ML20073H250
| ML20073H250 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 09/29/1994 |
| 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 | |
| RTR-REGGD-01.086, RTR-REGGD-1.086 P-94084, NUDOCS 9410050171 | |
| Download: ML20073H250 (4) | |
Text
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P.O. Box 840 16805 WCR 191/2; Platteville, Colorado 8%51 conver,co sa2ci.m40 September 29,1994 Fort St. Vrain P-94084 U. S. Nuclear Regulatory Commission A'lTN: Document Control Desk Washington, D. C. 20555 ATTN:
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
REFERENCES:
1.
NRC Ixtter, Pittiglio to Crawford, dated June 15,1994 (G-94113) 2.
PSC letter, Warembourg to Austin, dated September 21, 1994 (P-94080)
Dear Mr. Austin:
This letter submits a proposed modification to the facility release criteria in Regulatory Guide 1.86 for use during decommissioning of the Fort St. Vrain Nuclear Station (FSV).
During preparations for the final site survey, Public Service Company of Colorado (PSC) has evaluated samples from various facility surfaces and has determined that the Regulatory Guide 1.86 acceptance criteria for removable surface contamination, when adjusted to account for the hard to detect nuclides tritium and iron-55, would result in extensive sampling, analysis, and decontamination efforts that are not justified by the low risk associated with tritium and iron-55. PSC proposes that the release criteria for tritium and iron-55 removable surface contamination be increased in a manner similar to that previously approved for total surface contamination (Reference 1).
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P-94084 September 29,1994 Page 2 Using the current limits of Regulatory Guide 1.86 for removable surface contamination and taking into account the presence of hard to detect nuclides, PSC would have to demonstrate that the sum of the activities from tritium, iron-55, and the typical readily i
2 detectable beta-gamma emitting nuclides is less than 1000 dpm/100 cm. PSC does not have the capability to directly measure iron-55 at the activity levels present and it is not i
feasible to analyze all samples for these three distinct types of nuclides for all affected i
plant areas. Therefore, site specific guideline values (i.e., activity limits) for affected areas must be determined by reducing the acceptance criteria for those nuclides which can be detected by instrumentation that will be used during the final survey, based on known tritium and iron-55 activities in selected samples.
To illustrate the impact of tritium and iron-55 on removable activity limits at Fort St.
Vrain, PSC has determined site specific removable activity limits for eight representative FSV samples, using the guidance of Appendix A to Draft NUREG/CR-5849. Samples were taken from the reactor building and radioactive waste processing areas and are considered representative of contamination that was deposited during power operation and decommissioning, and are representative of conditions expected at the time of final survey. The samples were analyzed in accordance with 10 CFR 61, and were decay cortected through December 1995. The resulting removable activity limits for typical be'.a-gamma emitters, after adjusting for the presence of tritium and iron-55, are as i
follows:
Removable # yActivityLimit 2
Samole Source (dpm/100cm )
i 1.
PCRV Smear 138 2.
Hot Service Facility Smear 561 l
3.
Fuel Handling Machine Smear 142 4.
Liquid Waste Resin Sample 820 5.
PCRV Activated Concrete 78 6.
Graphite Dust 359 7.
PCRV Access Flange 112 8.
PCRV Shield Plug 228 i
l i
The average of the above limits is 305 dpm/100 cm, which would be used as the site-I 2
specific guideline value for removable activity for beta-gamma emitting nuclides in affected surfaces and structures within the FSV reactor building and radioactive waste processing areas. PSC considers that the reduction of the removable beta-gamma activity 2
2 limit from 1000 dpm/100 cm to 305 dpm/100 cm, along with action levels set at an 2
appropriate fraction of the limit (as low as 50% or 153 dpm/100 cm ), would result in 1
l y.
P-94084 September 29,1994 Page 3 significantly more extensive 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. In light of the reduced risk from tritium and iron-55 cited in SECY-94-145, PSC does not consider these efforts warranted.
PSC proposes that the acceptable levels of removable surface activity for tritium and 2
iron-55 in affected areas be increased to 20,000 dpm/100 cm, and that allowable activities be determined using a unity equation and the guidance of Appendix A to Draft NUREG/CR-5849.
Using the above samples, this would result in a site specific guideline value for removable activity for beta-gamma emitting nuclides of 750 dpm/100 cm. PSC considers that this modification to the removable activity limits is consistent 2
with ALARA principles, as follows:
2 From a dose standpoint, the total dose that could result from 750 dpm/100 cm beta-gamma emitters plus the maximum activity of tritium and iron-55 that 2
satisfies the unity equation, is less than the dose from 1000 dpm/100 cm of cobalt-60, which would be allowed by Regulatory Guide 1.86.
2 PSC estimates that raising the site specific guideline value from 305 dpm/100 cm to 750 dpm/100 cm would decrease the areas requiring further investigation by 2
greater than 50% and would almost eliminate the areas requiring remediation.
Although it is almost impossible to quantify the economic savings that would result from this modification of the limit, it clearly reduces resource requirements for decontamination with no reduction in public health and safety.
This factor of 20 increase from the Regulatory Guide 1.86 limit is conservative with respect to that approved for total surface contamination, in Reference 1.
For unaffected structures and plant systems outside of the reactor building and radioactive waste processing areas, tritium and iron-55 are not considered to be nuclides of concern.
No measurable beta-gamma emitters have been found in these areas during operational and decommissioning surveys, and it is reasonable to assume-that there are no measurable levels of tritium or iron-55 nuclides present either. Therefore, for these surfaces, the acceptance criteria for average total activity and removable contamination will not be adjusted to account for hard to detect nuclides. In the event that beta-gamma contamination is verified in excess of 25% of the guideline, the surface will be reclassified as affected (" suspect" for building surfaces and structures) and resurveyed as required.
j
P-94084
- September 29,1994 Page 4 PSC requests that the limits for removable surface activity for tritium and iron-55 be modified as discussed above, and that these revised limits be approved as part of the Final Survey Plan for Site Release. Although approval of this request is not required to support release of the FSV repowering area, as described in Reference 2, PSC requests approval of this request by December 1,1994, to support procedure development, training, and planning efforts for the final survey. In addition, timely approval of the modified removable surface activity limits is requested to support our ongoing efforts to decontaminate plant piping systems and equipment.
If you have any questions regarding this submittal, please contact Mr. M. H. Holmes at (303) 620-1701.
Sincerely, T)ts v>. Lvawbotin by 654GF Don W. Warembourg Decommissioning Program Director DWW/SWC cc:
Regional Administrator, Region IV Mr. Robert M. Quillin, Director Radiation Control Division Colorado Department of Health
.