RA-02-008, Transmittal of Special Report, Transuranic and Other Hard to Detect Radionuclides in Maine Yankee Sample Media.

Transmittal of Special Report, Transuranic and Other Hard to Detect Radionuclides in Maine Yankee Sample Media.

ML020440651
Person / Time
Site:	Maine Yankee
Issue date:	01/16/2002
From:	Williamson T Maine Yankee Atomic Power Co
To:	Document Control Desk, NRC/FSME
References
-RFPFR, MN-02-002, RA-02-008
Download: ML020440651 (46)

Text

MaineYankee 321 OLD FERRY RD. - WISCASSET, ME 04578-4922 January 16, 2002 MN-02-002 RA-02-008 UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk Washington, DC 20555

Reference:

(a) License No. DPR-36 (Docket No. 50-309)

(b) Maine Yankee Letter to USNRC dated August 13, 2001, "Revision 2, Maine Yankee's License Termination Plan," MN-01-032, Proposed Change No. 210, Supplement No. 2 (c) Maine Yankee Notice and Motion Before USNRC Atomic Safety and Licensing Board dated August 31, 2001, "Notice of Settlement and Joint Motion to Terminate the Proceeding," ASLBP No. 00-780-03-OLA

Subject:

Transmittal of Special Report: "Transuranic and Other Hard to Detect Radionuclides in Maine Yankee Sample Media" In support of the NRC review of Maine Yankee's License Termination Plan (LTP) submitted via Reference (a), Maine Yankee is providing the attached report, "Transuranic and Other Hard to Detect Radionuclides in Maine Yankee Sample Media," dated December 12, 2001. This information was requested by the NRC Staff during a routine LTP review status conference call held October 31, 2001.

The subject report was developed and finalized during the "Technical Issue Resolution Process"' (TIRP). The TIRP was completed in December 2001 in fulfillment of the agreement reached between Maine Yankee and the State of Maine, as described in Reference (c). The overall goal of the TIRP was to reach resolution for LTP issues related to: (1) nuclide characterization data variability and (2) transuranic and other "hard to detect" nuclides present in site nuclide mixtures. The subject report provided the basis for the resolution of TIRP Problem Statements 3 and 5, as defined in Item A.2 of Reference (c).

Also, as requested by the NRC Staff, Maine Yankee provides as Attachment 2 a copy of the G.

E. Chabot to P. J. Dostie letter, dated November 12, 1998, which is referenced in the Maine Yankee LTP (Table 2-3).

If you have any questions regarding this report, please contact me.

Sincerely,

ýa44,t4ýý-

Thomas L. Williamson, Director

ý co Nuclear Safety and Regulatory Affairs

U.S. NUCLEAR REGULATORY COMMISSION Document Control Desk Page 2 Attachments:

1. "Transuranic and Other Hard to Detect Radionuclides in Maine Yankee Sample Media," dated December 12, 2001.

2. Dr. George E. Chabot to Patrick J. Dostie, letter dated November 12, 1998 cc: Mr. H. J. Miller, NRC Regional Administrator, Region I Mr. M. K. Webb, NRR Project Manager Mr. R. Bellamy, NRC Region I Mr. J. T. Greeves, NRC Director, Division of Waste Management Mr. R. Ragland, NRC Region I Mr. C. L. Pittiglio, NRC NMSS Project Manager, Decommissioning Mr. R. A. Gramm, NRC Section Chief, Project Directorate IV Mr. P. J. Dostie, State of Maine, Division of Health Engineering Ms. P. Craighead, State of Maine, Nuclear Safety Advisor Mr. R. Shadis, Friends of the Coast - Opposing Nuclear Pollution

ATTACHMENT I Transuranic and Other Hard to Detect Radionuclides in Maine Yankee Sample Media' December 12, 2001

'This report was generated during the Technical Issue Resolution Process (TIRP) and provides the basis for resolution of TIRP Problem Statements 3 and 5. The TIRP problem statements are defined in Item A.2 of the agreement reached between Maine Yankee and the State of Maine, as documented in the Notice and Motion Before the USNRC ALSB, dated August 31, 2001, ASLBP No. 00-780-03-OLA. The report was approved as "Appendix B" to the TIRP Participant Consensus Aggreement, dated December 13, 2001.

APPENDIX B TIRP Participant Consensus Agreement Transuranic and Other Hard To Detect Radionuclides In Maine Yankee Sample Media Source Term Radionuclides important to decommissioning all originate from one source-the reactor. The radionuclides are created by either fissile (fuel) material interactions or are produced by fission or activation reactions. No matter what the type of plant-derived nuclides important-to decommissioning, they all originate in the reactor either from fission products or by reactor generated activation products. There is only one nuclide spectrum of origin and that spectrum is determined by the reactor or the reactor coolant system (RCS) corrosion products that are activated in the reactor. Site specific factors such as the Nuclear Steam Supply System design, fuel enrichment, fuel burnup, etc., all contribute to the unique nuclide spectrum fo a given reactor. In order to be found on building surfaces or in soil or sediment, the radionuclides must cross one or more of the fission product barriers after production. The barriers are: 1), the fuel pellet matrix, 2) the fuel clad, and 3) the reactor coolant system. The buildings in which the systems are contained provide another barrier to the release of radionuclides to soil.

Plant derived nuclides important to decommissioning include activation products. Materials outside the RCS, such as structural concrete, are activated through neutrons which escape the reactor. Those nuclides (e.g., Eu-152, Eu-154) are usually entrained within the sub-structure of the associated material and are not subject to substantial release from that material nor are they available to re-contaminate other surfaces. Metallic elements (e.g., Fe, Co, Mn) may be released into the RCS as a result of component wear and become activated as they pass through the reactor. Co-60 created by activation of cobalt in stainless steel components is an example of such an activation product. Co-60 is found in both ionic and particulate forms in the RCS. Each form of Co-60 displays its own unique contamination pathway once released from the RCS.

H-3 is also a byproduct of the fission process. It is also produced by the neutron interaction of low Z materials such as Li-6 and the subsequent decay of the Li-7. Once produced, H-3 follows the reactor coolant water contamination pathways.

These activation products are not the primary subject of this report.

Production of Hard to Detect (HTD) Nuclides The transuranic nuclides (TRUs) originate within the fuel pellets when heavy nuclei within the fuel absorb a neutron and/or decay to become one of the TRUs. The amount of a given TRU nuclide is determined by its parent's abundance in the fuel and its cross section. Pu-241 is the December 12, 2001 Page 1

APPENDIX B TIRP Participant Consensus Agreement most abundant TRU (12.4%, yield based on total Ci in fuel). The other TRU products such as Am-241 (0.08%), Cm-242 (0.07%), Pu-240 (0.06%), and Pu-239 (0.04%), are produced at much lower rates as indicated. When fission occurs, U-235 atoms are split into fission fragments. The fission fragments are of unequal atomic mass with one fragment having a mass of approximately 90 g/mole and the other approximately 135 g/mole. The total amount of a given fission fragment produced is related by its atomic mass according to the fission yield curve (Att. 1). The total amount of any fission product can be determined for a given fuel enrichment, power level, operating time (burnup) and decay time following reactor shutdown. The most abundant fission nuclides include Cs-137 (11.75%), Ba-137m (11.12%), Cs-134 (9.56%), Sr-90 (7.71%), and Kr 85 (0.87%).

Just because a radionuclide is produced in the fuel doesn't mean that it is available outside the fuel pellet. The three barriers prevent the release of reactor-produced nuclides outside the reactor operating systems. The first two barriers are of primary interest from a decommissioning standpoint. In order for the fission fragments or TRUs to be released, they must cross the fuel matrix and clad barrier. Typically, fission fragments and TRU nuclei recoil or migrate only a matter of 5-10 microns within the fuel pellet when produced (UCID-20926, "Spent Fuel Performance Data", LLNL, 8/87). This short recoil distance prevents nearly all the fission fragments and TRUs from being released from the fuel unless they originate very near the outer surface or near a pellet crack. Radionuclide movement across the UO2 matrix to the space between the pellet and the clad predominantly occurs with the nuclides which are in a volatile state. Only a few nuclides or their decay products have volatile species: bromine-krypton rubidium and iodine-xenon-cesium chains.

When the fuel clad has defects (which can range from pin-hole leaks to exposed pellets), it is easier for radionuclides residing in the gap space to enter the coolant. However, the volatile nuclides increase the most because they can traverse the fuel matrix, and in some cases the clad, with the greatest ease. These are the nuclides that are found in the coolant in greatest abundance, along with activation products like Co-60 or N-16. Large, non-volatile nuclides like the TRUs remain trapped to a large extent within the fuel pellet even when clad defects are present.

NUREG-1353, "Regulatory Analysis for the Resolution of Generic Issue 82, Beyond Design Basis Accidents In SFPs",1989 (Att. 2), estimated that with a complete destruction of fuel clad, 100 per cent of Cs, Rb, Kr, Xe, and I are released while only 0.2 per cent of the Sr-90 is released and 1E-4 per cent of TRUs are released for the relatively low temperature accidents referenced in the NUREG (Att. 2). Sandia Lab reported (Table 1) similarly small fractional values for non volatile nuclides released from fuel at temperatures as high as 530 degrees C ("Method for Determining Spent Fuel Contribution to Transport Cask Containment Requirements", NTISDE 93004596, 1992.) These references give an idea of how little of the non-volatile nuclides are released from the fuel matrix even with clad failures at elevated temperatures. The Maine Yankee FSAR presented similar values (Att. 3) for radionuclides expected to be seen in the December 12, 2001 Page 2

APPENDIX B TIRP Participant Consensus Agreement reactor coolant with one per cent clad failure (Table I).

