ML20078K723
| ML20078K723 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/17/1994 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Joseph Austin NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| P-94105, NUDOCS 9411250104 | |
| Download: ML20078K723 (12) | |
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P.O. Box 840 16805 WCR 191/2; Platteville, Colorado 80651 cenver.co 80201.o840 November 17, 1994 Fort St. Vrain P-94105 U. S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D. C. 20555 A'ITN:
Mr. John H. Austin, Chief Decommissioning and Regulatory Issues Branch Docket No. 50-267
SUBJECT:
Proposed Modification of Removable Surface Contamination Release Criteria for Tritium and Iron-55, Supplemental Information
REFERENCES:
1.
PSC Letter, Warembourg to Austin, dated September 29, 1994 (P-94084) 2.
NRC I2tter, Pittiglio to Crawford, dated June 15,1994 (G-94113)
Dear Mr. Austin:
This letter provides additional information in support of Public Service Company of Colorado's (PSC) request to modify the removable surface contamination release criteria for tritium and iron-55, previously submitted via Reference 1.
This additional information was requested during a telephone conversation on October 27,1994, between PSC and Messrs. Dave Fauver and Clayton Pittiglio of your staff.
In the Reference 1 submittal, PSC proposed that the removable surface contamination criteria for hard to detect nuclides tritium and iron-55 be increased from 1000 dpm/100 2
2 cm to 20,000 dpm/100 cm. This request is similar in principle to the modification to fixed surface contamination criteria for tritium and iron-55 approved by the NRC in Reference 2. As discussed in Reference 1, PSC requests an increase to the removable surface contamination criteria for these two nuclides so that the level of cleanup and survey activities during Fort St. Vrain decommissioning will be reasonable and consistent with the levels of detectable S - y emitters that contribute the majority of dose consequence rather than tritium and iron-55 which have little dose consequence.
PDR ADOCN 0500o267
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9411250104 941117 W
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November 17, 1994 Page 2 The attachment to this letter describes the extent of cleanup or housekeeping activities that PSC plans to perform, in addition to the equipment or facility decontamination efforts associated with facility decommissioning, to reasonably clean up removable contamination. The housekeeping protocol described in the attachment ensures that members of the public who may visit the site will not leave with significant amounts of dust, dirt, or contamination on their clothing. Also, a cost / benefit analysis is provided 2
which shows that the extra efforts required to meet a 1000 dpm/100 cm removable 2
surface contamination limit, as compared to PSC's proposed 20,000 dpm/100 cm limit for tritium and iron-55, would result in a conservatively calculated dose reduction of only 0.82 person-mrem / year, and would cost approximately $5.6 million. PSC considers that these efforts are not justified and that this minimal dose reduction is not " reasonably achievable."
PSC requests that the modified removable surface activity limit requested in this letter and in Reference 1 be approved by December 31,1994. This accelerated approval is required to support current decontamination efforts of plant systems and equipment and to define the Site Specific Guideline Values. In addition, this approval date is requested to support the ongoing efforts of procedure development, training, and planning efforts for the Final Survey.
If you have any questions regarding this information, please contact Mr. M. H. Holmes at (303) 620-1701.
1 Sincerely,
$ /YlWwin Don W. Warembourg Decommissioning Program Director DWW/SWC i
Attachment cc:
w/ attachment j
Regional Administrator, Region IV j
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Mr. Robert M. Quillin, Director
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Radiation Control Division Colorado Department of Health
Attachment to P-94105 Supplemental Information to PSC's Proposed Modification to Removable Surface Contamination Release Criteria for Tritium and Iron-55 This attachment provides additional information in support of PSC's proposed modification to removable surface contamination criteria for tritium and iron-55, as discussed with the NRC on October 27,1994. Included are a clarification of the request, f
a discussion on PSC's planned cleanup activities or housekeeping protocol, an analysis of the dose consequences from the remaining contamination if our request is approved, an estimate of the costs for a comprehensive plant decontamination and increased final survey efforts that would result if our request is not approved, and a discussion of possible public perceptions of this proposed increase in contamination limits.
i Clarification of the Limit Reauest In Reference 1, PSC requested that the removable surface contamination release criteria for tritium and iron-55 be adjusted in a manner similar to that previously approved for total surface contamination (Reference 2). The specific limits requested were 20,000 2
dpm/100 cm for tritium and for iron-55. The purpose for requesting this adjustment for tritium and iron-55 is to allow PSC to derive a reasonably measurable site specific guideline value for removable activity based on detectable # - y emitters which contribute
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the majority of dose consequence rather than tritium and iron-55 which have little dose consequence.
