ML20203F466
| ML20203F466 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 01/30/1998 |
| From: | Bangart R NRC OFFICE OF STATE PROGRAMS (OSP) |
| To: | Erickson J WASHINGTON, STATE OF |
| References | |
| NUDOCS 9802270321 | |
| Download: ML20203F466 (7) | |
Text
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UNITED STATES S ro h
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8 WASHINGTON, D.C. 30606 4001 o,,,,,+
January 30, 1998 Mr. John L Erickson, Director Division of Radiatier Protection Department of Heau Airdustrial Center 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 intemek for disposal at the U.S. Ecology site in Hanford, Washington. Prior to beginning a review of this transportation package application, it is our intent to address *he waste classification of the waste shipment. Under the 10 CFR 20 waste manifesting requirements, a waste generator must classify 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-Class C (GTCC), but is proposing to classify the wastes by averaging the reactor intemals with the pressure vessel. The cere 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 (BTP) on Concentration Averaging and Encapsulation riated January 17,1995. However, PGE indicated that the one-piece shipment of the RV with the intemals would ediow 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 1S4 to154 person-rem to 67 person-rem (out of 591 person-rem estimated for the entireTrojan decommissioning),
and would reduce waste shipments from 44 to 1.
PGE also provided a pathway anatysis performed by U.S. Ecology, which was previously submitted to the State of Washington. 'I nis pathway analysis addresses groundwater impacts and doses from direct exposure. Other intruder pathways such as construction and resident-fs; V scenarios are not addressed, nor in there a justification for assuming that the package will rernain 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 will 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 BTP on Concentration Averaging and Encapsulation). The evaluation should include a comprehensive and defensible pathway analysis that includes all relevant pathways.
'1e draft BTP on a Performance Assessment Methodology fur Low Level Radioactive Waste r-e 3"
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I J. Erickson,
Disposal Facilities could be used as guidance for this analysis. The draft BTP has been available for public comment and review and does not represent a final agency position.
We request that your staff ask U.S. Ecology, in coordination with PGE, to perform a comprehensive and defensible pathways analysis to demonstrate the suitability of the proposed wastes for disposal at the Hanford disposal site. Specifically, the analysis shoud be based on intruder-construction and intruder resident farmer scenarios carried out for the time frame proposed in the BTP.
If the waste package is acsumed to be intact for a period greater than 500 years, justification needs to be provided. The draft *BTP on a Performance Assessrnent Methodology for Low-Level Radioactive Was'.e Disposal Facilities" could be used as guidance. Sections 3.2.2, 3.2.3, 3.3.4, and 3.3.5 of this BTP provide guidance on the time frames for the performance assessment, use of engineered barriers, and evaluation of waste forms for the performanc2 assessment.
After your review of this information, if you conclude that the reactor vessel with intenals is suitable for disposal under the State of Washington't.agulations, we will consider allowing thc shipment to be classified under the attemative averaging provisions of the BTP on Concentration Averaging and Encapsulation. We are also willing to provide any technical assistance you mry desire for the review of the submitted pathway analyses.
Sinc 4 rely, the I
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Richard L. 81 igart, Director, Office of State Programs /
Attachment:
As stated
O m-J. Erickson
. Disposal Facilities could be used as guidance for this analysis. The draft BTP has been
. available for public comment and review and does not represent a final agency position.
We request that your staff ask U.S. Ecology, in coordination with PGE, to perform a comprehensive and defensible pathways analysis to demonstrate the suitability of the proposed wastes for disposal at the Hanford disposa: site. Spec!fically, the analysis should be based on intruder-construction and intruder resident farmer scenarios carried out for the time frame proposed in the BTP, if the waste package is assumed to be intact for a period greater than 500 years, justification needs to be provided. The draft 'BTP on a Performance Assessment Methodology for Low-Level Radioactive Waste Disposal Facilities
- could be used as guidance.- Sections 3.2.2,3.2.3, i
3,3.4, and 3.3.5 of this BTP provide guidance on the time frames for the performance assessment, use of engineered barriers, and evaluation of waste forms for the performance assessment.
After your review of this information, if you conclude that the reactor vessel with internals is suitable for disposal under the State of Washington's regulations, we will consider allowing the shipment to be classified under the alternative averaging provisions of the BTP on Concentration Averaging and Encapsulation. We are also willing to provide any technical assistance you may desire for the review of the submitted pathway analyses.
Sincerely, Oria'.e RM C/
ntCHM.D L. Bidar.itT Richard L. Bangart, Director l
Attachment:
As stated i
DISTRIBUTION File Center LLDP r/f DWM r/f NMSS r/f MMasnik/NRR PUBLIC
- D'.ohaus/OSP SDroggitis/OSP CHaughney SShankman/SFPO EEcston/SFPG RChapell/SFPO MFederline ACNW RJohnson MBell e,
DOCUMENT NAME: S:\\DWMV.1DR\\TCJiTROJLET.WPD
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Disposal Facilities could be used as guidance for this analysis. The draft BTP, as been rvailable for public comment and review and does not represent a final agenc,' position.
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- We request that your staff ask _U.S. Ecology, in coordination with PGE, to pe L
rform a -
l comprehensive asnd defensible pathways analysis to demonstrate the suitability of the proposed wastes for disposal at the Hanford disposal site. Specifically, the analysis'should be based on i
intruder-construction and intruder resident-farmer scenarios carried out for the time frame proposed in the BTP.
i if the waste package is assumed to be intact for a period greater thorn 500 years, justification needs to be provided. The draft *BTP on a Performance Assessmerd Methodology for-Low-Level Radioactive Weste Disposal Facilities
- could be used as guidance.- Sectinns 3.2.2, i
3.2.3,3.3.4, and 3.3.5 of this BTP provide guidance on the time frames for the performance assessment, use of engineered barriers, and evaluation of waste forms for the performance f
assessment.
/
k After your review of this information, if you conclude that the reactor vessel with internals is suitable for disposal under the State of Washington's regulations, we will consider allowing the-shipment to be classified #@r the ahomative averaging provisions of the BTP on i
Concentration Averaging W incapsulation. We are also willing to provide any technical assistance you may desire for the review of the submitted pathway analyees.
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Sinceroly, s
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I Richard L. Bangart, Director Office of State Programs
Attachment:
As stated i
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- Disposal Facilities could be used as guidance for this analysis. The draft BTP has been l
available for public comment and review and does not represent a final agency position.
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We request that your staff ask U.S. Ecology, in coordination with PGE, tdperform a~
comprehensive and defensible pathways analysis to demonstrate the suitability of the proposed J
wastes for disposal at the Hanford disposal site. Specifically, the analysis should be based on
- intruder-construction and intruder resident farmer scenarios carried out for the time frame.
i proposed in the BTP.
if the waste package is assumed to be intact for a period greater i an 500 years, justification needs to be provided. The draft "BTP on a Performance Assessment Methodology for Low-Level Radioactive Waste Disposal Facilities" could be used as guidance. Sections 3.2.2,3.2.3,3.3.4,
- and 3.3.5 of this BTP provide guidance on the time frames for the performance assessment, use of engineered barriers, and evaluation of waste forms for the performance assessment.
I i-After your review of this informgtion, if you conclude that the reactor vessel with intemals is -
l suitable for disposal under the State of Washington's regulations, we will consider allowing the j
shipment to be classified under the altomative averaging provisions of the BTP on Concentration -
Averaging and Encapst lation. We are also willing to provide any technical assistance you may desire for the review of the submitted pathway analyses.
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, Sincerely, l
- Richard L. Bangart, Director l
Office of State Programs
Attachment:
As stated DISTRIBUTION: File Center LLDP r/f DWM r/f NMSS r/f-MMasnik/NRR PUBLIC PLohaus/OSP SDroggitis/OSP CHaughney SShankman/SFPO n
!' EEcstan/SFPO RChapell/SFPO MFederline ACN'N RJohnson MBell DOCUMENT NAME: S:\\DWM\\LLDP\\TCJ\\TROJLET.WPD 3
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We request that your staff ask U.S. Ecology, in coordination with PGE, to perform a,
comprehensive and defensible pathways analysis to demonstrate the suitability of the proposed wastes for disposal at the Hanford disposal site. Specifically, the analysis should be based on intruder-construction and intruder resident farmer scenarios carried out for the t'me frame
. proposed in the BTP.
4 If the waste package is assumed to be intact for a period greater than 500 years, justification.
needs to be provided. The draft "BTP on a Performance Assessment Methcdology for Low-Level
- Radioactive Waste Disposal Facilities" could be used as guidance. Sections 3.2.2,3.2.3,3.3A and 3.3.5 of this BTP provide guidance on the time frames for the performance assessment, use of engineered barriers, and evaluation of waste forms for the performance aseessment.
1 l
After your review of this information, if you conclude that the reactor vessel with intemals is i
~ suitable for disposal under the State of Washington's regulations, we will consider allowing the l
shipment to be classified under the altemative averaging provisions of the BTP on Concentration Averaging and Encapsulation; ",';
c'r t. ",gTulrovide any tecncm ; ::-2**== ; r n-,
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l Richard L. Bangart, Director Office of State Programs
Attachment:
As stated
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BTP has been available for public comment and review and does not represent a final agency position.
We request that your staff ask U.S. Ecology, in coordination with PGE, to perform a comprehensive and defensible pathways analysis to demonstrate the suitability of the proposed wastes for disposal at the Hanford disposal site. Specifically, the analysis should be based on intruder-construction and intruder resident farmer scenarios carried out for the time frame proposed in the BTP.
If the waste package is assumed to be intact for a period greater than 500 years, justification needs to be provided. The draft "BTP on a Performance Assessment Methodology for Low-Level Radioactive Waste Disposal Facilities" could be used as guidance. Sections 3.2.2,3.2.3,3.3.4, and 3.3.5 of this BTP provide guidance on the time frames for the performance assessment, use of engineered barriers, and evaluation of waste forms for the performance assessment.
