ML20199F814

Requests Approval of Staff Approach for Resolving Issues Re Waste Classification of Plant Rv

ML20199F814
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	10/21/1997
From:	Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
SECY-97-244, SECY-97-244-01, SECY-97-244-1, SECY-97-244-R, NUDOCS 9802040122
Download: ML20199F814 (67)

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POLICY ISSUE (Notation Vote) W 3Yf October 21.1997 SECY-97-244 EQB: The Commissioners FROM- L. Joseph Callan Executive Director for Operations

SUBJECT:

WASTE CLASSIFICATION OF THE TROJAN REACTOR VESSEL PURPOSE:

To request approval of the staff's approach for resolving issues related to the waste classification of the Trojan reactor vessel (RV).

BACKGROUND:

On July 25,1997, staff transmitted SECY-97-164, to the Commissioners, suggesting an approach for reviewing a request from Portland General Electric Company (PGE), to allow a one-time,2 million curie shipment of the Trojan Nuclear Plant (RV), including its irradiated intemais, by barge, to the low-level radioactive waste disposal site in Hanford, Washington.

The recommended approach is to consult with the State of Washington - an Agreement State - as to whether the RV, with its intemais, is suitable for disposal at the U.S. Ecology low-level waste site in Hanford, Washington. If the State of WasUngton determines that the  :

RV, with its intemals, is not suitable for disposal at the aforementioned site, staff will terminate

its review and retum PGE's application. If the waste classification of the RV with intemals is appropriate for the U.S. Ecology site, staff will notice rc

aipt of PGE's application in the Federal Reoister and perform a review of the transportation package.

On August 26,1997, the Commission issued a Staff Requirements Memorandum approving the above recommended approach. This Commission Paper addresses the first step in the above approach - the waste classification of the RV and its suitability for disposal.

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The Commissioners -24 j On June 18,1997, PGE provided information to NRC on questions related to the waste classification of the proposed shipment (Anachment 1). This information confnmed staff information that the core baffle plates, core former plates, and the lower core plate in the RV would be classified as Greater-Than-Class-C (GTCC) wastes if classified separately, in the submittal, PGE averaged the intemals together with the pressure vessel and other activated components and concluded the waste would be classified as Class C waste.

-In the June 18,1997, submittal, PGE provided a dose analysis of the shipment, performed by U.S. Ecology, to demonstrate that the RV, with its intomals, is suitable for disposal at the Hanford site. In the June 18,1997, submittal, PGE indicated that the one-piece shipment of the n RV, with the intemals, would: (a) allow contact handling of the shipment; (b) would result in 39 to 44 fewer waste cans requiring storage at the Trojan site until a GTCC waste disposal site is -

developed; (c) reduce contamination control problems; (d) reduce occupational exposures from  :

134 to 154 person-rems to 67 persori ems (out of 591 person-roms estimated for the entire Trojan decommissioning); and (e) rwce waste shipments from 44 to 1, L

L . Under the Low-Level Radioactive Waste Pclicy Amendments Act of 1985, the Federal i Government is responsible for developing and operating a disposal facil;ty for commercially

generated GTCC. The U.S. Department of Energy (DOE) accepted this responsibility. Note 4 I that.10 CFR 61.55(s)(2)(iv) requires that waste not generally acceptable for near-surface i disposal be disposed of in a geologic repository, as defined in 10 CFR Part 60, unless

proposals for disposal of such waste in a disposal site licensed pursuant to 10 CFR Part 61 are approved by the Commission.

DISCUSSION:

Under 10 CFR 20.2006 and 10 CFR Part 20, Appendices F and G, low-level waste shipmentsa -  ;

- must be classified in accordance with 10 CFR 61.55. The 10 CFR Part 61 waste classification

~ system is used to define low-level wastes that are generally suitable for disposal at a near-

' surface disposal facility, such as the Hanford site. Wastes with a classification of GTCC are -

generally not suitable for disposal at a near-surface disposal site.: Under 10 CFR 61.58, the

- Commission may approve other provisions for the classification and characteristics of waste, on

- a specific basis, if, after an evaluation of the specific characteristics of the waste, disposal site,  ;

or method of disposal, it finds reasonable assurance of compliance with the performance

, objectives in Part 61, Subpart C. The NRC role in this case is to ensure that PGE properly l classifies its waste in accorda nce with 10 CFR 61.55. The State of Washington would be responsible for ensuring that We wastes are suitable for disposal at the U.S. Ecology site. The State of Washington has not yet made a final decision on the acceptability of the Trojan RV,

- with its intemais, for disposal at the U.S. Ecology site.

In PGE's June 18,1997, submittal, PGE classified the waste by averaging the pressure vessel and the intemals together. The " Branch Technical Position on Concentration Averaging and L Encapsulation" (BTP) was intended to provide guidance to licensees on how to determine the

! concentrations of nuclides in the wastes by averaging over the volume or mass of the wastes.

> Section 3.3 of the BTP provides specific guidance on averaging activated materials such as reactor intemais. The guidance includes a series of ratio tests that are intended to limit intruder L

l

p The Commissioners - .

doses from ind'/idual r actlyated metal shipments. Staff evaluated the waste classificatic n of the :

core baffle plates, core former plates, and the lower core plate, and found that these components substamially exceed the ratios by up to a factor of over 1600 (see Attachment 2).

The BTP also accep+s the use_of attemative provisions, in Section 3.9. This section would accept attemative averaging approaches provided it is demonstrated, with reasot,able ~

assurance, that the performance objectives of Subpart C of Part 61 are met aad, thus, the

. wastes are acceptable for near-surface disposal. The physical form of the wastes, the specific characteristics of the disposal site, and the method of disposal would be considered in specific ,

evaluations performed under the aforementioned Section 3.9.

The pathway analyses provided in PGE's June 18,1997, submittal was prepared by U.S.

Ecology and submitted to the State of Washington. it addrssses only groundwater impacts and direct exposure in a very general manner, Other intruder impacts from construction or residential scenarios were not addressed in a comprehensive manner. U.S. Ecology assumed .

that the pressure vessel would remain intact in the analyses, even though the critical nuclides -

Carbon-14 (half-life 5700 years), Nickel-59 (half-life 80,000 years), and Niobium-94 (half-life 20,000 years) - will be present well beyond 10,000 years.' Nickel 63 (hatf-life 92 years) la also .

a critical nuclide for waste classification purposes. Staff reviewed the groundwater analysis and performed an independent assessment. Staff agrees that there will be no groundwater hazard from the proposed disposal.~ Staff did not, however, prepare a comprehensive analysis of other'-

- intruder impacts. Accordingly, staff concluded that a more complete analysis is needed, if the shipment and disposal of the Trojan RV, with intomals, is implemented, the action could set an important precedent for other reactors that will be decommissiored in the future. The staff considers that its proposed approach to resolve the waste classification issues for the Trojan case should also be applicable to the review of similar requests and should ensure  !

compliance with public health and safety objectives. Note that in the Environmental impact Statement supporting the development of Part 61, and in the Generic Environmental Impact -

' Statement supporting the 1988 decommissioning rulemaking, the staff did not consider the _ >

environmental impacts of disposing of intact reactor pressure vessels with all the intomals in the same shipment.- For these studies, it was assumed that the GTCC intomals would be removed and disposed of at a future disposal facility designated for GTCC wastes. Note also that in previous communications with the public, staff indicated that GTCC reactor intemals would not be expected to be disposed of at near-surface disposal facilities, but at the DOE GTCC disposal facility.- More recently, reactor licensees have more fully investigated the practicality of one-piece RV shipment ar,d disposal, and have concluded that cost arid occupational exposure savings can be acheved. To address future cases, staff would request licensees to submit case-specific analyses of the disposal impacts, of the pressure vessel and intemais. Waste

classification averaging approaches would be accepted for those cases where it is demonstrated, in a comprehensive and defensible manner, that disposal impacts are consistent with the performance objectives in Part 61 and, thus, that the wastes are within the envelope of the prior environmental reviews. Further, an Agreement State, such as the State of Washington, would continue to have authority and responsibility for assuring that wastes are suitable for disposal at a facility it licenses.

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The Commissioners '

If a comprehensive and defensible pathway analysis is submitted, the Trojan RV would be defined as Class C waste by the Commission using its authority under 10 CFR 61.58. As *

[ Class,C waste, the State of Washington would be free to accept the RV for disposal. However, Washington would not be required to do so Under the Low-Level Radioactive Waste Policy Amendments Act of 1985, States are obligated to take only Class C wastes as defined effective ,

i LJanuary 26,1983,' The Act does not, however, prohibit States from accepting waste defined as i ' Class C after January 26,1983. Thus, Washington State (or any Agreement State facing a ,

similar request from a reactor licensee) could decide not to accept the RV for disposal under the Act. Nonetheless, in view of the precedent-setting nature of this issue, disposal of the

pressure vessel of reactors in decommissioning could become an issue in Agreement States

including States with low-level waste disposal sites under development.

- in a March 31,1997 submittal, PGE requested authorization to ship its reactor vessel with its F intomals intact. The licensee needed authorization by November 1997 in order to reserve a

' shipping slot in August 1996. Because of the identification of waste classification issues and

the extensive package certification caseload, this review schedule was not feasible and the

licensee was so advised, it is currently estimated that it could take 6-9 months to complete the

NRC waste classification review and for Washington State to determine the acceptability of the package for disposal at the Hanford low level radioactive waste disposal site. If this finding is

[ favorable, following the general approach outlined in SECY 97-164, the transportation review -

! would begin and is estimated to take 9-12 months. Based on this sequential review schedule,

, PGE could receive approval in November 1999 in time to reserve e sh,pping siot in summer -

2000. Staff will partially mitigate the delay caused by this sequential review by beginning tht:

- technical transportation review with contractor support.- However, extensive staff resources will l . not be expended until favorable resol;; tion of the waste disposal issue.

The staff requests that the Commission approve the following general approach for resolving the Trojan RV waste classification issue:

Transmit the letter in Attachment 3 to the State of Washington, requesting that U.S.

!' Ecology coordinate with PGE and provide a more comprehensive and defensible -

pathway analysis of the Trojan RV disposal impacts. Based on this information, if the analysis demonstrates that the disposal of the RV will be in conformance with the -

performance objectives of Part 61, the waste would be classified as Class C, in l_ accordance with 10 CFR 61.55,10 CFR 61.58, and the altemate averaging provisions of the BTP, 4

COORDINATION-c This paper has been coordinated with the Office of the General Counsel (OGC). OGC has no

_ legal objection to this paper. The Office of the Chief Financial Officer has reviewed this

Commission Paper for resource implications and has no objections. The Office of the Chief

, Information Officer did not have to review this paper as there were no information technology or

]

information management implications identified.

i.

' The Commissioners =5 -

, RECOMMENDATION That the Commission:

Acprove the general approach outlined above to transmit the letter in Attachment 3 to the State

- of Washington requesting that a more comprehensive and defensible pathway analysis be

_ performed for the Trojan RV.

L-J eph Callan Executive Director for Operations Attachments: As stated Commissioners' comments or consent should be providea directly to the Office of the Secretary by COB Thursday, November 6, 1997.

Commission Staff Office comments, if any, should be submitted to the Commissioners NLT october-30, 1997, with an information copy to the Office of the Secretary.

If the paper is of such a nature that it requires additional review and comment,  ;

the Commissioners and the-Secretariat should be apprised of when comments may be expected.

DISTRIBUTION:

Commissioners OGC OCAA OIG OPA-OCA ACNW-

-CIO CFO EDO SECY

i y W W MCol1@y Stephen M. Quennoz Trojan Site Executive June 18,1997 VPN 048-97 Trojan Nuclear Plant Docket 50-344,72-017 License NPF-1 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Engonse to NRC Reauest for Additional Information - Reactor Vessel Package This letter transmits the PGE response to the NRC request for additional information dated May 19,1997. Although the licensing action related to the PGE Safety Anilysis Report for the Reactor Vessel Package is primarily a transportation issue, pursuant to 10 CFR 71, this response is being provided on both the Trojan Part 50 and 72 Dockets for information.

The PGE response to the NRC questions is packaged in the following four Attachments.

• Attachment I provides a restatement of the NRC questions, follow:d by the PGE response.

• Attachment II provides Calculation RPC 97-018 which is' referred to in the rewnse to Questions 1 and 2.

• Attachment III contains the US Ecology letters to the State of Washington.

• Attachment IV is the State of Washington letter documenting the results of the Department of Health, Division of Radiation Protection review of the US Ecology information.

In accordance with the NRC instructions in the May 19,1997 letter, this response includes the requested executed oath or affirmation.

71760 Columbia River Highway. Rainier. OR 97048 503/556-3713 1

STATE OF OREGON, )

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COUNTY OF COLUMBIA )

i I, C. P Yundt, beir.g duly sworn, subscribe to and say that I am the General Manager Plant Support and Technical Functions for Portland General Electric Company, the licensee herein; that I have full authority to execute this oath; that I have reviewed the foregoing; and that to the best of my knowledge, information, and belief the statements made in it are true.

Date Em e /T 1997 (JD/h40 IC. P. Yundt, General Manager Plant Support and Technical Functions Portland General Electric Company On this day personally appeared before me, C. P Yundt; to me known to be the individual who executed the foregoing instrument, and acknowledged that he signed the same as his free act.

GIVEN under my hand and seal this , I day of XL ,1997.

7 0FFICIAL SEAL 9

PAT SCIIAFFRAN NOTARYPUBUC-OREGON // A Aj_

m c0MMGSON DFW Y 1999 Notary Public i r the State of n Residing at d>$I//A22 My commisIio'n' expires 7M7 /FJ

PGE Response to NRC Request for Additional Information

' Safety Analysis Report for Reactor Vessel Packare NRC OUESTION:

1.

