally, the bulk of the activities involving workers handling wet, comaminated items will occur after this 2-month period.

38-15

Attachment to P-92014 January 9,1992 C. Evaluation of Worst Case Accident Condhas The maximum credible accident involving the PCRV Shield Water System would be the rupture of the system, resulting in the liquid release of the entire content of the flooded PCRV. Although such a release is considered to be improbable, this accident scenario has been postulated and was analyicd in Section 3.4.7 of the PDP [2]. This accident scenario was analyzed to bound the offsite radiological consequences that could result from the installation and operation of the PCRV Shield Water System during normal operation and from potential accidents.

This accident scenario conservatively assumed that the theoretical maximum amount of tritium (100,000 Curies) is transferred to the PCRV shielding war from the graphite blocks, resulting in a tritium concentration in water of 62.* pCi/cc.

Furthermore, it is also assumed that the entire inventory of the l'CRV water spills into the Reactor Building sump / keyway and Hoods the basement floor to a height of two feet.

The dose to an individual standing at a point on the Emergency Planning Zone (EPZ) boundary 100 meters from the Reactor Building as a result of this accident was calculated to be 34.8 mrem whole body and lung dose for a two hour period. The radiological consequences are well within the 25 Rem whole body and 300 Rem to any specific organ guidelines established in 10 CFR 100.

D, Occupational .11adiation Exposure from Accjdents Additional concerns also exist relating to personnel exposure as a result of accidents and routine operations, contamination of concrete with- tritiated water, and for disposal of tritiated water. An evaluation of occupational radiation exposures for three accident scenarios involving the PCRV Shield Water System were made. These accident scenarios were:

Scenario 1: A worker falls from the work platform into the Dooded PCRV.

Scenario 2: A worker is exposed to the tritium-contaminated water in the Reactor Building sump following a catastrophic rupture of the PCRV Shield Water System.

38-16 i

.=- -- -

Attachment to P-92014 January 9,1992 Scenario 3: A worker is exposed to a small spill of tritiated water as a result of a routine maintenance mishap.

Each scenario was evaluated for two tritium inventories in the PCRV. These are the maximum theoretical value (100,000 Curies) and the expected value (500 Curies).-

The assumed Co-60 water inventory of 1000 Curies in both cases is twice the amount of Co-60 expected to enter the water. No credit is taken for removal by the demineralizer systems to maintain the Co-60 level at a small fraction of this value.

Accident Scenario 1: Worker Falls into Flooded PCRV In this scenario, a worker falls through the work platform access openings into the flooded PCRV Personnel exposures were evaluated based on the assumption that the worker is rescued and dried within: (1) 10 or (2) 60 minutes of the fall. The Reactor Building and work platform ventilation systems are assumed inoperable at the time of the accident.

For this accident scenario, individual exposiires due to tritium are assumed to occur as a result of (1) inhalation of air (86*F at 70% rela *.ive humidity); (2) absorption through the skin as a result of 100% wetting; and (3) ingestion of 10 ml of water.

Exposure due to Co-60 is assumed to occur as a result of (1) external irradiation from the water itself, and (2) ingestion of 10 ml of water.

- For the Scenario 1 accident involving the maximum theoretical value (100,000 Curie case), the following exposures were determined for the 10 minute and 60 minute immersion cases, respectively:

10 minute exposure 60 ndnute exposure (Rem) (Rem)

Route !ritium Col 0 tritium Cm@

Inhalation = 0.01563 ---

0.0937 ---

Wetted Skin 0.02253 ---

0.0981 ---

Swallowing 0.03900 0.198 0.0390 0.198 External --

- 0.734- --

4.402 TOTAL 1.009 Rem 4.832 Rem (Whole Body) 38-17

Attachment to P-92014 January 9,1992 For the Scenario 1 accident involving the expected value (500 Curie), the corresponding totals are 0.932 Rem (WBE) for the 10 minute exposure and 4.602-P.em (WBE) for the 60 minute exposure.

Accident jSegnuio 2f Worker exposed at the Reactor Building Sump after a postulated Loss of PCRV Shield Water.

This exposure is postulated to occur following complete loss of all water in the PCRV water to the Reactor Building sump. An emergency worker is sent to the area of the Reactor Building sump containing the entire contents of the PCRV and remains in the ' vicinity 'of the flooded sump area. The exposure is due both to direct shine (beta and gamma from the water), as well as inhalation of the tritiated water vapor.

The dominant contributor to exposure is Co-60.