After your review of this information, if you conclude that the reactor vessel with intemals 'e suitat'e for disposal under the State of Washington's regulations, we will consider allowing the shipment to be classified under the attemative averaging provisions of the BTP on Concentration Averaging and Encapsulation. We are also willing to provide any technical assistance you may desire for the review of the submitted pathway analyses.
I Sincerely, Richard L. Bangart, Director Office of State Programs DISTRIBUTION: Filo Center LLDP r/f DWM r/f NMSS r/f MMasnik/NRR PUBLIC PLohaus/OSP SDroggitis/OSP CHaughney SShankman/SFPO EECstsn/SFPO RChapell/SFPO MFederline ACNW RJohnson MBell DOCUMENT NAME: S:\\DWM\\LLDP\\TCJiTROJLET.WDD OFC LLDF( f LLDS-NRR DWM NMSS OGC OSP TCJoMon/bg Jhckey SWeiss JGreeves CPaperiello WReamer RBangart NAME DATE 1/%98 1/D/98 il /98 il /98 1/ /98 1/ /98 1/ /98 OFFICIAL RECORD COPY
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Stephen M. Quennoz Trojan Site Executive June 18,1997 VPN 048 97 Trojen Nuclear Plant Docket 50-344,72-017 License NPF-1 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Response to NRC Request 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 licensir.g action related to the PGE Safety Analysis Report for the Reactor Vessel Package is prirnarily a transportation issue, pursuant to 10 CFR 71, this response is being provided on both the Trojan Part 50 and 72 Dockrts for information.
The PGE response to the NRC questions is packaged in the following four Attachments.
Attachment I prosides a restatement of the NRC questions, followed by the PGE response.
Attachment Il provides Calculation RPC 97-018 which is referred to in the reyonse to Questions I and 2.
Attachment III contains the US Ecology letters to the State of Washington.
Att chment 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 effirmation.
71760 Columbia River Highway. Rain er. OR 97048 503'556 3715 M i C' ) 0 2-l l bu Qe S7yb
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VPN 048 97 June 18,1997 Page 2 or2 If there are any questions related to the response to these NRC requests for additional information, please contact Mr. H. R. Pate, at ( 03) 556 74E0.
(
Sincerely,
( l f} r
.j Stephen M. Quennoz Trojan Site Executive Attachments c:
L H. Thonus, NRC, NRR M. T. Masnik, NRC, NRR D. G. Reid, NRC, NMSS R. A. Scarano, NRC Region IV S. F, Shankman, NRC E. Fordham, WDOH B. Bede, US Ecology David Stewart-Smith, OOE A. Bless, OOE
~
- STATE OF OREGON,
)
)
)
. COUNTY OF COLUhtBIA
)
I, C. P Yundt, being duly sworn, subscribe to and say that I am the General hianager Plant Support and Technical Functions for Ponland 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 fue
/I
,1997 (r%h4nk
.'C P. Yundt, General hianager Plant Support and Technical Functions Ponland General Electric Company P
On this day personally appeared before me, C. ') 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
'8 day of '
/46
,1997
/
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7 0FFick SEAL 9,c0 mms 0Nomsmve MTSCHAFFRAN NOTARY PUBLic#EGON
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COMMtSSION NO 045755 u
Notary Pubhc i r the State of n
Residing at 2Dht?A'ik' hiy commissio'n' expires '7/J7 /6G
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ATTACilMENT I l
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i PGE Response to NRC kequest for AdditionalInformation Safety Analysis Report for Reactor Vessel Package NRC OUESTION Provide the basis for the statement that the reactor vesselinte nals comprise 340 cubic fe 1.
of Greater Than Class C Waste This statement originally appeared in your submittal dated January 30,1997 entitled " Reactor Vessel and Internals Removal Plan" Specifically, 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 ft' of greater than Class C radioactive waste from the reactor vessel internals is included in the conservative estimates of waste volume projections Also, Table 2 2-4 of PGE-1061 includes th 340 A' of greater than Class C waste for the reactor vesselinternals In estimating the v greater than Class C waste for the reactor vessel internals, it was assumed that the be segmented and packaged for disposal with the high-level spent nuclear fuel The esti volume (340 n')is calculated based on the gross container volume to be shipped and burie is therefore, greater than the actual segmented volume of the GTCC material in the internal (approximately 92 R')
The following sub-components comprise the reactor vesselinternals Core Baffle - The core bame consists of a series of axial plates which are attached to the core barrel by the core formers The core bame assembly provides lateral support for the fuel assemblies en the core periphery, a< we l as serving as a flow baffle by directing i
cooling water up through the core region and limiting bypass flow. The bame cor-ist rectangular plates 1 125 inches thick,154.94 inches long and of varying widths The core bame pla es weigh a total of 26,644 pounds. The core bame plates are Type 304 sta steel with a density of 500 pounds per cubic foot. Therefore the volume of the core bame is approximately 53 3 A' The core bsme, if segmented, would be classifice as greater than Class C waste.
Core Formers - The core formers are basically structural support members, providing the form for the core bime plates and attaching these plates to the core barrel. Core formers are located at several different elevations along the longitudinal axis of the reactor core At each of eight elevations, the formers consist of four units, for a total of thirty-two I
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 n' The core formers, if segmented, would be classified as greater than Class C waste Core Battel The core barrel consists of two major sections, the upper and lower core banels The barrelis a right circular cylinder with a nominalinside dir, meter of 148 inches and a nominal wall thickness of 2 38 inches in the active core region The activation analysis model includes all of the lower core barrel (61,850 pounds) and a portion of the upper core barrel (14,280 pounds) Th core barrelit Type 304 stainless steel The volume of the core barrelis approximately 152.3 n' The core barrelis not classified as greater 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 i= to reduce the neutron flux to the vessel wall at locations when the core is closest radially x 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 circumfe ence The four sets of th.rmal pads weigh a total of 20,950 pounds The thermal pads sre Type 304 stainless steel. The volume of the thermal pads is approximately 419 n' The thermal pads are not classified as greater than Class C waste Lower Core Plate The lower core plate support; the fael assemblies from underneath, 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 n' 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 syport contains the core support columns Additionally, this region contains columns which spon 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 Their volume is approximately 10.2 n' The lower core suppon columns are not classified as greater than Class C waste.
Lower Core Support - The lower core support is a massive piece of metal which suppo 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 an overall weight of 60,000 pounds The volume of the lower core support is approximately 120 A' The lower cora support is not classified as greater than Class C waste.
2
+
Region Below Core Suppcen and Above Upper Tie Plate - This region below the lower core support contains fifty six instrument tubes and suppon columns The total mass of i
Type 304 stainless steel in this region was estimated to be 4,072 pounds The volume is approxirnately 81 ft' 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 fl' The Upper Core Plate is not classified as greater than Class C waste.
Upper Core Support Columns - Above the upper core plate and attached to it, are the upper core support columns These columns provide support for the control rod assemblies moved ia and out of the core, as well as suppon for various pressure and temperature instrumentation There tre fony eight support columns and sixty one guide tubes in the region bet ren the upper core plate and the upper suppon 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 ft' The upper support columns are no' classifica 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 Repon, dated March 31, 1997, the activation radioauivity was determined thiough calculations performed by TLG Services, Inc (TLG)in support of the Trojan Nuclear Plant Radiological Site Characterization Report These calculations consisted of one dimer.sional neutron transport and point neutron activation analyses of the rea, r vessel and its internals These calculations were performe j 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 ORIGEN2 computer codes, obtained through the Oak Ridge National Laboratory's Radiation Shielding Information Center (RSIC). Reduction of the output from these programs and ancillary calculations were performed using the MISNOUT and 02 READ computer codes, written by TLG, and the Microsoft 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 neutrun flux beyond the boundaries of the reactor core. This was accomplished using the ANISN one-dimensional neutron transport computer code with five radial and axial geometric models The results of the 3
1
radial 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 subsequently collapsed into two-energy group formats (fast and thermal) and into a series of ORIGEN2 point activation / depletion calculations Additionalinput to the ORIGEN2 alculations included material compositions and historical plant performance data The radionuclide actisities for each component are presented in Table 1 of Calculation RPC 97 018 which is provided as Attachment 11 for information NRC. QUESTION:
2.
Provide the methodology and results of the waste classification for the pressure vessel with the internals intact. Describe how the waste classification conforms to the recommendations in Section 3.3 of the Branch Technical Position on Concentration Averaging and Encapsulation (BTP), dated Ja,1uary 17,1995. Demonstrate that the waste classification considered each of the internal components as a separate entity "
the averaging.
PGERESPONSE:
The Reactor Vessel and Internals will be packaged as one integral component containing neutron activated metals. The analysis to <letermine the waste classificat!on v/as performed in accordancs with the general requirements of BTP Section 3.3 which states:
"For neutron cctivated 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 component (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 part of the Site Characterization Repon that supported the Trojan Decommissioning Plan. A copy of the PGE analysis that determined the waste classification, PGE Calculation RPC 97-018, Revision 0, is provided as Attachment II.
The Branch T;:chnical 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 Subpart C of 10 CFR Part 61. US Ecology performed ground-water 4
i
a-4 and direct exposure dose analyses to support the' disposal approval process tv the State of i
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 3
package, therefore, satis 0cs the attemative provisions of Section 3.9 of the BTP. A summary I
discussion of the bounding performance objective from Subpart C of 10 CFR Part 61 follows.
I For activated metals, the intruder scenarios represent the worst case dose pathway. Since the
-intact vessel could not conceivably be handled by an inadvenent intruder, the intruder discovery scenario is the most appropriate Assuming this scenario, several observations are r
. appropriate:
i i-
.1.
The only long lived gamma emitter present is Nb-94 which is present in its highest t
concentration in the core bafne within the pressure vessel. The core bafne contains 2.23 curies of Nk 94 which is 68% of'he total Nb 94 activity of 3.29 curies. The l
baf0e 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 in the context of the appropriate intruder discovery scenario, the objectives of 10 CFR Pan 61 Subpart C are clearly j.
satisfied.