Provide the basis for the statement that the reactor vessel internals comprise 340 cubic feet i of Greater Than Class C Waste. This statement originally appeared in your submittal dated January 30,1997 entitled " Reactor Vessel and Internals Removal Plan",

Specif'Uly, identify the internals in the reactor vessel, the displaced volumes of each internal component, the radionuclide activities for each component, and describe how the radionuclide activities were calculated.

PGE RESPONSE:

The Trojan Nuclear Plant Decommissioning Plan (PGE-1061), Section 2.2.7 DECOMMISSIONING RADIOACTIVE WASTE PROJECTIONS, indicates that 340 A) of greater than Class C radioactive waste from the reactor vessel intemals is included in the conservative estimates of waste volume projections. Also, Table 2.2-4 of PGE-1061 includes the 340 A' of greater than Class C waste for the reactor vessel internals. In estimating the volum greater th'an Class C waste for the reactor vessel internals, it was assumed that the inter

~ be segmented and packaged for disposal with the high-level spent nuclear fuel. The estima volume (340 A') is calculated based on the gross container volume to be shipped and buried, is therefore, greater than the actual segmented volume of the GTCC material in the internals (approximately 92 A8).

The following sub-components comprise the reactor vessel internals:

Core Bame - The core bame consists of a series of axial plates which are attached to the core barrel by the core formers The core baffle assembly provides lateral support for the fuel assemblies en the core periphery, as we'l as serving as a flow bame by directing cooling water up through the core region and limiting bypass flow. The bame consists o rectangular plates 1.125 inches thick,154.94 inches long and of varying widths. The core bame plates weigh a total of 26,644 pounds. The core baffle plates are Type 304 stainless steel with a density of 500 pounds per cubic foot. Therefore the volume of the core baffle is approximately 53.3 A2. The core bame, if segmented, would be classified as greater than Class C waste.

s, providing the

< Core Formers - The core formers are basically structural support meS form for the core bame plates and attaching these plates to the core barrel. Core formers are located at several different elevations along the longitudinal axis of the reactor cort.

At each of eight elevations, the formers consist of four units, for a total of thirty-two 1

pieces. The formers weigh a total of 12,740 pounds. The core formers are Type 304 ,

stainless steel. The volume of the core formers is approximately 25.5 ft'. The core formers, if segmented, would be classifad as greater than Class C waste.

• Core Barrel - The core barrel consists of two major sections, the upper and lower core barrels. The barrelis a right circular cylinder with a nominal inside diameter of 148 inches and a nominal wall thickness of 2.38 inches in the active core region. The activation analysis niodel includes all of the lower core barrel (61,850 pounds) and a portion of the upper core barrel (14,280' pounds). The core barrelis Type 304 stainless steel. The volume of the core barrel is approximately 152.3 fR The core barrelis not classified as Ereater than Class C waste.

- Thermal Pads - The thermal pads are located at four azimuthal angles, attached to the outside of the lower core barrel. These thermal pads each consist of two pieces and are axially centered on the reactor core midplane. Their purpose is to reduce the neutron flux to the vessel wall at locations where the core is closest radially to the wall. The thermal pads are 2.75 inches thick and 149.7 inches long. They cover approximately 135' azimuthal, or about 37.5% of the circumference. The four sets of thtanal pads weigh a total of 20,950 pounds. The thermal pads are Type 304 stainless steel. The volume of the thermal pads is approximately 41.9 fR The thermal pads are not classified as greater than Class C waste.

- Lower Core Plate - The lower core plate suppons the fuel assemblies from underneath, e contacting the fuel assembly bottom nozzles. The plate is 2.00 inches thick and 146.66 inches in diameter. The plate weighs 6.700 pounds and is Type 304 stainless steel. The volume of the lower core plate is approximately 13.4 fR The lower core plate, if segmented, would be classified as greater than Class C waste

. Lower Core Support Columns - The region below the core support plate and above the core support contains the core support columns. Additionally, this region contains columns which ~.mport the travel path ofinstrumentation which is inserted into the reactor core. The weight of these columns is estimated to be 5,109 pounds. The columns are Type 304 stainless steel. Thi.r volume is approximately 10.2 fB The lower core support columns are not classified as greater than Class C waste.

- Lower Core Suppon - The lower core support is a massive piece of metal which supports the entire weight of the reactor core, and some of the internal components. The support rests on radial supports welded to the reactor pressure vessel. The lower core support is Type 304 stainless steel and has a diameter of 151,75 inches, a thickness of 20 inches, and f an overall weight of 60,000 pounds. The volume of the lower core support is approximately 120 fP The lower core support is not classified as greater than Class C waste.

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Region Below Core Support and Above Upper Tie Plate - This region below the lower core support contains finy-six instmment tubes and support columns. The total mass of Type 304 stainless steel in this region was estimated to be 4,072 pounds. The volume is approximately 8.1 n'. This is not classified as greater than Class C waste.

Upper Core Plate - The upper core plate serves as the locating guide for the upper fuel assembly nozzles. The plate is 3.00 inches thick,147.25 inches in diameter, and weighs 7,980 pounds. The upper core plate is Type 304 stainless steel and has a volume of approximately 16 fR The Upper Core Plate is not classified as greater than Class C ,

waste.

Upper Core Support Columns - Above the upper core plate anci attached to it, are the upper core support columns. These columns provide support for the control rod assemblies moved in and out of the core, as well as support for varioes pressure and temperature instrumentation There are forty-eight support columns and sixty-one guide tubes in the region between the upper core plate and the upper support assembly. The mass of Type 304 stainless steelis estimated at 11,569 pounds. This mass does not include all of the mass of the support tubes and guide columns which extend well beyond the upper bound of the analysis models. The volume of the upper core support columns is approximately 23 n' The upper support columns are not classified as greater than Class C waste.

The reactor vessel internals components that, if segmented, would be classified as greater than Class C waste are the Core Baffle, Core Formers, and the Lower Core Plate. The combined volume is approximately 92 ft' As discussed in the Trojan Reactor Vessel Package - Safety Analysis Report, dated March 31, 1997, the activation radioactivity was determined through calculations performed by TLG Services, Inc. (TLG) in support of the Trojan Nuclear Plant Radiological Site Characterization Report. These calculations consisted of one-dimensional neutron transport and point neutron activation analyses of the rea< ..r vessel and its internals. These calculations were performe3 to estimate the neutron-induced radionuclide inventory. The calculations were performed using the FISSPEC and 02 FLUX computer codes, written by TLG, and the ANISN and OPJGEN; computer codes, obtained through the Oak Ridge National Laboratory's Radiation Shielding Infomtation Center (RSIC). Reductior. of tiie output from these programs and ancillary calculations were performed using the ANISNOUT and 02 READ computer codes, wntten by

TLG, and the Microson EXCEL computer code.

The neutron-induced radionuclide inventories were estimated using a two-step analytical approach. The first step was to determine the magnitude and spectrum of the neutron flux beyond

( the boundaries of the reactor core. This was accomplished using the ANISN one-dimensional neutron transport computer code with five iadial and axial geometric models. The results of the 3

adial transport calculations were normalized against plant-specific neutron flux data obtained ,

from an available reactor vessel neutron fluence surveillance capsule report.

The ANISN outputs were s6sequently collapsed into two-energy group formats (fast and thermal) and into a series of ORIGEN2 point activation / depletion calculations. Additional input to the ORIGEN2 calculations included material compositions and historical plant performance data. The radionuclide activities for each component are presented in Table 1 of Calculation RPC 97-018 which is provided as Attachment 11 for information.

NRC OfIERTION-

2. Provide the methodology and results of the waste classification for the pressure vessel with the internals intact. Describe how the waste classif; cation conforms to the recommendations in Section 3.3 of the Branch Technical Position on Concentration Averagine and Encapsulation (BTP), dated January 17, 1995. Demonstrate that the waste classification considered each of the internal compenents as a separate entity in the averaging.

PGF RESPONSE:

The Reactor Vessel and Internals will be packaged as one integral component containing neutron activated metals, The analysis to determine the waste classification was performed in accordance with the general requirements of BTP Section 3.3 which states:

"For neutron activated materials or metals, or components incorporating radioactivity in their design, the waste classification volume or weight should be taken to be the total weight or displaced volume of the material, metal, or companent (i.e., major void volumes subtracted from the envelope volume)."

The activity was averaged over the entire metal volume of the component. The volume utilized does not include voids filled with grout, shielding, closure plates, or impact limiters.

The waste classification was performed using the activation analysis done as pan of the Site Characterization Repon that supported the Trojan Decommissioning Plan. A copy of the PGE analysis that determined the waste 'assification, PGE Calculation RPC 97-018, Revision 0, is provided as Attachment II.

The Branch Technical Position Section 3.9 provides " Alternate provisions" for packaging large intact components. Under the alternatives provision, the licensee's obligation is to demonstrate, to the NRC or Agreement State, that land disposal of the object will meet the performance objectives in Subpan C of 10 CFR Pan 61. US Ecology performed ground-water 4

and direct exposure dose analyses to suppon the disposal approval process oy the State of Washington. The State of Washington, as an Agreement State, has determined that the waste classification of the Trojan waste appears to be consistent with the NRC BTP. The Trojan package, therefore, satisfies the alternative provisions of Section 3.9 of the BTP. A summary discussion of the bounding performance objective from Subpart C of 10 CFR Part 61 follows.

For activated metals, the intruder scenarios represent the worst case dose pathway. Since the intact vessel could not conceivably be handled by an inadvertent intruder, the intruder discovery scenario is the most appropriate. Assuming this scenario, several observations are appropriate:

1. The only long lived gamma emitter present is Nb 94 which is present in its highest concentration: in the core baffle within the pressure vessel. The core baffle contains 2.23 curies of NN94 which is 68% of the total Nb-94 activity of 3.29 curies. The baffle will be shielded by low density cellular concrete (LDCC), the vessel wall, and steel plate.

2. The estimated exposure rate on the surface of the intact vessel after 500 years is less

, than 0.02 mR/hr. When this dose rate is considered !n the context of the appropriate intruder discovery scenario, the objectives of 10 CFR ' Pan 61 Subpan C are clearly satisfied.

The required analyses were submitted to the Washington State Depanment of licalth for review and are provided as Attachment 111. The Washington Stai: Lepanment of IIcalth reviewed the analyses and concluded that "the waste classifica!ien of the Trojan waste appears to be consistent with the Nuclear Regulatory Commission's January 17,1995 Final Branch Technical position on Concentration Averaging and Encapsulation." The letter from the State of Washington Depanment of IIcalth is provided as Attachment IV. The documents referenced above demonstrate that tht. Peactor Vessel with the intemals installed, configured as described in the PGE Safety Andysis Repen. : stisfy the requirements of 10 CFR Pan 61 Subpan C perfonnance objectives. Based on the submittals and approval by the washington  ;

Depanment of Ilealth, the reactor vessel package with the internals installed meets the

, requirements of the Branch Technical Position, Section 3.9 Alternative provisions.

Funher Waste Disposal Performance Objectives Discussion NUREG 0782 Draft Environmental Impact Statement on 10 CFR 61 " Licensing Requirements for Land Disposal of Radioactive Waste" listed four basic performance objectives that should be achieved in waste disposal. These are:

1. Protection of the inadvertent intruder.

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2. Assure long term stability to climinate the need for long term maintenance after '

operations cease.

3. Protect public health and safety over the long term.

4. Assure safety during the short term operational phase.

The degree to which the disposal of the Trojan Reactor Vessel and Internals would meet these performance objectives was considered in the decision to pursue this alternative. Specifically, protection of the inadvertent intruder, long term site stability, and safety during the short term operational phase were all enhanced by the unit disposal alternative as opposed to the segmentation alternative, in particular, the occupational and radiological safety advantages in handling a single package and the minimal impact on long term site perfomiance objectives as compared to handling multiple high dose rate liners weighed especially heavily in this decision. The discussion of the domirant performance objective, protection from an inadvertent intruder, follows to demonstrate that the performance objectives of 10 CFR Part 61 are satisfied by the burial of the reactor vessel and internals as a single package.

Intruder Assessment Scenarios Inadvertent intrusion ass':mes that an individual, or group of individuals, intrudes into the waste either accidentally or without realizing that there is a potential hazard. The former case is considered most likely but is assumed to be quickly recognized by the individual with minimal resulting exposures. More significant exposures are expected to occur if the intruder does not realize that there is a potential hazard. This could occur if there is a breakdown in institutional controls.

There are two possibilities for inadvertent intruder exposures to low level radioactive wastes.

These include the Intruder-Construction scenario and the Intruder-Agriculture scenarin.

Population exposures are also considered based upon waste that is uncovered and brought to the surfaa being tr::n: ported offsite by surface water and wind.

The Intruder-Construction scenario assumes that some time after the end of operations at the facility, institutional controls break down and an intruder inadvertently constructs a house on the disposal facility. The intruder is assumed to dig a three meter deep foundation that is 10 m by 20 m in dimension at the bottom. Exposures are assumed to occur through the suspension of centaminated dust via inhalation and direct exposure, consumption of food 3 grown nearby upon which airborne contamination is assumed to have settled, and via direct gamma exposure to the waste during excavation.

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'!he Intmder Agriculture scenario assumes that an individual inadvertently lives on and consumes food grown on the disposal facility. Farming is a surface activity and generally does not involve disturbing the mil for more than a few feet. As long as the cap of one or two meters is maintained over the waste then it is unlikely that agricultural activities would ever contact the waste. To implemer.t this scenario at the end of the institutional control period, however, a porion of the soll excavated during the intruder construction activity is assumed to be backfilled around the house foundation. The remainder is assumed to be utilized in the agricultural scenario. The house is assumed to be located at the center of a 50 m circle which includes the agricultural area.