Table 1-NUCLIDE FRACTIONS RELEASED FROM FUEL FSAR Sandia NUREG-1353 Coolant Fuel Release Fuel Release Nuclide Fxn at 530 degrees C Fxn in SFP H-3 0.00045 NR NR Fe-55 NR NR NR Co-57 NR NR NR Co-60 0.0005 NR NR Ni-63 NR NR NR Sr-90 5.50E-07 5.00E-03 2.OOE-03 Cs-134 0.016 NR NR Cs-137 0.079 0.01 0.01 Sb-125 NR NR NR Pu-238 6.001E-08 2.50E-04 I .00E-06 Pu-239/40 6.00E-08 2.1OE-05 1.00E-06 Pu-241 NR 7.50E-03 1.00E-06 Am-241 6.OOE-08 8.00E-05 1.00E-06 Cm-243/44 6.OOE-08 2.OOE-04 1.00E-06 Sr FXN 5.50E-07 5.00E-03 2.OOE-03 TRU FXN 2.40E-07 8.09E-03 1.00E-06 Sr/Cs 6.96E-06 0.5 0.2 NR = not reported Effect of Clean-Up Systems During operation, the reactor coolant system was constantly being run through filters and demineralizers to remove radionuclides. This reduced the amount of radionuclides released to the environment and kept operating personnel exposure ALARA. The filters removed larger particulates and the demineralizers removed ionic constituents including TRUs. The higher valence elements like Sr and the more insoluble elements like Co, along with the TRUs, tended to be preferentially removed during cleanup of the reactor coolant system (Table 2). The Maine Yankee FSAR projected a cleanup systems removal factor of 10,000 for most radionuclides.

Waste stream data are presented only to show the relatively low amount of Sr and TRUs compared to other nuclides even when such nuclides are concentrated in the cleanup systems.

The use of these ratios is qualitative. Radionuclide ratios in waste streams are not representative of ratios in coolant leaks, on building surfaces, or in soils.

December 12, 2001 Page-31

APPENDIX B TIRP Participant Consensus Agreement Table 2-NUCLIDE FRACTIONS - MY Part 61 Data 1990 1996 1998 1998 1998 Resin Resin Chem Decon RCS Filter SFP Filter Nuclide Fxn Fxn Resin Fxn Fxn Fxn H-3 0.0044 0.0044 <0.01 0.0102 0.0097 Fe-55 0.089 0.0903 0.1015 0.2577 0.182 Co-57 0.0014 0.0007 0.0004 NR NR Co-60 0.23 0.195 0.8182 0.6263 0.5247 Ni-63 0.065 0.319 0.0695 0.0916 0.2746 Sr-90 0.0026 0.0016 0.0001 NR NR Cs-134 0.0319 0.1099 NR NR NR Cs-137 0.1096 0.2544 <0.01 <0.01 0.0014 Sb-125 0.004 0.0143 NR 0.0071 0.0044 Pu-238 5E-07 NR NR NR 0.0001 Pu-239/40 4E-07 NR NR NR 0.0001 Pu-241 0.00015 0.0002 0.0004 NR 0.0028 Am-241 0.000001 NR NR NR 0.0001 Cm-243/44 0.000002 NR NR NR 0.0001 Sr FXN 0.0026 0.0016 0.0001 NR NR TRU FXN 0.000155 0.0002 0.0004 NR 0.0032 Sr/Cs 0.024 0.0063 <0.01 NR= no data reported for theses nuclides. Short-lived nuclides were not included so the fractions don't add to unity.

The low level of TRUs found on component surfaces was tracked during operation and was found to slowly decrease over time until, late in plant life, little or no TRUs were detected on RCS surfaces. (Beta-gamma to alpha ratios remained fairly high, in the 500:1 to 5000:1 range.)

The overall reduction in TRU activity late in plant life was a result of having few fuel clad problems, better crud removal processes, and RCS cleanup over time. Part 61 samples of the radioactive waste streams (Table 2, Att. 4) did show TRUs to be present in resin and filter media because they were concentrated by them and the more soluble Sr was found only in the resin.

The Part 61 TRU fraction to Cs-137 ratios in resin were very low (0.001 to 0.0008) and Sr-90 to Cs-137 was about 0.02. However, waste stream data cannot be directly, quantitatively compared to sample media data because of selective concentration effects.

Several plant cleanup campaigns were waged in response to industry initiatives to reduced personnel exposure and contamination events. Decontamination solutions used during plant decon operations also tended to remove the high valence ions from floors and walls just like the demineralizers as a result of chelating agents contained in the solutions.

December 12, 2001 Page 4

APPENDIX B TIRP Participant Consensus Agreement Even when released to the environment, radionuclides were limited in how far they traveled by their solubility in water (TRUs and Co are relatively insoluble) and the retention by the soil (Kd effect) which is quite high for TRUs. This tended to limit the spread of any spills or leaks to only the very soluble nuclides like Cs-137. The amount of radioactivity released to the soil from the RWST leak was limited by the routine filtration and demineralization of reactor cavity water at the end of refuelings prior to its return to the RWST.

Plant Areas With Potential HTDs Noting that the reactor, either by fission or activation, was the source of all radionuclide contamination, HTDs were analyzed for using radiochemical techniques on samples taken from buildings or areas which were potentially exposed to reactor coolant water or which were directly activated by the reactor neutron fluence. These areas included the buildings within the radiological Restricted Area (RA): Containment, PAB, RCA, Fuel Bldg., and Spray Bldg. These buildings contained the reactor coolant pumps and piping, the reactor water cleanup systems (filters, demins, evaporators, etc.), rad waste processing and storage areas, and the fuel storage pool. These would have been the areas repeatedly contaminated by leaks, spills, and routine operational activities. In addition to these buildings, outside areas subjected to infrequent leaks or spills such as the RWST and radwaste storage areas were also sampled.

These potential areas were also the areas in which tens of thousands of radiological surveys were taken during plant operation. The reactor coolant and waste processing systems were the systems which were sampled and analyzed as required by the operating license and radwaste shipping regulations for the very nuclides of current concern. The RA soils were the soils sampled and analyzed in response to leaks or spills. Operational surveys provided a very complete picture of those areas on site which were likely sources of residual radioactivity.

Characterization Data The historical information, including operational survey data, was used to target areas for site characterization surveys. The likely areas were sampled along with the not so likely areas.

Easily accessible surfaces were surveyed along with hard to reach areas. Sample material included concrete cores from both floor and wall surfaces and at depths of up to 24 inches.

Sumps, drains, trenches and pits were sampled. The samples showing the highest total activity were sent for HTD analysis in order to give the best chance of detecting the very low activity nuclides like TRUs. Concrete samples from floors (generally the most contaminated areas) and sumps were especially targeted for HTD analysis. Cores from each of the buildings of concern were analyzed. Additional cores from these buildings were obtained and analyzed in order to provide even more data and the assurance that key areas of concern were adequately sampled.

December 12, 2001 Page 5

APPENDIX B TIRP Participant Consensus Agreement Soils sent for HTD analysis were taken from the areas of highest soil contamination. Several samples of soil from each spill location were composited to ensure representativeness.

The nuclide suite that was analyzed for was constructed after consulting several literature sources to ensure completeness. The lowest practicable MDCs were requested for each of the analytical techniques used. Repeat analyses were performed until acceptable sensitivity was achieved. The initial round of data analysis used MDC values as though the nuclides were present at those concentrations. MDCs were reduced until it could be shown that nuclides which were not detected and which were eliminated from the mixture were not a significant dose contributor.

Non-detects were removed only after it was determined that their dose contribution was less than 10% of the total for the media of concern. This does not mean that such nuclides may not be present at much lower concentrations, but rather that they are not significant from a dose perspective below the level to which they were analyzed and eliminated from the mixture.

The radionuclide fractions were determined using the average of the fractions method to reduce the impact of sample activity differences. Each media addressed by the LTP has its own unique radionuclide fraction (Table 3). The nuclide fractions for the various media are shown to be similar to the HTD to fission and activation product fractions for liquid effluents.

Table 3-HTD Nuclide Fractions For Various Media Media TRU Nuclide TRU Dose Sr-90 Nuclide Sr-90 Dose Fraction (LTP) Fraction (LTP) Fraction (LTP) Fraction (LTP)

Concrete 4E-6 6E-10 6E-4 IE-7 Soil 5E-5 (MDA) 2E-4 7E-3 (MDA) 5E-2 Ground 3E-4 (MDA) 1E-3 9E-4 (MDA) 4E-2 Water MY Liquid Effluent Releases Year TRU to Fission/Activation Activity Sr-90 to Fission/Activation Fractions Activity Fractions 1988 2E-4 to 3E-4 3E-3 to 2E-4 1977 2E-6 to 6E-7* 3E-4*

• May denote high fission/activation product activities.

The range of values for Sr-90 in contaminated concrete has been examined. The ratio of Sr to Cs has been determined as shown in the following table.