For the Scenario 2 accident involving maximum theoretical values (100,000 Curies),

~t he worker would be exposed to an external dose late of 2.2 Rem / hour and the worker would receive an internal exposure of 0.097 Rem due to inhalation of tritiated water vapor. The total 1-hour exposure would be 2.3 Rem (WBE) for the maximum (100,000 Curies) tritium inventory scenario.

For the Scenarie ' accident involving expected values (500 Curies), the total exposure would be approximately 2.2 Rem external and 4.85 E(-4) Rem internal exposure due to tritium inhalation, for a total of 2.2 Rem (WBE).

.- Accident Scenario 3: Small Snill of Tritiated Water In this accident scenario, it is postulated that a worker stands I foot from a two

. gallon spill of PCRV shield water (from a clarifying pump). It is assumed that the worker becomes contaminated with 1 E(-5) (0.001%) of the spill [12], resulting in exposure from (1) direct (external) exposure from the spill; (2)' internal Committed Effective Dose Equivalent (50 yr CEDE) due to Co-60; and (3) internal Committed

_ Effective Dose Equivalent (50 yr CEDE) from tritium. As noted for Accident Scenario 2, the dominant contributor to exposure is Co-60.

. For the Scenario 3 accident involving maximum theoretical values (100,000 Curies),

the v/orker would receive an external exposure of 87.4 mrem and an internal exposure of 10.6 mrem, primarily due to the Co-60. The total exposure would be 38-18

~

% F LAttachment to P-92014 January 9,1992 -

s.

98 mrem (WBE) for the maximum (100,000 Curies) tritium inventory scenario.

For Be Scenario 3 accident involving expected values (500 Curies), the total exposure is approximately 97.7 mrem (WBE).

E. Accident Scenario Conclusions; Since these exposures are postulated to occur- as a result of accidents, the occupational exposure limits of 10 CFR 20 are not applicable. However, none of the accident ' scenarios identified above will result in life-threatening exposures to workers, eveni with the most conservative assumptions as to radionuclide

concentrations in the water. Fort St.- Vrain decommissioning operating procedures will minimize the likelihood of these accidents.-

VL EVALUATION OF TRITIATED CONCRETE AT TIIE AMES LABORATORY RESEARCil REACTOR Based on the decommissioning experience gained from the Ames Laboratory  ;

Research Reactor (ALRR), the NRC identified a concern for potential contamir.ation T

oflarge volumes of concrete with tritiated water, The ALRR operated for about 12

-years and had a tritium concentration in the water of 1,8 Ci/ liter (1800' Ci/cc) at

^ shutdown [13,14]l.This represents a mu:h longer potential exposure time and a much higher tritium concentration than the maximum that could exist at Fort St. Vrain, Thus :the~ analysis- described below predicts -an impact of worst case concrete

. contamination of only a small fraction of the Ames case,-

In order to assess.the effect on Fort St. Vrain decommissioning of spills of tritiated 1 water on concrete, the worst case of a total release of the water in the PCRV shield water system'was considered. This release was postulated to occur with the water containing the maximum possible concentration of tritium (62,4 pCi/cc) permitted by the ; Decommissioning Technical Specifications (15), Such a release is very Jimprobable, but could be postula:ed to occur as a result of a major break in the water system."It was further assumed that de water would enter the sump and remain there

for one month until repairs could be effected and the tritiated water pumped back into the PCRV.

L p

l 38-19 l

L l -

1' l-t'

~_. -- .- ..

Attachment to P-92014 January 9,1992 The concrete in the lower portion of the Fort St. Vrain Reactor Building sump has been exposed to water during operations and will continue to be exposed during decommissioning. The - sump can be periodically wetted down during decommissioning as necessary to prevent dryout and inhibit possible diffusion into dry concrete. Therefore, the concrete can be considered to be sMurated with water.

The primary mechanism for tritium to enter the concrete as a result of exposure to tritiated water will be Sy diffusion in accordance with Fick's law of diffusion.

Diffusion coefficients for water into concrete vary with parameters such as aggregate size, porosity, and hydrostatic pressure, and therefore only an approximate analysis is possible. The values of the diffusion coefficient for unpainted concrete were measured at about 1.0 E(-5) cm2 /s [16] for small samples. Another set of measurements was made on an actual concretc wall and Door of a heavy water research reactor [17]. -These measurements gave a much lower value of 3 E(-7) cm2/s. Using the conservative higher value, calculations were performed to estimate the tritium distribution in the concrete as a function of time. It was assumed that the surface of the concrete was washed with clean water after removal of the tritiated water, and that the surface was kept wet to allow migration of the tritium back out of the concrete by exchange with the surface moisture. Results of the diffusion calculations are shown in Figures 38-6 and 38-7. These calculations indicate that one year after the exposure, the tritium concentration in the water of hydration throughout the concrete will be less than 2% of the 62.4 pCi/cc maximum. The surface concentration will be even lower ar.d the tritium in the top centimeter of concrete will average less than 0.015 pCi/cc.

