ML20203B972

Partially Deleted Final Response to Request Re Current Public Exposure Stds Approved by DOE for Amount of Radioactivity in Metal & Pending Decisions to Lower Exposure Stds to 10 Mrems Per Year.Responsive Documentation Encl

ML20203B972
Person / Time
Issue date:	01/28/1999
From:	Lopez A NRC OFFICE OF ADMINISTRATION (ADM)
To:	Eikenburg S AFFILIATION NOT ASSIGNED
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ML20203B938	List: IA-99-111, Partially Deleted Final Response to Request Re Current Public Exposure Stds Approved by DOE for Amount of Radioactivity in Metal & Pending Decisions to Lower Exposure Stds to 10 Mrems Per Year.Responsive Documentation Encl ML20203B935 ML20203B972
References
FOIA-99-111 NUDOCS 9902110093
Download: ML20203B972 (2)

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Department of En:rgy lT '

Washington, DC 20585 -

..- January 28,1999 Ms. Sue Eikenburg Re: FOIA Request No. 9810160001 l

Dear Ms. Eikenburg:

i This is in final response to your request for information pursuant to the Freedom ofInformation l l

Act (FOIA), 5 U.S.C. 552. Your request asked for information about (1) current public exposure standards approved by the Department of Energy (DOE) for the amount of radioactivity in metal, and (2) pending decisions to lower the exposure standards to 10 millirems per year. J A search was conducted of the files of the DOE IIeadquarters Office of Environment, Safety and ,

Health, the ofIice most reasonably expected to possess documents responsive to your request. l The search located the responsive documents identified on the enclosed listing. With the exception of two documents that originated at the Nuclear Regulatory Commission (NRC), the documents are provided to you in their entirety. The two documents, attacfiraents to Enclosure 7, have been transmitted to the NRC for a disclosure determination and a direct response to you.

For further assistance, you may contact Russell Powell, FOIA/ Privacy Act Officer, Mail Code T6D8, NRC, Washington, DC 20555 or telephone (301) 415-7169.

The above referenced number has been assigned to your request and you should refer to it in any future correspondence with the Department concerning this matter. If you have any questions, please contact Joan Ogbazghi of my staff on (202) 586-3595.

I appreciate the opportunity to assist you.

Sincerely, L 't Abel Lopez, Director FOIA/ Privacy Act Division OfTice of The Executive Secretariat

Enclosures:

As listed cc: Russell Powell FOIA/ Privacy Act Officer Nuclear Regulatory Commission 9902110093 990208 PDR FOIA EIKENBU99-111 PDR ,

Q40&Hoo93 oft 3

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Enclosures for FOI A Request No. 9810160001

1. DOE Order 5400.5, Radiation Protection of the Public and the Environment, February 8, 1990.

2. DOE Order 5400.5 (Chg 1), Radiation Protection of the Public and the Environment, June 5, 1990.

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3. DOE Order 5400.5 (Chg 2), Radiation Protection of the Public and the Environment, January 7,1993.

4. Draft DOE Standard, Applying the ALARA Process for Radiation Protection of the Public and Environmental Compliance with 10 CFR Part 834 and DOE 5400.5 ALARA Program Requirements, Volume 1 - Discussion, April 1997.

5. Draft DOE Standard, Applying the ALARA Process for Radiation Protection of the Public and Environmental Compliance with 10 CFR Part 834 and DOE 5400.5 ALARA Program '

Requirements, Volume 2 - Examples & Case Studies, April 1997.

6. Draft DOE Standard Handbook for Controlling Release for Reuse or Recycle of Non-Real Property Containing Residual Radioactive Material, June 1997.

7. Memorandum from Raymond F. Pelletier, Director, Office of Environment Policy and Assistance, to Distribution,

Subject:

Application of DOE 5400.5 requirements for release and control of property containing residual radioactive material, with attachment, reference documents, and examples, November 17,1995.