December 12, 2001 Page 6

APPENDIX B TIRP Participant Consensus Agreement Table 4-Sr-90/Cs-137 Ratios Source Ratio Original 7 Core 0.002 Average New 6 Core Ave 0.008 11 Core Ave (no 0.0038 pipe tunnel) 13 Core Ave 0.005 LTP Value 0.005 Original 7 Core 0.047 (MDA)

Maximum O/A Trench 0.01 PAB Pipe Tunnel 0.049 Two anomalous areas stand out with respect to the Sr to Cs ratio: the O/A Trench and the PAB Pipe Tunnel. Both areas are significantly impacted by the liquid waste stream. The O/A Trench captured all water released to the floor of the Containment building and routed it to the Containment sump. The PAB Pipe Tunnel held the pipes which carried the liquid waste water being processed by the filters and demineralizers in the PAB. Both areas had standing water and boron encrustations during plant operation. The question is whether there are other areas with similar high Sr to Cs ratios which have not been examined.

A total of 8 cores have been obtained from the Containment building floors and only the O/A Trench and PAB Pipe Tunnel stand out with respect to detectable Sr/Cs ratio. The Spray building has the same source term as the Containment building, the RCS, and the Sr to Cs ratio is similar to the 13 Core mean value. The RCA area was used for resin storage, waste water demineralization, and solid waste storage. The Sr/Cs ratios from the RCA and Drumming Room are similarly low. The SFP cooling system operated at lower temperatures than the Containment or PAB systems and any leakage drained to the Fuel building sump. The spent fuel pool itself will be checked when the pool is emptied and the liner is removed. Other cubicles in the PAB (evaporator, aux charging, sump area) have been sampled and the Sr/Cs ratios were found to be similar with the average values.

December 12, 2001 Page 7

APPENDIX B TIRP Participant Consensus Agreement One potential area which contained high temperature liquids, had high surface contamination during operation and has not been sampled, is the Letdown Heat Exchanger cubicle. This is an area of approximately 2.5 m by 2.5 m by 3 m tall located in the PAB basement. Because of its small size, it was not specifically sampled. This is the one area that stands out as perhaps needing to be examined in light of the Sr to Cs ratios found in areas with high temperature liquids and standing boron. This process will be used to determine whether additional attention is needed in the' letdown area or any other similar area.

With respect to the dose implications of the Sr to Cs anomalies, the dose per dpm and dose per dpm detectable beta were determined for Sr/Cs ratios as high as ten times the LTP value. At this level, the annual dose rate for contaminated concrete increases from 0.556 mrem/y to 0.968 mrem/y (Att. 5). Buried and embedded pipe dose rate increase from 0.0064 mrem/y and 0.0206 mrem/y to 0.0334 mrem/y and 0.0364 mrem respectively. At a surface DCGL of 18,000 dpm/100 cm2, the total increase in annual dose from a ten-fold increase in Sr-90 would be 0.45 mrem/y and the total annual dose rate would not exceed the 10 mrem/y criteria, nor would the drinking water dose rate exceed 4 mrem/y.

In the case of TRUs, there exists a gamma emitting nuclide that falls within the TRU family.

Am-241 has an easily detected gamma photon and a relatively long half-life. Am-241 can be detected at very low MDC values and is usually present in a fixed ratio to the remaining TRUs.

These characteristics make it an ideal marker for TRU activity if present. Am-241 was not detected in any samples except the O/A trench samples which showed the presence of the other TRUs as well. If, during FSS, Am-241 is detected in soil, an investigation will be performed to determine the source. Such an investigation could include performing HTD analyses on the soil samples and deriving separate DCGLs, if necessary.

Historical Differences At Other Power Plants Related To HTDs There were some operational events that occurred at other reactors that affected their nuclide fractions. Some of those events are described below to point out differences between Maine Yankee's nuclide fractions and those at other facilities.

Connecticut Yankee had significant fuel failures following the transition to zircalloy clad fuel.

(Beta-gamma to alpha ratios at CY were very low (approaching 1). Maine Yankee's beta-gamma to alpha ratios have remained in the range of 500 or 5000 to 1). The problems at CY were exacerbated by the physical destruction of fuel pellets and the release of TRUs into the RCS and into buildings as reported in 1979-1980. System leaks from the RWST left contaminated soil with high measurable dose rates around the RWST. Repeated failures of the service water line allowed liquid effluents to be released directly into soil beneath the drumming room.

December 12, 2001 Page 8

APPENDIX B TIRP Participant Consensus Agreement CY has detected measurable amounts of TRUs in concrete samples on site. Soil samples have not yet shown TRUs, but sampling has been limited. Containment foundation sump water has shown detectable Sr-90. (Maine Yankee containment foundation sump water has been analyzed with MDCs below the levels at which CY detected Sr-90 and no Sr-90 has been detected.)

Trojan had a history of fuel clad failures and a known incident in which a fuel pellet was believed to have been crushed between the reactor vessel flange and the head. The Containment Building was discovered to have water intrusion between the liner and the concrete foundation which degraded the liner and allowed contamination of the concrete. (Neither of these problems occurred at Maine Yankee.) TRUs were detected in concrete at low levels.

Conclusions Residual radioactivity was vigorously looked for in those areas targeted by operational history and site characterization plans, and the results provided no surprises. Hard to detect radionuclides, such as TRUs or Sr-90, have only been detected in small amounts and in very isolated locations (O/A trench, PAB pipe tunnel). Those media which do not contain the HTDs have been carefully analyzed to ensure that those nuclides are not present at a level that would result in a significant dose contribution. Variations in nuclide fraction between the various media are a reflection of the specific characteristics of a given media such as contamination history, decontamination history, waste stream clean up, ionic form of the nuclide, Kd, solubility, and half life. In spite of these differences, the remaining major contributors to dose are the major fission nuclide, Cs-137, and the major activation nuclide, Co-60. In addition to being the most abundant in terms of activity, they also have some of the highest dose factors. Thus the combination of activity and dose factor for these two nuclides tends to dwarf the other nuclides in comparison.

The Maine Yankee data are not unique. NUIREG/CR-4289, "Residual Radionuclide Contamination Within and Around Commercial Nuclear Power Plants", published in 1989, reported the same nuclides of significance and similar ratios between nuclides as found in Maine Yankee media. The TRUs and Sr-90 (Table 5) were detected at very low fractions of total activity by several orders of magnitude (1E-3 to 1E-4 or less for both Maine Yankee and NU*REG/CR-4289 plants, Att. 6). The small fractions of TRUs and Sr-90 were attributed to the same processes and historical events (e.g., demineralization, decontamination, etc.) as those found at Maine Yankee.

December 12, 2001 Page 9

APPENDIX B TIRP Participant Consensus Agreement Table 5-HTD NUCLIDE FRACTIONS, LTP and Literature Values LTP NUREG-4289 LTP NUREG-4289 Concrete Concrete Soil Soil Nuclide Fxn Fxn pCi/g Range pCi/g H-3 0.0236 NR 5 to 6 NR Fe-55 0.00481 0.1229 <5 NR Co-57 0.000306 NR <0.2 NR Co-60 0.0584 0.0676 0.5 to 1.7 0.006 to 377 Ni-63 0.355 0.00295 2 to 5 NR Sr-90 0.0028 0.000145 <0.8 NR Cs-134 0.00455 0.19 <0.15 NR Cs-137 0.55 0.227 60 to 99 0.05 to 260 Sb-125 NR NR <1.0 NR Pu-238 <0.00005 0.0000086 <0.009 3E-5 to 0.17 Pu-239/40 <0.00002 0.0000047 <0.010 6E-4 to 0.23 Pu-241 <0.002 NR <1 NR Am-241 <0.00006 0.00000535 <0.011 NR Cm-243/44 <0.00002 0.00000799 <0.0037 NR Sr FXN 0.003 0.0001045 <0.8 NR TRU FXN <0.002 1.47E-05 <0.034 3E-5 to 0.23 Sr/Cs 0.005 6.36E-4 <0.013 NR = no data reported Summary While the reactor produces HTDs and TRUs, the fuel matrix acts as an effective barrier against the release of significant levels of HTDs and TRUs.

Operating history plays a large role in the level of HTDs and TRUs released. Plants with severe fuel failures or pellet damage have significant HTD and TRU issues. While Maine Yankee experienced fuel failures in the early years of operation, timely and effective measures were taken to limit the impact of fuel clad failures and to reduce the level of radionuclides in the coolant. Maine Yankee was operated conservatively with respect to fuel clad or primary to secondary leakage which resulted in fewer problems.

Other operational activities such as reactor coolant cleanup, building decontamination, system chemical decontamination (Table 6) have all had an impact on reducing both the nuclides present and their respective fractions in the media presented in the LTP. The December 12, 2001 Page 10

APPENDIX B TIRP Participant Consensus Agreement system nuclide levels, dominated by the metal oxides, are similar to reported values for other reactors.

There are limited HTD data available for power reactors. The NUREGs cited above were published in the mid 80's and early 90's so they covered the very early, smaller reactors.

Large reactor decommissioning experience is limited to Ft. St. Vrain, a very dissimilar reactor, and Shoreham, a reactor which operated only briefly. CY, Trojan and Maine Yankee are the large power reactors currently well along in decommissioning and some of the significant fuel performance and operational differences between MY and the other reactors have been pointed out above. Operating plants don't perform the types of sampling, such as concrete coring, or analyzing soils for HTDs and TRUs that are necessary for decommissioning which makes finding comparable data from conservatively operated plants difficult. None the less, Maine Yankee data are reasonable, representative and appropriate for decommissioning.

The Letdown Heat Exchanger Cubicle will be included with the other anomalous areas (i.e., O/A Trench, PAB Pipe Tunnel) which require the use of a unique DCGL.