Based on the diffusion analysis a maximum concentration of 0.02 times 62.4 pCi/cc or 1.25 pCi/cc of tritium in the water in the concrete will remain. For typical concrete density and water fraction, this results in a maximum of 0.11 pCi/gm.

Evaluation of the impact of exposures to such residual radioactive contamination is

~

contained in the report " Residual Radioactive Contamination from Decommissioning"

[18]. This report contains the results of analyses to support the technical basis for translating contamination levels to annual dose. For tritium in concrete, the worst case exposures are concluded to occur as a result of the " building renovation scenario". The value derived for tritium is 2.9 E(-7) total effective dose equivalent (TEDE) in mrem per pCi/gm for an exposure of 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> (Table 3.1.of [18]).

Therefore, based on 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> exposure, the Fort St. Vrain exposure for this adivity would be 0.032 mrem, which is well below the allowable maximum of 10 mrem /yr.

38-20

Attachment to P-92014 January 9,1992 It is therefore concluded that exposure of the concrete to tritiated water at or below the limit of 62.4 pCi/cc will not result in the necessity to remove any contaminated concrete, nor will it create any problems in the decommissioning schedule. The case considered above will bound cases of smaller spills that can be cleaned up in shorter times and that will expose more limited volumes of concrete.

VII. CONCLUSIONS The dominant source of tritium that will remain in the graphite after defueling was generated during plant operations by neutron capture by Li 6 impurities in the HLM graphite. 13ased on the maximum specification for lithium impurity in the IILM graphite and the actual power history, the activation analysis predicts that the theoretical maximum amount of tritium that can be in the PCRV Shield Water System is approximately 100,000 Curies.

Testing performed by the British has shown that a total of 0.5% of the contained

-tritium leaches into the water after 100 days. In addition, tritium leach testing performed by the French on unimpregnated graphite in distilled water has also shown that the maximum amount of tritium that leaches into the water after 90 days is -

0.3 %. The 500 curies that are estimated to be leached into the PCRV water is based on British tec' data that has since been substantiated by independent testing performed by the French.

The preceding evaluations have also shown that even if the entire inventory of tritium

.(100,000 curies) is released into the PCRV water, the radiological consequences to the public and decommissioning workers from normal and postulated accidents scenarios would not exceed the guidelines established in 10 CFR 100. Finally, this tritiated water could be disposed of by solidification if the preferred release method of effluent discharge release is determined to be not _ viable. Allowing for this extreme case, decommissioning can proceed and will be accomplished within the decommissioning cost estimate previously submitted to the NRC. In addition, with considerations for the worst credible accident and this extreme case, decommissioning will also be accomplished without undue risk to the safety of the public.

38-21

Attachment to P-92014 January 9,1992 BEFERENCES

1. Fort St. Vrain Updated Final Safety Analysis Report, Rev. 9, July 22,1991.

2. PSC ~ letter, - Crawford to -Weiss, cated November 5, 1990, " Proposed Decommissioning Plan for the Fort St. Vrain Nuclear Generating Station", (P-90318) (Revised July 1991, P-91217).

3. 1.F. White, et.al., " Assessment of Management M(xles for Graphite from Reactor Decom*nissioning", EUR 9232 en, Commission of the European Ccmmunities.

4. E.D. Hespe, " Leach Testing of Immobilized Radioactive Waste Solids",

Atomic Energy Review 2, p.195, (1971).

5. J.R. Costes, et.al., " Conditioning of Graphite Bricks from Dismantled Gas-Cooled Reactors for Disposal", I ow itad Intermediate Waste Management 1, ASME, New York,1989, p 497-501,

6. Supplement to the Environmental - Post Operating Licensing Stage, for the Fort St. Vrain Nuclear Generating Station, July 10,1991 (P-91278).

7. Applicant's Environmental Report, Operating License Stage, Fort St. Vrain Nuclear Generating Station, Supplement 1, October 1971.

8. USNRC Regulatory Guide 1.21, " Measuring, Evaluating and Reporting _

Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light Water Cooled Nuclear Power Plants", Rey, 1, June 1974.