The required analyses were submitted to the Washington State Department of Health for review and are provided as Attachment III. The Washington State Department of Health reviewed the analyses and concluded that "the waste classification of the Trojan waste appears to be consistent with the Nuclear Regulatory Commission's January 17,1995 Final Branch Technical position on Con:entration Averaging and Encapsulation." The letter from the State of Washington Department of Health is provided as Attachment IV. The documents referenced above demonstrate that the Reactor Vessel with the internals installed, configured as I
described in the PGE Safety An:!ysis Repon, : atisfy the requiremems of 10 CFR Pan 61 Subpart C performance objectives. Based on the submittals and approval by the wasHngton Department of Health, 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 disposa!. These are:
l 1.
Protection of the inadvenent intruder.
4 5
4
--c
3.
Assure long term stability to climinate the need for long term maintenance after operations cease.-
4 3.
Protect public health and safety over the long term.
4.-
' Assure safety during the short term operational phase.
i The degree to which the disysal of the Trojan Reactor Vessel and Internals would meet these performance objectives was coicidered in the decision to pursue this alternative Specifically, protection of the inadvertent intruder. loag term site stability, and safety during the shon term operational phase were all enhanced by tN unit disposal alternative as opposed to the segmentation alternative. In panicular, the occupational and radiological safety advantages in handling a single package and the minimal impact on long term site performance objectives as compared to handling multiple high dose rate liners weighed especially heavily in this decision. The discussion of the dominant performance objective, protection from an inadvenent intruder, follows to demonstrate that the performance objectives of 10 CFR Pan 61 are satisfied by the burial of the reactor vessel and internals as a single packaFe-Intruder Assessment Scenarios inadvenent intrusion assumes that an individual, or group of individuals. intrudes into the waste either accidentally or without realizing that them is a potential hazard. The former case is considered most likely but is assumed to be quickly recognized by the individual with minimal resultin; exposures. More signihcant 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 inadvenent intruder exposures to low level radioactive wastes.
These include the intruder-Ccnstruction scenario and the Intmder-Agriculture scenario.
I Population exposures are also considered based upon waste that is uncovered and brought to the surfu being tr::= ported cffsite 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 contaminated dust via inhalation and direct exposure, consumption of food grown nearby upon which airborne contamination is assurred to have settled, and via direct gamma exposuit to the waste during excavation, p
i 6
l.
l l'-
e:
+
u
-l The intruder Agriculture scenario assumes that an indi ridual inadvertently lives on and -
consumes food grown on the disposal facility. Farming is a surface activity and generally does L
not involve disturbing the soil 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 implemeu this scenario at the end of the institutional.:ontrol period, however, a porion of the so!) excavated during the intruder-construction activif y is assumed to be backfilled around the house foundation. The remainder is assumed to be uti.Nd in the '
agricultural scenario. The house is assumed to be Ivealed 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 ' ' W by fallout from the contaminato 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, i
42 feet 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 ".. intruder scenarios analyzed contain one very large assumption -
that the soil' waste mixture in which construction or agriculture takes place ir more or less indistinguishable from dirt." That is, the weste has decomposed to the point that the intruder does not know he is contacting waste. This assumption is necessary since without it, the scenarios cc.ild 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 i
exposure pathway associated with the Intruder Construction and Intruder Agriculture scenarios
= is the direct gamma 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 4
4 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.
7
- To assess 'the direct gamma exposure pathway, the dose rates on the exterior of the reactor J
i 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 intruder digging down to, but not actually contacting, the-reactor vessel as it lays on its side in tlk disposal trench, The intruder would then construct his house on the vessel.
Vessel external dose rates were projected at time of shipment and in increments over the -
subsequent 500 years, it is important to note that at the projected time of shipment, the vesset will contain an estimated 1.15E6 Curies of Co-60. Due to transponation regulations, the external dose 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 i
rate for long tenn disposal guidance is provided in the 1995 revision to the US NRC Branch Technical Position on Waste Classification This document assigns a limit of 0.02 mRih, on j_
the surface of the disposal container as the acceptable dose rate from encapsulated sealed
^
sources anC activated metals 500 years after disposal. This limit is met by the reactor vessd and internal package 100 years following disposal.
i General Discussion of the Merits of the Two Options For the Reactor Vessel and Internals l
l Removal Project l
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 L
liners; or leaving the internals in the Reactor Vessel, filling the reactor vessel with a low l
l.
density cellular concrete, and then disposing of one large package. Segmentation of the i
Reactor Vessel Internals would most probably use a plasma arc torch (or equivalent l
equipment) to cut up the subcomponent parts that comprise the internals. The cutting would be conducted in the water filled refueling canal in Containment utilizing underwater tooling.
t
(
The internals would be ct.. mto pieces sr"-il enough to be placed in shipping casks for transport to the disposal facility. The liners would be retuoved from the transport cask at the disposal facility and placed in vertical disposal caissons, engineered concrete barriers, or o
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 L
Segmentation would result in some internal parts being classified as greater than class C waste c
(GTCC). A total oi39 to 44 GTCC storage cans would be generated from the segmentation 1'
l=
E 8
i.
f 1
h l
u I
Processi nese cans would remain at Trojan until the US DOE develops a disposal option. A byproduct of the p'asma arc method of segmentation is " dross" (i.e., welding slag and fine I
' particles) generated during the cutting process. Dross consists of the molten and vaporized; l
metal from the cutting process tht has condensed and solidified in the refueling canal water.
i his materir.1 is " caught" in buckets loc'ated below the piece being sectioned or is removed
'l from the refueling canal water via mcchanical 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.
e-Contamination control is one of the radiological hazards associated with this approach. The tasks of filling intermediate and high 3ctivity liners with waste and placing the liners in the transporta' ion casks are performed underwater in the refueling wanal 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 lE5 to > IE6 l
dpm/100 cm. Hot particles created during reactor operation or during the setmenting 2
- 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 transponation cask and placing it in the disposal trench or in a concrete caisson. This ly 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. - High 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 3
performed many times with minimal radiological impacts because each evolution was carefully plannd r.nd executed, the potential, however, exists for significant radiological hazards to oc
(
created in the event of an equipment malfunction or handling mishap.
Additional disadvantages are also p.eser.t with respect to long term facility closure, Cask liners filled with reactor internal sections (that are not classified as GTCC) contain void spaces which are undesirab!c from a waste form perspective. Void spaces in waste packages may lead to I
- channeling of *ainfall percolating through the waste and may lead 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 i
9 r
..w-
~~..a
.~
par < %s,it is not clear at what size they would be considered to be an inherently stable waste l
fort.: ;ch as irradiated metal. In the event ofinadvertent intrusion into the waste at some time after 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 planning, however, these issue > 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 significant 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 vcssel. 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 e,,ternal surface of the reactor vessel v ill also have to meet DOT contamination limits, consequently external contamination levels will be suffich;ntly low to prevent any spread of contamination during handling and dispsal.
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 ves:el 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 operathmal 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% unused capacity at the end of the facility lease in 2063. There is only one other nuclear power plant located within the Northwest and Rocky 14ountain Compacts and continued waste receipts beyond 2063 are not curre tly planned. The trench will be constructed with a long access ramp to minimize the grade u2 vessel transporter must negotiate. The acc:ss ram, will terminate at the bottom of the disposal trench. The vessel will be transported to the US Ecology site and down into the 10
disposal trench as single package. The packay will then be unloaded into its final resting 1
position.
Conclusion PGE has performed an extensive assessment of the storage and disposal options for_ the reactor vessel and inteinals. 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 protectmg 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 i
with regulatory guidance and can be safely buried at the licensed US Ecology radioactive waste disposal facility, i
1 I
I1
h 9
ATTACilMENT 11 1
QA RECORD WHEN COMPLETED ***
RC _L,,i_.1_
CFP cru rwa 7.s 004 Admi n se rvi c e s Letter Nwnter wA Systee t%naker u
A mber of Pages 1
Occument Date calc. Reference a TROJAN CALCULATION COVER SREET Sheet I
Cont'd on Sheet 2-.
Title __. Ef* Tet Vfssel to A s re eldss' ric ard_ Av a tu s h Calculation No.
RPc at 7-oe s Trojan Nuclear Plant Structure 6 cra b ear Supersedes Calculation Nc.
O 62 /W Quality-Related Yea /No system Cossponent fewrit s/65s ti Statust I Final Interim References (FMR/DPMR, SPEER, MR, FSC, etc.)
Bas Been Changed by or Revision has been Responsible Affected Identify Change Def erred by (Identify Supervisor /Date Document No. Vehicle: (MR, DPMR, DCP, /CF, Memo
(.TL, e tc. )
(Deferrals Only)
SPEER, PSC, etc.)
Calculation Objective Te D ct e a -.~ 5 The proprc usr e class,teu ted oF 7%
Esac rue dessrI s.)orr aII.Ts odrtu ee svk wMea1rs as cas dw r.
Revision Description 4
J Rev.
No.
Preparer Date Verified By Date Approved By Date O
M.e i,el Ms f d'bsk, Tbtrk 1 b'
w i
j TPP 18-9
~
Revision 2 Fase 1 of 1 Page 11 of 11 jOf0 bS
. - ~.
- =_
PORTLAND GENERAL ELECTRIC CALCULATION SHEJ'
?
Calculation No. 2k 97 0II Revision O Sheet 2
of 7
Sl21l41 Prepster Aike:l k1aa batd
- Date Verifier L
Date c,/n /4 7
.. 7 Table of Contents jdge Sheet No.
Calculation Cover Sheet..................................................................................... 1 Table of Cootents...............................,............................................................
2 Objective.......................................................................................................
3
..:ceptance Criteria..........
3 l
Sununary....................................................................................................
3 Assumptions. Design inputs, and M etixxiology..........................................................
3 Resulu..................................................,.................................................
4 References.....................................................................................................
4 Body of Calculation.........................................
4 RP 310 Form;:
i
!. Form RP 73
- 2. Form RP 78 Attachments:
1.QPRO Spreadsheet
l PORTLAND GENERAL ELECTRIC CALCULATION SHEET Calculation No. N 9ho'I Revision 0
Sheet 3
of I
l'repater AdhrI Mat.sout Date
$ln /U Verifier Date f /M 4 7
/
RFACTOR VF.Url WASTE CLASSITICATION ANALYSIS DNECTION The objec% of this calculation is to determine the proper waste classificatiot,f the Reactor Vessel with all its internal sub components as one unit.