Table 6-SYSTEM INTERNAL ACTIVITY MY 1990 MY 2001 NUREG-CR-4289 DAW Pipe Internal Pipe Nuclide Fxn Fxn Fxn H-3 0.0011 <0.01 NR Fe-55 0.272 0.0889 <0.01 to 0.90 Co-57 0.0034 NR 0.24 to 0.43 Co-60 0.219 0.6635 0.01 to 0.46 Ni-63 0.101 0.2403 0.0003 to 0.19 Sr-90 0.0015 NR 1 E-6 to 7E-5 Cs-1 34 0.0028 NR NR Cs-1 37 0.0553 <0.01 3E-6 to 0.02 Sb-125 0.0129 NR NR Pu-238 0 0.0001 NR Pu-239/40 0.0002 0.0001 NR Pu-241 0 0.007 NR Am-241 0.0002 0.0001 NR Crn-243/44 0.0002 0 NR Sr FXN 0.0015 NR 1E-6 to 7E-5 TRU FXN 0.0006 0.0073 11E-5 to 0.001 NR = data not reported. Short-lived nuclides were not included so the fractions don't add to unity.

December 12, 2001 Page 11

APPENDIX B TIRP Participant Consensus Agreement Attachment I Fission Yield Curve December 12, 2001 Page 12

i ia 10-2 I I 1 I!

F I I I I I 1 I I

-  !, 233u 2 35 U

2 39 Pu*

10-4 __

70 82 94 106 U18 130 142 154 166 Moss number Figure 17-3. Fission-product yields showing the minimum at symmetry that is typical of thermal fission.

APPENDIX B TIRP Participant Consensus Agreement Attachment 2 NUIREG-1353 Release Fractions December 12, 2001 Page 13

data and on fraction of each radionuclide release was determined by BINL based on available Table 4.2, Ref.

engineering judgment, and is provided in Table 4.8.1 (from NUREG/CR-4982, 10).

Table 4.8.1 Estimated Radionuclide Release Fractions During a Spent Fuel Pool Accident Resulting in Complete Destruction of the Fuel Cladding Release Fractions(1 )

Value Uncertainty Range Chemical Family Element or Used Isotope iKr, Xe 1.0 0 Noble gases 1_0 0.5 1.0 Halogens 1-129, 1-131 1.0 0.1 1.0 Alkali metals Cs, (Ba-137m) Rb 0-02 .002 0-2 Chalcogens Te, (1-132) 2x10- 3 lxl0-4_ lx 10-2 Alkali earths Sr, (Y-90), Ba (in fuel) 1.0 0.5 1.0 Sr, Y-91 (in cladding) 0.1 0.1 1.0 Transition Co-58 (assembly hardware) 2 0.12 0.1 1.0 Elements Co-60 (assembly hardware)( ) 0.1 0.1 1.0 Y-91 (assembly hardware) 0.01 lxi0-3 lxIO-1 Nb-95, Zr-95 (in fuel) 1.0 0.5 1.0 Nb-95, Zr-95 (in cladding) lxI0-6 Ix10-8 1IxI-5 Miscellaneous Mo-99 Sx 10-6 lx 0-4 2x10-5 Ru- 106 0.5 1.0 1.0 Sb-125 lx10-8, lxl0-5 La, Ce, Pr, Nd, Sm, Eu 1xl 0-6 Lanthanides lxl0-6 lx10 IxlO-5 Transuranics Np, Pu , Am, Cm Notes: (1) Release fractions of several daughter isotopes are determined by their precursors, e.g., Y-90 by Sr-90, Tc-99m by Mo-99, Rh-106 by Ru-106, 1-132 by Te- 132, Ba-137m by Cs-137, and La-140 by Ba-140.

(2) Release fraction adjusted to account for 100% release of the small amount of Co-60 contained in the Zircaloy cladding.

4-38

APPENDIX B TIRP Participant Consensus Agreement Attachment 3 MY FSAR Coolant Activities For Some Radionuclides With 1% Fuel Clad Defect December 12, 2001 Page 14

Table 9.3-1 Reactor Coolant Fission and Corrosion Product Activities with One Percent Failed Fuel Rods Specific Activity Coolant Inventory Nuclide lic/cc curies H3 Fission 0.119 36.7 Activation in Coolant 0.322 99.0 CEA's 0.008 2.5 Total 0.449 138.2 Br 84 4.25 10-2 13.1 85m Kr 1.40 432.0 85 Kr 3.59 1104.0 87 Kr 0.76 233.0 88 Kr 2.44 751.0 Rb88 2.36 728.0 Rb89 5.88 x 10-2 18.1 Sr89 2.83 x 10-3 .871 Sr 90 1.46 x 10-4 4.49 x 10-2 y90 9.67 x 10-4 0.298 Sr 9 l 2.64 x 10-3 0.814 Y91 0.271 83.4 M 099 2.12 651.0 1 10-3 Ru 03 2.32 x 0.716 Rul06 2.06 x 10-7 6.35 x 10-5 Te129 1.58 x 10-2 4.88 1129 4.02 x 10-8 1.24 x 10-5 1131 2.29 704.0 Xe 1 3 1 m 2.09 643.0 i132 0.848 261.0 Te132 0.195 60.0 1133 3.82 1177.0 XeI33 206.0 6.34 x 104 Cs1 3 4 4.34 1336.0 134 10-2 7.32 Te 2.38 x 1134 0.560 172.0 1135 2.10 648.0 35 Xe 1 7.24 2230.0 136 Cs 0.145 44.5 Cs 1 3 7 21.1 6497.0 38 Xe1 0.334 103.0 Cs138 0.643 198.0 1

Ba 40 3.47 x 10- 3 1.07 140 3.25 x 10-3 La 1.00 10-4 Co 6 0 5.19 x 0.160 10-5 Fe 5 9 2.13 x 6.56 x 10-3 10O-3 Co58 4.66 x 1.435 Mn54 10-5 2.75 x 8.48 51 10-3 Cr 3.80 x 1.17 10)-7 Zr 95 9.35 x 2.88 x 10-4 9-26

APPENDIX B TIRP Participant Consensus Agreement Attachment 4 Radionuclide Fractions in Radwaste December 12, 2001 Page 15

Sample Report 10/18/2001 Revision Date " 01/0612000 Report Date :

User Sample ID 96-R-5 Description TK-109 Resin Activity Units uCi/gm Sample Date 11116/1996 Nuclide Activity LLD  ;%Abundance Scaling Nuclide Scaling Factor iI 0.440 % Cs-137 1.725E-02 H-3 2.380E-01 IFalse C-14 5.590E-02 ..False 0.100 % Co-60 5.274E-03 Mn-54 3.890E-01 False 0.720 % Co-60 3.670E-02 Fe-55 4.900E+00 False i 9.030.% Co-60 4.623E-01 Co-57 3.530E-02 False j 0.070 0/ Co-60 3.330E-03 Co-58 9.600E-02 False 0.180% Co-60 9.057E-03 Co-60 1_060E+01 t! False j 19.540% N/A N/A Ni-63 1.730E+01 False 31.8900% Co-60 1.632E+00 S-90 8_440E-02 False 0.160 % Cs-137 6.116E-03 Tc-99 1.000E-02 True I <LLD> Cs-i 37 7.246E-04 Sb-125 J 7.740E-01 False 1.430 Cs-I 37 5.609E-02 1-129 2.700E-04,! True <LLD> Cs-137 1.957E-05 Cs-134 5.960E+00 False 10.990 % Cs-137 4.319E-01 Cs-1 37 1.380E+01 False 25.440 % N/A N/A Ce-144 1.400E-01 True <LLD> NIA N/A Pu-238 2.640E-04 False 0.000 0/0 Ce-144 1.886E-03 Pu-239 1.530E-04 False 0.000  %/ Ce-144 1.093E-03 Pu-241 1.140E-02 False 0.020 Y/o Ce-144 8.143E-02 Am-241 1.640E-04 False 0.000 0 Ce-144 1.171E-03 Cm-242 2.730E-05 False 0.000 YO Ce-144 1.950E-04 Cm-243 1.540E-04 False 0.000 % Ce-144 1.100E-03

Sample Report Report Date : 10/1812001 Revision Date : 10/18/2001 User Sample ID 98-F-4 IL Description SFP Filter FL-2 Activity Units uCilgm Sample Date 08/3111998 Nuclide Activity LLD '%Abundance Scaling Nuclide Scaling Factor

__I H-3 5.180E-03 [ False 0.970 % Cs-137 6.727E+00 C-14 4.400E-03 True 1 <LLD> Co-60 1.577E-02 Fe-55 9.680E-02 False 18.200 Y/ Co-60 3.470E-01 Co-60 2.790E-01 -False j 52.470 % N/A N/A Ni-63 1.460E-01 False 27.460 % Co-60 5.233E-01 Tc-99 5.600E-03 ITrue <LLD> Cs-137 7.273E400 Sb-125 2.350E-03 False I 0.440 0/ Cs-137 3.052E400 1-129 3.OOOE-04 I True <LLD> Cs-137 3-896E-01 Cs-137 7.700E-04 False 0.140% N/A N/A Ce-144 2.200E-03 I True <LLD> N/A N/A Pu-238 3.230E-05I

_ False 1 0.010 0/ Ce-144 1.468E-02 Pu,239 3-400E-05 False _ _0.010%_ Ce-144 1.545E-02 Pu-241 1-470E-03 False 0.280% Ce-144 6-682E-01 Am-241 6.020E-05 False 0.010% Ce-144 2.736E-02 Cm-242 6.100E-06 False 0.000 % Ce-144 2.773E-03 Cm-243 6.440E-05 False 0.010 % Ce-144 2.927E-02