9. Fort St. Vrain Offsite Dose Calculation Manual, dated August 29,1990.

10. PSC letter, Crawford to Weiss, "PSC Response to NRC RAI On the Fort St.

Vrain Proposed Decommissioning Plan", dated December 6,1991 (P-91423),

11. USNRC Regulatory Guide 8.32, " Criteria for Establishing a Tritium Bioassay Program", Rev. O, July 1988 38-22

. . . . . - _ . . - . - .-. . , - _ _ - ~ . . .

a Attachment to P-920141 u January 9,1992 : ,

12. .Brodsky, Allen, _"IIcalth Physics", 39(6):pages 992-1000,1980.

n 13._ : B,W. Link and R.L. Miller, " Evaluation of Nuclear Facility Decommis'sioaing

. Projects, Summary . Report, : Ames laboratory Research Reactor",

~_

' NUREG/CR-3336,' July 1983.

14. M.A ; Langsan, " Interim Overview / Certification for the Ames Laboratory Research Reactor Facility", letter, Mar. 8,1981. 3

.15. - PSC letter, Crawford to Weiss, dated August 30,1991, " Decommissioning

• ' Technical Spec.ifications" (P-91278) -

16. ' G.G. Eichholz, et. al., " Tritium Penetration Through Concrete", Waste Management 2, p. 27-36,1989.

= 17. _ ' S.4 Numata,'et. al., " Diffusion of Tritiated Water Vapor into Concrete",

Fusion Technology 19, p.140-145,1991. -

_ , 18.o W.EJ Kennedy, Jr; and R. A. Peloquin, " Residual . Radioactive Contamination'-

from Decommissioning",; NUREG/CR-5512,. Draft Report for Comment,

~

January.1990.

1 m

[ jie

,0 r

i a

p 38-23

. - . - . - - - - -. . - -. - ,, - - . - . . - _ . . . . . -. u ~

Attachment to P-92014 January 9,-1992

' TABLE 38-1

. GRAPIIITE PROPERTIES COMPARISON TABLE Large Side Boronated Side Removable . Core Britiso Parsmeter Reflecter Snacer Blocks Reflector Support Blocks Test Sample Remarks Type of Graphite IILM llLM H-451/H-327 PGX (Reactor Grade) 2 samples from Magnox reactor Density (g/cc)

Unirradiated 1.8 1.8 1.72-1.77 1.76 1.82 Irradiated 1.8 1.8 1.77-1.77 1.76 1.7 0.08 0.75 0.12-0.53 ** 1.5 ** -A/V ratios not significant due to Surface to Volume Ratio (ent')

very small tritium curie content Total Mass (g) 1.83 E8 6.11 E7 1.5 E8 8 E7 680 To'al Volume (ce) 1.015 E8 3.395 E7 8 E7 4.5 E7 376* * - Actual sample size: 2 samples tested Total Tritium Content (Ci) 82.558 11.532 3500 17 Tritium Concentration ( Ci/cc) 813 340 <0.01 6.6 10.7* * - Measured value 2.2 E5 Bq/g Major Impuritie;(ppm)

Li <2 <2 < 0.1 <2 <0.05 Fe 2000 2000 < 20 1900 10 Co 0.2 0.2 < 0.01 0.2 0.02 Flux History (EFPD) 890 890 s890 890 =3550 Thermal Flux (n/cm%ec) 3.8 E13 < 3.8 E13 < 3.8 E13 < 3.8 E13 3.4 E13 Maximum Temperature (*C) 300 - 500 300 - 500 400 - 700 700 Primary Coolant He He He He CO2 38-24

p u, ..,

a, ,,;

M e L E

u -.

- \ SIDE REFLECTOR BLOCK f

sq

'k {HLM GRAPHITE)

%,,g ,/YS ' '

j REMOVABLE HEXAGONAL

fff,. TOP & SIDE REFLECTOR BLOCK

%-g_. ./ ~

(H 327Ai 451 GRAPHITE)

M l

, l_ kk ,

, .. - DEFUELING ELEMENTS kN '

BOTTOM REFLECTOR BLOCKS

1. # (H 327/H 451 GRAPHITE)

I ,, ,6 EI g# y[ BORONATED SIDE

1. SPACER BLOCK

~

%  % M- . N '] (HLM GRAPHITE)

f. j L

)

q l hff( ' CORE BARREL

-s. '

(CARr$0N STEEL)

KEYED CORE .O O' ^' # s SUPPORT BLOCK 'O O D* o O \

(PGX GRAPHITE)"(O h o - O O o oO,-- / f HEXAGOr4AL REFLECTOR BLOCKS O '- t WITH PASTEtt OY CANS

=g [ l

_ j (H 327 GRAPritTE)