ACCEPTANCE CRTIERIA:
None SUMMAny:
The purpose of this calculation is to aid in the planning for the eventual shipping for burial of the reactor vessel. This calculation can be used as a guideline when an actual shipping date is determined. The activities will need to be dec.y corrected when a date is known.
The approach to determine the waste class will be in accordance with 10CFR 61 and the US Ecology buria] Esense.
The setivity is obtained from the activation analysis done as part of the $lte Characterization in rupport of the Trojan Decommissioning Plan and RPC 96 008, " Reactor Veasel and laternals Surface Area Contamination'. The calculation will be done under the guidelines of RP 310, Rev 2.
ASSUMFilONS DFAIGN INPtTTS. AND METHODOLOGY Both the Attivation and corrosion activity represent activity decay corrected to 11/97.
In accordance with section 1.1 of the Branch Technical Position, the rewtor vessel and core internals is censidered to be one cornponent containing neutron activated metals inentpvating radie.ctivt in its design thus allowing concentrating averaging over the displaced volume of the material.
The displaced volume is the mass of the metals only as identified in Table of TLO's calculation titled
- RVAIR Weight and (. O.* as received unde' PGE letter No.102 97L. This does not include any closure plates, impact limiters, or thielding.
The majority of the activat.oa is located in the region between the upper and lower core plates including the vessel cladding and walls.
The 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 Thimbles, currently in place in the vessel, has nd been included in the overall total activity but is assumed to be less than.05% of the total activity.
_______ A
PORTLAND GENERAL ELECTRIC CALCULATION SHEET Calculation No. kN T7-o'i Revision d
$ beet N
of T
Preparer _
Mne est H vabest. '
pate
$bs k1 ikm Date MM 7 Verifier
/
Displaced Volume of Metal = 2622.2 ft'or 74,266.?'i8 cc ( reference TI.O's ' Weight and C G Calculation)
Material Weight = Displaced Volume x 7.86 gms/cc or $83,733.653 gms Activation Activity = 2.007 curies ( See Art.
I, taken from table 4.7.30 of the Site Characterization Report)
Corrosion or S i.l ace area Activity = 155.2 curies ( from Table I of RPC 96403 )
RESULTS:
7hc reactor ves.el and sub<om,mnents are Clus C wt..te.
Table I results =.328 of Clus C limit
- Table 2 results =.303 of Clus C limit -
SNM = 1.55 gms of Pu '
Total Pu = 5.25 curies "
RENNCES;
- 1. RPC 96408. Reactor Vessel & Internals Surface Area Activity
- 2. RVAIR 102 97L, POE Tracking No of TLO'. 7.VAIR Weight and C.O. Calculation
- 3. RP 310 Rev 2, Determination of Radioactive Matettal Shipping and Wute Clusificatbns.
- 4. Table 4.7.3o of the Trojan Site Characterization Repon BODY OF CALCULATION:
- 2. Attachment 1. QPRO Spreadsheet of Table 4.7.30
)
m-si is i mm1s mins--i a
Feb RP 73 '
Sh4pment Number N/A Pope 4 of a Package Number Reedor Vessel
- 12. Determeaten of Concentreten Motonal Volume =
7 43E+07 cc /
MotenalWeght
=
$ HE*08 pm /
D E
A B
C lootope She Date Cone Cone Adwy loolope isotope (mCl)
<5 yr T 1/2
>$ yr T 1/2 (uCvoc)
(uCVoc)
(Aa1E3yD (AstE3FD H-3 811E+01 <
100E 03 H 3 (A) 4 44E+06 -
8 00E+00 C 14 1.1 SE +02,
1.SSE 03
' C 14 (A) 218E +06,
2 93E+00 Mn64 867E+01 -
1.1SE 03 i
MS$4 (A) 216E+08 2 91E+01 Fe SS 2 77E+04 3 73E 01 Fe-S$ (A) 897E+08,
9 39E+03 Co40 9 92E+04 <
134E+00 Co40 (A) 116E+09,
1SSE+04 Nt>M (A) 3 29E+03 4 42E 02 Nr$9 ;A) 9 53E+0$,
126E+01 St>145 187E+03 2 2$E 02 Stw125 (A) 4 03E+02 5 42E 03 i Eu152 (A) 149E+04 2 00E 01
- Co 144 4 HE+01 8 2SE 04 F
G H
i N.43 2 00E+04 2 89E 01 Conc.
Cone Gram NF43 (A) 157E +08 212E +03 lootope Fedor of
- Sr 90 9 24E+02,
1.24E 02 (nCvgm)
(pmtmCi) lootope
< Tc w(A) 7.06E +02 9 SOE 03 (Ar1E8VE (AsO)
- TRU 181E+02 -
2 43E 03 3 09E 01 1.12E 03 1.43E 01 6 SE 05 4 ME 03
- Pu 238 835E+01
- P&239/40 9 33E+01 1.26E 03 1.80E 01 1.80E 02 1.49E+00 Pu 3 3 07E+03 883E-02 899E+00 9 00E 08 4 87E 02
~ Cm 242 168E 02 210E 07 2 87E 05
- Totals 2 01E+09 /
942E+03 1.78E +04 1.55 Sum A SumB Sum C Sum H (rnCI)
(uCvoc)
(uCVoc)
(gme. of Pu)
Attechr' ant 1 Propared by Mephapi Murdock RP 110 Page 22 of 37 Checked by d ,,,,
Rev 2
/
Pope 51 of 42
- 1. MeheelMurdock VerVy that the is e true computer generated copy of RP 310, Rev 2
Stupmerd Numter it'A Pepe lC S Paceage Numter Reactor Vessel t3, Evatusten of results for Donal See Lenasters and possible DOT /NRC Form 741 Shement
- s. Determene total preme of SNM 1 55
+
0
+
0
=
1 51 Grame of Grame of Grams of Grams of Pu U 235 U 213 SNM If SNM > or e 1 gram, notdy Radweste Supervoor.
- b. Du.31 See TRU is TRU wth T 1/2 >$ yrs except Pu 238, Pu 236/240, Pu-241 and Cm 242 from Column F.
Surtel She TRU.
3 09E 01 Lenst <10 nCvem (ncvem)
- c. Wests Casse TRU e TRU wfth T 1/2 >$ yrs and PU 238 and Pu 239/240 from Column e.r Wes's class TRU.
3 09E41
+
143E 01
+
160E 01 e
812E 01 (TRU)
(Pu238)
(Pu 23E740)
(nCvem)
- 14. Sum as motopes with half Irfe <S yrs-042E+03 Lima <7 UCvec (uCvec)
IS. Sum As motopes with half 14e >$ yrs 1.76E +04 Lima <1 uCvec (WCvec)
Nole. If any of the above twnste are enceeded, notty the Redware Specialet Prepared by Michael Murdock Checked by bm
/
Rp310 Page 23 of 37 Rey 2 Page $2 of 82 i M e melMoreock Verify that the to e true computer generstod copy of RP 310, Rev 2
~
FORM RP 78
$hipment No N/a
~,..
DETERulNATION OF Packspo No Rea7 g sg,
WA$TE CLAS$WJCAT10N Class A Class B Qp,-
A B
C D
Tab 6e 1 Isotope Conc Lrna Quotent Lma Ovotent
' ^ >
49 +
lsotopes (uCvec)
(uCvec)
(AS)
(uCvec)
(AC)
' 4/
6 C 14 165E 03 8 00E41 1 D4E 03 (b)
_tryf*'
dj C 14 (A) 2 93E+00 8 00E+00 3 67E 01 o) 8 DW e 3 67E 02 Tc0 9 SOE 03 3 00E 01 317E 02 (b) 3 00E +00 317E43 N 69(A) 126E+01 2 205*01 S 83E 01 (ti) 2 20E +02,
6 83E 02 Nt>D4(A) 4 42E 02 2 00E 02 2 21E+00, (td
+ 00E41
_F *1E41 TRU (a c) 612E41 100E*01 612E 02 (b) 100E+02
$ 12E 03 Pu 241 (a) 8 89E+00 3 SOE+02 2 48E-02 (b) 3 W+03 2 48E 03 Cm442 (a) 2 67E45 2 00E+03 134E 06 (b) 200E +04 134E49 Sum of Tat >6e 1 Quotents 3 28E+00 3 28,E-01 NOTE TRU wth T.1/2 >$ yre o all TRU escept Pu 238, Pu 239/240, Pu 241 and Cm.242 Class A Class B Class C A
B C
D isotopc
. Table 2 Cone Lma Ovotent Lme Ovotent Lmet Owotent isotopes (uCver)
(uCvec)
(A/D)
(wCvec)
(A/C)
(uCvec)
(A/D)
H3 6 00E +00 4 00E +01 150E 01 (d)
N/A (d)
N/A Cc40 155E + 04 7 00E+02 2 21E +01, (d)
H'A (d)
N!A N.63 2 69E 01 3 50E*00 7 69E 02 7 00E +01 7 00E+02 3 85E 04 N.- 63 (A) 212E *03 3 50E +01 7 00E+02 7 00E+03 3 03E 01 St 90 124E 02 4 00E 02 311E 01 150E +02 7.00E +03 178E 06 C&137 100E +00 4 40E +01 4 60E+03 NucJades w/
(T.1/2,c5 yr) 9 d2E+03 7 00E+02 135E +01 (d)
N/A (d)
N/A Sum of Tabte 2 Ouotents 3 61E+01 3 03E 01 a. Unas are in nCvgm
- 3. If Class A hme a exceeded, the weste e Class C or greater
- . TZU o THU with T 1/2 >S yrs
- Pu 238 + Pu 231V240
$.NoLma 90TE lt c not necessary to het on the mandest any nuclide whose Class A quotent a less than 0 01 except C 14 Tc 99,1129, and H 3.