The exposure pathways associated with this scenario include:

1. Inhalation of contaminated dust suspended due to tilling activities,

2. Direct gamma exposure from standing in the contaminated cloud,

3. Consumption of food (leafy vegetable ' ' Td by fallout from the contaminated cloud

4. Consumption of food grown in the contaminated soil

5. Direct gamma exposure When assessing exposures from inadvertent intrusion, the physical form of the Trojan reactor vessel must be considered, The Trojan reactor vessel is a right cylindrical carbon steel vessel, 42 fect 6 inches tall and 17 feet 1 inch in diameter that will weigh approximately 950 tons when disposed. Additionally it will be disposed at least 5 m (16.5 feet) below grade. As stated in NUREG 0782, " ,, intmder scenarios analyzed contain one very large assumption -

that the soll' waste mixture in which constmetion or agriculture takes place it more or less inaistinguishable from dirt." That is, the waste has decomposed to the point that the intruder does not know he is contacting waste. This assumption is necessary since without it, the scenarios could not happen.

Given the physical size and composition of the reactor vessel (i.e., 5 to 8 inch thick carbon steel vessel walls, stainless steel internals, and void spaces filled with grout) the only credible exposure pathwc) P.ssociated with the Intruder Construction and Intruder Agriculture scenarios is the direct gammt exposure pathway. This results from the fact that the reactor vessel is not, nor will it degrade, into a form that is indistinguishable from dirt. There is no credible means by which the activity contained within the activated metal of the internals could become tilled

- up or mixed with soil such that jt could become suspended in air or that vegetables could be grown in it. Consequently, only the direct gamma exposure pathway is considered.

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To assess the direct gamma exposure pathway, the dose rates on the exterior of the reactor ,

vessel and internals, as they will be disposed, were modeled using the MICROSHIELD

- computer code. Assessments of dose rates at this point is conservative and appropriate in that it would be representative of an intmder digging down to, but not actually contacting, the l reactor vessel as it lays on its side in the disposal trench. The intruder would then constmet  !

his house on the vessel.

Vessel external dose rates were projected at time or shipment and in increments over the subsequent 500 years. It is important to note that at the projected time of shipment, the vessel will contain an estimated 1.15E6 Curies of Co-60. Due to transponation regulations, the external done rate will be limited to 200 mR/hr at the surface of the vessel. Consequently, the vessel, as shipped and disposed must provide sufficient radiation shielding to attenuate the radiation from Co-60 and other nuclides to acceptable levels. The acceptability of this dose rate for long term disposal guidance is provided in the 1995 revision to the US NRC Branch l Technical Position on Waste Classification. This document assigns a limit of 0.02 mRih4 on the surface of the disposal container as the acceptable dose rate from encapsulated scaled  ;

sources and activated metals 500 years after disposal. 31s limit is met by the reactor vessel j '

and intemal package 100 years following disposal.

General Discussion of the Merits of the Two Options For the Reactor Vessel and Internals Removal Project There are two alternatives for disposal of the Trojan Reactor Vessel and Internals. These alternatives are segmentation, placing pieces in individual steel liners, then disposing of the liners; or leaving the internals in the Reactor Vessel, filling the reactor vessel with a low density cellular concrete, and then disposing of one large package. Segmentation of the Reactor Vessel Internals would most probably use a plasma-arc torch (or equivalent equipment) to cut up the subcomponent pans that comprise the internals. The cutting would be conducted in the water filled refueling canal in Containment utilizing underwater tooling.

• ihe internah would be ct., into pieces strll enough to b( placed in shipping casks for transpon to the disposal facility. The liners would be removed from the transport cask at the disposal facility and placed in venical disposal caissons, engineered concrete barriera, or radioactive waste disposal trenches depending on the isotopes and activity contained in each liner. A study of the segmentation indicated that it would result in as many as 44 liners for  :

disposal at the US Ecology site in Richland, Washington.

Disposing of the Internals By Segmentation Segmentation would result in some internal pans being classified as greater than class C waste (GTCC). A total of 39 to 44 GTCC storage cans would be generated from the segmentation 8

pi cess. These cans would remain at Trojan until the US DOE develops a disposal option. A byr duct of the plasma arc method of segmentation is " dross" (i.e., welding slag and fine panlues) generated during the cutting process. Dross consists of the molten and vaporized metal from the cutting process that has condensed and solidified in the refueling canal water.

This material is " caught

• in buckets located below the piece being sectioned or is removed from the refueling canal water via mechanical filtration. The majority of this material is not expected to be classified as greater than class C waste (GTCC). This dross and filters would be sent for disposal in shipping cask liners.

Contamination control is one of the radiological hazards associated with this approach. The tasks of filling intermediate and high activity liners with waste and placing the liners in the transportation casks are performed underwater in the refueling canal or spent fuel pool Due to the high radiation levels associated with these liners, hands-on decontamination of external surfaces is not possible. Contamination levels on the liner can range from lES to > IE6 dpm/100 cm'. Ilot particles created during reactor operation or during the segmenting process will also be present. Consequently, there is substantial potential for the spread of contamination when the liner is lifted from the cask at the disposal facility or when cleaning gasketed cask closure surfaces.

Disposal of low, moderate, and intermediate liners involves unloading the liner from the transportation cask and placing it in the disposal trench or in a concrete caisson. This operation is generally performed utilizing long handled tools with the unloading technician remaining in as low a radiation area as possible. Typical personnel radiation exposures associated with unloading these shipments ranges from 50 to 150 mR for each cask.

Unloading high activity casks is a far more demanding task due to the extreme radiation levels involved and the severe consequences that would result from a mishap during the unloading process. liigh activity liners are unloaded from the shipping cask vertically and are disposed of in concrete caissons constructed in the disposal trench. While these operations have been performed many times with minimal radiological impacts because each evolution was carefully plannC t.nd executed, the potential, however, exists for significant radiological hazards to oe created in the event of an equipment malfunction or handling mishap.

Additional disadvantages are also p.esent with respect to long term facility closure. Cask liners filled with reactor intermi sections (that are not classified as GTCC) contain void spaces which are undesirable from a waste form perspective. Void spaces in waste packages may lead to channeling of rainfall percolating through the waste and may 1:ad to waste slumping. The disposal caissons are designed to minimize the effects of void spaces inside the disposal liners, however, elimination of void spaces is preferable with respect to the waste form. The dross generated by segmentation is also problematic in that it consists of solid particles ranging in size from macroscopic slag down to fine particles 0.5 microns in diameter. While they would be solid 9

1

particles, it is not clear at what size they would be considered to be an inherently stable waste ,

form such as irradiated metal. In the event ofinadvertent intrusion into the waste at some time aAer institutional controls have lapsed, the form of the dross and the pieces of segmented internals is such that it could be relatively easily transported to the surface and dispersed.

Potential hazards associated with segmentation can be controlled and minimized through proper radiological controls plarming, however, these issues must be recognized as distinct and significant disadvantages to this method of disposal.

Reactor Vessel and Internals Removal As a Single Package Disposal of the reactor vessel with internals in place as a single package has sigelficant '

operational and disposal advantages.

The Reactor Vessel and Internals will be received at the site as one package. The exterior of the reactor vessel is the outside of the shipping container. Consequently, external dose rates will be less than 200 mR/hr on contact with the vessel. Since the external dose rates will be less .s;:. 200 mR/hr on contact, the radiological hazeus associated with off loading will be minimal. The external surface of the reactor vessel will also have to meet DOT contamination limits, consequently external contamination levels will be sufficiently low to prevent any spread of contamination during handling and disposal.

The reactor vessel will be filled with Low Density Cellular Concrete (LDCC) to minimize void spaces. The LDCC will be pumped into the vessel and will be allowed to set up. Additional grout will then be pumped into the vessel to fill void spaces to the maximum extent practicable. Since the unit will be disposed by itself in a single trench, it will not have any effect on other waste at the site. The structural strength of the reactor vessel will preclude changes in the trench such as slumping. Since the internals will not be cut up, what would have become GTCC waste and dispersible dross will remain inside a heavy walled vessel.

Overall, disposing of the Reactor Vessel and Internals as a package provides distinct advantages in terms of operat;onal safety and waste form.

A separate disposal trench will be constructed at the US Ecology site to receive the Trojan reactor vessel. Constructing a separate disposal trench does not impact the life of the disposal site since it is expected that the facility will have at least 50% umced capacity at the end of the facility lease in 2063. There is only one other nuclear power plant located within the Nonhwest and Rocky Mountain Compacts and continued waste receipts beyond 2063 are not currently planned. The trench will be constructed with a long access ramp to minimize the grade the vessel transporter must negotiate. The access ramp will terminate at the bottom of the disposal trench. The vessel will be transponed to the US Ecology site and down into the 10

disposal trench as single package. The package will then be unloaded into its final resting position.

Conclusion PGE has performed an extensive assessment of the storage and disposal options for the reactor vessel and internals. The selected option of transportation and disposal of the reactor vessel-and internals as one integral component (containing neutron activated metals and stabilized by low density cellular concrete) is the best available solution from the perspective of protecting the public health and safety. The analyses performed and summarized in this response demonstrate that the radioactive material in the singular reactor package is classified consistent with regulatory guidance and can be safely buried at the licensed US Ecology radioactive waste disposal facility.

11 4

ATTACIIMENT 11 l

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

• • • QA RECORD WREN C(ftPLETED ***

7.8 oo,P - . s. ~ u . ,

,--. Letter Number n System Number ma leumber of Pages 1 Document Date Calt. Reference #

TROJAN CALCUIATICH COVER SEEET Shoet I cont'd on Shoet .$:

Title REAc1pt Pfisti fantre sa leess'e r,'c ari..) AAtatu u t Trojan Nuclear Plant calculation No. ftPc 47-ort Structure 6urabarar Supersedes Calculation No. 0 6E/W _ Quality-Related Yan %

System Component Aret. Wasal Status: / Final Interia References (FMR/DFMR, SPEER, MR, FSC, etc.)

Bas Been Changed by Or Revision has been Responsible Affected Identify Change Deferred by (Identify Supervisor /Date Memo , CTL, e tc . ) (Deferrals only)

Document No. Vehicles (MR, DFMR, DCF, .*CT, SPEER, PSC, etc.)

Calculation Objective y berea4ws The pr. pre. unsr s closeteuarsJ of 74c sisse rn e. Veu rt ve rk all ers a r ttH

• t sek.wMe.m os ea s 04 r.

Revision Description Rev. Approved By Date No. Preparer Date Verified By Date

<"brk , Ylf ?

O- Mduln4st* Sbs /ss w l t

~ TPP 18-9 Attachment 2 Revision 2 Page 1 of 1- Page 11 of 11

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

PORTLAND GENERAL ELECTRIC CALCULATION SIIEET ,

Revision O 2- of i Calculation No. EPC O 0I1 , Sheet .

Preparer #I. Mad Mua nce d

• Date Nt 7 /f t Verifier V . Y. in,. Date c.l n lo i Table of Contents Sheet No.

section 1

Calculaten Cover Sheet. . . . ... . . .. . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2 T able of Contents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3 Objective.......................................................................................................

• 3

. .cceptance C rite ria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3 Summary.......................................................................................................

3 Assumptions , De sign inputs , and M ethodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4 Results..........................................................................................................

4 References......................................................................................................

4 Body o f C alcul a tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

RP 310 Forms:

1. Form RP 73

2. Form RP 78 Attachments:

1.QPRO Spreadsheet

- PORTLAND GENERAL ELECTRIC CALCULATION SHEET Calculation No. kN' 97-0'I Revision 0 Sheet 3 of I preparer VldLuI MuLhot4 Date $lHl9i Veritier iin Date rlnh >

/ '

REACTOR YESSEL WASTE CLASSD'ICAT10N ANALYSIS ORHrnON The objective of this calculation is to determine the proper waste classincat'an of the Reactor Vessel witn all its internal sub-components as one unit.

. ACCEPTANCE CRHERIA None

SUMMARY

r De purpose of this ca..ulation is to aid in the planning for the eventual shipping for burial of the reactor vessel, nis calculation can be used as a guideline when an actual shipping date is determined. De activities will need to be decay corrected when a date is known.

De approach to determine the waste class will be in accordance with 10CFR 61 and the IIS Ecoiogy burial license.

The activity is obtained from the activation analysis done as part of the Site Characterization in support of the Trojan Decommissioning Plan and RPC 96408, " Reactor Vessel and laternals Surface Area Contamination". The calculation will be donc under the guidelines of RP 310. Rev 2.

ASSUhurnONS. DESIGN INPtHS. AND ME1110DOLOGY Both the Activation and corrosion activity represent activity decay corrected to 11/97.

In accordance with section 3.1 of the Branch Technical Position, the reactor vessel and core internals is considered to be one component containing neutron activated metals incorporating radioactive in its design thus allowing concentrating averaging over the displaced volume of the material.

De displaced volume is the mass of the metals only as identified in Table of TLO's calculation titled

'RVAIR Weight and C. G.* as received under PGE letter No.102 97L. This does not include any closure plates, impact limiters, or shielding.

De majority of the activation is located in the region between the upper and lower core plates including the vessel cladding and walls.

Tbe surface contamination is considered to be distributed over all the reactor vessels' internal surfaces.

Although < 1% of the total activity, the surface contamination will be considered in this calculation.

The contribution from the incore Flux Dimbles, currently in place in the vessel, has not been included in the overall total activity but is assumed to be less than .05% of the total activity.

PORTLAND GENERAI, ELECTRIC CALCULATION SHEET .

Calculation No. kk 97* o 'I Revision d Sheet N of Y Preparer Acharl M utb ac. C ' Date _ S I'1 /f 7 Veritter na Date chth?