Two reference documents have been transmitted to the Nuclear Regulatory Commission for a disclosure determination. The documents are a letter dated October 25,1995, from Michael F. Weber, NRC, to Andrew Wallo, DOE, and a Policy Issue dated May 27,1994, for the Commissioners from James M. Taylor.

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O k UNITED STATES <

l f j NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 30886 4001 I o i

%***** October 25, 1995 i l 1

4 i Dr. Andrew Wallo

{ U.S. Department of Energy l Forrestal Building j EH-0g 1

! Washington, DC 20545 l 4

Dear Dr. Wallo:

l In response to your facsimile request, we have performed a review of the

Department of Energy's (D0E) suidance to its field offices on the 4

implementation of DDE.releast criteria set forth in Order 5400.5. As you 1' l

know, the release of any contaminated material to an unlicensed party involves technical and policy concerns that are not completely addressed by existing i j Nuclear Regulatory Commission guidance. l l

l

Recognizing the incomplete nature of existing regulations and guidance, we

agree that DOE field offices should coordinate with the Agreement States or

NRC, as well as other applicable regulatory authorities (e.g., State j

permitting agencies) to ensure that the site specific release limits and the survey and. review protocols are appropriate and acceptable. We also believe .

your application of the Al. ARA process is reasonable for establishing i

authorized limits for material either sent to .a non-DDE landfill or 1 transferred to the public. However, as you know, a 15 mres/yr dose constraint 1 i is currently included within our proposed rule on radiological criteria for l l decossaissioning. Your upper bound constraint level of 25 ares /yr for release i

of property to a landfill or the public could be viewed as being inconsistent  ;

with this proposed value, even though we recognize that the calculated l . radiological impacts from actual. releases are typically well below this value.

l With regard to the collective dose constraint imposed on the release of I

+ surface-contaminated property, we were not able to make a judgement, without i further infomation on assessment methods, on the appropriateness of using the  ;

i 10 person-ren as a threshold value, below which only qualitative screening i

would be required.

3 The case-by-case decisions that NRC has made in releasing land and structures i for unrestricted public use have been typically related to NRC's Site ,

i Deconsnissioning Management Plan sites. Release criteria have been those in

! the " Action Plan to Ensure Timely Cleanup of Site Decommissioning Management  ;

' Plan Sites" (57 fB 13389-133g2) or have.been justified as being consistent with these criteria (See SECY-g4-145 enclosed). These criteria include j surface contamination guidelines in Regulatory Guide 1.86. In the few i

instances where radioactive material' has been or is being considered for -

j transfer to unlicensed entities on a case-by-case basis, impact analyses have

been perfomed and accepted by appropriate approval authorities. These j analyses generally indicate extremely small (i.e., less than I mrem / year) i annual individual exposures and minimal collective exposures.

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At Wallo As you know, the criteria in NRC's proposed decomissioning rule, .when finalized, will replace the above guidance, but only for release of lands and structures. Both EPA and NRC have then indicated that a proposed recycle rule will be developed.

As with the "decomissioning" rule, the applicable dose criteria will be the subject of considerable debate. It is likely that the criteria in the decomissioning rule could influence the selection of criteria for this recycle rule.

I hope this response has been. responsive to your request.

Sincerely, l WA .

. Michael F. Weber', Chief Low-Level Waste and Decomissioning Projects Branch '

Division of Waste Management Office of Nuclear Material Safety and Safeguards t

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POLICY ISSUE May 27, 1994 nf0MadOn) StCY-94-145 f.QB:' The Commissioners 18Q5: James M. Taylor Executive Director for Operations

SUBJECT:

INCREASE OF TRITIUM AflD IRON-55 UNRESTRICTED USE LIMITS FOR SURFACE CONTAMINATION AT SHOREHAM AND FORT ST. VRAIN PURPOSE: s To inform the Comission of the staff's decision to increase certain unrestricted use limits for surface contamination on buildings, structures, and equipment for the decommissioning projects at Fort St. Vrain Nuclear Station (FSV and Shoreham Nuclear Power Station (Shoreham). This increase applies only)to fixed contamination from iron-55 (Fe-55) and tritium (H-3),

and is' a modification to one of the clean-up criteria that the Comission directed the staff to consider in the." Action Plan to Ensure Timely Cleanup of Site Decommissioning Plan Sites" (Action Plan) (i.e., Table 1 of Regulatory Guide 1.86 (RG 1.86)) ,

SUMMARY

Shoreham and FSV are currently decomissioning their facilities with the goal of releasing the buildings, equipment, and grounds for unrestricted use in the short term (i.e., the.DECON option). The major radionuclide identified in contaminated structures and systems during the characterization of these facilities ns cobalt-60 (Co-60). However, during the dismantling of the facilities, both licensees identified concrete containing unexpectedly high concentrations operations.

of H-3 and Fe-55 in areas subjected to neutron radiation during Shoreham also identified Fe-55 in steel used as a linar for the bioshield concrete. The total (fixed plus removable) surface contamination

Contact:

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415-6625 StCY NoIE:

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! levels of H-3 and Fe-55 on the concrete and steel exceed the current j

unrestricted use limits. Smaller concentrations of europium-152 (Eu-152) and Co-60 have also been identified on the activated material. Shoreham estimates

{ that compliance with the current surface contamination limits for the H-3 and Fe-55 in the activated material will require the shipmen l additional cubic meters (2600 cubic feet) of slightly co,t of up toconcrete ntaminated 73

and steel, above that required to comply with the Co-60 and Eu-152 surface contamination limits alone, to a licensed low-level waste facility at a cost of up to $1 million. FSV has estimated that compliance with the current surface contamination limits would require the removal and shipment of up to 260 additional cubic meters (9300 cubic feet) of activated concrete, at a cost of up to $4.5 million. Both licensees submitted requests to increase the unrestricted use limits for H-3 and Fe-55 surface contamination based on their conclusions that the risks from H-3 and Fe-55 are very small, at the current limits, and that application of the current limits to the activated material at their facilities is not in accordance with the as low as is reasonably achievable (ALARA) principle.

The staff evaluated the risk from H-3 and Fe-55 surface contamination relative to the other radionuclides listed in RG 1.86. The risks from Fe-55 and H-3 surface contamination at their respective RG 1.86 limits were found to be )

lower than the average risk from the other nuclides by at least a factor of  ;

i 400. The risks are lower because Fe-55 emits only low-energy X-rays and H-3 '

emits, only low-energy beta radiation. A detailed discussion of the relative risks, and additional background information, is contained in the enclosure.

Based on the magnitude of the disparities between risks, the staff concluded that it would be ALARA to increase the surface contamination limits for H and Fe-55 at FSV and Shoreham, and that the magnitude of the increase could be ,

l limited to ensure that the risks from H-3 and Fe-55 remain consistent with the risks from the other nuclides in RG 1.86. The staff contemplated raising the total average surface contamination limits for if-3 and Fe-55 from the current '

5,000 dpm/100 cm' limit to 2,000,000 dpm/100 cm'. At this ley?l,' the risk l

from Fe-55 would be equal to the average risk from the other nuclides in RG 1.86, and the risk from H-3 would be less than the average. tiowever, in i consideration of ALARA, the staff selected 200,000 dpm/100 cm' as the new l total average surface contamination limit for H-3 and Fe-55 at FSV and '

Shareham.

RG 1.86 also contains maximum limits that are 3 times the average limit.