Prepared by Mchael_Murdoc6 ^ ~
Checked by
@ /Tu;;
Attachment i
/
RP 310
- age 36 of 37 Revacn 2 Page 65 of 82 1 MichaelMurdock Venty that this e a true computer generated copy of RP 310. Rev 2
i l
PORTLAND GENERAL ELECTRIC CALCUALTIOf% SHEET l
Calastation No.
RPC974IS Revision Sheet 8
of 3
l Preparer MichaelMurilock #"
Dese OV27M7 i
Verifier
,J Dale f/x fn
_f i
l Table 1 REACTOR VE55El
't i
CURSE CDP. TENT FIVE VEAltS AFTER $8tl1TimoWN 14CTlvATION Osel.V3 i
)
SUIKOh4POMENT3 H-3 C-14 56-825 hee-54 Es-152 fc-55 Ce40 Ni-59 Me43 P4 94 Tc 99 Teests e
CoveBeme 2336E+02 1.IS6E+42 3329E41 1632E+03 763tE43 3 7J2E+05 7 248E+05 5329E+62 368eE*04 2 225E*90
$ Oett41 11871+06 Case Fesumers 8735E+01 7 018E+01 5.100E42 2110E+02 1 190E-06 2 293E+05 2.482E+05 2 6t eE+02 4 75tE+04 5 es2E41 7650s 42 5 lavl +05 1meer Case Bessel 5 465E+91 I.IB3E+91 1579E-C3 I633E+02 7 367E+00 3 733E+04 3555E+04 6 589E+01 9 656E+03 2 40eE41 6 502E42 1 32_tE +05 Upper Case Bervel 5151E44 1 Il5E44 I 400E4B I $3*E-03 7 415E45 3 SIBE4B 8 063E4I 6144E44 8629042 2279E-06 5 762E47 52471300 l
Thesensi pads 1303E+98 2.832E+99 2.003E44 3 989E+0I 2.997E+00 8933E+03 2056E+C4 1367E+08
- 2. 893E+03 5 786E42 146*E42 31788 +04 l
VesselClad 5.796E+00 I265E+0e 6 640E-05 6 790E+00 5358E-OI 4 IJ2E+03 6165E+03 7 574E+00 1021E+03 I 4 t4042 2 Seet 43 11341+04 Vesect well 4198E+90 6122E43 1332E42 3041E+00 II42E+00 IIIJE+03 3 92SE+02 I 33eE4I I577E+0I I 75tE43 4 445t 43 2 2355 +03 Lawer Case Plane 2 910E+41 9392E+00 4323E43 8Ol7E+08 3 785E-0I 3022E+04 51431+04 4 757E+08 7130E+03 I 35tE45 2 934t'42 S WI +04 Lauer Core say Cet 8758E+49 1.529E+98 6465E45 4 487E+00 4 546E-Of 6 052E+03 7 4 89E+03 1136E+01 ISO 4E*03 8 4560-02 1733E43 5 Seed +04 1meer Case Sep 20 lee 42 4356E43 I.630E-IO
- 5. IOJE-02 7.771E43 1386E+0! 2 919E+08 2 488E42 3 422E+90 7907E45 2 0108 -05 4 6558.+0I llelse 14= Cove Sep 7.563E46 1.492E46 0SeeE+00 3390E46 2.207E-06 4 940E43 6 062E-03 9376E46 I 2398-03 I 160E48 13966-09 1227E42 Upper Case Fleer 7.512E+90 I673E+90 8291E44 I335E+01 I037E+00 5377E+03 9 900E+03 9 612E*95 t323E+03 2_Se6E42 5 25tE43 I 6721+04 Upper Core See Col IJ99E+49 2.744E41 6327E47 6 737E41 3 272E-0I 9 077E+02 I829E*03 I 754E+80 2.276E+02 2207E43 2 7985.-04 2 2641 +03 Toeses 4 457E+02 2.I?OE*G2 4 020E4I 2.16t E*03 I487E+01 6973E+05 Il49E*06 9532E+02 1373E+95 3 286E+00 7 056E4B 2 007t +06 NOTE. 1. The acerney is tese the Sise Chasecaeressesse Itcyert Table 4130 2_ Nededes emissed wese; Ar-39.Co 41.Co-45.5e-II9sn, esed Te-125se der en sech a saisit conewbesses and ese a recaer in = esse cleansisemme W
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ATTACilMENT 111 l
I l
g,-,.,, m USEcology samen twmso-ce u hfarch 12,1996 Gary Robenson Head of Waste Management Section State of Washington Depanment of Health j
Division orkadiatu-Protecticn Airdustrial Center. Building 5 PO Box 47827 Olympia, Washington 98504 7S:7
Dear Mr. Robenson:
US Ecology and Penland General Electne would like to take this opponunity to address some of the issues pennining to the disposal of the Trojan Reactor Vessel at the Low-Level Radioactive Waste Disposal Site near Richland Washington. We are specifically l
requesting the Depanment's concurrence on the waste classification of the reactor vessel.
The reactor vessel package will consist of the vessel internals and the reactor pressure vessel as one component. The vessel will be cenified as an NRC Type B package. We believe this package meets the recuirement for classification as a Class C Stable waste form. Void r. paces will be filled to the maximum extent possible with a concrete based grout. Thesu items are disc.issed in detail below in addition to issues of concern raised at our January 25 presentation.
Stability Washington Administrative Code 246 250 050 specifies the stability requirements for radioactive waste disposal. Additionally, NUREG 0782 (Draft EnvironmentalImpact 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 must maintain their physical dimensions and consistency under the condi ions of compressive t
load, radiation, and biodegradation expected to be encountered in disposal. This is to preclude slumping, collapse or other failing of the trench cap; the need for active long-term maintenance and the ability to predict long term performance. Under the section
" Form of the Waste as Generated", activated steel from nuclear r(actors is given as one of the examples of waste that meets stability.
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Page 2 ef 3 Arthur L Palmer !!!, CHP to Gary Robenson, Head of Waste Management Section March 12,1996 With regard to the Trojan React:r Vessel, the physical dimension and consistency 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 moie than ader, sate s ength to prevent any deforrnation.
Radiation effects as far as maintaining physical dimensions and consistency, are not a concern since the vessel was designed to withstand much higher radiation fic!ds when the reactor was in operation. Biodegradation is also not a concern because the vessells steel.
Since the proposed disposal method meets criteria for stability and void reduction, due to the waste being irradiated metal and 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 '
. '.s creposalis 3 090 curies. The breakdown of the individual subcomponents within the vessel are premted in the table below.
Subcomponent I
Activity (Cl)
% of Total Activity Core BaHle 1
2 230 68%
Core Former 1
0.568 l
17 %
Lower Core Plate 1
0.135 4%
Remainder 1
0.360 11 %
TOTAL l
3 290 l
100 %
Under the proposal where the reactor vesseli,ternals are segmented and shipped in individualliners the total niobium content would be 0.360 curies. It should be noted th bb94 is generally considered to be an extemal dost hazard. Due to the thickness of the reactor vessel walls and external shlelding combined with the shielding pr vided by the grout used to fill the void spaces, extemal dose rates on the outside of the vessei ut be less than 200 mr/hr. The combined shielding is necessary due to the package containing an estimated 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 provide 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 of 3 Arthur J. Palmer !!!, CHP to Gary Robertson, Head of Waste Management Sectien March 12,1996 Source Term An individual pathway analysis, to determine the impact of the increase in the site' source term with the disposal of this component, hrs been completed. This analysis entitled " Trojan Reactor Vessel Dose Analysis"is enclosed as Attachment I to this
.Y911h The grouting process of the Trojan reactor vessel will be the same process us Trojan steam generators. The steam generators were filled with a low densit containment through several fill and vent connections. Approxirnately 2 days later th generators were moved out of containment and inspected for additional vo from settling. An approximately 5% void was retilled using the same grout prior t closure clthe generaters-License Variance A License variance request will be submitted to exceed the possession limit of sec (60,000 curies) of the site license %%I019 2. The variance will be temporary to cove the reactor vessel shipment only. The variance request will be for approximately 2,100,000 curies.
Waste Classl0 cation 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 information is based on the latest data available but is subject to revision onc removed and can be better characterized, Any revision to the fmal turie content w likely be in the downweid direction sir" the present calculations assume the most conservative assumptions.
Your timely review and comments with this matter would greatly be appreciated contact me at (800) 567 2372 if t can be of any further assistance.
Sincerely Arth J. Palmer !!!, CHP Chief Radiological Control A Safety Officer
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TABLE OF CONTENTS
- 1. GROU N D WATE R P ATHWAY DO S E AN ALY SIS.................................................... 1 1.1 P U R P O S E................................................................. ~... m
..m..m.....~m.1
. ~.......... m.. m... 1 1.2 APPROACH.......
..............................m....m.......m.
1.3 DATA.............................................................................m.......m......~mm.m..m1
- 1. 3.1 S o u r c e T e rm............................................................
- 1. 3. 2 A s s u m p t i o n s...........................................................,,...........
- 1. 4 A N A L Y S l S.........................................................
.........m.....2
..............m.3 1.5
SUMMARY
- 2. DIRECT EXPOSURE DOSE ANALYSIS.............,
............... 3
2.1 INTRODUCTION
............3
- 2. 2 C 0 N C E P T U AL M O D E L...................................................
2.3 APPROACH................................................................................4 2.3.1 Selection and Justification of Mocel..................................................... 4
- 2. 3. 2 A s s u m pt i o n s..............................................................................
2.4 DATA.................................................................................................5 2.4.1 Ar e a I G e o m e t ry........................................................................
2.4. 2 Cla d ding and C oye r T hickne s s.................................................................... 5 2.4. 3 Wa st e and C oy e r D e n s ity.....................................................................
2,4. 4 S o u r e e An a l y s i s.......................................................................
- 2. 4. 5 D a t a S u mm a ry............................................................................
- 2. 5 M O D E L S I M U LAT I O N S......................................................