/

Displaced Volume of Metal = 2622.2 ft' or 74,266.3'i8 cc ( reference TI.O's

• Weight and C. O.

Calculation)

Material Weight = Displaced Volume x 7.86 gms/cc or $83,733,653 gms Activation Activity = 2.007 curies ( See Att. 1 , taken from table 4.7.30 of the alte Characterization R8 Port)

Corrosion or surface area Activity = 155.2 curies ( from Table 1 of RPC 96408 )

RESULTS:

The reactor vessel and sub components are Class C waste.

Table I results = .328 of Class C limit "

Table 2 results = .303 of Class C timit -

SNM = 1.55 gms of Pu '

Total Pu = 5.25 curies -

REFERENCES:

1. RPC E008, Reactor Vessel & Internals Surface Area Activity

7. RVAIR 102-97L, PGE Tracking No. of TLO'. Ta' AIR Weight and C.O. Calculation

3. RP 310 Rev 2 Determination of Radioactive Material Shipping and Waste Classifications.

4. Table 4.7.30 of the Trojan Site Characterization Report SODY OF CALCULATION:

1. RP Form 73 & 78 applicable pages of RP 310

2. Attachment 1. QPRO Spreadsheet of Table 4.7.30

Form MP 73 SNpment Number N/A

' Page 4 of 6 Pedago Nutrder Reackw Vessel i 12 Determenshon of Concorfraten Metenet Volume e 7.43E+07 oc / MalenalWeepht e 6.84E+00 pm /

D E A B C lectope SNp Date Conc Conc AetMtv lootope lootope (mCO <5 yr T 1/2 >5 yr T 1/2 (uCVoc) (uCVoc)

(Ax1E3pD (Ax1E3W H-3 a.11E+0t , 1.09E43 H-3 (A) 4 deE+06 - 6 00E+00 C 14 1.18E+02 , 1.SSE43 C 14 (A) 2.10E+06 , 2.93E+00 Mn 54 857E+01 - 1.1SE43 Cn-64 (A) 2.10E+06 , 2 91E+01 Fe-S$ 2.77E+04 . _3 73E41 Fe66(A) 8 97E+06 - _9,39E +03 Co40 9 92E+04 , 1.34E+00 Co40(A) 1.16E+09 , 1.65E+04 Nb.94 (A) 3 29E+03 , 4 42E 02 N C (A) 9$3E+05 - 1.28E+01 Sb 125 167E+03 , 2 25E 02 tb 125 (A) 4 03E+02 $ 42E 03 Eu162 (A) 1.49E+04 2 00E 01 Co-144 4 64E+01 - 6 26E44 F G H 1643 2.00E+04 2.69E 01 Conc. Conc Grom N43 (A) 1.57E+08 212E+03 lootope Factor of Sr 90 9 24E+02 , 1.24E 02 (nCVgm) (grWmCQ looto;e Tc W (A) 7.06E+02 9 60E43 (Ax1E6)E (AxO)

TRO 1.41E+02 - 2.43E43 3.09E 01 Pu-238 8.36E+01 - 1.12E 03 1.43E 01 6.6E45 4.84E 03 Pu-239/40 9 33E+01 - 1.26E 03 1.60E 01 1.60E42 1.49E+00 Pu-241 607E+03 6 83E 02 8 69E+00 9.60E 06 4 87E 02 Cm-242 1.66E 02 . 210E 07 2.67E45

. Totals . 2 01E+09 / 9 42E+03 1.76E+04 1.55 Sum A Sum B Sum C Sum H (mCO (uCvec) (uCVoc) (gms.of Pu)

Attachment 1 - Prepared by M$. hapl Murdock RP110 Page 22 of 37 Checked by M;',,m Rev.2

/ '

Page $1 of 82

1. Mchael68urdock Venty that this is e true corwuter generated copy of RP 310, Rev 2 l

Form RP 73 Shtsnent Hunwr N/A Page 6 of $ Packapa Number Reactor Vesse'

13. Evaluatlan of resuRs for Bunal See Limnations and posstde DOT /NRC Form 741 Shipment.

a. Determ6ne total grams of SNM.

1.65 + 0 + 0 = 1.65 Orams of Grams of Orems of Grame of Pu U 235 U.233 $NM If $NM > or a 1 gram, notify Radweste Supervsor.

It Burlet Sne TRU is TRU wth T.1/2 >$ yrs ercopt Pu 238 Pa 239/240, Pu-241 and Cm-242 from Column F.

Surtal Gne TRU: 3 09E41 LMt <10 nC#gm (nCvgm)

c. Weste Class TRU 6s TRU wth T 1/2 >5 yrs and PU-238 and Pu 239f240 from Column r'.

Waste Ctssa TRU.

3 09E41 + 1.43E 01 + 1.60E 01 = 612E 01 (TRU) (Pu-238) (Pu 239/240) (nCvgm)

14. Sum all isotopts wth half ido <5 yts.

942E+03 Lima <7 UCvcc (uCvec)

16. Sum AN isotopes wtth half ide >5 yrs:

1.76E+04 Lima <1 uCVcc (uCVcc)

Note: If any of the above laruts are noeeded, nottry the Radweste Specialst Memel Murdock Prepared by Checked by w '

/

Attachment 1 Rp 310 Page 23 of 37 Rev 2 Page 52 of 42

l. MchaelMurdock Venfy that thrs is a true computer generated copy of RP 310, Rev 2

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

h0RM RP 78 ' Shipment No. MIA DETERMINATION OF Package No Reactor Vessel WASTE CLAS WICATION Class A Class B Class C A B C D Tath i lootopic Conc. Uma Ouotent LimR Ouotent Lima Ouotant lootopes (uCVoc) (uCrec) (AS) (uCycc) (AC) (uCvec) (A/D)

C 14 1.55E 03 8 00E 01, 1.94E 03 (b) 800E+00 1.94E 04 014 (A) 2 93E+00 800E+00 3 67E 01 tc) 8 00E+01 3 67E42 To 99 _ _ 9.50E 43 3 00E41 3.17E 02 (b) 3 00E+00 3.17E 03 C 59 (A) 1.2SE41 2.20E +01 6 63E 01 (b) 2.20E+02 6 83E 02 Nb44 (A) _4_42E-02 2 00E 02 2.21E+00 (b) 2.00E 01 2.21E 01 TRU (a,c) 6.12E41 100E+01 6.12E 02 (b) 1.00E+02 6.12E 03 Pu-241 (a) 880E+00 3.90E+02 2 44E42 (b) 3 50E+03 2.46E 03 Cm-242 (a) 2 67E-06 2.00E+03 1.34E 08 (b) 2.00E +04 1.34E 09 Sum of Tath 1 Quotants 3 28E+00 3 28.E 01 NOTE: TRU wth T 1/2 >$ yrs e all TRU except Pu-238, Pu-239/240, Pu-241 and Cm 242 CussA C4ss D CWss C A D C D isotopic lable 2 Conc. LimM Ouotent Lima Ouotent Lima Quotent lootopes (uCVcc) (uCycc) (A/B) (uCVcc) (A/C) (uCvec) (A/D)

H3 600E+00 4 00E+01 1.50E 01 (d) N/A (d) N/A Co-60 1.55E+04 7.00E+02 2 21E+01 (d) N/A (d) N/A N>63 2 69E-01 3 50E+00 7.69E 02 7.00E+01 7.00E+02 3 BSE 04 Ne 62 (A) 2.12E+03 3 50E+01 7 00E+02 7.00E +03 3 03E 01 St90 1.24E 02 4 00E 02 311E 01 1.50E+02 7.00E+03 1.78E 06 C&137 1.00E+00 4 40E+01 4 60E+03 Nochdes w/

(T 1/2.<5 yr) 9 42E+03 7.00E+02 1.35E +01 (d) N/A (d) N/A Sum of Table 2 Cuotents 3 61E+01 3.03E 01 a . Unas aro 6n nCV9m b . lf Class A hmat is exceeded, the waste is Class C or greater,

c. TRU c TRU wth 7 1/2 >$ yrs + Pu-238 + Pu-23nr240.

d . No Lima.

NOTE: It O not necessary to hst on the mandest any nucide whose Class A quotent is less than 0 01 except C 14, Tc 99,1129. and H-3.

Prepared by h9chael Murdoc6 Checked by @/h ,

Attachment i / RP 310 Da0e 36 of 37 Rowson 2 Page 65 of 82 1 Mchael Murdock Venfy that this as a ifue computer Generated copy of RP 310, Rev 2

PORTLAND GENERAL ELECTRIC CALCUALTION SHEFT RPC 97-Ol8 Revision 0 Sheet 8 of 8 Csiculation No.

Mir*aelMurdock M Dale 05/27/97 Preparer Verifier ,f ,,

Oate ,5-/m /, ,

( ,

Table 1 REACTOR VESSEL CURIE EON TENT FIVE YEARS AFTER SittTa:7WN (ACITWAT10% ONLY)

Ma-54 Es-I52 Fe-55 Co40 Ni-59 N643 7494 Tc-99 Totals SimC)MPONENTS 11-3 C-14 Sb-125 3320E45 1.632E+03 7 631E43 3.712E+05 7 24EE+05 5329E+02 8688E+04 2.225E+00 5.091E4' 3.187E+06 Core Bame 2336E+02 1.186E+02 2.Il0E+02 1.180E45 2293E+05 2.412E*05 2 614E+02 4 78tE*04 5 682E45 7 610E42 5Is9E+05 Core Formers 8 755E*01 7.010E+01 5.108E42 1379E43 1633E+02 7 867E+00 3.733E+04 8 555E+04 6319E+0i 9.l~4E+03 2.409E41 6102E42 13231.+05 tower Core Banel 5 465E+01 1.183E+Cl 1339E43 7.415E45 3318E45 8 063E45 6144E44 8 629E42 2170E-06 5.7621- T I247E+00 Upper Core Barrel 5.151E44 1.IISE44 1.488E-08 2.003E44 3 919E+0i 2.997E+00 8 933E*C3 2054E+C4 1367E+0! 2.193E+03 5 786E-02 1.46 % 2 3178E+04 Thermalpads 1303E+0i 2 832E+00 6.640E45 6.790E+00 5358E41 dl32E+03 6165E+03 7374E+00 1.02tE+03 1.414E42 2.584F43 Il34E+04 VesselClad 5.786E+00 1265E+00 1.813E+03 3 928E+02 IJ3*E45 1877E+08 1.753 E43 4 445E43 2 238t+03 Vessel Wall 4.198E+00 6.222E43 IJ32E42 8041E+00 1142E+00 4323E-03 8.017E+01 3.715E-01 3 022E+04 5.843E+04 4 757E+01 7 530E+03 1351E45 2E4E42 8895E+04 lower Core Plate 2.918E+0i 9392E+00 4546E41 6 052E+03 7 419E+03 1.136E+0! 1304E+03 1.456E42 I 733E43 I$001+04 tower Core Sep Col 8758E+00 IJ29E+00 5.465E45 4 417E+00 7.77tE-03 1386E+01 2.919E+0! 2.490E42 3 422E+00 7.907E45 2 010E-05 4 65SE*01 lewes Core Sep 2.018E42 4356E43 1.630E-10 5103E42 0 000E+00 3 390E-06 2107E-06 4.940E43 60CE43 9376E-06 Il39E43 1.168E-08 1396E 09 Il27E-02 Delow le= Core 5mp 7.563E-06 1.492E46 5377E+03 9 988E+03 9 612E*00 1323E+03 2.586E-02 5 851E-03 1.672E+04 Upper Core rlate 7.582E+00 1.673E+00 Il91E44 1,555E+0! I.037E+00 3272E-01 9 077E+02 Ii29E+03 1.754E* 30 2.276E+02 2_207E-33 2.700E44 2169E+03 Upper Core Sup Col IJ99E+00 2.744E41 6327E47 6787E41 1.487E+el 6.973E+05 ll49E+06 9332E+02 1573E+05 3286E+00 7 056E-01 2007E+06 Totals 4 457E+02 2.178E+02 4.028E41 2.16IE+03 NOTE: 1. The activity is %n the Site Characterushoe Report Table 4.730

2. Nucl' ales omitted were; Ar-39.Co-41.Co-45.Se-Il9m. and Te425m due se such a small contirbation and not a factor la wasse c*assification l

. e

e ATTACitMENT 111 l

l

v s ^8 m.emens wasNayionC313 Mwsn.te11

)

USEcology 6R Amerecee italogy tempsny March 12,1996 Gary Rober. son Head of Waste Management Section State of Washington 1 Depanment of Health Division of Radiation Protection Airdustrial Center, Building 5 PO Box 47827 Olympia, Washington 98504 7827 .

Dear Mr Robenson:

US Ecology and Ponland General Electric would like to take this opponunity to address some of the issues penaining to the disposal of the Trojan Reactor Vessel at the Low. +

Level Radioactive Waste Disposal Site near Richland Washington We are specifically requesting the Depantnent's concurrence on the waste classification of the reactor The reactor vessel package will consist of the vesselinternals and the reactor pressure vessel as one component. The vessel will be eenified as an NRC Type B package. We believe this package meets the recuirement for classification as a Class C Stable waste

' form. Void spaces will be filled to the maximum extent possible with a concrete based grout. These items are discussed in detail below in addition to issues of concern ra our January 25 presentation.