Accordingly, the maximum limit for total H-3 and [e-55 surface contamination was raised by a factor of 3 to 600,000 dpm/100 cm The existing limits for removable H-3 and Fe-55 surface contamination were i also evaluated. Increasing the removable contamination limits is not i considered ALARA, since standard remediation techniques are capable of lowering removable contamination to levels below the current limit. ,

Therefore, the removable limits for H-3 and Fe-55 will remain at the current level of 1000 dpm/100 cm2 t

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The Comissioners The primary bases for the staff's decision to increase the surface contamination limits for H-3 and Fe-55 were 1) that the risk from the increased limits are consistent with the risk from other nuclides, and 2)

ALARA. However, to provide additional information, dose assessments were performed assuming that, contamination is present at the increased limits At 200,000 dpa/100 car , the estimated doses from H-3 and Fe-55 surface contamination are 0.01 and 1.1 E-03 mSv (1.0 and 0.11 arem)/y, respectively.

These potential doses will decline as Fe-55 and H-3 decay with half-lives of .

. 2.6 years and 12.2 year,s, respectively.

. The above calculations assume that the dose from the activated concrete is from the inhalation and ingestion of material resuspended or removed from contaminated surfaces, whicle are considered the most probable exposure J pathways. Although considered unlikely, the staff also estimated the potential dose assuming that the activated concrete is removed from the buildings and disposed without restrictions. For this case, the groundwater pathway, and the other exposure pathways in the residential farmer scenario, were evaluated. The resulting potential doses from the unrestricted disposal of the material at Shoreham and FSV were estimated as 0.012 mSv/y (1.2 mrem /y) and 0.019 mSv/y (1.9 mrem /y), respectively. These modeled doses decline

, rapidly with time following the maximum due to the conservative assumption that all of the H-3 is leached in the first year. For example, the fourth and sixth year doses at both facilities are less than SE-03 and 2E-04 mSv/y (0.5 ~

and 0.02 arem/y), respectively.

The modified Fe-55 and H-3 limits discussed in this paper are based 'on ALARA considerations specific to the FSV and Shoreham decomissioning projects.

. However, volumetric contamination of activated materials at decomissioning power reactors is a generic issue and the staff anticipates that other decomissioning reactors will make similar requests for exceptions to RG 1.86 for H-3 and Fe-55, and that such requests would likely be approved. This' l issue should be addressed generically in the guidance developed to implement  ;

the rulemaking on Radiological Criteria for Decomissioning. This guidance would supersede or modify RG l.'86. The staff expects to notify Shoreham and FSV that the limits have been increased no later than June 3, 1994. This early date is associated with the time required to remove and ship low-level waste, if necessary, to Barnwell prior to the June 30, 1994, deadline. A meeting with the Comission assistants to discuss this matter would be useful.

COORDINATION:

The Office of the General Counsel ~has reviewed this paper and has no legal objection.

. / 1 y or -

. ecutiv Director for Operations

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Enclosure:

As stated .

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.'l ENCLOSURE ADDITIONAL BACKGROUND AND TECHNICAL INFORMATION In the Action plan, the Commission directed the staff to consider existing guidance, and ALARA, when detennining a site's suitability for unrestricted use

'pending the final rule on " Radiological Criteria for Decomnissioning." Table 1 of- RG 1.86, which contains surface contamination limits for buildings, structures, and equipment, was listed in the Action Plan as one of the existing criteria to consider. Accordingly, the staff approved the limits in RG 1.86 as the unrestricted use criteria for surface contamination at FSV and Shoreham.

1 The experience, to date, in applying the RG 1.86 surface contamination limits as unrestricted use criteria has been positive. In general, the limits are practical and ALARA, and have been applied by Nuclear. Regulatory Commission {

licensees undertaking deconnissioning. However, on December 23,1993, the Public

{

Service Company' of Colorado (PSC) requested that NRC increase the surface l contamination limits for Fe-55 and H-3 at FSV. The Fe-55 and H-3 contamination ' I identified at FSV resulted from neutron activation and is primarily located in the concrete comprising the former prestressed concrete reactor vessel. PSC <

contends that the surface contamination limits for these two nuclides should be i increased since: 1) the potential health and safety risk from these nuclides is very low; 2) the relative risk from these nuclides is very low. compared to the '

other nuclides of concern at FSV, predominantly Co-60 and Eu-152; and 3) compliance with' existing limits for these nuclides would cost an additional $4.5  ;

million above the cost required to comply with the surface contamination limits <

for Co-60 and Eu-152. . ,

In addition, on April 22,'1994, Shorehar requested that NRC consider revising the release criteria for Fe-55 and H-3 because of the recent identification of ,

concrete and steel, from the biological shield, that contains elevated levels of  ;