t PSC understands and has actively pursued the ALARA philosophy throughout the decommissioning process. The decommissioning team has spent considerable effort i
examining the issues impacting various cleanup and housekeeping options as they relate 2
to ALARA and to the proposed modified limit of 20,000 dpm/100 cm, such as the following:
Dose consequences of removable activity; e
The relative activities of nuclides expected to be present at the time of license termination (i.e., the source term), radionuclide halflives and the variability of the source term in various plant areas / systems; The current radiological conditions of the facility; t l l
The derived guideline value and its associated action level (50%) for investigation surveys; The potential for beta attenuation caused by the presence of dirt and debris; Safety concerns associated with an extensive housekeeping effort (i.e.
scaffolding, asbestos cable trays, electrical panels and wiring, high work, confined spaces, additional dose consequences, etc.) in the absence of any direct correlation between surface dust and removable contamination except in affected-suspect areas; and Cost and schedule impacts due to extensive plant decontamination required to demonstrate compliance with very low removable surface contamination 2
guideline values (e.g.,305 dpm/100 cm as identified in Reference 1).
Eacilily_Housekeepine Although PSC's housekeeping practices are not actually considered part of decontamination efforts, they are and will be generally effective in cleaning up removable contamination and in preventing workers or other individuals in the facility from carrying contamination away from the site. General housekeeping activities have been a part of facility dismantlement and iecommissioning activities to this point and we intend to continue these efforts throughout the physical decommissioning process.
Routine housekeeping is normally limited to areas and surfaces routinely accessed.
Facility housekeeping is a recognized responsibility of the licensee and is important for the general health and safety of workers as well as proper facility maintenance. The need to remove dust and debris which might otherwise impact the ability to survey surfaces during the termination surveys is also recognized. In all cases, surfaces in excess of the site specific guideline criteria will be decontaminated or remediated. PSC intends to use the following housekeeping protocol:
Affected Areas (> 50% of Guideline Value)
Horizontal surfaces will be swept, vacuumed or wiped down using a wet wipe and mild detergent, as appropriate. Exceptions to this requirement may be taken where concern for the safety of personnel is determined to be significant as a result of cleanup activities.
l 2-
Affected Areas (s 50% of Guideline Value)
Readily accessible horizontal surfaces will be swept, vacuumed or wiped down using a wet wipe and mild detergent, as appropriate.
Exceptions to this requirement may be taken where concern for the safety of personnel is determined
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to be significant as a result of cleanup activities. Readily accessible will be considered 1 meter horizontally and 2 meters vertically from catwalks, walkways, stairways, landings, or other locations designed for routine personnel access.
Unaffected agu Dust or debris which prevents the ability to assess the underlying surface will be o
famoved from the survey point.
PSC believes the housekeeping protocol outlined above is appropriate based on:
The execution of a well defined housekeeping program by the licensee which meets the ALARA requirement of " reasonably achievable".
The need to minimize health and safety risks to decommissioning personnel.
Pose Conseouences of Proposed Limit Adiustment 2
Revising the removable surface contamination limit from 1000 dpm/100 cm to 20,000 2
dpm/100 cm for tritium and iron-55 will have somewhat higher dose consequences for individuals in the Reactor Building after license termination. This limit change will 2
generally not result in acces:ible surfaces being released with 20,000 dpm/100 cm of removable tritium or iron-55, as shown in the calculations in Attachment A, due to the mix of isotopes found in the facility. It should be pointed out that if tritium is identified in the absence of other radionuclides, it would most likely be internal to system piping or components and would not be accessible. In such a case, however, the full guideline 2
value of 20,000 dpm/100 cm will be applied and this contamination level could remain.
The dose consequences of removable activities at the values under consideration.are summarized in the following table. The calculated Total Effective Dose Equivalent (TEDE) values report the annual TEDE which would be attributable to the removable activity component of the total residual using the methodology and dose factors contained in NUREG-5512 (Volume 1, October 1992). Although PSC has no current plans for use of the reactor building after license termination, the " Occupancy Scenario" and associated default values were used for this estimate of the potential dose consequences. This is a conservative scenario assumption and assun;es a total of 2000 person-hours per year spent in t'he reactor building. Our decommissioning workers currently have far greater access to the facility than is envisioned after license termination, and our experience has been that reactor building access and routine maintenance activities have been made generally without picking up contamination; in fact, most reactor building activities can be performed in street clothes, unlike conditions at most water reactors. The removable surface contamination value used in the dose calculations was conservatively assumed to be uniformly distributed over all surfaces. This is conservative in that it does not take credit for cleaned surfaces, as described in the housekeeping protocol discussion above.
DOSE CONSEQUENCES OF REMOVABLE ACTIVITIES Scenario Derived Limit Value Calculated TEDE
- 2
- 1. Removable Guideline 1000 dpm/100cm 1.84E+0 person-mrem /yr.
(No H-3 or Fe-55) detectable # - y 2
- 2. Removable Guideline 305 dpm/100cm 5.69E-1 person-mrem /yr.