________________________ _______________ I______________ ___________________________ ___________________________

Sample Report Report Date : 10118/2001 Revision Date : 10118/2001 User Sample ID 98-F-2 IL Description FL35A FILTER Activity Units uCitgm Sample Date 02/09/1998 Nuclide Activity LLD  !%Abundance Scaling Nuclide Scaling Factor H-3 7.350E-03 I False t 1.020 % Cs-137 4.594E+00 C-14 Mn-54 1.100E-02 2.380E-03 1Fas I True False j

1

<LLD>

0.330 %

Co-60 Co-60 2.434E-02 5.265E-03 Fe-55 1.860E-01 -False 25.770% Co-60 4.115E-01 Co-60 4.520E-01 False 62.630 % N/A N/A Ni-63 6.610E-02 False 1 9-160 % Co-60 1A462E-01 Nb-95 2-090E-03 False 0,290 % Cs-137 1.306E+00 Tc-99 2.700E-03 True <LLD> Cs-137 1.688E+00 Sn-113 7.OOOE-04 False 0.100% Co-60 1.549E-03 Sb-125 5.110E-03 False 0.710% Cs-137 3.194E+00 1-129 2.600E-04 True <LLD> Cs-137 1-625E-01 Cs-137 1.600E-03 True <LLD> N/A N/A

_________________I_________________ ___________ _______________I___________________ ___________________

Sample Report Report Date - 10/1812001 Revision Date : 1011812001 User Sample ID 98-R-2 IL Description DECON RESIN Activity Units uCi/gm Sample Date 03125/1998 Nuclide Activity LLD 1%Abundance Scaling Nuclide Scaling Factor H-3 I 1.200E-03 True I <LLD> Cs-137 1.319E-01 C-14 i 2.900E-03 True <LLD> Co-60 2.923E-04 Mn-54 5.900E-02 False 0.490 % Co-60 5-948E-03 Fe-55 1.230E+00 I False 10.150 % Co-60 1.240E-01 Si Co-57 4.440E-03 I False 0.040 % Co-60 4.476E-04 Co-58 6.140E-02 False 0-510% Co-60 6.190E-03 Co-60 9920E+00 False 81.820% N/A N/A I I Ni-63 8.430E-01 False 6.950 % Co-60 8.498E-02 Sr-90 7.190E-04 j False 0.010 % Cs-137 7.901E-02 Tc-99 4.900E-03 True <LLD> Cs-137 5385E-01 129 3.800E-04 i True <LLD> Cs-137 4.176E-02 Cs-137 9.100E-03 I True I <LLD> N/A N/A Ce-144 { 2.500E-02 True <LLD> NIA N/A Pu-238 8.200E-05 False 0.000 % Ce-144 3.280E-03 Pu-239 9.350E-05 False 0.000 % Ce-144 3.740E-03 Pu-241 4.870E-03 False 0.040 % Ce-144 1.948E-01 Am-241 1.470E-04 False 0.000 % Ce-144 5.880E-03 Cm-242 3.640E-05 False 0.000 % Ce-144 1.456E-03 Cm-243 9,170E-05 False L 0.000 % Ce-144 3.668E-03 i I

Sample Report Report Date : 10/18/2001 Revision Date : 10/18/2001 User Sample ID 98-R-4 IL Description DECON CLEANUP RESIN Activity Units uCi/gm Sample Date 03/25/1998 Nuclide Activity LLD %Abundance Scaling Nuclide Scaling Factor H-3 2.200E-04 True <LLD> Cs-137 4.314E-01 C-14 1.100E-03 I True <LLD> Co-60 3.741E-02 Mn-54 2.300E-04 I False 0.090% Co-60 7.823E-03 Fe-55 2.540E-02 False 9.770 % Co-60 8.639E-01 Co-58 3.510E-04 False 0.140% Co-60 1.194E-02 Co-60 2.940E-02 False 1 11.310 % N/A NIA Ni-63 1.020E-02 False I 3.920 % Co-60 3.469E-01 Zr-95 I 4.980E-03 False 1.920 % Cs-137 9.765E+00 Nb-95 7.810E-03 False 1 3.000% Cs-137 1.531E+01 Tc-99 i 8.900E-05 True <LLD> Cs-137 1.745E-01 Ag-110m i 3.380E-04 False 0 130 % Co-60 1.150E-02 Sb-124 3.620E-04 False I 0140 % Co-60 1.231E-02 Sb-125 1.809E-01 False 69.580 % Cs-137 3.547E+02 1-129 2.500E-05 True <LLD> Cs-137 4-902E-02 Cs-137 5.100E-04 True <LLD> N/A NIA

Sample Report Report Date : 10118/2001 Revision Date : 10118/2001 User Sample ID 01-D-1 IL Description RCP Internals Activity Units uCilswipe Sample Date 0110812001 Nuclide Activity LLD i%Abundance Scaling Nuclide Scaling Factor H-3 2.600E-03 True <LLD> Cs-137 5-417E-01 C-14 2.900E-03 True <LLD> Co-60 6.954E-03 Fe-55 5.590E-02 False 8.890 % Co-60 1.341E-01 Co-60 4.170E-01 False 66.350 % N/A N/A Ni-63 1.510E-01 -False 24.030 % Co-60 3-621E-01 Tc-99 2.800E-02 True <LLD> Gs-137 5-833E+00 1-129 4.OOOE-04 True <LLD> Cs-137 8.333E-02 Cs-137 4.800E-03 True <LLD> N/A N/A Ce-144 1.400E-02 True <LLD> N/A N/A Pu-238 4.600E-05 False 0.010 % Ce-144 3.286E-03 Pu-239 6.100E-05 False 0.010 % Ce-144 4.357E-03 Pu-241 4.390E-03 False I 0.700% Ce-144 3.136E-01 Am-241 9.400E-05 False 0.010% Ce-144 6.714E-03 Cm-243 1.960E-05 False 0.000 % Ce-144 1.400E-03

A*AAA4 DATA BASE

SUMMARY

FOR WASTE TYPE 14"*44 .............. AS OF 04/23/91 .............

WASTE CHARACTERISTIC FILE (WASDAT.DAT)

SUMMARY

INFORMATION CALCULATION INFORMATION

TE

DAW-U-NA WASTE VOLUME(FT3): 23.8 6.7

'SICAL FORM: SOLID DENSITY(LBS/FT3):

WASTE WEIGHT(LBS): 160.0

ýMICAL FORM: DEP.METAL OXIDES

'KAGE TYPE: POLY HIC(37:4LL) LIMIT FACTOR:

28.0 ACTIVITY FACTOR: I1ýý

KAGE VOL(FT3)

'KAGE WT(LBS): 410.0 SUMMA Y OF WASTE SAMPLE DATA FILE *A*AA**4

,LIDE SAMPLE ABUNDANCE SCALING DOSE DATA (PERCENT) FACTOR (PERCENT)

-51 1.79E-02 2.94 1.84E-01 .13

.27 1 .67E-02 .34

-54 1.63E-03 14.90 9.33E-01 .00

-55 9. 1OE-02

.04 2.71E-03 .01

-57 2.64E- 04 15.20 9.48E-01 22.62

-58 9.24E-02 9.75E-02 16.00 1.00E+00 60.74

-60

.22 1.39E-02 .41

-59 1.36E-03 32.50 2. 03E+ 00 .00

-63 1 .98E-01

.41 2.57E-02 .89

-124 2.51E-03

.43 2.69E-02 .27

-125 2. 62E-03

.13 8.31E-03 .11

-65 8. 1OE-04

-110m 9. 57E- 04 .16 9.82E-03 .60 &wa.

.04 3.25E-03 .00 3 2.46E-04

.05 2.94E-03 .00 14 2.87E-04 lt414C4.

.37 2.95E-02 .00

-90 2. 23E-03

.39 3.13E-02 .43

-95 2.37E-03

.65 5.21E-02 .75

-95 3.94E-03 jrP t/A ,vP

.04 3.31E-03 .00

-99 2.50E-04 5.59E-04 .09 7.39E-03 .06

-103

.49 3.97E-02 .00

-106 3.OOE-03 2.30E-05 .00 3.04E-04 .00 129 5.13E-03 .84 6.79E-02 2.03

-134 7.56E-02 12.40 1.00E÷00 10.60

-137 2.99E-03 .49 1.OOE+00 .02

-144 1.70E-04 .03 2.25E-03 .00

-89 I . 67E-05 .00 5.59E-03 .00

-238 3.91E-05 .01 1.31E-02 .00 239/40 5.80E-03 .95 1.94E+00 .00

-241

-241 1.76E-05 .00 5.89E-03 .00

-242 I. 21E-05 .00 4.05E-03 .00

.00 9.70E-04 .00 243/44 2.90E-06 AVERAGE GAMMA MEV: .6586

APPENDIX B TIRP Participant Consensus Agreement Attachment 5 Sr-90 Detectability and Dose December 12, 2001 Page 16