H-3 and Fe-55 as a result of neutron activation. Shoreham asserts that applying l

.the current unrestricted release criteria for Fe-55 and H-3 would cost up to 51  :

million without appreciable decrease in potential risk to public ' health and safety, and that Fe-55 and H-3 pose significantly lower risk than other nuclides in RG 1.86.

In response to the FSV and Shoreham requests, the staff reviewed the technical bases for Table I of RG 1.86 to determine if the limits for H-3 and Fe-55 are  !

inconsistent with the other nuclides in Table 1. The RG 1.86 limits were developed using a semi-quantitative evaluation of relative risk based on the  !

maximum permissible concentrations (MPC) for air and water listed in 10 CFR Part  !

20. The starting assumption in . developing the limits was that licensees should '

not be expected .to lower surface contamination below the existing environmental background levels caused by fallout from the atmospheric testing of nuclear devices. The predominant radionuclide found in the environment as a result of  :

atmospherictestingigstrontium-90(Sr-90);thebackgroundlevelofSr-90was i about 1000 dpm/100 cm in 1974, when RG 1.86 was published.

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The surface contamination 2

limits in Table 1 of RG 1.86 were selected using the i 1000 dpm/100 cm environmental background level for Sr-90 as the baseline. Using j

the ratio of Sr-90 MPC's to the NPC's for the various nuclides as a measure of i ' relative risk, the RG 1.86 surface contamination limits were generally set at one

ofthree]evels,i.e., 100, 2000, or 5000 dpm/100 cm'. The upper limit of 5000 l dps/200ca' was set, in part, to limit direct radiation and, .in part, as a level that seemed readily attainable. Note that the direct radiation exposure from j both Fe-55 and H-3 is essentially zero. The ability to measure the contamination

using standard industry instrumentation and methods was also considered in t setting the RG 1.86 limits. -

4 The developers of RG 1.86 were aware that disparities existed between the risks j

• from the various nuclides at their respective RG 1.86 limits, but issued the 4

guide as a matter of practicality, realizing that a method for more closely

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estimating the risk from surface contamination would not be accepted as a i consensus in the near term. Before the FSV and Shoreham cases, i

i there has not been a compelling reason to evaluate more closely the magnitude of the disparities. Note that the inconsistencies in RG 1.86 are being addressed generically in the rulemaking on " Radiological Criteria for Decomissioning."

j To determine if the RG 1.86 limits for H-3 and Fe-55, i.e., 5000 dpm/100 car ,

3 pose relative risks that are significantly lower than other nuclides, the staff

i. evaluated the risk from each of the nu:lides specifically listed in RG 1.86, as well as Co-60, Cs-137. Eu-152, C-14, Ni-59, Ni-63, H-3, and Fe-55.

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The risk from each nuclide was estimated by multiplying the RG 1.86 limit by the nuclide specific dose' factors for surface contamination developed for' the building occupancy scenario in NUREG/CR 5512. "Resid"al Radioactive Contamination j from Decommissioning," October 1992. For example, the NUREG/CR-5512 dose factor for Jr-90 surface contamination is 1.51E-05 mSv/y (1.51E-03 ares /y) per dpg/100 car and the RG 1.86 surface contamination limit for Sr-90 is 1000 dpm/100 car. Multiplying these two values results in a dose of 1.51E-02 mSv(1.51 arem)/y. Note that the NUREG/CR-5512 dose factors are currently the staff's best estimate of dose from surface contamination. These dose factors were developed to support the ongoing rulemaking on " Radiological Criteria for Decommissioning."