2 (H-3, Fe-55 @ 1000dpm/100cm )
detectable # - y 2
(1000 dpm/100cm all isotopes) 2
- 3. Removable Guideline 750 dpm/100cm 1.39E+0 person-mrem /yr.
2 (H-3,Fe-55 @ 20,000dpm/100cm )
detectable # - y 2
(2443 dpm/100cm all isotopes) 2
- 4. H-3 @ 20,000dpm/100cm 20,000 dpm/100cm 1.13E-2 person-mrem /yr.
Tritium Only
- 5. Fe-55 @ 20,000dpm/100cm 20,000 dpm/100cm' 1.02E-1 person-mrem /yr.
2 Fe-55 Only
- Based on the relative nuclide activities present in the Fort St. Vrain waste streams, and based on 2000 person-bours access per year.
Five calculated dose scenarios are presented above. The first scenario assumes that 1000 2
dpm/100 cm detectable # - y removable contamination remains with no dose contribution due to tritium or iron-55. The second scenario assumes that the removable contamination limit for all isotopes, including tritium and iron-55, remains at 170 2
dpm/100 cm. This scenario results in a FSV site specific guideline value of.,s5 2
dpm/100 cm detectable S - 7 The third scenario assumes that the removable 2
contamination limit fcr Pitium and iron-55 is raised to 20,000 dpm/100 cm, as requested 2
in this letter, which results in a FSV site specific guideline value of 750 dpm/100 cm detectable # - 7. The dose consequence associated with the difference between scenarios two and three, i.e., the additional dose resulting from approval of PSC's request, is approximately 0.82 person-mrem /yr. Scenarios four and five are presented to portray i
a the' dose associated with the potential presence of only tritium or iron-55 at the requested 2
removable contamination limit of 20,000 dpm/100 cm,
As can be seen from this table, the 0.82 person-mrem dose consequence of the limit 2
adjustment is below the 1.84 person-mrem dose based on 1000 dpm/100 cm detectable
- - y (Scenario One) which would be allowable per Regulatory Guide 1.86, and is a small fraction of the 10 mrem /yr basis for total residual contamination (Reference 2).
Note that NUREG-5512 does not differentiate between removable and fixed surface activity and that the FSV site-specific guideline value for total residual surface contamination will not change with approval of this limit adjustment for removable activity. Hence, the calculated TEDE value for total residual (fixed plus removable) activity will not change with approval of the limit adjustment for tritium and iron-55.
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_ Cost Analysis for C9_moliance with the 305 dom /100 cm Site Specific Guideline Vl
- it E 1
In Reference 1, PSC stated that the reduction of the removable contamination beta-2 2
I gamma activity guideline value froi.1000 dpm/100 cm to 305 dpm/100 cm, along with action levels set at an appropriate fraction of the guideline value (as low as 50% or 153 2
dpm/100 cm ), would result in significantly more investigation surveys, more extensive decontamination activities, and would also result in greater need for laboratory counting equipment for routine surveys to monitor the effectiveness of decontamination. We estimate that the costi associated with the additional efforts could be as much as
$5,600,000 and that the decommissioning could be extended three months. Attachment B contains a breakdown of this cost estimate.
Based on biased surveys performed prior to and during decommissioning, most of the Reactor Building greater than 2 meters above floors or deck grating surfaces was initially classified as affected non-suspect.
This classification was based on a removable 2
contamination limit of 1000 dpm/100 cm detectable beta-gamma and action levels set at 50% of the limit. Without an adjustment in the tritium and iron-55 limits, these areas would require reclassification as affected suspect if samples indicate readily detectable 2
beta-gamma removable activity greater than 153 dpm/100 cm. Given this lower limit and current plant survey data, the cost assessment assumed that the majorky of the Reactor Building must be classified as affected suspect and that a Wrough decontamination would be required to achieve unconditional release. The cost estimate in Attachment B includes costs for:
Decontamination of the entire Reactor Building. The costs shown are those above and beyond the housekeeping activities described previously in this letter.
Additional survey requirements required by the change in classification from affected non-suspect to affected suspect.
Additional laboratory equipment (e.g. liquid scintillation, gamma spectrometry) and laboratory technician support to count the additional smears.
Additional radwaste generated by the decontamination efforts.
Miscellaneous support for decontamination and additional surveys. This includes scaffold erections, temporary lighting and ventilation, extended operation of mobile laundry facilities and safety engineering support.
Overhead staff and facility costs to support the additional three month effort.
j Public Perception PSC understands that increasing the limit for removable surface contamination from 1000
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2 to 20,000 dpm/100 cm for tritium and iron-55 could be perceived as an undesirable increase in the amount of contamination that could be carried away from the Fort St.