SlahilltX Washington Administrative Code 246-250 050 specifies the stability requirements for radioactive waste disposal. Additionally, NUREG-0782 (Draft Environmental Impact Statement on 10CFR61, Volume 2) provides the same guidance for stability for disposal purposes. Specifically section 5.5.2.4 of the NUREG states that stable waste forms m maintain their physical dimensions and consistency under the conditions of compressi load, radiation, and blodegradation expected to be encountered in disposal. This is to preclude slumping, collapse or other failing of the trench cap; the need for active lo term maintenance nr.hhe ab lity to predict long term performance. Under the section

" Form of the Waste as Generated", activated steel from nuclear r(actors is given as the examples of waste that meets stability, p ,,, ,,,,

. - -..y + -

Page 2 0f 3

. Arthur J. Palmer Ill, CHP to Gary Robenson, Head of Waste Management Section March 12,1996 With regard to the Trojan React:r Vessel, the physical dimension and cotisistency will be maintained under the compressive load of shallow land burial due to the inherent construction of the vessel. The carbon and stainless steel along with the concrete grout will form a solid structure with moir than aderpate s ength to prevent any deformation.

. Radiation effects as far as mainta'.ning physical dimensions and consistency, are not a concern since the vessel was designed to withstand much higher radiation fields when the reactor was in operation. Blodegradation is also not a concem because the vesselis steel.

Since the proposed disposal method meets criteria for stability and void reduction, due to the waste being irradiated metal ano void spaces being filled with a concrete grout, we do not believe a specific topical report should be necersary.

Niobium 94 Total Nb 94 contained in the reactor vessel un * .:.!t troposalis 3;90 curies. The breakdown of the individual subcomponents within the vessel are pres:nted in the table below.

Subcomponent i , Activity (Cl)  % of Total Activity Core Bame 1 2.230 68 %

Core Former 1 0.568 17 %

Lower Core Plate 0.135 4%

Remainder 0.360 11 %

TOTAL l 3.290 100 %

Under tht proposal where the reactor vesselinternals are segmented and shipped in individualliners the total niobium cor. tent would be 0.360 curies. It should be noted that Nb-94 is genera!!y considered to be an extemal dose haurd, Due to the thickness of the reactor vessel walls and external shielding combined with the shielding pr:vided by the grout used to fill the void spaces, external dose rates on the outside of the vessei el be less than 200 mr/hr. The combined shielding is necessary due to the package containing an ertimated 1 E' curies of Co-60 at time of shipment. Given that the package will adequately reduce the radiation levels from the 1 E' curies of Co 60, it will be more than sufficient in minimizing intruder dose in the future from Nb-94.

Disnosal Trench To nrovide for the segregation of class A wastes, any future trenches must maintain total separation of stable and unstable waste forms between trenches. Presently, the cpen Class C trenches do not have sufficient space to efficiently dispose of the reactor vessel. Trench 12 will be constructed as the new stable trench which will have the space to accept the reactor vessel. A portion of the trench will be constructed with a ramp to allow the heavy

- haul trailer access to the trench bottom. The ramp will have a slope of about 4-6% slope with no area greater than 10% slope.

Page 2 ef 3 Anhur L Palmer III, CHP to Gary Robertson, Head of Waste Management Section March 12,1996 Source Term An individual pathway analysis, to determine the impact ot' the increase in the site's  ;

source term with the disposal of this component, has been. completed. This analysis, entitled " Trojan Reactor Vessel Dose Analysis"is enclosed as Attachn.ent I to this letter.

Y.91d1 The grouting process of the Trojan reactor vessel will be the same process used on the Trojan steam generators. The steam generators were filled with a low density grout in containment through several fill and vent connections. Approximately 2 days later the generators were moved out of containment and inspected for additional voids resulting from settling. An approximately 5% void was refilled using the same grout prior to f.aal '

closure olthe generators.

License Variance A License variance request will be submitted to exceed the possession 1;mit of section 6.A.

(60,000 curies) of the site license %%1019 2. The variance will be temporary to cover the reactor vessel shipment only. The variance request will be for approximately 2,100,000 curies.

Waste Classification Attached is a final curie content and waste classification for the vessel. As you can see from the attached calculation sheets provided by PGE, the waste will be Class C, This information is based on the latest data available but is subject to revision once the vesselis removed and can be better characterized. Any revision to the final durie content will most likely be in the downw4 d direction sir " the present calculations assume the most conservative assumptions.

Your timely review and comments with this matter woulo greatly be appreciated. Please contact me at (800) 567 2372 ifI can be of any further assistance.

Sincerely Anh J. Palmer III, CHP

- Chief Radiological Control

& Safety Officer v

4 9 e TROJAN REACTOR VESSEL DOSE ANALYSIS 4

9

w p M f w ~ua US Ecology TABLE OF CONTENTS .

1. GROUN D WATER P ATHWAY DO SE AN ALYSIS .....................................................

1,1 PURPOSE.............................................................................................................1 1.2 APPROACH..........................................................................................................1 1.3 DATA...................................................................................................................1

1. 3.1 S o u r c e T e rm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1. 3. 2 A s s u m pti o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1.4ANALYSlS...........................................................................................................2 1.5

SUMMARY

............................................................................................................3 2.- DIR E CT EXP O S U R E D O S E AN ALY S IS . . ... .. . . . . . . .... . .. .. . . .... . ... .. . . .

2.1 1 N T R O D U C T I O N -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. 2 C O N C E P TU AL M O D E L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.3 APPROACH..........................................................................................................4 2.3.1 S election and J ustification of Mod eI . ............. .... .... ....... .................................. 4

2. 3 . 2 A s s u m p t i o n s , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . .

-2.4 DATA...................................................................................................................5 2.4.1 Are a l G e o m e try . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.4.2 Clad ding and C ove r Thick ne s s . . . .. . . . ... . . . . .. ... ... . . .. . ... . ... .. . .. ... .. . . . . .. .. . . . . .. .. .. ..

2,4. 3 Wa st e a nd C ove r D e nsity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.4. 4 S ou r c e An aly sis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. 4 . 5 . D a t a .S u m m a ry . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. 5 M O D E L S IM U LATI O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2.6

SUMMARY

............................................................................................................7 March 11.1996 i

Trojan Rx Vessel US Ecology

1. GROUND WATER PATHWAY DOSE ANALYSIS 1.1 PURPOSE This calculation examines the expected dese via the ground-water pathway attributed to the disposal of the Trojan Reactor Vessel as a single component at the US Ecciogy Low Level Radioactive Waste Disposal Facility in Richland, Washington.

1.2 APPROACH The calculation follows the methodology used in the Dose Analysis for the 1996 Closure Plan. The models used in the analyses are described in that plan. A simplified description of the analysis is as follows.

infiltrating water (from The metal reactor vessel is buried intact in the trench.

precipitation) that migrates through the cap ccmes in contact with the reacter vessel,

~ne leachate migrates downward to the and leaches radionuclides from the vesses.

water table, where it is diluted in the uppermost aquifer. A hypothetical well withdraws the water for a subsistence farming family, that irrigates a vegetable garden and pastureland for a dairy cow. The primary dose for this pathway is via ingestion.

1.3 DATA The data used in this analysis are presented in Table 1; support for this data is presented in other calculations and in the Closure Plan, and referenced in Table 1.

1.3.1 Source Terrn The activity inventory for the analysis comes from Portland General Eiectric. The reactor vessel activity is presented for November 1,1997, 5 years post shutdown. The activity is separated into surface contamh-ilon and activation, with the activation totaling 2.01x10' Curies (Ci), and surface contamination totaling 3S7.9 Curies.

The development of the source term for the analysis follows the method used in the Closure Plan. Only long lived, high activity isotopes are expected to remain in any significant quantity after migration through the vadose zone. The cut-off for isotopes was for half-lives equal or greater than 0.1 times the travel time through the vadose zone, and activities greater than 1 Curie. Additionally, Sr-90, an isotope which is easily uptaken by humans, was included.

Ten radionuciides were selected from the activity inventory for the source term for the ground water pathway analysis. Several other isotopes were included in the analysis for comparison with the Closure Plan. The source term for the reactor vessel is presented in Table 2.

March 11,1996 1

Trojan Rx VesnF US Ecology 1.3.2 Assumptions .

The assumptions used in this analysis are as follow.

The peak concentrations of leached constituents in the leachate are assumed This to reach the ground water at the same time, and to reach the well at the same time.

assumption is conservative because it maximizes the exposure value to the hypothetical individual.

There is no time delay associated with the leaching. This assumption does not account The for the gradual leaching and removal of radionuclides from the reactor itself.

assumption is that the activity is distributed uniformly, and can be leached uniformly from the reactor vessel. The assumption is conservative because the radionuclides will leach from the vessel slowly, thereby decreasing the amount of radionuclide available for transport.

Solubility and distnbution coefficients frem the Closure Plan will be used. ine leaching concentration of radienuclides from the metal reactor vessel are likely to be lower than the solubility of the racionuclides. but sinct. ca cliier values could be located, these conservatively larger values will be used.

The assumptions described above represent an upper bounding condition on the expected dose attributed to the disposal of the Trojan reactor vessel via the ground- ,

water pathway.

1.4 ANALfSIS The transport of the source term isotopes through the vadose zone was modeled using the TRANSS program. Two infiltration rates were modeled,0.2 inches / year and 0.05 inches / year. These two cases simulate the infiltration of the area surrounding the Facility (natural conoitions, ss if the final cap was completely ineffectual), and the conditions over the Facility with the final cap in place and functioning as designed, respectively. Tijese infiltration rates are the result of natural and expected precipitation at the Facility.

The maximum concentration for each radionuclide was selected for the input to the aquifer. Mixing and dilution occurs in the aquifer. The flow through the aquifer is greater than the recharge rate from infiltrating water, therefore dilution of the loachate occurs. The dilution factor is dependent on the infiltration rate, and is c:ri:ulated to be Multiplying the about 0.003 for 0.2 inches / year, and 0.0007 for 0.05 inches / year.

leachate at the ground-water table by the dilution factor yields the expected l concentration in the hypothetical well.

l 1

Well concentrations less than 1x10* PC!/L are eliminated from further analysis. This left seven radionuclides for dose analysis. The isotopes for dose analysis are listed in Table 3.

l March 11,1996 2

. 1.5

SUMMARY

The well concentration becomes the input to the PRESTO il computer code for dose analysis. The results of the dose analysis are presented in Table 4. These output data are presented on the fourth page of the PRESTO il output, under the selected individual dose equivalent (in mrem / year). The doses are small, and are less than 1 mremlyear to any organ.

The PRESTO Il model did not identify dose factors for Nb 94 in its internal database.

The dose for Nb 91 was calculated by the equations upon which the PRESTO Il model is based. The niobium dose is checked in another calculation. The dose calculated from Nb 94 was far less than 0.01 mrem / year.

2. DIRECT EXPOSURE DOSE ANALYSIS

2.1 INTRODUCTION

This calculation examines the direct gamma exposure pouible from the disposal at the Richland Facility of the intact Trojan Reactor Vessel The reactor vessel is to be shielded by a soil cover. The exposure of an individual while outside man-made structures was examined.

2.2 CONCEPTUA!. MODEL The conceptual model for the direct exposure calculation is that an intruder could be For tnis exposed to direct gamrna radiation from the waste buried in the trench.

analysis, consideration is given to the following potential scenarios:

1. At closure, a Facility operator / maintenance person may receive direct exposure while standing directly over the trench;

2. After closure, a person standing at the Facility boundary may receive direct exposure from a capped trench;

3. At some future date after closure, an inadvertent intruder rnay receive direct exposure by intermittently passing over the Facility atea and receive direct exposure via this pathway; and

4. At some future date after closure, an inadvertent intruder may receive direct exposure by living within a structure constructed into the waste and/or cover (e.g. the basement scenario).

The Facility is located in the area identified as the Central Plateau. The f'ndings of the Hanford Future Site Uses Working Group has identified this area as the location for waste storage from cleanup activities from the rest of the Hanford Reservation. One 3

March 11.1996

US Ecciogy - ,

pathway, direct exposure via a basement construction ,

the Facility. The Central Plateau region, in which the Facility is located, is anticio to be the waste storage area for the cleanup activities to be performed at the Hanfor reservation. As such, the Facility area will have restricted access as a result of the Therefore, the basement construction scenario was larger scale storage area.

eliminated on the basis of tho following assumptions, e Institutional control for the Central Plateau area will eliminate long term access (e.g. no residential or commercial construction) to the Facility; and The elimination cf on site construction will eliminate the possibility of people living over the Facility.

Therefore, the bounding (highest exposure) case for the above scenarios is the first case involving the Facility enarator/ maintenance person (recepter), j The conceptual model for this pathway censiders the radiation source to be a large The receptor is solid mass, since the source is the waste in the burial trenches.

assumed to be standing on top of the trench cap, which serves as a shield to direct i

radiation.

The effects of cover thickness en the direct exposure dose rate were '

examined, The analysis performed for the Facility was undertaken to provide conservative radiological impacts to a hypothetical maximally exposed individual. The

~

assumptions and data are considered extremely conservative for the conditions at the Facility, and are discussed in the fcilowing sections.

2.3 APPROACH 2.3.1 Selection and Justification of Model t

The computer program MICROSHIELD (version 4.1) was used for this analysis. The MICROSHIELD program allows a user to input a source geometry and shield thickness, density and material type. The data and calculation methods used in MICROSHIEL are d:.:umented and the program is widely acccpted.

2.3.2 Assumptions The following assumptions were used for modeling the direct exposure pathway for the disposal of the Trojan Reactor vessel at the Richland Facility:

1. The reactor vessel will be disposed as a single, intact unit.

2. The reactor vessel activity is shielded by steel and concrete to reduce

! Side cladding

! surface dose rates to less than 200 mrem / hour.

thicknesses were estimated to reduce the reactor surface activity to levels I

L below this limit. These thicknesses were then included as side and end cladding of the reactor vesselin the disposal modeling.