2.6
SUMMARY
............................................................................................................7 March 11.1996 l
4
ito,on F.x vcsse US Ecology
- 1. GROUND. WATER PATHWAY DOSE ANALYSIS 1,1 PURPOSE This calculation examines the expected ecse via the ground water pathway attributed to the disposal of the Trojan Reactor Vessel as a singte component at the US Ecciogy Low level Racioactive 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.
The metal reactor vessel is buned intact in the trench.
Infiltrating water (from precipitation) that migrates through the cap comes in centact with the reacter vnsel, j
and leaches radionuclides from the vessee.
ne leachtte migrates downward to the water table, where it is oiluted in the uppermost aquifer. A hypc'hetical well withdraws the water for a subsistence farming family, that irrigates a vegetable garden and pastureland for a dairy cew. The primary deso for this pathway is via ingestion.
1.3 DATA 4
The data used in this enalysis 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 Term The The activity inventory for the analysis comes from Ponland General Eiectric.
reactor vessel activity is presented for November 1,1997,5 years post shutdown. The activity is separated into sudace contamir. don and activation, with the activation totaling 2.01x10' Curies (Ci), and surface contamination totaling 357.9 Curies.
The deselopment 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 radionuclides 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
[
1.3.2 Assumptions The assumptions used in this analysis are as follow.
The peak concentrations cf leached constituents in the leachate are assumed to reach This i
the ground water at the same time, and to reach the well at the same time.
assumption is conservative because.t maximizes the exposure value to the hypotheticalindividual.
There is no time delay associated with the leaching. This assumption does not account The for the gradual leaching and removal cf radionuclides from the reactor itself.
l assumption is that the activity is distributed uniformly, and can be leached uniformly from the reactor vessel. The assumption is conservative because the radiorwijdes will leach from the vessel slowly, thereby decreasing the amount of radionudide available for transport.
Solubility and distributien coefficients frem the Closure Plan will be used, i ne leaching concentration of racienuclides frem the metal reacter vessel are likely tu te lower than the so!ubility of the racicnuclices, but s;rn a caer values ceuld be focated. thFat conservatively larger values will be used The assumptions described aboye represent an upper councing condition on the expected dose attributed to the disposal cf the Trojan reactor vessel via the ground-water pathway.
1.4 ANALYSIS The transport of the scurce term isotepes tnrough the vadese zene was modeled using the TRANSS program. Two infiltration rates were moceled,0.2 inches / year end 0.05 These two cases simulate the infiltration cf the area surrounding the inches / year.
Facility (natural conoitions, as if the final cap was completely ineffcctual), and the l
conditions over the Facility with the final cap in place and functioning as designed, l
l respectively. These infiltration rates are the result of natural and expected precipitation l
at the Facility.
The maximum concentration for each radionuclide was selected for the input to the aquifer. Mixing and dilutiori occurs in the aquifer. The flow through the aquifer is i
greater than the recharge rate from infiltrating water, therefore dilution of the leachate occurs. The dilution factor is dependent on the infiltration rate, and is calculated to be about 0.003 for 0.2 inches / year, and 0.0007 for 0.05 inches / year, Multiplying the leachate at the ground water table by the dilution factor yleids the expected concentration in the hypothetical well.
-Well4mcentrations less than 1x10* PCl/L are eliminated from further analysis, This left t, van radionuclides for dose analysis. The isotopes for dose analysis are listed in Tab a 3.
i l
March 11,1996 2
%en m.cuci vs Ecology 1.5 5UMMARY The well concentration cecomes the input to the PRESTO ll 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 mrem / year to any organ.
The PRESTO Il model did not identify dose factors for Nb 94 in its internal database.
The dose for Nb 94 was calculated by the equations upon which the PRESTO il model is based. The niobium dose is checked in another calculation. The dose calcu from Nb 94 was far less than 0.01 mrem / year.
2 DIRECT EXPOSURE DOSE ANALYSIS
2.1 INTRODUCTION
This calculat;on examines the cirect gamma exposure possible from the disposal at the Richland Facility of the intact Trojan Reactor Vessel. The reactor vessel is to be snielded by a soil cover. The exposure of an individual while outside man made l
structures was examined.
2.2 CONCEPTUAL MODEL The conceptual model for the direct exposure calculation is that an intruder could be For this exposed to direct gamma raciation 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 stancing 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 may receive direct exposure by intermittently patsing over the Facility area 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 findings of the Hanford Future Sito Uses Working Group has identified this area as the location for waste storage from cleanup activities from the rest of the Hanford Reservation One March 11,1996 3
pathway, direct exposure via a basement ccnstructicn in the v.aste, was eliminated as a.
result of the assumption of post cicture institutionci control to be exercised outside of the Facility. The Central Plateau region, in which the Facility is located, is anticloated to be the waste storage area for the cleanup activities to be performed at the Hanford 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 the following assumptions.
Institutional centrol fcr the Central Plateau area will eliminate leng 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 lcr the above scenarios is the first case involving the Facility rarator/ maintenance person (receptor).
The cenceptual model for this pathway censicers the radiatien source to to a large The recepter is solid mass, since !" scurce is the waste m the turial trenches.
assumed to be stancing en tcp cf the trench :ap, wnich serves as a shield to direct radiation.
The effects of cover thickness en the cirect exposure cose rate were Tne analysis performed fer the Facility was undertaken to provide examined.
conservative radiological impacts to a hypcthetical maximally exposed individual. The assumptions and data are censidered extremely censervative for the cenditions at the Facility, and are discussed in the following sections.
2.3 APPROACH 2.3.1 Selection and Justification of Model The computer program MICROSHIELD (versien 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 MICROSHIELD are d;.:tmented and the prog.am 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 vesselwill be disposed as a sing'e, intact unit.
- 2. The reactor vessel activity is shielded by steel and concrete to reduce surface dose rates to less than 200 mrem / hour.
Side cladding thicknesses were estimated to reduce the reactor surface activity to levels below this limit. These thicknesses were then included as side and end cladding of the reactor vesselin the disposal modeling.
4 March ti.1996
US Ecology
- 3. A t:tal thickness of 16.5 fect of soil will cov r the react:r vessel.
Sensitivity analyses were performed that examined the exposure and dose for changes in sal cover thickness in two foot increments.
- 4. Radionuclide activities of the reactor waste 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 wns calculated.
- 6. The trench caps will be maintained such that major soil crosion is repaired. Because of the Facility's location and the soil cor.ditions, the soil erosion potential is assumed to oe minimal.
2 4 DATA 2.4.1 Areal Geometry The cap area is 3.64x10' square feet (rectan0ular 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 area 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 vesse 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 vt 2.6 g/ctn' for the scurce material.
2.4.2 Cladding and Cover Thickness The thickness of the waste is equivalent to the reactor vessel dimensions.
The p evious Closure Plaq' 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 thickness over the source of about 16.5 feet. Additional soil backfill (reportedly 1 to 10
'US Ecology,1990, Site Stabilization and Closure Plan for t.ow l.evel Radioactive Waste Management Facility, Richland, Washington.
March 11,1996 5
vo e..a3 ft) may be placed Cver the trenches prior to the cap placement, to level the Facility. '
This additional material was not considered in this c)lculation, as its depth is unknown.
The side cladding thicknesses were estimated to produce a design exposure rate of 200 mrem / hour.
The disposal design identified eight inches of steel shielding.
Concrete cladding was added to reduce the exposure rate. A sensitivity analysis was performed using MICROSHIELD for different coacrete thicknesses; a concrete thickness of six inches yielded an exposure of *.32 mrad / hour. Therefore, eight inches 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 Cover Density The model requires the shield material be defined. The shleid materialis the earthen The side and end cladding (steel and concrete) is included as part of the cover.
shielding described above.
MICROSHIELD contains attenuation information for concrete, but not for soil. The shield (cover) was simulated as concrete, but the density was reduced to match the cover soil density cf about 94 pounds per cubic foot (lb/ft')
(1.5 g/cm'). The waste was simulated as it:r but the density was reduced to mate.
the censity of the source (about 2.6 g/cm').
2,4.4 Source Analysis The total activation activity in the Reactor Vessel is 2.01x10' curies, and censists of 11 isotopes, listed in Table S.
2.4,5 Data Summary The following summarizes the data used in the analysis of the direct exposure pathway.
- 1. Soil used for trench backfill and trench cap materialis sand with a density sand *, and matches soil densities meP=ured at the Facility'g of 94 lb/ft' (1.5 g/cm'). This is an average dry unit wei I
- 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 B feet and a thickness cf 40 feet.
I i
Lambe, T.W., and R V. Whitman,1969, Soil Mechanics, John Wiley & Sons, New Yo*, p. 31, Table 2.2 2
l
'Sergeron, M.P., et al,1987, Geohydrology of a Low Level Radioactive Waste Olsposal Facility, Richland, Washington, Battelle, Pacific Nonhwest Lab., p. 39.
March 11,1996 6
l l
US Ecology
- 4. The waste has an average density of about 160 lb/ft' (2.6 g/cm ).
The
)
8 average waste density is calculated from reactor vessel density, volume and void ratios.
S. The source term waste activity (summed over 11 isotopes) is about 2.01 mi!! ion Cl. The waste is assumed to be disposed of intact within the t
trench.
- 6. The trench cap is modeled as 16.5 feet thick soove the waste.
2.5 MODEL SIMULATIONS The model was run for a referen;e value of 16.5 ft of shleid (TROJAN 4.MS4 and TROJAN 4.ASC). A sensitivity analysis of shield thlCKness Was performed for Valueh of 0 to 16 feet in 2 feet increments to examine the results of pcssible cap removal.
2.6
SUMMARY
The results of it.is analysis incicate that the oitect exposure route is not a pathway that requires additional analysis. Reported averago ex'.arnal 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 mR/hr to 0.075 mR/hr). The trench backfill and cover design reduce the amount of radiation to below background levels.
The dose rate for an individual standing'over the intact 16.5 foot trench ca estimated to be 2.3x10"' mrad /hr (5.5x10' mR/hr (6.2x10"' mR/ day). This value was compared to the minimum average gamma exposure rate for 1971 1972 at the Hanford Reservation 100 Area of 0.15 mR/ day (ERDA,1975) and determined to be insignificant.