IMPACT OF SR-90 FRACTION ON DETECTABILITY AND DOSE ATTACHMENT 5 12/12/01, Rev 4 Sheet 1 of 2 DETECTABILITY Sr/Cs Ratio=0.005 Normalized Sr/Cs Ratio=0.02 Normalized Sr/Cs Ratio=0.05 (max ratio) Normalized Concrete Ave Beta Apparent Concrete Ave Beta Apparent Concrete Ave Beta Apparent Nuclide LTP Fxn Energy Beta Eng Nuclide Fxn Energy Beta Eng Nuclide Fxn Energy Beta Eng H-3 0.0236 0.00568 0,000134 H-3 0.0236 0.00568 0.000134 H-3 0.0236 0.00568 0.000134 Fe-55 0.00481 0 0 Fe-55 0.00481 0 0 Fe-55 0.00481 0 0.000000 Co-57 0.000306 0 0 Co-57 0.000306 0 0 Co-57 0.000306 0 0.000000 Co-60* 0.0584 0.096 0.005606 Co-60* 0.0584 0.096 0.005606 Co-60* 0.0584 0.096 0,005606 Ni-63 0.355 0.017 0.006035 Ni-63 0.355 0.017 0.006035 Ni-63 0.355 0.017 0.006035 Sr-90* 0.0028 0.196 0.000549 Sr-90* 0.012 0.196 0.002352 Sr-90* 0.026 0.196 0.005096 Y-90* 0.0028 0.935 0.002618 Y-90* 0.012 0.935 0.01122 Y-90* 0.026 0.935 0.024310 Cs-134* 0.00455 0.157 0.000714 Cs-134* 0.00455 0.157 0.000714 Cs-134* 0.00455 0.157 0.000714 Cs-137* 0.55 0.171 0.09405 Cs-137* 0.5408 0.171 0.092477 Cs-137* 0.5268 0.171 0.090083 Sum

_ 0.109707 _ Sum 0.118539 Sum 0.131979 detectable beta DOSE-Total Activity Sr/Cs Ratio=0.005 Normalized Sr/Cs Ratio=0.02 Normalized Sr/Cs Ratio=0.05 (max ratio) Normalized Concrete Nuclide Concrete Nuclide Nuclide Concrete Nuclide Nuclide Concrete Nuclide LTP Fxn Dose Fac Dose Nuclide Fraction Dose Fac Dose Nuclide Fxn Dose Fac Dose H-3 0.0236 3.34E-05 7.88E-07 H-3 0.0236 3.34E-05 7.88E-07 H-3 0.0236 3.34E-05 7.88E-07 Fe-55 0.00481 5.85E-07 2.81 E-09 Fe-55 0.00481 5.85E-07 2.81E-09 Fe-55 0.00481 5.85E-07 2.81 E-09 Co-57 0.000306 2,48E-06 7.59E-10 Co-57 0.000306 2.48E-06 7.59E-10 Co-57 0.000306 2.48E-06 7.59E-1 0 Co-60* 0.0584 5.99E-05 3.50E-06 Co-60* 0.0584 5.99E-05 3.50E-06 Co-60* 0.0584 5.99E-05 3.50E-06 Ni-63 0.355 1.15E-06 4.08E-07 Ni-63 0,355 1.15E-06 4.08E-07 Ni-63 0,355 1.15E-06 4.08E-07 Sr-90* 0.0028 6.14E-04 1.72E-06 Sr-90* 0.012 6,14E-04 7.37E-06 Sr-90* 0.026 6.14E-04 1.60E-05 Cs-134* 0.00455 3.33E-05 1.52E-07 Cs-134* 0.00455 3.33E-05 1.52E-07 Cs-134* 0.00455 3.33E-05 1.52E-07 Cs-1 37* 0.55 2.25E-05 1.24E-05 Cs-137* 0.5408 2.25E-05 1.22E-05 Cs-137* 0.5268 2.25E-05 1.19E-05 I_ Sum 1.89E-05 _Sum 2.44E-05 _Sum 3.27E-05 mrem/y/dpmr __________________________mrem/y/dpm mrem/y/dpm

ATTACHMENT 5 Sheet 2 of 2 DOSE-Detectable Beta Sr/Cs Ratio=0.005 Normalized Sr/Cs Ratio=0.02 Normalized Sr/Cs Ratio=0.05 (max ratio) Normalized Ave Fxn Nuclide Concrete Ave Fxn Nuclide Concrete Ave Fxn Nuclide Concrete Nuclide beta det. Dose Fac Dose Nuclide beta det. Dose Fac Dose Nuclide beta det. Dose Fac Dose H-3 0.038326 3.34E-05 1.28E-06 H-3 0.038326 3.34E-05 1.28E-06 H-3 0.0383264 3.34E-05 1.28E-06 Fe-55 0.007811 5.85E-07 4.57E-09 Fe-55 0.0078114 5.85E-07 4.57E-09 Fe-55 0.007811 5.85E-07 4.57E-09 Co57 0.000497 2.48E-06 1.23E-09 Co-57 0.0004969 2.48E-06 1.23E-09 Co-57 0.000497 2.48E-06 1.23E-09 Co-60* 0.094842 5.99E-05 5.68E-06 Co-60* 0.0948416 5.99E-05 5,68E-06 Co-60* 0.094842 5.99E-05 5.68E-06 Ni-63 0.57652 1.15E-06 6.63E-07 Ni-63 0.57652 1.15E-06 6.63E-07 NI-63 0.57652 1.15E-06 6.63E-07 Sr-90* 0.019488 6.14E-04 1.20E-05 Sr-90* 0.042224 6.14E-04 2.59E-05 Sr-90* 0.004547 6.14E-04 2.79E-06 0.007389 3.33E-05 2.46E-07 Cs-1 34* 0.007389 3.33E-05 2.46E-07 Cs-134* 0.0073892 3.33E-05 2.46E-07 Cs-134*

Cs-137* 0.878259 2.25E-05 1.98E-05 Cs-137* 0.8555232 2.25E-05 1.92E-05 Cs-137* 0.8932 2.25E-05 2.01E-05 3.08E-05 Sum 1.623 Sum 3.96E-05 Sum 1.623 Sum 5.31E-05 sum 1.623 Sum mrem/y/dpm mrem/y/dpm rnrem/y/dprn

[Sr/Cs=0,05 ISr/Cs=0.02 Total Annual Dose For Sr/Cs Ratio Increase**

Normalized Normalized Material 0.005 0.02 0.05 1.203014 1.08 Contain Con 0.556 rnrem 0,713 mnrem 10.968 mrem Increase in detectability Change in dose 1 dpm 1.72E+00 1.29 Emb. Pipe 0.0208 " 0.0267 " 10.0364 "

Change in dose 1 dpm beta 11.72E+00 1.29 Bur. Pipe 0.0064 " 0.0168 " 10.0334 "

Sr-90/Cs Ratio By Core and 18,000 dpm Contam. Conc. Dose Rate (mrem/y) **At a DCGL of 18,000 dpm/100 cm2.

101FLI 01FL3 01FL4 01FL6 01FL8 02FL2 02FL5 I11.63E-03 1.61E-03 6.62E-03 1.53E-031 5.92E-04 4.12E-04 <4.66E-2 9.05E-01 4.28E-01 7.52E-01 414E-01 4.59E-01 4.14E-01 9.31E-01 12-C03 12-C04 12-C05 13-C01 Mean Std Dev Range 1.34E-03 1.68E-03 1.51E-031 6.68E-03 6.39E-03 1.35E-0214.12E-4 to <4.66E-2 4.92E-01 4.86E-01 I 4.60E-01 I 5.99E-01 5.79E-01 O/A Trench Tunneli oPAB 1.O 2 4.0-E--021 NOTE: Unless otherwise noted, St-90 values presented in this attachment include contributions from Y-90.

APPENDIX B TIRP Participant Consensus Agreement Attachment 6 NUREG/CR-4289 Excerpts December 12, 2001 Page 17

PNL-5429 esidual Radionuclide Contamination Jithin and Around Commercial uclear Power Plants and sessment of Origin, Distribution, Inventory

-commissioning nuscript Completed: October 1985

!e Published: February 1986

'pared by H. Abel, D. E. Robertson, C. W. Thomas, A. Lepel, J. C- Evans, W. V. Thomas, C. Carrick, M. W. Leale cilic Northwest Laboratory

hland, WA 99352 repared for ivision of Engineering Technology ffice of Nuclear Regulatory Research

.S. Nuclear Regulatory Commission

!ashington,D.C. 20555 RC FIN B2299

4.0. RESULTS AND DISCU'SION the seven and analytiCal program for exists in:

The results of the insampling this program indicate a wide range reactor sites examined at the

1) the relative radionuclide composition of coT aminated materials inventory of the residL3l radionuclide reactor sites, 2) the distribution and 3) the tLtal inventory present.

stations, throughout the generating program are dis resulting from the field and ana Artical where The observations compared to previous assessments, and are related, dis cussed below; are A more complete parameters and history.

possible, to site operating process is presented in Section cussion of the inventory construction upon a estimated for the seven plants are based 4.2. The inventories uncertainty in number of samples and data points. The greatest limited process; if inventory estimates is associated with the sampling these existed corrosion product deposition large variations in radioactive from our limited number of throughout plant systems, extrapolations uncertainties. Thus, the samples would be subject to corresponding degree of herein should be used with some inventory estimates presented estimates are accurate to within caution- We feel that our inventory other inventory studies.(6,1O)

+/-50%, based upon comparisons with 4.1. RESIDUAL RADIONUCLIDE COMPOSITIONS of contaminated materials ob The relative radionuclide composition The range in com reactor sites ranged considerably. 1) the elapsed at the seven factors including:

positions was influenced by numerous served

2) rated generatin capacity; time from the last reactor operations; systems; 4) reactor type operating

3) materials of construction for the and corrosion control; chemistry e.g. PWR, BWR, dual cycle; 5) coolant episodic equipment failure and 7)

6) fuel integrity during operations; liquids; and 8) the specific source for the leakage of contaminated or steam system, rad waste samples, e.g. primary system, secondary facility.