I 4

March 11,1996 i

---

m__._.___ . _ _ _

_ m __-_ _ _ . _ , _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _

US Ecology Trojan Rx Vessel

. 3. A total thickness of 16.5 feet of soil will cover the reactor vessel.

Sensitivity analyses were performed that examined the exposure and dose for changes in soil cover thickness in two foot increments.

4. Radionuclide activities of the reactor weste as of 5 years after shutdown will be used. The activation activity is 2.01x10' Curies, compared to the 357.9 Curies of surface contamination. The contribution to the dose from the surface contamination are negligible, and was not included in the analysis.

5. The radiation exposure is to the maximally exposed individual, and is based on a person who is standing on the trench cap in the center of the Facility. An hourly exposure rate and dose was calculated.

~

6. The trench caps will be maintained such that major soil erosicn is repaired. Because of the Facility's location and the soil conditions, the soil erosion potential is assumed to be minimal.

24 DATA 2.4.1 Areal Geometry The cap area is 3.64x10' square feet (rectangular shape). This area includes the plan view area of the trenches, the areas between the trenches, and the areas to the sides of the trenches. This area is the cap ares used in HELP model studies performed for Closure Plan.

The source geometry that most closely matches the Facility conditions is that of a cylindrical source with side and end cladding, and with end shields geometry. In this situation, the waste is isolated by a horizontal shield (e.g. the cap), The reactor vessel volume is 7,951 ft' (2.25x10' cm'), with a void volume of 5295.3 ft* and a displaced volume of 2655.7 ft'. Estimated dimensions are about 16 foot diameter and 40 feet high. The total weight of the reactor vessel was estimated to be 5.91 x10' g, which yields an average density of 2.6 g/cm' for the source material.

2.4.2 Cladding and Cover Thickness The thickness of the waste is equivalent to the reactor vessel dimensions. The previous Closure Plan' indicates that 8 feet of soil will be placed over the new trenches, in which the reactor vessel will be placed. The soil cap design will be added on top of the trenches and soil cover. The cap is about 8.5 feet thick. This yields a total cover th'.ekness over the source of about 16.5 feet. Additional soil backfill (reportedly 1 to 10

'US Ecology,1990. Site Stabilization and closure Plan for Low-Level Radioactive Waste Management Facility, Richland, Washington.

March 11,1996 5

___ - ' ' ^ ^ - - - - - - - - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

ft) may be placed over the trencher prior to the cap placement, to level the Facility. .

This additional material was not considered in this calculation, as its depth is unknown.

j The side cladding thicknesses were estimated 'o produce a design exposure rate of

" inches of steel shielding.

200 mrem / hour. The disposal design identiiled e'  ;

Concrete cladding was added to reduce the exposure : ae. A sensitivity analysis was '

performed using MICROSHIELD for different concrete thicknesses; a concrcte thickness of six inches yielded an exposure of 132 mrad / hour. Therefore, eight !nches of steel and six inches of concrete were used as the side and end cladding for the trench disposal exposure.

2.4,3 Waste and C ver Density i

The model requires the shield material be defined. The shield material is.the earthen cover. The side and end cladding (steel and concrete) is included as part of the shielding described above. MICROSHIELD contains attenuation information for concrete, but not for soil. The shield (cover) was simulated as coners'1, but the density was reduced to match the cover soil density of about 94 pounds per cubic icot (Ib/ft')

(1.5 g/cm'). The waste was simult,ted as ircr, but the density was reduced to match the density of the source (about 2.6 g/cm').

2,4,4 Source Analysis Th1 total activation activity in the Reactor Vessel is 2.01x10' curies, and consists of 11 isotopes, listed in Table 5.

2,4,5 Data Summary The following summarizes the data used in the analysis of the direct exposure pathway.

1. Soil used for trench br,rkfill and trench cap material is sand with a density of 94 lb/ft' (1.5 g/cm'). This is an average dry unit wei sand', and matches soil densities mescured at the Facility'g .

2. The trench area is 3.64 x 10'ft'.

3. The waste in the trenches was modeled as a large cylinder that had a radius of 8 feet and a thickness of 40 feet.

'Lambe, T.W., and R V. Whitman,1969, Soil Mechanics, John Wiley & Sons, New York, p. 31. Table 3.2

'Bergeron, M.P., et al,1987, Geohydrology of a Low. Level Radioactive Waste Disposal Facility, Richland, Washington,- Battelle, Pacific Northwest Lab., p. 39.

March 11,1996 6

Trojan RxVessel

.Us Ecology . ,

i

4. The waste has an avetsge density of about 160 lb (2.6 g/cm'). The

. average waste density is calculated from reactor ve:..el density,_ volume

' and void ratios.

5, The source term waste activity (summed over 11 isotopes) is about 2.01 million Cl. - The waste is assumed to t;? dir. posed of intact within the trench.-

6. - The trench cap is modeled as 16.5 feet thick aoove the waste.

2.5 MODEL SIMULATIONS The model was run for a reference value-of 16.5 ft of sNeld (TROJAN 4.MS4 and TROJAN 4.ASC). A sensitivity analysis of shield thickness v.as performed for values of 0 to 16 feet in 2 feet increments to examine the results of possible cap removal.

2.6

SUMMARY

The results of this analysis indicate that the direct exposure route is not a pathway inat requires . additional. analysis. Reported average external gamma exposure--rates for areas around the Hanford Reservation range from 0.15 mR/ day to a maximum of 1.8 mR/ day (0.0063 tr WJr,r to 0.075'mR/hr). The trench backfill and cover design reduce the amount of radiauon to below background levels.

' The : lose rate for an individual standing over the intact 16.5-foot trench cap was estimated to be 2.3x10'" n. rad /hr (5.5x10'" mrad / day) or an exposure rate of 2.6x10" i mR/hr (iaxiC ' mR/ day). This value was compared to the minimum average gamma exposurt, 9K tu 1971-1972 at the Hanford Reservation 100 Area of 0.15 mR/ day (ERDA,1975) artd determined to be insignificant.

4:

t i

March 11,1996 7 I

- ~

\. --- --_--____ 70@% MW1mr------- ----- - - - -----

t US Ecology TABLE:

DATA Pararneter Value i

Trench depth 45 ft 265 ft Vadose zone thickness Vadose zone corosity 0.30 100 ft Saturated zone thickness Saturated zone corosity 0.10 Infiltration (surrounding area) 0.2 in/yr Infiltration (through cac) 0.05 inlyr 8

Trench, intertrench area , 3.64 x 10 scuare feet March 11,1996 8

Trojan Rx Vessel

- US Ecology TABLE 2 SOURCE TERM isotope Activity Half Life 2.54E-01 458 Am 241 C-14 2.18E+02 5.730 Ni-59 9.53E+02 8.00E+04 Ni-63 1.57E+05 92 Nb-94 3.29E+00 2.00E+04 Pu 238 1.92E-01 86.4 Pu-239/Pu 240 2.15E-01 24.390/6.600 Pu-242 1.08E-03 3.79E+05 Sr-90 2.13E+00 27.7 Tc-99 7.06E-01 2.12E+05 March 11,1996 9 e

' Trojan Rx Vessel

' US Ecciogy

• TABLE 3 RADIONUCLIDES FOR DOSE ANALYSIS Infiltration = 0.2 inlyr infiltration = 0,05 in/yr sotope Hypothetical Well Hypothetical Well Concentration Concentration (pCl/L)

(pCi/L) 3.90E+0: 5.65E+00 -

C-14 -

7.2 E 14- 4.03E-16 Nb-94 ,,_

3.21 E-03 5.73E-05 Ni-59 i

Pu-239/ 1,78E-07 2.29E-09 Pu-240 5.49E-07 4.96E-08 Pu-242 2.44E-15 1.19E-17 Sr-90 2.65E-01 - 5.80E-02 Tc-99 .

March 11,1996 10

~

,- -US Ecology Trojan Rx Vessel

.- TABLE 4

SUMMARY

OF RESULTS Organ PRICH22. PRS PRICH23. PRS Infiltration Rate Infiltration Rate 0.2inlyr 'O.05 in/yr Body 0.261 0.038 Red Merrow 0.555 0.080 Thyroid 0.160 0.024 Note: C-14 is major contributing radionuclide March 11,1996 11

~

,. Trojan Rx Vessel US Ecology TABLE 5 .

l DIRECT EXPOSURE ACTIVITY isotope Activity ICuries)

H-3 6.55E+02 C-14 2.18E+02 Sb-125 4.04E-01 .

Mn 54 2.16E+03 Eu-152 2.54E+01 Fe-55 6.57E+;5 Co-60 1.15E+06 Ni-59 9.53E+02 Ni-63 1.57E+05 Nb-94 3.29E+00 Tc-99 7.06E-01 TOTAL 2.01E+06 l

March 11,1996 12

ATTACHMENT IV

h . .-.

f, TATE OF WASHINGTON  ;

~~

DEPARTMENT OF HEALTH i DIVISION OF RADIATION PROTECTION AirdustrialCenter. Bldg. 3

• P.O. Bot 47827

• Olvmpia. Washington 98504 7827 June 10,1996 Art Palmer, Chief Radiological & Safety Officer US Ecology, Inc.

120 Frankli1 Road Oak Ridge, Tennessee 37830

Dear Mr. Palmer:

This is in response to your letters dated March 12 and April 17, 1996, requesting the department's review on the waste classification for the Trojan reactor vessel.

We have reviewed your submittals and have determined that the waste classification of the Trojan waste appears to be consistent with the Nuclear Regulatory Comminion's January 17,1995 Final Branch Technical position on Concentration Averaging and Encapsulation.

As a result, the classification of this waste appears to conform to your state of Wmbington radioactive Materials license WN-1019-2, and WAC 246-249-040. Please be advised, however,it is the generator's responsibility to ensure compliance with waste classification and waste form. It is requested that if any of the data used for waste calculations change significantly, that the revhed numbers be submitted to the department.

If you should have any questions, do not hesitate to cmEact me.

Sincerely, b

Mikel J. Eisen Radiation Health Physicist ec: WDOH - Richland, WA l

I

-a-. o 1

i Amenean Ecology Corooration 120 Frankhn Rosa *et 615/483-8768 oak Rxsge.TN 37830 Fax: 615/483-8078 hdean Fe%

April 17,1996 Mr. Mikel J. Elsen Radiation Health Physicist Washington Department of Health P.O. Box 47827 Olympia, Washington 98504-7827

Dear Mr. Elsen:

This is provided in response to you letter of April 2,1996 transmitting your comments regarding the disposal of the Trojan Reactor Vess : Each ofyour comments is reprinted below followed by our response.

Comment 1: "FF' hen does PGE expect tofmd out if the NRC willissue a C of C on the reactor vessel?"

Response: PGE met with the U.S. NRC on January 31,1996 to discuss the proposed shipment of the reactor vessel and internals. PGE proposed several i alternatives for shipment of the package. The NRC's preference was to license it as a Type B package. PGE discussed the requirements for a Type B package and the ability to meet the requirements. Overall, it appears that the NRC will license the package if properiy designed and the shipment is completed in a well controlled manner. Initial conceptional design meetings will be held with the NRC in May. The final Safety Analysis Report will be submitted late 1996. It is expected t'ae Certificate of Compliance will be issued late 1997.

Comment 2: "Please supply the waste classifcation calculation. It was not submitted with your request. Will the acrimtion analysisfor the vessel with internals be analytically venfied with any sampling? A sample may be able to be compared to the activation analysis to venfy the waste classification results.

Response: The waste classification calculation was inadvertently omitted from the copies distributed. A copy is attached with this submittal. US Ecology regrets the error and apologizes for any inconvenience it may have caused.

.p

@n.

s s

m Mr. Mikel J. Eisen

• April 17,1996 Page 2 The two approaches historically used to characterize routinely generated activated metals components included; (1) direct sampling ofindividual components coupled with radiochemical analysis of samples. and (2) the use of activation analysis computer programs. Both approaches rely on the gross radioactivity method in the BTP and have only been employed to define Part 61 scaling factors. These scaling factors are then used in conjunction with radiation levels to quantify the base radionuclide Co-60 to which the scaling factrs are applied. The direct sample approach to scaling factor determination peaked in usage about 1987, and has not been used at all since 1992. This is based on its time and cost, the difficulty with the representative sampling of routine components, and the uncertainties in the analytical results. Ni-59 was always scaled from measured Ni-63 and Nb-94 concentrations were always definw as LLDs.

The direct sampling method has never been used to determine scaling factors for reactor vesselinternals. This is due to the difficulty of obtaining representative samples from internals and the very high Co-60 concentrations in components which approach Class C limits which prevents accurate radiochemical analysis of the samples. The conccatrations of significant radionuclides in internals components varv as a function of the i base metal's nickel and contaminant content and the integrated flux in the base metal. Empirical data, in the form of piece specific radiation surveys from Yankee Rowe internals. indicate that concentrations varied by five orders of magnitude from component to component. Additionally, the concentration variations in a single component were found to be one to two orders of magnitude. Some direct sampling analysis has been performed at PNL on nonfuel bearing components under a government contract, and the reported comparisons between sample results and acti.ation analysis results were good.

The waste classification analysis for Trojan is based on a detailed one dimensional neutron transport and point neutron activation analysis and the material properties of the component parts. These calculations were performed using TLG Services, Inc. FISSPEC and O2 FLUX computer codes and ANISN and ORIGEN computer codes obtained through the Oak Ridge National Laboratory's Radiation Shielding Information Center.

Ancillary calculations were performed using TLG's ANISNOUT and O2 READ computer codes.

p ,

@ n.

s .

Mr. Mikel J. Elsen April 17,1996 Page 3 The one dimensional ncutron transpon model was normalized with data obtained from a Westinghouse Elec ric Corporation report on a reactor vessel radiation surveillance specimen which was removed from the plant in 1990.