March 11,1996 7
TABLE 1 DATA 4-Parameter Value Trench depth 45 ft Vadose zone thickness 265 ft Vadose zeno ocrosity 0.30 Saturated zone thickness 100 ft Saturated zene ocrosity 0.10 Infiltratien (surrounding area) 0.2 in/yr Infiltrat..n (through cao) 0.05 in/yr l
3.64 x 10'souare feet Trench, intertrench area i
March 11,1996 8
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US Ecology TABLE 2 l'
SOURCETERM isotope Activity Half Life Am 241 2.54E 01 458 4
4 C 14 2.18E+02 5,730 i
j Nl.59 9.53E+02 8.00E+04 Ni63 1.57E+05 92 i
Nb.94 3.29E+00 2.00E+04 l
1.92E 01 86.4 Pu 238 Pu 239/Pu 240 2.15E 01 24.390 /6.600 Pu 242 1.08 E.03 3.79E+05 Sr.90 2.13 E+ 00 27.7 Tc.99 7.06E 01 2.12E+05 1
-I March 11,1996 9
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TABLE 3 RADIONUCLIDES FOR DOSE ANALYSIS isotope Infiltration = 0.2 inlyr Infiltration = 0.05 in/yr Hypothetical Well Hypothetical Well Concentration Concentration (pCl/L)
(pCl/L)
C 14 3.90E+01 S.65E+00 Nb 94 7.2 E 14 4 03E 16 Ni59 3 21E 03 5.73E 05 Pu 239/
1.78E 07 2.29E 09 Pu 240 Pu 242 5.49E 07 4.96E 08 Sr 90 2.44E 15 1.19E-17 Tc 99 2.65E Oi 5.80E-02 March 11,1996 10
US Ecology Trojan Rx Vesset TABLE 4
SUMMARY
OF RESULTS Organ PRICH22. PRS PRICH23. PRS Infiltration Rate Infiltration Rate 0.2 in/yr 0.05 in/yr Body 0.261 0.038 Red Marrow 0555 0.080 Thyroid 0.160 0.024 Note: C 14 is major centributing racienuclide March 11,1996 11 l
TABLES l
DIRECT EXPOSURE ACTIVITY Isotope Activity ICurles)
H.3 6.55E+02 C.14 2.18E+02 Sb.125 4.04 E-01 Mn.54 2.16E+03 Eu 152 2.54E+0i Fe.55 6.E7E+;5 Co 60 1.15E+06 Ni59 9.53E+02 N163 1.57 E+05 Nb.94 3.29E+00 Tc.99 7.06E-01 TOTAL 2.01 E+ 06 March 11,1996 12
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ATTACilMENT IV
tTATE OF WASHINGTON DEPARTMENT OF HEALTH
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olVislON OF RAotATION PROffCTION AordustrialCenter, Elds,3 e P.O. Boe dist?
- Olsmpia. Hashington v830Ain2?
June 10, If 96 Art Palmer, Chief Radiological & Safety Officer US Ecology, Inc.
120 Frank 1h Road Oak Ridge, Tennessee 37830
Dear Mr. Palmer:
- n11s is in response to your letters dat:d 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 Commission's January 17,19). atal Branch Technical position on Concentration Averaging and Encapsulation.
As a result, the classification of this waste appears to conform to your state of Washington radioactive Materials license WN 1019 2, and WAC 246 249M.
Please be advised, however, it is the generator's responsibility to ensure compliance with waste classificstion and waste form. It is requested that if any of the data used for waste calculations change signiScantly, that the resed numbers be submitted to the department, tr you should have any questions, do not hesitate to contact me.
Sincerely, fb Mikel J. Elsen Radiation Health Physicist cc:
WDOH Richland, WA
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Amrean Ecq Crovaten O lis463B?tt 120FfennhnRose Fan 616 483 6398 osa Mege, TN 37630 WWN Apnl 17,1996 Mr. Mikel J. Eisen Radiation Health Physicist Washington Depanment of Health P.O. Box 47827 Olympia, Washington 98504 7827
Dear Ah. Elsen:
r This is provided in response to you letter of April 2.1996 transmitting your comments regarding the disposal of the Trojan Reactor Vess d. D:h of your comments is reprinted below followtd by our response.
Comment 1: "If 7:en does PGE expect tofmd out of the NRC uitiisne a C of C on the 1
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 alternatives for shipment of the package. The NRC's preference was to i
license it as a Type B package. PGT bussed 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 properly designed and the shipment is completed in a well controlled manner. Initial conceptional design meetings will be held with the NRC in May. The fmal Safety Analysis Repon will be submitted late 1976. It is expected Qe Certificate of Compliance will be issued late 1997.
Comment 2: "Please apply the waste classifcanon calculation. It was not submined wnh pur request. Will the acnvanon analystsfor the vessel with internals be analytically venfied wnh any samphng? A sample may be able to be compared to the actnation analysis to venfy the waste classsfication 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 N
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Mr. %kel1 Elsen Apnl 17,1996 Page 2 The two approaches historically used to characterit.e routinely generated activated metals components incluced,(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 rad.vactivity method in the BTP and have only been employed to defme Pan 61 sealing factors These scaling factors are then used in conjunction with radiation levels to quantify the base radionuclide Co 60 to which the scaling factors are applied The direct sample approach to scaling factor deterrrunation peaked in usage about 1987, and has not been used at all since 1992 This is based on its time and cost, the difTiculty with the representative sampling of routine components. and the uncertainties in the anai -"al results Ni 59 was alwrvs scaled from measured Ni 63 2.nd Nb 94 concentrations were always definu as LLDs The direct sampling method has never been used to deternune scalir g factors for reacter 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 w hich prevents accurate radiochenucal analysis of the samples Tae concentrations of significant radionuclides in internals components vary as a function of the base metal's nickel and contaminant content and the integrated Cux m 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 congarisons between sample results and activation 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. FlSSPEC and 02 FLUX computer codes and ANISN and ORIGEN computer codes obtained through the Oak Ridge National Laboratory's Radiation Shielding Infonnation Center.
Ancillary calculations were performed using TLG's ANISNOUT and 02 READ computer codes.
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Mr. Mikel L Elsen April 17,1996 Page 3 The one dimensional neutron transport model was normalized with data obtained from a Westinghouse Elec*ric Corporation repon on a reactor vessel radiation surveillance specimen which was removed from the plant in 1990 Based on the above, PGE does no; 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 NRC BTP gross radioactivity method of characterization and was employed during the Shoreham and Yankee Fowe projects.
Comment 3.
"Is the Pu 24) that is used m your pathu ay analysis decayed mio Am-241?
Our results mdicate 2.5 Ci u here the proposal shows 0.25 Ci ofAm-24).
The difference could be m the Pu-241 imtialacnvirv re.g.,11. ~ Co. Plecse show how the valuefor Am-241 was arrived at. "
tesponse:
The value of 0 25 Ci of Arn 241 documented in our waste classiiication report was based on the fractional percentages of the sample results taken i
from a S'O tube in 1994 The re ults 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. Combined 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 activity 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-241 was 3.69 x 10' and 5.41 x 10'* pCi/L for recharge rates of 0.2 in/yr
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j Mr. Mikel J. Elsen April 17,1996 Page 4 and 0 05 irtyr. respectively Because Am 241 is adsorption controlled.
multiplymg the increased activ.ty rat.s times the concentration at the water ta'ule should yield an approximatim of the concentration at the water table with the new source valut. These values are 7 38 x 1023 and 1.08 x 10'"2 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.~'. ^ 25 curies of Am-241. The activity shift still maintains the waste as Class C.
Over time the Pu 241 activity d< creases 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 ;71 euries. The sum of the fractions at that time would actually decrease.
Comment 4 "In the patinsav analysis at is assumed that the package 's exterior dose is only 200 mr hr. It'nh one of the steam generators. a 430 mr hr hot spot wasfound and enclosed before shipping, ll'ith this m mmd, perhaps the PA should use i R hr as an nunal exposure before takmg any shieldmg mio accoun. This dose rate is the most conserunive and should not impact the outcome qf 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 Transportation shipping requirements However, for purposes of evaluating this possibility, the PA 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 frori. the PA results for the 200 mR/hs cases. That is, multipiying 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.
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Mr. Mike.! J Eisen April 17,1996 Page 5 Soil Cover Thickness (ft)
Dose Rue in Air Dose Rate in Air (mrad'hr) E 200 rrAhr (mrad!hr) G 1 R/hr 2
05 1
3 416 x 10' 1705 x 10 2.5 l
3 628 x 10" 1.184 x 10*
45 I
3 654 x 10" 1.827 x 10 2 65 I
3 552 x 10 1776 x 10" 4
Sf I
3 361 x 10" 1
1.681 x 10*
10 5 l
3.120 x 10
l 1.560 x 10
j 12 5 I
3 862 x 10"'
1931 x 10"#
i 14 5 1
3 583 x 10"'
I 1792 x 10";
16 5 I
I314 x 10"'
I 1657 x 10*
i Comment 5; "Willany shielding be u elded onto the usseP is this weight used in the waste classificanon process'"
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
' Comment 6: "Since this os not a rounne shipment, proceduresfor the handhng and disposal of this uaste should be developed and submitted to the department for review. Addinonally, the proposed trench set-up (e.g. whe. c the ramp (s), thefinal vesselplacement, and backfilhng should be addressed. If the proposed trench is different that what is contained in the March 6, 1991 Comprehensive Facihty Unh: anon Plan. Document 200-DOC-001.
Rev. 3 the department must approre 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.
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Mr. Mikel L 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 cenerators and pressurizer as well as the Pathfmder reactor vessel. Since the dose rate on the reactor vessel is expected to be less than 200 mR/hr, 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 depurrment 's opmion that the basement construction scenario bc exannnedin the patinvay analysis.
Response
The casement scenario has. in etTe:t, been.un as a part of the sensitivity analysis performed for the Direct Exposure Dose Analysis. This sensitivity analysis examined the etTect 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.