Vrain site, as for example, by brushing against facility surfaces. We consider that the housekeeping protocol described above will reasonably remove contamination from normally accessible areas so that members of the public who may visit the site do not Icave with significant amounts of dust, dirt, or contamination on their clothing.
i In addition, the dose consequences associated with the proposed guideline values are consistent with the concept that remaining contamination levels should be
" indistinguishable from natural background", as was championed by members of the public during the enhanced rulemaking process for decommissioning. The increased dose represents less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> worth (less than 0.2%) of the approximately 450 mrem / year (including Radon) natural background level in Colorado. Discussions with the State of Colorado have indicated that analyses of proposed rtions based on pathways analyses, such as has been done in this case, generally are acceptable to the public.
Summary The above discussion shows that PSC's planned housekeeping protocol for facility cleanup will ensure that the amount of removable surface contamination accessible to members of the public during normal activities will be insignificant.
A pathways analysis using the methodology provided in NUREG-5512 conservatively determined that the additional public dose resulting from approval of the 20,000 dpm/100 2
cm removable surface contamination limit (or conversely, the dose savings associated
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witit the 1000 dpm/100 cm 1 mit) is only 0.82 person-mrem / year. This analysis is 2
conservative in that: it is based on a conservative building occupancy scenario whereas PSC has no plans for use of the reactor building after license termination; it does not take credit for cleaned surfaces, but rather assumes that remaining contamination is uniformly distributed over all surfaces; and it does not take into account the additional dose that would be received by workers performing additional cleanup activities.
This requested change in the limit for removable surface contamination will have no impact on the calculated TEDE value for total residual activity, determined in accordance with NUREG-5512.
2 The cost to implement the 1000 dpm/100 cm limit is estimated at $5.6 million, and the associated cleanup and survey tasks are estimated to require an additional three months.
PSC does not consider that an expenditure of $5.6 million to save a conservatively estimated 0.82 person-mrem is reasonable nor justified. In addition, this level of expenditure would not be warranted using the industry standard cost / benefit guidance in 10 CFR 50, Appendix I, of $1000 per total body person-Rem.
2 PSC believes that the proposed 20,000 dpm/100cm Euideline value for H-3 and Fe-55, combined with the cleaning and housekeeping protocol outlined above represents a programmatic and reasonable effort to reduce contamination levels to "As Low As Reasonably Achievable" during Fort St. Vrain decommissioning.
I Attachment A FSV Site Specific Guideline Value for Removable Activity 2
Based on a 20.000 dpm/100cm Limit for Fe-55 and H-3 The site-specific guideline value for detectable S y removable activity was calculated 2
to be 750 dpm/100cm using a derivation of equation A-2 in Draft NUREG/CR-5849, Appendix A.
I Site-Specific Guideline Value =
fi + 52 f.
+...
Where:
Site-Specific Guideline Value The limit for surface contamination for
=
nuclides normally detected during field measurement, adjusted for H-3, Fe-55 and other radionuclides not readily detectable.
F=
Detectable nuclide fraction.
f=
Fraction of the total activity contributed by each nuclide, i
G=
Guideline value for each nuclide. (From Regulatory Guide 1.86 or as i
specified by the NRC),
As discussed in Reference 1, eight samples were taken in the years 1993 and 1994 for the purposes of characterizing FSV waste streams. The samples were analyzed in accordance with 10 CFR 61 and were decay corrected through December 1995. A site-1 2
specific guideline of 750 dpm/100 cm was calculated using the above formula and average isotopic values from the eight samples. The isotopic fractions for H-3, Fe-55 and detectable #
y emitters are listed below:
.518 = Fraction of total removable activity attributable to Fe-55
.146 = Fraction of total removable activity attributable to H-3
.307 = Fraction of total removable activity attributable to detectable m
beta-gamma emitters
.g.
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Attachment A (Continued)
FSV Site Specific Guideline Value for Removable Activity 2
R. ped on a 20.000 dpm/100cm Limit for Fe-55 and H-3 2
Although approval of the 20,000 dpm/100 cm limit for H-3 and Fe-55 could hypothetically result in release of accessible surfaces with these contamination levels, this will not typically be the case. Based on the isotopic fractions presented above, approval 2
of the 20,000 dpm/100 cm for H-3 and Fe-55 would more realistically result in the release of surfaces with the removable H-3 and Fe-55 levels as calculated below:
(750) x (.518) = 1265 dpm/100cm2 Fe-55
(.307)
(750) x f*I40) = 357dpm/100cm2 H-3
(.307) i Attachment B Breakdown of the $5.6 Million Cost to Comply 2
with 305 dpm/100 cm Site Specinc Guideline Value Additional analytical equipment and analyses
$250K Extensive decontamination of Reactor Building
$1,200K Area reclassi6 cation and additional surveys
$1,896K Additional radwaste
$140K Miscellaneous support
$465K Staff support
$750K PSC staff, facility costs during additional 3 months
$900K Total
$5,601K 1
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