Based on the above, PGE does not intend to obtain samples from the Reactor Vessel or Internals. The waste characterization will be based on the activation analysis and radiation surveys. The radiation surveys will be used to quantify Co-60 content and the activation analysis based scaling factors will be applied to Co-60 quantities. This is fully consistent with the h%C BTP gross radioactivity method of charactenzation and was employed during the Shoreham and Yankee Powe projects.

Comment 3: "Is the Pu-241 that is used in3 ourpathway analysis decayed into Am-241?

Our results indicate 2.3 Ci where the proposal shows 0.23 Ci ofAm-241.

The di(ference could be in the Pu-241 imtial acurity (e.g.,11. 7 Ci). Please show how the valuefor Am-241 was arrived at. "

Response: The value of 0.25 Ci of Am-241 documented in our waste classification report was based on the fractional percentages of the sample results taken i from a S/G tube in 1994. The results were not decayed to November 1997 as were the other activation analysis results.

The isotope Pu-241 has a short halflife (13.2 years) relative to the travel time through the vadose zone. Combin:d with a large distribution

,ccefficient, the isotope will not migrate an appreciable distance from the reactor vessel, but will decay to Am-241 and U-237. Because the decay chain is almost exclusively to Am-241, this will yield a maximum actisity of Am-241 of 0.571 curies at approximately 60 years. The decay product of 0.571 curies of Am-241 from Pu-241 compares to the 0.254 curies of Am-241 identified in the source term from the reactor, and used in the ground-water pathway analysis. The change from 0.254 to 0.571 curies represents and increase of about 2 times.

Am-241 has a larger distribution coefficient value and longer halflife than Pu-241. The estimated leachate concentration at the water table for Am-4 241 was 3.69 x 10 " and 5.41 x 10"" pCi/L for recharge rates of 0.2 in/yr pyn. .r 1

s . l

x - . . - _

7___ c .. ..

l r

Mr. Mikel J. Elsen '

April 17,1996 Page 4 and 0.05 in/yr, respectively. Because Am-241 is adsorption-controlled, multiplying the increased activ.ty rat.s times the concentra*. ion at the water table should yield an approximation of the concentration at the water table with the new source value. These values are 7 38 x I0'S and 1.08 x 10'*

pCi/L for the two recharge rates. These concentrations are negligible.

With regard to waste class calculation at the time of shipment, the sum of fractions is estimated to be 0.335 due to 11.7 curies of Pu-241 and o 25 curies of Am-241. The activity shift still maintains the waste as Class C.

Over time the Pu-241 activity decreases and the Am-241 increases. At 60 years the Pu-241 activity is <0.1 times the original and the Am-241 activity reaches a peak of approximate. . v..di curies. The sum of the fractions at that time would actually decrease.

Comment 4: "In the pathway analysis it is assumed that the package 's exterior dose is only 200 mr:hr. Wuh one of the steam generators, .1430 mrihr hot spot wasfound and enclosed before shipping. With this m mind, perhaps the PA should use i R hr as an mitial exposure before taking any shielding into account. This dose rate is the most conservative and should not impact the outcome of the analysis. "

Response: It is expected that the reactor vessel and internals will be shielded to less than 200 mR/hr to meet Depanment of Transponation shipping requirements. However, for purposes of evaluating this possibility, the P A results may be scaled directly to the increase in package dose rates.

.Specifically, the 1 R/hr case may be evaluated by extrapoladon linearly and directly from the PA results for the 200 mR/hr cases. That is, multiplying the results of the 200 mR/hr cases by a factor of 5 yields the results for the associated 1 R/hr cases. A table of the base case and the sensitivity analysis cases for the 200/hr package and the and the associated 1 R/hr package are provided below.

@Reewwd Pacer

. Mr. Mikel J. Eisen April 17,1996 Page 5 i

~

Soil Cover Thickness (ft) Dose R:te In Air Dose Rate in Air (mrad /hr) @ 200 mR/hr (mrad /hr) G 1 R/hr 2

0.5 3.416 x 10' 1.705 x 10 2.5 l 3.628 x 10" 1.184 x 10*

4.5 l 3.654 x 10'* 1.827 x 10 2 6.5 l 3.552 x 10

• 1.776 x 10" S' l 3.361 x 10" 1.681 x 10*

10 5 3.120 x 10" l 1.560 x 10

12.5 3.862 x 10"' i 1.931 x 10"*

14.5 I 3.583 x 10"3

~

1.792 x 10a2 16.5 l 3 314 x 10"' l 657 x 10"'

Comment 5: "Will any shielding be welded onto the vessel? Is this weight usedin the

~ waste classipcationprocess?"

Response; Shielding may be welded onto the reactor vessel to reduce radiation levels on the exterior of the package as necessary. However, the weight of the shielding will not be used in the waste classification process.

9 Comment 6: "Since this is not a rounne shipment, proceduresfor the handling and disposal of this waste should be developed and submitted to the department for review. Additionally, the proposed trench set-up (e.g. whe:c the ramp (s), chefinalvesselplacement, and backfilling should be addressed. If the proposed trench is different that what is contained in the March 6,

~1991 Comprehensive Facility Utili:ation Plan, Documem 200-DOC-001, Rev. 3, the department must approve the change. "

Response: At this time, we are requesting the Department's concurrence with the waste classification of the PGE reactor vessel. We presently anticipate that the reactor vessel will de disposed ofin Trench 12. This is consistent with the proposed use of Trench 12 in the CFUP of March 6,1991. Waste placement, ramp location and backfilling are not addressed in the CFUP.

Only trench configurations (i.e., maximum dimensions and slopes) are addressed.

@n wr 1 .

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1 Mr. Mikel J. Eisen April 17,1996 Page 6 A number oflarge components have been disposed at the Richland site without special site handling and disposal procedures These have included the Trojan steam generators and pressurizer as well at the Pathfinder reactor vessel. Since the dose rate on the reactor vessel is expected to be less than 200 mRihr, we believe that this unit will be able to be appropriately disposed within the existing site procedure framework. As the project progresses ifit becomes necessary to develop additional site procedures, these will be forwarded to the Department for review and approval.

Comment 7: "It is the department 's opmion that the basement construction scenario &

exammedin the patinvay analysis.

Response: The basement scenario has. in erTe:t, been .un as a part of the sensitisity analysis performed for the Direct Exposure Dose Analysis. This sensitivity analysis examined the effect of varying the trench cap thickness from 0.5 to 16.5 feet in two-foot thick increments. The results of these analysis are presented in response to Comment 4. Since radium is not present in the source term, the reactor vessel disposal does not need to be evaluated for radon gas.

\ Comment 8: "If 7aat is the processfor vers.Gmg that the void spaces arefilled with grout? "

Response: The grouting process will be developed to provide the best engineering assurance that the package is being completely filled. This willinclude

~

filling the vessel until a positive vent of Low Density Cellular Concrete is obtained from each vent connection. Any small voids that form during .c.e filling process, will fill by gravity flow and by the thermal expansion driving force as the grout heats up during cure. Any voids created at the top / vents due to the above will be filled. A final visual inspection will be completed from each top vent to ensure the package is filled. In addition, a time-weighted average density and total weight of LDCC constituents will be used to estimate the volume ofinjected grout and verify it is greater than or equal to the available intemal free volume. This process will ensure the veid spaces are filled with grout.

hRecycseti Peer 3 .

A Mr. Mikel J. Elsen April 17,1996 Page 7 Comment 2: "You have indicated that the vessel will have 357.9 curies ofsurface contamination. How will the contcnunation on the vessel be handled so that it willnot be spread?"

Response: The 357.9 curies of surface contamination represents contamination .

contained within the reactor vessel on reactor vessel walls and irternals.

The outside of the reactor vessel will be below Department of Transportation shipping limits for surface contamination. We expect the outside of the package to be essentially free from removable radioactive cor.tamination.

We appreciate the Department's review of our request and trust these replies ;

appropriately respond to your comments. PGE is presently incurring unrecoverable costs associated with this project; consequently, we appreciate your continued timely consideration and evaluation of this proposal.

Sincerely, 1 i

/M(JL 1 .

's Arthur J. Palmer, CHP Chief Radiological Control and Safety Officer Att.

'OW &f

I ,

PORTLAND GENERAL ELECTRIC CALCULATION SIIEET f

Revision o Sheet __ 1 Cont'd on Sht Calculation No. N/A

-. Date t '13/96 Preparer Michael Murdock Veri 6er

[, , o . Date _ - /-/ac 1

/

REACTOR VESSEL WASTE CLASSIFICATION ANALYSIS INTRODUCTION The intent of this analysis is to determine the waste classification of the Trojan Reactor Vessel with its sub-components (internals) as one complete package. The approach will be to ae the radioactivity, as analyzed for and, identified in the Trojan Decommissioning Plan along with the most current revision of the Radiation Protection Manual Procedure RPMP 4," Determination of Radioactive Material Shippir.g And Waste Classifications". The anslysis will be performed combining both the surface contamination activity and the neutron activated activity to calculate over the envelope volume of the package.

REVIEW CRITERIA The analysis will be reviewed and checked for accuracy and regulatory conformance but wilf not be documented or follow the same format as an approved PGE calculation.

RESULTS 4 The vessel results in being (Class C) waste.

4 The package contains 3.56 gms of plutonium.

+ Table 1 results .335 4 Table 2 results .299 e.

- - - - - - - - ~ _ -_ _ _ . __ ___

PORTLAND GENERAL ELECTRIC CALCULATION SHEET- ,

- Calculation No, N/A Revision o Sheet 2 Cont'd on Sht Preparer Michael Murdock Date 2/1 w 6 hfuu Verifier Date - eAJE COMPONENT ASSUMPTIONS -

+ The reactor vessel and core internals is considered to be one component containing neutron activated metals incorporating radioactivity in its design thus allowing-concentration averaging over the displaced volume of the material.

4 The "envelone volume" is considered to be the reactor vessel (including the head) and the reactor core internals minus the void space in accordance with section 3.3 of the BTP. (Mass of metal only)

+ The majority of the activation is locatea la rue region between the upper and lower i core plates including the vessel cladding and walls.

+ The surface contamination (although <1% of the total activity) is considered to be -

distributed over all the reactor vessel internal surfaces.

i+ The activity contribution of the Incore Flut Thimbles, currently in place in the vessel, has not been calculated as of yet or been included in the total source term but is assumed to be less than .05% of the total activity.

COMPONEh i DBfENSIONS

+ Burial / Envelope Volume = 7951 ft*

3 - (Does not include any additional steel shielding or penetration closures)

+ Vold Volume = 5295.3 ft*

(With internals)

+ Displaced Volume = 2655.7 ft* or (75,201,049 cc)

(Envelope volume minus major void volumes)

+ Material Weight = 591,080,249 gms

[ Displaced Volume x 7.86 gms/cc]

?L ,

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PORTLAND GENERAL ELECTRIC CALCULATION SHEET l

mA Sheer i Cont'd on Sht Calculation No. . Revision _. 6 -

Preparer u m iun e Date m o^

Verifier _ 4 >w Date frat

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REACTOR VESSEL ACTIVITY (5 years Post Shutdown,11/1/97) 1I I SURFACE CONTAMINATION ACTIV/. TION NUCLIDE (Curies) (Curies)

H-3 1.87E-01 6.55E+02 C-14 2.65E-01 2.18E+02 Sb-125 3.86E+00 4.04E-01 Cc-144 1.08E-01 Mn-54 1.98E-01 2.16E+03 fu-152 2.54E+01 Fe-55 6.39E+0i 6.97E+05 Co-60 2.29E+02 1.15E+06 Ni-59 9.53E+02 Ni-63 4.60E+01 - 1.57E+05 Nb-94 3.29E+00 Sr-90 2.13E+00 Tc-99 7.06E-01 Pu-238 1.92E 01 Pu-239/240 2.15E-01 Pu-241 1.17E+01 Cm-242 3.62E-05 Cm-243 8.29E-02 Cm-244 7.87E-02 Am-241 2.54E-01 Pu-242 1.08E-03 TOTAL 357.9 Curies 2.01E+06 Curies i

rg ,

4

-f

- Form RP 73 SNPment Number N/A Page 4 cf 5 Packagi Number RV & Int L12. Determinaten of Concentration

- Matrial Volume '7.52E+07 cc Material Weight 5.91E+08 gm O E

-A -B C tsstop3 SNp Date Conc- Cone i Actmty ,

lootope - Isotope

- (mci) '(uCl/cc) - (uCl/cc)

(Ax1E3)/D (Ax1E3)/D

<5 yr T 1/2 .