system, or the spent fuel storage 4.1.1 Piping and Hardware - Generic Observations for the total plant inventor The relative radionuclide compositions and hardware estimated for the ies of contaminated piping, equipment, 4.1. These inventories include seven sites sampled are shown in Table deposited on corrosion film and crud only the radioactive contamination and do not include the highly surfaces of the various plant systems, vessel.

activated components of the pressure The most abundant radionuclides in samples two to three months old 5 4 Mn, 5 5 Fe, 5 8 Co, 6 0 Co, and 63Ni. Zinc-65 or older generally included film in relatively high concentrations in BWR corrosion 8 pu, 239- 2 4 0pu, was presentTraces of transuranic radionuclides, including 23 samples. Other long 2 4 1Am, 242Cm, and 2 4 4 Cm, were also observed in the residues.

a waste management standpoint, such lived radionuclides of interest from below their limits of detection; ld 99 and 1291, were frequently as 9 49 Nb, wasTc, always extremely low concentrations. The 9 0 Sr, and 0Sr present in 9

TABLE 4.1. Long-lived Residual Radionuclide Compositions in Total Plant Inventories* at Seven Nuclear Generating Stations Composition In Percent of Total Activity Decay Corrected to Shutdown Date or Sampling Date Dresden-1 Monticello Indian Point-I Turkey Point-3 Rancho Seco Radionucl ide Pathfinder Humboldt Bay 4

3 0.9 1 4 0.4 Mn-54 1 67 31 28 0.36 90 28 Fe-55 43 24 Co-57 24 18 1.0 46 11 15 Co-60 6 0.004 0.02 2 X 10-4 0.09 19

.Ni-59 0.04 2 0.1 0.03 0.2 5 rNi-63 11 1 0.09 19 84 Zn-65 98.6** 0,0007 0.0008 <0.01 0,007 0.002

<1 X 10-4 0.004 <0.004 Srý-90 <0.1 00008 <0.004

<0. 004 <0.003

<3 X 10-6 <0.005 Nb-94 4 X 10-5 8 X 10-5 8 X 10- 5 0.008

<1 X0-5 3 X 10-4 4 Tc-99 Ag-110m <0.003 <I X 10-5

<I X 10-5 <1 X 10-6 2 X 10-5

<3 X 10-6 0.4 1-129 0.04 2 0.5 3 X I0-4 0.5 <0.04 Cs-137 0.2 Ce-144 1 0.002 0.006 0.001 0.005 0.1 0 008 TRU*** 3 X 10-6 Total Plant 4460(g)

Inventory 1070(e) 57(a) 2 3 5 0 (c) 448(d)

(Cur ies) vessel and internals, contaminated

• Excludes highly activated metal components of the reactor pressure concrete, and soils, radionuclide

• • At time of station shutdown, 65 Zn was the overwhelming contributor to the total residual 13 years between shutdown and our sampling and inventory at Pathfinder, but had all decayed during the in this table was The relative radionuclide composition and total inventory analysis program. 0 shortly after shutdown.

by NSP reconstructed from our data and a measured 65Zn/6 Co ratio made including 238,239.240pu,

• • • Transuranic alpha-emitting radionuclides with half-lives greater than 5 years, 2Al4P~m 2 4 3 Am, and 244Cm.

9 9 Tc, and 1291 were generally very soluble and did not accumulate in any 9 4Nb may be Although significant degree in corrosion product deposits.

in the reactor vessel by activa produced in relatively large quantities vessel com tion of trace niobium impurities in stainless steel pressure product deposits were ponents(l,2,13), its concentrations in corrosion of of the extreme insolubility extremely low. This may be the result 6 Ru, 1 4 1 -144Ce, and as 103,10 niobium. Occasionally, radionuc]ides such liOmAg were detectable in the corrosion films.

was As shown in the table the relative radionuclide composition 6 0 Co were the two most abundant radio quite variable. However, 55Fe and These two radio nuclides in all cases except Monticello and Pathfinder.

at Humboldt nuclides constituted over 95% of the estimated inventories Dresden Unit One, and Bay and Turkey Point. At Indian Point Unit One, Rancho Seco they accounted for 86, 84, and 60%, respectively, of the total 5 5 Fe and 6 0 Co accounted for the majority estimated inventory. Although of the inventory (greater than 60% at five of the seven stations), the The ratio relationship between the two radionuclides was quite variable. 5 5 Fe and 6 0 Co consti where of 5 5Fe to 6 0 Co at the six generating stations tuted the majority of the inventory ranged from 15 to 1 at Humboldt Bay 6 5 Zn constituted 90% of to 0.36 to I at Pathfinder. At Monticello, where 5 5 Fe to 6 0 Co ratio was even lower, 0.09 to 1.

the total inventory, the 5 This large variability was presumably due to differences in: 1) the parent

• element composition of construction materials used in the pressure vessel and primary coolant loop, 2) differences in the water chemistry which controlled the corrosion and deposition of these radionuclides, and 3) differences in operating history which affect the production ratios since the radionuclides have an approximate factor of two difference in half-life.

2 3 8 pu, 2 3 9 - 2 4 0 pu, 2 4 1 Am, 2 4 2 Cm, and The transuranic radionuclides (

2 4 4 Cm) constituted percentages of the total inventory ranging from 3 X 10-6% at Pathfinder to 0.1% at Dresden Unit One. The value observed at Pathfinder was somewhat anomalous due to its short operational period prior to shutdown, along with the fact that there were no detected fuel

• failures during this brief period. Considering the other six stations, the total transuranic component of these inventories ranged a factor of o , fifty, from 0.002% at Indian Point Unit One to 0.1% at Dresden Unit One.

5.7 The largest ran es, as a percentage of the total inventory, were noted for 6 5 Zn and 6 %Ni, which showed ranges of 1090 and 630, respectively.

L), This wide range was related to the composition of the materials of construc tion used in the primary systems of the reactors. The large amounts of 6 5 Zn observed at Monticello, Dresden Unit One, and Indian Point Unit One were the result of the use of admiralty brass heat exchangers (29% zinc).

These were replaced with stainless steel at Dresden Unit One, but a large L i residual corrosion product inventory was still present in the operating systems at the time of our sampling. Pathfinder also utilized admiralty 6 5 Zn accounted for brass heat exchangers, and at the time of shutdown over 98% of the total radioactivity. Humboldt Bay power plant also initi ally utilized admiralty brass heat exchangers which resulted in generating relatively large amounts of 6 5 Zn during early reactor operations.

11

6 3 Ni was due to a relatively high percentage abun The wide range in low abundance at dance observed at Rancho Seco (19%) and an unusually 6 3 Ni in the Rancho Seco inven Monticello (0.04%)- The large component of (80% nickel) in the tory was due to the more extensive use of inconel 3 primary system of this reactor. The low percentage of 6 Ni at Monticello use of nickel alloys since was typical of newer BWRs which make minimal the relatively more oxidi.

they are subject to higher rates of corrosion in coolant loop. Excluding Rancho Seco ing environment of the BWR primary 6 3 Ni ranged a factor of 50, and Monticello, the relative abundance of Unit-3 to 5% at Dresden from 0.1% of the total inventory at Turkey Point Unit One.

corrosion films were The highest concentrations of radionuclides in Table 4.2 lists observed in the systems exposed to the primary coolant.

for a number of the the concentration ranges and average concentrations fron primary radionuclides observed in corrosion films attached to piping measurements These data were synthesized from individual coolant systems.

on appropriate samples given in Appendix C. Large variations in residual piping existed between radionuclide concentrations on the primary coolant in the table should the various stations, so the average value presented be considered with appropriate reservations.

9 0 Sr and The residual concentrations of fission products, such as are di 1 3 7 Cs and the transuranic radionuclides (Pu, Am, Cm), on piping history of the rectly related to the fuel integrity during the operating fuel element failures were relatively plant. At power stations where Dresden Unit One), the concentra frequent or severe (e.g. Humboldt Bay, radionuclides in the corrosion tions of fission products and transuranic stations where the fuel remained rela films were higher than observed at tively free from failures.

BWRs con Secondary coolant loops in PWRs and condensate systems in than observed in primary tained much lower radionuclide concentrations Typically, radionuclide concentrations were loop or feedwater samples.

in the secondary system samp1t.

approximately two orders of magnitude lower and average concentrations obserw Table 4.3 contains concentration ranges in the corrosion films from nonprimary loop samples. The actual measure-f ments from which this table was constructed are given in Appendix C. Theýý,

average concentration values presented in the table should be used with table typically cover several reservations since the ranges shown in the orders of magnitude.

upon reactofi The piping and hardWare data from the seven sites, based below.

type, are discussed in detail 4.1.2. PWR Piping and Hardware Contaminated piping and hardware samples were obtained from three PW1s-Indian Point Station Unit One, Rancho Seco Nuclear Generating Stati and Turkey Point Station Units 3 and 4. Indian Point Station Unit One is an early design Babcock & Wilcox reactor, Rancho Seco is a more recent 12

j]

in Corrosion Concentration Ranges of Radionuclides Reactor Coolant(a) to Primary Films on Piping Exposed Average 1)

Concentration S(pCi/cm fiaiflife (r) (pCi/cm1 ) Range Concentration Radionucl ide 1.2 0.028 - 4.4 0.854 33 Mn-54 0.039 - 149 2.7 6.4 Fe-55 0.16 - 23 5.27 6.6 x 10-3 Co-60 <5 x 10 2.6 x 10-2 75,000 0.80 Ni-59 0.003 - 1.3 100 5.6(b)

Ni-6 3 0.0005 - 5.8 3.2 x 10-3 0.67 Zn-65 <3 x 10 8.4 x 10-3 2.2 x 10-4 28.5 Sr- 90 <I x i0 5.0 x 10-4 1.5 x 10-4 20,000 Nb-94 4.5 x 1) 5.6 x 10)-4 1.4 x 10-6 2.13 x 105 Tc-9 9 <1 x 10 4.3 x 10-6 1.57 x 107 0.030 1-.129 0.019 - 0.046 2.06 0.056

.Cs-134 0.003 - 0.17 1.1 x 10-3 30.2 Cs-137 2.4 x 10 4.5 x 10-3 9.8 x 10-4 87.8 Pu-238 1.5 x 10 4.4x 10-3 1.8 x 10-3 24,400 Pu-239,240 1.8 x 10 8.3 x0-3 7.2 x 10-2 433 Am-241 7.2 x 10 2.4 X 10-1 6.8 x 10-4 0.447 Cm-242 2.2 x 10 2.5x 10-3 18.1 Cm-244 or shutdown for terminated reactors (a) Decay corrected to date of units. Excludes data from Pathfinder sampling date for operating atypically low.