The estimated doses for the 24 nuclides evaluated range from about 0.8 mSv/y (80 arem)/y for uranium to about 2E-05 mSv (2E-03 mrem)/y for I-133. The result.ing doses for Co-60, Fe-55, and H-3 are 0.14, 2.5E-04, 2.8E-05 mSv -(14,

,2.5E-02, and 2.8E-03 mrem)/y, respectively. The average dose for the 24 auclides -

evaluated was about 0.1 mSv (10 mrem)/y.

Comparing the relative risks from Co-60 to b'oth Fe-55 and H-3, it is seen that the risk from Co-60, at the RG 1.86 limit, is 5000 times greater than the risk from H-3 and 560 times greater than the risk from Fe-55. Because of the magnitude of these differences, and the estimated cost of compliance with the existing limits at FSV and Shoreham, the staff believes that it is appropriate to consider the 's.srface contamination liuits for H-3 and Fe-55 separately.

To determine the total surface contamination limits, the staff considered the average dose from the 24 nuclides at their respective RG 1.86 limits, and ALARA.

In addition, to maintain simplicity in implementation, the limits for both H-3

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and Fe-55 were both set-at equivalent levels, using the dose from Fe-55 as the basis. This results in a more conservative dose for H-3. The estimated average dose from nuclides at the RG 186 Timits, as evaluated above, is 0.1 mSv (10 arem)/y, which translates to a limit of

. 2,000,000 dpm/200 caf. In consideration of ALARA, the limit finally selected for total surface contpination from Fe-55 and fi-3 at FSV and Shoreham was 200,000 dpm H-3 doses ar/200 cm . At 200,000 dpm/200 caf, the estimated Fe-55 and e 0.01 and 1.1 E-03 mSv (1.0 and 0.11 arem)/y, respectively. These potential doses will decline as Fe-55 and H-3 decay with half-lives of 2.6 years and 12.2 years, respectively.

The 200,000 dpe/I'00 cs' limit discussed above applies to average contamination i

levels.

limit. RG 1.86 also contains maximum limits that are 3 times the average Accordingly, the maximum limit for H-3 and Fe-55 total surface contamination was raised to 600,000 dpm/100 cm'.

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The existing limits for removable H-3 and Fe-55 surface contamination were also evaluated. Iricreasing the removable contamination limits is not considered ALARA l

since standard remediation techniques are capable of lowering removable contamination to levels below the current limit. Therefore, the removable limits L for' H-3 and Fe-55 will remain at the current level of 1000 dpm/200 cm2 .

The above dose calculations assume thalthe exposure from the activated concrete is through the inhalation and ingestion of material resuspended or removed from contaminated surfaces, which are considered the most probable dose pathways.

However . although considered unlikely, the staff also evaluated the potential dose as,suming that the activated concrete is removed from uthe b' ildings and disposed without restrictions... The potantial dose was estimated using the RESRAD environmental pathway and dose assessment code (ANL/EAD/LD-2). The resulting dose is considered conservative since,1) the activity in the concrete and steel -

was assumed to be imediately available for uptake by plants and animals, and 2) all of the H-3 is assumed to leach from, the concrete in the first year and migrate to groundwater. The resulting maximum potential doses fiom groundwated, plus the other exposure pathways in the residential farmer scenar'io, at Shoreham and FSV are 0.012 mSv/y (1.2 mrem /y) and 0.019 mSv/y (1.9 mrem /y , respectively.

The modeled dose declines rapidly with time following the max)imum due to the conservative assumption that all of the H-3 is leached in the first year. For example, the fourth and sixth year doses at both facilities are less than SE-03 and 2E-04 mSv/y (0.5 and 0.02 mrem /y), respectively. -

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