> 5 yr T 1/2 H3 1.87E+02 2.49E 03 -

~ H 3 (A)- 6.55E+ 05 . 8.71E +00

' C 14 2.66E+02 3.52E 03

.C 14 (A) 2.18E + 05 2.90E+00 Mrt 54 ' 1.98E+02 2.83E 03 Mn 54 (A) 2.16E+ 06 2.87E+01 1 Fe 55 6.39E+04 8.50E 01 -

Fo 55 (A) 6.97E + 08 9.27E+03 -

Co 60 2.29E +05 3.05E+ 00

. Cs 60 (A) 1.15E + 09 1.53E +04

- Nb-94 (A) ~ 3.29E+03 4.37E 02 --

s Ni-59 (A)-- 9.53E+ 05 1.27E +01 Sb 125 =- 3.86E+03 5.13E42 '

' Sb 125 (A) 4.04E+02 ' 5.37E-03

- Eu152 (A) 4.54E+04 ,

3.38E 01 Co 144 - 1.08E + 02 1.44E 03 - ,

Fe 55 6.39E+04 8.50E 01

- Ni-63 4.60E+04 - 6.12E 01 F G H Ni 63 (A) ... 1.57E + 08 2.09E+03 Conc. ' Conc Gram St90- 2.13E+03 - 2.83E-02 Isotope Factor of l Tc 99 (A) .-- 7.06E +02 9.39E 03 (nCl/gm) - (gm/mCl) lootope I129- ' (Ax1E6)/E (AxG)

TRU- 4.16E+02 5.53E 03 7.03E-01 '

' Pu 238 ' - 1.92E + 02 ' 2.55E 03 3.25E-01 5.8E 05 1.11E 02 Pu 239/40 2.15E +02 2.86E 03 3.64E 01 1.60E-02 3.44E+00 Pu 241 - 1.17E+04 1.56E 01 1.98E+01 9.60E-06 1.12E 01

. Cm-242 3.62E 02 4.81E 07 6.12E 05 Pu 2421 1.08E+ 00 ,

1.44E-05 1.83E 03 9.30E+03 1.74E +04 3.56 Sum B Sum C Sum H

. (uCl/cc) (uCl/cc) (gms. of Pu)

- Attachment 1 - RPMP4 Pcg)22 of 35 Rev.17 Page 47 of 93

1. - Verify that' lis is a true computer generated copy of RPMP 4 Rev 17.

di!I/ra7 t ,

n

• 1 ShiPm:nt NumD t N/A i

Forrn D.P 73 Package Number RV & Int Pap 5 of 5 ,

13. Evaluation of results for Burial Site Limitations and possible DOT /NRC Form 741 Shipment.

a. Determine total grams of SNM:

3.56 + 0 + 0 = 3.56 Grams of Grams of Grams of Grams of Pu U-235 U 233 SNM If SNM > or = 1 gram, notty Radwaste Supervisor,

b. Burial Site TRU is TRU with T 1/2 > 5 yrs except Pu 238, Pu 239/240, Pu 241 and Cm-242 from Column F.

Burial Site TRU; 7.03E 01 Umit <10 nCi/gm (nCi/gm)

c. Waste Class TRU is TRU with T 1/2 > 5 yrs and PU-23G and Pu-239/240 frem Column F.

Wasto Class TRU:

7.03E 01 + 3.25E 01 = 1.39E + 00

_ 3.64E-01 (Pu-238) (Pu 239/240) (nCl/gm)

(TRU) 14 Sum allisotopes with half life <5 yrs:

9.30E+03 Umit <7 UCi/cc s (uCl/cc) 15 ,, Sum Allisotopes with half life >5 yrs:

' 1.74E + 04 Umit <1 uCi/cc (uCl/cc)

Note: If any of the above limits are exceeded, notty the Radwaste Supervisor.

Prepared by Michael Murcock Checked by Ag4ny

/

Attachment 1 RPMP4 Pagi23 of 35 Rev i7 Page 48 of 93

1. Michael Murdock Verify that this is a true computer generated copy of RPMP 4, Rev 17.

s .

FORM RP 7e WORK SHEET FOR DETERMINAT)ON OF WASTE Ct.ASSIFICATION Class A Class 8 Class C , , _ _

A B C D looeopic

' Conc. . Umit Quotsent Umit Quotient UmR Quo 6ent

~ T(b'.31 (A/C) (uCl/re) (A,"J)

Isotopes (uC1/cc) (uCi/cc) (A/B) - (uCl/cc) 8.00E+ 00 3.63E 01 (b) 8.00E + 01 3.83E 02 C-14. 2.90E + 00 3.00E41 - 3.13C42 (b) - 3.00E + 00 3.13E 03 Tc 99 9.39E 03 5.77E 01 tb) 2.20E + 02 5.77E42 Ni59 1.27E + 01_ 2.20E + 01 4.37E 02 2.19E + 00 (b) 2.00E 01 2.19E 01 Nb94 2.00E 02_

1.39E 01 1.00E + 02 - 1.3f,E 02 TRU (3.c) 1.39E +00 1.00E+ 01 _

1.98E+ 01 3.50E + 02 5.68E 02 (b) 3.50E + 03 5.esE 03

" Pu-241 (a) 6.12E 05 2.00E + 03 3.06E 08 (b) LO0E+ 04 3.06E 09 Cm 242 (a)

Sum of Table 1 Quotients 3.35E + 0c 3.35E 01

' NOTE: TRU with T 1/2 > 5 yrs = all TRU except Pu 238, Pu 239/240. Pu 241 and Cm 242.

Class A ~us B Class C A B C D lootopic Conc. Umit Quotent Umtt Quotient Umit Quotient

. Table 2 -

lettop:s (uct/cc) (uCl/cc) (A/B) (uC1/cc) (A/C) (vCVcc) (A/0)~

8.71E +00 4.00E + 01 2.18E 01 (d) N/A (d) N/A

H-3-'

2.19E + 01 (d) N/A (d) N/A Co40 1.53E +04 7.00E + 02 Ni43 2.00E + 03 3.50E + 01 5.97E + 01 7.00E + 02 7.00E + 03 ' 2.99E 01 7.08E 01 1.50E + 02 7.00E + 03 - 4.04E 06 St90 2.83E 02 - 4.00E 02

~*-

100E + 00 4.40E + 01 4.60E + 03 0.00E + 00 Cs 137

. Nuclid:s w/

(T 1/2,5 yr) 7.00E + 02 - 0.00E + 00 (d)- N/A- (d) N/A' Sum of Table 2 Ouotients 8.25E + 01 2.99E-01 i

e . Units are in nCl/gm -

b .11 Class A limit is escoeded, the weste is Class C or greater.

c .TRU = MU with T 1/2 >5 yrs + Pu-238 + ru-239/240.

d . N3 Umn.

~ NOTE: It is not necessary to list on the marutest any nuctice whose Class A quotient is less than 0.01 except C 14.Tc 99,l-129 and H 3.

Prepared by Michael Murdock Checked by uu

- Attachment 1.

Pt93 34 of 35' RPMP 4 Revision 17 Page 59 of 93 i Michael Murdock Verify that this is a true computer generated copy of RPMP 4, Rev 17

's .

Trojan Pressure Vessel Internals -

Waste Classification Waste classification Limits C-14 80 Ci/m 3 3

Ni-59 220 Ci/m Ni-63 7000 Ci/m 3 Nb-94 0.2 Ci/m 3 Note thctLthe Branch Technical Position on Concentration Averaging and Encapsulation (BTP) allows the use of averaging of activated metal components if gamma-emitting nuclides are less than 1.5 times the average for thoea. nuclides in the container, and if non-gamma-emitting nuclides are less than 10 times the average for those nuclides in the container.

The BTP also recommends that the displaced volumes (i . e . , no void volumes) be used to compute the wacte classification.

Pressure Vessel and Internals Average PV&I volume - 8341 ft' (234 m ) from Reference 1; 3

PV&I nuclide inventories from Reference 2; total activity 2,200,000 Ci C-14 218 Ci Ni-59 953 Ci Ni-63 157,000 Ci Nb-94 3.29 Ci PV&I average concentrations C-14 0.932 Ci/m 3 Ni-59 4.07 Ci/m' Ni-63 670 Ci/m 3 Nb-94 0.0141 Ci/m core Baffle Plates Baffle plate volume - 53.3 ft' (1.49 m 3) from Reference 3; Baffle plate nuclide inventories from Reference 2; total activity 1,190,000 Ci C-14 119 Ci BTP ratio 89 Ni-59 533 Ci BTP ratio 88 Ni-63 86900 Ci BTP ratio 87 Nb-94 2.22 Ci BTP ratio 110 ATTACHMENT 2

e Baffle plate concentrations C-14 79.9 Ci/m'; just below the Class C limit Ni-59 358 Ci/m'; exceeds Class C limit and factor of 10 Ni-63 58300 Ci/rc. ; exceeds Class C limit and f actor of 10 Nb-94 1. 51 Ci/m'; exceeds Class C limit and factor of 1.5 Core Former Plates Core former plate volume - 25.3 ft 3 (0.708 m3 ) total of 8 plates from Reference 3; Core former plate nuclide inventories from Reference 2; total activity 519,000 Ci C-14 70.1 Ci BTP ratio 110 Ni-59 261 Ci BTP ratio 91 Ni-63 47,800 Ci BTP ratio 100 Nb-94 0.568 Ci BTP ratio 57 Core former plate concentrations C-14 99.0 Ci/m 2 ; exceeds Class C limit and f actor of 10 3

Ni-59 369 Ci/m ; exceeds Class C limit and f actor of 10 Ni-63 67,500 Ci/m 3; exceeds Class C limit and factor of 10 Nb-94 0.802 Ci/m 3 ; exceeds Class C limit and f actor of 1.5 Lower Core Plate The lower core plate volume - 13.4 ft ( 0 . 3 8 m') ' f rom Reference 3; The lower core plate nuclide inventories from Reference 2; total activity is 89,000 Ci.

C-14 29.2 Ci BTP ratio 82 Ni-59 47.6 Ci BTP ratio 31 Ni-63 7130 Ci BTP ratio 28 Nb-94 13.5 Ci BTP ratio 2500 Lower core plate concentrations:

C-14 76. 8 Ci/m'; just below the Class C limit Ni-59 125 Ci/m 3; about half of the Class C limit Ni-63 18,800 Ci/m 2; exceeds the Class C limit and factor of 10 Nb-94 35.5 Ci/m ; exceeds the Class C limit and factor of 1.5 l

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

i References

{

f Reactor-Vessel and Internals Removal' Plan; January ~30,

1. .

1997; i -S e c t i o n 1 .' 3 .:-

F-4 .-- 2. . Reactor Vessel Radiological Characterization-Information; 3 September 11, 1996.

-3. . Safety Analysis Report forl' Reactor-Vessel Packager RVAIR Weight-and Center of Gravity Calculation; Appendix 2-1; March-31, 1997.

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m<. ., o 6 -. , , , - - . . . .

Mr. John L. Erickson, Director Division of Radiation Protection.

Department of Health

' Airdustrial Centar Building #5 P.O. Box 47827 Olympia, WA 98504-7827

Dear Mr. Erickson:

. On March 31,1997, Portland General Electric Company (PGE) requested the U.S. Nuclear-Regulatory Commission to issue a Type B Certificate of Compliance under our transportation regulations to allow a one-time shipment of the Trojan Nuclear Plant's reactor vessel with its intomais for disposal at the U.S. Ecology site in Hanford, Washington. Prior to beginning a full -

review of this transportation package application, it is our intent to 9ddress the waste classification of the waste shipment and make sure of its suitability ; , ' disposal. Under the 10 CFR 20 waste manifesting requirements, a waste generator mutt claeify wastes in accordance with 10 CFR 61.55. It is our goal to ensure that the waste shipment is properly classified.

On June 18,1997, PGE submitted responses to several of our questions relating to the classification of the waste shipment (Attachment 1). PGE acknowledges that some of the intemals are Greater-Than-Claes C (GTCC), but is proposing to classify the wastes by -

averaging the reactor intemals with the pressure vessel. The core baffle plates, the core former plates, and the lower core plate substantially exceed the recommended ratios for classifying activated metals given in Section 3.3 of the Branch Technical Position of Concentration Averaging and Encapsulation dated January 17,1995. However, PGE indicated that the one-piece shipment of the RV with the intomals would allow contact handling of the shipment, would result in 39 to 44 fewer waste cans requiring storage until a GTCC waste disposal site is developed, would reduce contamination control problems, would reduce occupational exposures from 134 to 154 person-rem to 67 person-rem (out of 591 person-rom estimated for -

the entire Trojan decommissioning), and would reduce waste shipments from 44 to 1.

PGE also provided a pathway analysis performed by U.S. Ecology, which was previously submitted to the State of Washington. This pathway analysis addresses groundwater impacts

. and doses from direct exposure. Other intruder pathways such as construction and resident-farmer scenarios are not addressed, nor is there a justification for assuming that the package will remain intact over the hazard lifetime of the nuclides that are critical to the waste classification: C-14, Ni-59, Ni-63, and Nb-94.

The NRC staff wil! consider attemative approaches to waste nuclide averaging if it can be

'shown that the wastes will meet the performance objectives in 10 CFR Part 61 (see 10 CFR 61.58 and Section 3.9 of the Branch Technical Position on Concentration Averaging and Encapsulation). The evaluation should include a comprehensive and defensible pathway analysis that inclu#s all relevant pathways.' The draft Branch Technical Position on a Performance Assessment Methodology for Low-Level F adioactive Waste Disposal Facilities could be used as guidance for this analysis.

ATTACHMENT 3

s

=9 2

We request that your staff ask U.S. Ecology, in coordination with PGE, to perform a.

compshensive and defensible pathways analysis to demonstrate the suitability of the proposed we:tes for disposal at the Hanford disposal site. Specifically, the analysis should be based on intruder-construction and intruder resident-farmer scenarios carried out for a 10,000 year

- period.

t-if the waste package is assumed to be intact for a period greater than 500 years,]ustification .

e needs to be provided. The draft

• Branch Technical Position on a Performance Assessment L

Methodology for Low-Level Rad,oactive Waste Disposal Facilities

• should be used as guidance.

Sections 3.2.2, 3.2.3,3.3.4, and 3.3.5 of this Branch Technical Position provide guidance on the j time frames for the performance assessment, use of engineered barriers, and evaluation of l

. waste forms for the performance assessment. After your review of thi.:information, if you

conclede that the reactor vessel with intomals is suitable for disposal under the State of

Washington's regulations, we will consider allowing the shipment to be classified under the a;temative averaging provisions of the Branch Technical Position on Concentration Averaging

. and Encapsulation. We are also willing to provide any technical assistance you may desire for '

the review of the submitt:d pathway analyses.

Sincerely, i' R' chard L. Bangart, Director -

Office of State Programs I

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