\ - .- f g ,

,. j V

TRANLITION HEXAGONAL RTFLEC"OS. BLOCKS tri 327 GRAPHITE)

Figure 38-1 Core Graph'te Crmponents - Elevation V;ew

NOTE:

SHADED AREA REPRESENTS LIMITING RADIUS OF GRAPHITE BLOCKS THAT CAN BE REMOVED BY THE FUEL HANDLING MACHINE REMOVABLE HEXAGONAL N SIDE REFLECTOR BLOCK &

(H 327/H 451 GRAPHITE)

G l PERMANENT SIDE I REFLECTOR BLOCK (HLM GRAPHITE) iin.,

, jj/lpj

,.e g

Y>~<x w v

,;;j, ' N n

ly '

m. - T

.m \k< s Q :P _

, J 'wh,1; p' , ,.ff '

0 hI A

i 3

II ' NIi u s _

v//f,,lg9',${ \

E i>  ;

f I <

! fa ,$1$ {

f 4Y // f '

J :llllll1 f 2  ;

n fl

. . i_. e _

y 9 23 #/////t 9

/// ,

j' e yl u llll.f .D - 4:z j

.~  %$

^

' ' D,I 7 '! d . '.' a ,,

, W///

/~

r p 4g,71. . il

• ' /

/ '!,!!

't  ? b . L' y /

'I'lll .. y w:

DEFUELING ELEMENTS t ,,

' [t *

'l ,'f/'lllIfE~

II

'l yjj,,ik'hi' ,

i >

BORONATED SIDE CORE BARREL SPACER BLOCK (CARBON STEEL) l (HLM GRAPHITE)

CORE MAP Figure 38-2 Core Graphite Components - Plan View

.w

")

[ Figure 6.2 of Reference 3_ .

l 1- i- i i i i i I 1

Y

/ -- -

If r f _

/

/ -!

10f -

= 1

= =

i

~

gg3

]

i I

= = 3g

. _ _ __ __133 Ba -

4 10  :  : 13&Cs l'C  :

a 10 Co l i I I I i

-5 i 1 10 60. 80 100 120 140 160 18,Q - -l 0 20 40

. Time , d . .l Figure 38-3 Cumulative Fraction of Activity

~

Leached in Demineralized Water  :

.(1' bar, 25 C)  ;

-- - , -  ? '

% r ., -, . -

[ Figure 6.6 of Reference 3[

I- I I 3 10-I I 3 8 1 i

10-2

\ -

s

%'~ -

T 10-3

~~_' ~~__ -

v - _ . _ _ _

E '

$'10-'

= = =

g

.c u

O e --

l s 10 i

10

~

~ -- ~ 133Ba 3H- 2 1 -

% 134Cs l'C l  :  : 60Co -

i -7 I I 1 i i t [ _I 10 80 120 160 1 0 20 40 60 10 0 1EG 180 l

Time, d -

Figure 38-4 Leach Rate Curves l

I in Demineralized Water (1 bar, 25 C)

r 350 ASSUMPTIONS: 3 30 -

Blowdown Discharge at 2000 gpm with

~

Tritium Levels up to .001 uCi/cc. .

250 -

.O E -

j200 -

nC -

j150 -

E ~

i" 100 -

50 -

i i i i i i i i i .

i i i i i i i i i i i o

0 50 100 150 Days Figure 38-5 Tritium in the PCRV System

1 i  ! I i a

100 -

I TRITIUM IN PERCENT OF MAXIMUM i

CURVE A: AFTER 1 WEEK EXPOSURE CURVE B: AFTER 2' WEEKS EXPOSURE 80 CURVE C: AFTER 1 MONTH EXPOSURE CURVE D: 1 MONTH AFTER END OF EXPOSURE CURVE E: 2 MONTHS AFTER END OF EXPOSURE E

$ 60 -

5 c

?>

u g 40 ~

A '

C B

20 -

D E

O -

, , 4 0 10 20 30 40 50 Depth in Concrete (cm)

Figure 38-6 Tritium Distribution

10 i i i i i i ,

TIME AFTER CLEANUP 8 -

CURVE E: '2 MONTHS CURVE F: 6 MONTHS CURVE G.1 YEAR CURVE H: 2 YEARS

~

E

.g E

.t Y

c o 4 -

8 g F 2 -

G H

0 -

, i ,

0 10 20 30 40 50 60 70 Depth in Concrete (cm)

Figure 38-7 Tritium Distribution

_ _ _ - _ _ _ _ _ .