"What is the processfor ver#ing that the void spaces arefilled u oth grout? "
Response
The grouting process will be developed to provide the best engineering assurance that the package is being completely filled. This will include filling the vessel until a positive vent of Low Density Cellular Concrete is obtained from each vent connectior Any small voids that form durint.ne filling procen, 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 visualinspection will be completed from each top vent to ensure the packaga 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 internal free volume. This process will ensure the void spaces are filled with grout.
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April 17,1996 Page 7 i
Comment 9: "You have indicated that the vessel willhave 337.9 cunes ofsurface contannnanon. How will the contcnnnanon 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 internals The outside of the reactor vessel will be below Depanment of Transportstion 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,
/Vm?
N Arthur J. Palmer, CHP Chief Radiological Control and Safety Officer Att.
4 w.. w..
s o
PORTLAND GENERAL ELECTRIC CALCULATION SHEET Calculation No, N/A Revision 9
Sheet t
Cont'd on Sht Preparer -
wehiel Murdxk Date
'13/96 Verifier d[,,o
- Date 44c 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 use the radioactivity, as analyzed for and, identified in the Trojan Decommissioning Plan along with the most current revision of the Radia*!on Protection Manual Procedure, RPMP 4," Determination of Radioactive Material Shippirg 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 I
will not be documented or follow the same format as an approved PGE calculation.
RESULTS 9 The vessel results in being (Class C) waste.
+ The package contains 3.56 gms of plutonium.
+ Table 1 results. 335 1
+ Table 2 r.tsults.299
PORTLAND GENERAL ELECTRIC CALCULATION SHEET Calculation No, N/A Revision o
Sheet 2
Cont'd on Sht Preparer
% chiel Murdock Date 2/13 s 6 Verifer M$e Date
- Mk COMPONENT ASSitMPTIONS
+ 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) 4 The majority of the activation is located in the region between the upper and lower core plates including the vessel cladding and walls.
9 The surface contamination (although <1% of the total activity)is considered to be distributed over all the reactor vesselinternal surfaces.
"4 The activity contribution of the Incore Flux Thimbles, currentiv 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.
COMPONES i DTMENSTONS
+ Burint/ Envelope Volume = 7951 ft*
(Does not include any additional steel shielding or penetration closures) 4 Void V lume = 5295.3 ft*
(With internals)
+ Displaced Volume = 2655.7 ft* or (75,201,049 cc)
(Envelope volume minus major void volur.:es) 4 Material Weight = 591,080,249 gms (Displaced Volume x 7.86 gms/cc]
ft eh
PORTLAND GENERAL ELECTRIC CALCL1ATION SHEET Calculation No.
N'/ A Revision n
Sheet Cont'd on Sht Preparer u u uu e Date - w on Ikt ?au Date
- lrht Veriller.
/
REACTOR VE5SEL ACTIVITY (5 years Post Shutdown,11/1/97)
SURFACE
- DNTAMINATION ACTIVATION 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 Ce144 1.08E 01 bb54 1.98E-01 2.16E+03 Gu 152 2.54E+01 Fe 55 6.39E+ 01 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 Sr90 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 r
.m..
Fzrm RP,73
' $Npment Number N/A Pageecf5 P ck:ge Number RV & Int 4
a
- 12. Determinston of Concentraton 1
M: tend Volume 7.52E+07 cc Matenal Weight 5.91E+08 - gm 0
E A-B C
- setece
- SNp Cate Cone Cone Acuvity isotope Isotope (mci)
(uCl/cc)
(vCi/cc)
(Ax1E3)/D (Ax1E3)/D
<5 yr T in
> 5 yr T in H3 1.87E + 02 2.49E 03 H 3 (A) 6 55E+05 8.71 E + 00 1
C 14 -
2.65E + 02 3.52E 03 C 14 (A) 2.18E+05 2.90E + 00 Mn 54
_1.98E + 02 2.fDE 03 Mn 54 (A) 2.16E + 06 2.87E + 01 Fe 55 6 39E + 04 8.50E 01 Fo 55 (A) - 6 97E+08 9.27E + 03 Co 60 2.29E + 05 3.05 E + 00 l
Co 6') (A) 1.15E + 09 1.53 E + 04 Nb 94 (A) 3.29E + 03 4.37E 02 Ni 59 (A) 9.53 E + 05 1.27E + 01 a
Sb 125 3.86E + 03 5.1'IE 02 Sb 125 (A) 4.04E+02 5.37E43 Eu152 (A) 2.54E+04 3.38E 01 C:-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
1.57E + 08 2.09E + 03 Conc.
Cone Gram hi-(0 (A)..
Sr 90 2.13E + 03 2.83E 02 Isotope Factor of Tc 99 (A) 7.06E + 02 9 39E 03 (nCl/gm)
(gm/ mci)
Isotope l129 (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 239/10 2.15E + 02 2.86E 03 3.64E 01 1.00E 02 3.44E +00 1
) Pu 241 1.17E + 04 1.56E 01 1.98E + 01 9.60E 06 1.12E 01
~
Cm-242 3.82E-02 4.81E 07 6.12E 05 j
Pu 242 1.08E+ 00 1.44E 05 1.83E 03 9.SOE+03 1.74E + 04 3.56 4
SumB Sum C Sum H (uCl/cc)
(uCl/cc)
(gms. of Pu)
RPMP4
- Peg) 22 of 35 Rev.17 i
Page 47 of 93 1.
h.
Verify that tNs is a true computer generated copy of RPMP 4, Rev 17-4
..,_r
Tcrm RP 73 Shipment Numeer N/A Page 5 of 5 Package Number RV & Int
- 13. Evaluation of results for Bunal Site Lirrutations and possible DOT /NRC Form 741 Shipment.
- a. Determine total grams of SNM:
3.56-
+
0
+
0
=
3.56 Grams of Grams cf Grams cf Grams of Pu U 235 U 233 SNM If SNM > or = 1 gram, notty Radwaste Supervisor.
- b. Burial Site TRU ls TRU with T 1/2 > 5 yrs exceot Pu 238, Pu 239/240, Pu 241 and Cm 242 from Column F.
Burial Site TRU.
7.03E 01 Limit <10 nCi/gm (nCi/gm)
- c. Waste Class TRU is TRU with T.1/2 > 5 yrs anc PU 230 and Pu 239/240 from Column F Waste Class TRU:
1.39E + 00
_ _ 3 54E-01 7.03E 01
+
3 25E 01
=
(TRU)
(Pu 235)
(Pu 239/240)
(nCl/gm) 14, Sum allisotopes with half hfe <5 yrs:
9.30E + 03 Limit <7 UCi/cc (uCi/cc; 6
15, Sum Allisotopes with half hfe >5 yrs:
1.74 E + 04 Limit <1 uCl/cc (uCi/cc)
Note: If any of the above hmits are exceeded, nett/ the Radwaste Supervisor.
I Prepared by Michael Murdock Checked by 4ighou
/
RPMP4 Page 23 of 35 Rev 17 Page 48 of 93
- 1. Michael Medock Venfy that this is a true computer generated copy of RPMP 4, Rev 17.
4 4.
f FORM RP 78 WORK SHEET FO;l CETERMINATION OF WAST 1E Ct. ASS:FICATION Class A Class B Class C A
B C
D isotopic Tabu 1 Conc.
Umst Quotant Umst Quotent Umst Quot,ent Isotip:s (uCl/ce)
(WCi/cc)
(A/B)
(WCl/cc)
(NC)
(WCVcc)
(ND)
C 14 2 90E + 00 -
8 00E + 00 3 63E 01 (b) 8 00E + 01 3.63E 02 7c f 9 9 39E43 3.00E 01 3.13E 02_
(b) 3.00E + 00 3.13E 03 Ni 59 1 27E + 01 2 20E + 01 5 77E 01 (b) 2.20E + 02 5.77E-02 Nb 64 4 37E 02 2 00E 02 2.19E + 00 (b) 2.00E 01 2.1 BE-01 TRU f a.c) 1.39E + 00 1.00E + 01 1.39E 01 1.00E + 02 1.39E 02 Pu 241 (a) 1.98E + 01 3.50E + 02 5.66E 02 (b) 3.50E + 03 5.66E 03 Orn 242 (a) 6.12E 05 2 00E + 03 3 06E 08 (b) 2.00E + 04 3 06E 09 Sum of Table 1 Ouctents 3.35E 00 3.35E 01 NOTE: TRU with T.1/2 >5 yrs - all TRU except PW 238, PW 239/240. Pu 241 and Cm 242.
C: ass A
"'u s E Class C A
B C
D isotopic Table 2 Conc.
Umst Quotant Umrt Quotant U mit Quotent isst: pes (uCl/cc)
(WCl/cc)
(A/B)
(wCl/cc)
(A/C)
(uCi/cc)
(A/0)
H3 8.71 E + 00 4 00E+ 01 2.18E 01 (d)
N/A (d)
N/A Cs 60 1.53E + 04 7.00E + 02 2.19E + 01 (d)
N/A (d)
N/A Ni43 2.09E + 03 3.50E + 01 5.97E + 01 7.00E + 02 7.00E + 03 2.99E 01 St90 2.83E 02 4 00E 02 7.08E 01 1.50E + 02 7.00E + 03 4.04E 06 Cs 137 1.00E + 00 4 40E + 01 4.60E + 03 000E + 00 Nucides w/
(T 1/2.5 yt) 7.00E + 02 0.00E + 00 (d)
N/A (d)
N/A Sum of Table 2 Quoteents 8.25 E + 01 2.99E 01 a. Units are in nCl/gm b + 11 Class A limit is escoeded the waste is Class C or greater.
c.TRU = TRU with T 1/2 > 5 yrs + Pu 238 + Pu 239/240.
d. No Umrt.
NOTE: 11 is not necessary to list on the mandest any nuclide whose Class A quoteent is less than 0.01 Cacept C 14. Tc 99. I 129 and H 3.
Prepared by Michael Murdock Checked by nu Pa9e 34 of 35 RPMP 4 Revision 17 Page 59 of 93 i Michael Murdock Verity that this is a true computer generated copy of RPMP 4. Rev 17
-