Generating Plant, which was (b) Average value for two BWR units.

13

TABLE 4.3. Concentration Ranges of Radionuclides in Corrosion Hardware Films Internally Deposited in Piping and and Secondary Coolant Exposed to Liquid Radwastes Average Concentration Halflife LEI Concentration Range Radionuclide (pCi/cm2)

(pCi/cm2 )

0.854 2 - 4.70 X 105 160,000 (3)A Mn-54 1.0 x 106 (7) 2.7 710 - 7.1 x 106 Fe-55 42,000 (7) 5.27 64 - 2.74 X 105 Co-60 6.2 (5) 75,000 0.63 - 15 Ni-59 1,900 (6) 100 3 - 10,000 Ni-63 14,600

<0.4 - 27,400 (2) 0.67 It Zn-65 88 28.5 <0.009 - 260 (3)

Sr-90

<0.1 - 100 ns Nb-94 20,000

- 0.2 a Tc-99 2.13 x10 5 <0.05 420 b 2.77 9.3 -. 1,200 (3) a Sb-1?5

<0.0006 - 0.9 b 1-129 1.57 x 107 a

2.06 <0.3 - 1,900 T Cs-134 860 30.2 <0.6 - 4,000 (5) ti Cs-]37 0.78 <0.6 - 2,000 Ce-144 7.4 (7) 87.8 0.0014 - 51 Pu-238 3.6 (7) 0.0012 - 24 Oi Pu-239,240 24,400 (6) 41 7.3 TI 433 0.0009 -

Am-241 820 5.

0.447 0.0013 - 3,600 (5)

Cm-242 12 ri 18.1 0.0015 - 58 (5)1 ha Cm-244 or tr a) Numbers in parentheses indicate the number of ma reactor sites used for construction of the co average value presented. ti b) Average value of two BWR units.

14

ýi I IABLE 4.4. Concentration Ranges of Radionuclides Associated with Concrete from Highly Contaminated Areas Within Selected Nuclear Generating Stations i-,('

Average Halflife (yr) Concentration Range Concentrat ion Radionuclide (pCi/cm2) (pCi/cm2)

Mn-54 0.854 35 - 21,000 6,200 (5) a II:

Fe-55 2.7 2,200 - 830,000 200,000 (5)

Co-60 5.27 590 - 460,000 110,000 (5)

Ni-59 75,000 30 - 2,400 860 (3)

Ni-63 100 3,100 - 6,400 4,800 (2)

Sr-90 28.5 1.6 - 480 170 (4)

Nb-94 20,000 <3 - 50 Tc-99 2.13 x 105 0.27 - 2.4 1.6 (3)

Ru-106 1.0 <30 - 190 Ag-110M 0.686 59 - 3,600 I,800 (2)

Cs-134 2.06 70 -1.7 x 106 310,000 (6)

Cs-137 30.2 550 -2.0 x 106 370,000 (6)

Ce-144 0.78 26 -3.1 x 106 620,000 (5)

Eu-152 12.4 9 - 3,100 1,000 (3)

Eu-154 8.5 90 1,500 680 (3)

Eu-155 4.96 10 - 500 260 (2)

Pu-238 87.8 0.025 - 48 14 (4)

Pu-239,240 24,400 0.089 - 21 7.7 (4)

Am-241 433 0.10 - 30 8.7 (4)

Cm-242 0.447 0.06 - 1,800 880 (3)

Cm-244 18.1 0.05 - 52 13 (4)

Np-237 2.14 x 106 0.013 - 0.026 0.016 (3) a) Number of reactor units included to calculate the average value.

19

Bay bioshield showed the presence of both surface contamination 4.6). The I 52 Eu and

.itu neutron activation products (see Table activation, with the e primarily derived from in situ neutron after concentrations being 309 pCi/gm and 32 pCi/gm, respectively ion to 30 EFPY.

contamination Residues in Soils and Sediments areas of dionuclide contamination of soils within the exclusion lear power stations was typically limited to small patches of very centrations of radionuclides (see Appendix C). Areas included for ltion in this program focused upon locations within radiation con I areas and where spills of radioactive materials (mostly liquids)

urred. This was a cursory examination of areas of known or sus contamination and did not seek to address the extent of such con

-ion. As would be expected from the intent of the sampling, resid dionuclide contamination was usually present; however, the levels tamination were usually below a suggested residual radionuclide ination level which would result in an annual dose of 10 mreuVyr.(15) r, several small patches of soils and hqlding pond sediments at I of the stations contained 60Co and 1 3 'Cs concentrations up to

,l tens to hundreds of pCi/g (on a dry weight basis). Table 4.7 the ranges for the detectable gamma emitting radionuclides in exclu rea surface soils, and for 238Pu and 2 3 9 -240pu in Pathfinder and idt Bay exclusion area soils. The most abundant radionuclides were Illy 6 DCo and 1 3 7 Cs, with occasional trace amounts of other radio Jes including 134 CS 5 8 Co, 5 4 Mn, 1 0 6 Ru, 1 1 0 aAg, 12 5 Sb, and 1 4 4Ce.

DISTRIBUTIONS AND INVENTORIES OF RESIDUAL CONTAMINATION A significant part of this research program was the determination of ibutions and inventories of residual radionuclide contamination within arious operating systems at the PWR's and BWR's examined. This pro entailed utilizing the residual concentrations, as defined by the ing and analytical phases of the research program, to construct operat

.ystem and plant inventories. In order to construct operating system

.otal plant inventories of residual radionuclides, it was necessary

!termine total surface areas present in the various systems upon which

ontaminated corrosion films were deposited. The total surface area Pates were constructed by either: 1) using the surface area estimates the reference PWR or reference BWR studies, corrected by a scaling Dr, or, whenever possible, 2) through the use of the specific plant etrics and prints for the various systems and subcomponents. The onuclide distributions within the plant systems and the inventories ystem are described below, based upon reactor type. The plants were rated by reactor type since there were significant differences between operating systems of BWR's and PWR's, and thus significant differ s between the radionuclide distributions throughout the plant.

25

Soils U- 4 cm) r ,i, " (.1 ý I ý ý - ý II Turkey Point 5 Rancho Seco 6 Dresden3 Monticello 4 Pathf inderl Humbo 1dt _Lay Radionucl ide 0,03-0.34 <0.003-0.27

0. 02-0, 23 <0. 004 -<0. OZ

<0.005-0.06 0.45-5.5 4.0-45 0.012-11 54 Mn 1.3-161 0. 006-0 .45 6 <0.01-1,7 26-377 <0. 1-0.2 <0.03-<0.09 OCo <0.07-0,2 <0.07

<0, 1-0.36 <0.Oz-<0, 05 <0.006-0.75 106Ru <0.03 <0.06-22

<0.4-4.9 <0, 03-<I 0.01-0.95 1 2 5 Sb <0. 02 <0.004-0.16 0.28-5.5 1,5-6,1 <0.01-6.3 0.o05-4, 9

<0. 004 -<0.01 1,7-11 1 34 Cs 0.49-260 0.068-2.1 1 3 7 Cs 0,15-2,9 25-9 1 <0, 05-0.27 <0.02

<0.04-1.5 0,083-0,17

<0, 03 <0, 3-1. 3 N.M.

1 4 4 Ce N.M, N.M.

8.2-170 X 10-3 N.M. N.M.

2 3 8 pu 3-41 X 10-5 N.M.

9.5-230 X 10-3 N.M. N.M.

2 6-42 X 10-4 Z39- 40pu I. Fourteen soil samples

2. Five soil samples Four soil samples; highest observed contamination

3. of 15-30 cm in O-SS-25B was at a depth

4. Four soil samples

5. Six soil samples 6,, Seven soil samples N.M. Not measured

ATTACHMENT 2 George E. Chabot to Patrick J. Dostie, letter dated November 12, 1998 (Best Copy Available)

As referenced in the Maine Yankee LTP, Rev. 2, Table 2-3

George E. Chabot, Ph.D., CHP 2 Eugley Park East North Reading, MA 01864-1306 Tel: 978-664-5167 978-934-3288 Fax: 978-441-0934 E-mail: georgechaabo tI(auml.edu Nov. 12, 1998 Patrick J. Dostie State Nuclear Safety Inspector Office of Nuclear Safety Division of Health Engineering State of Maine Department of Human Services Augusta, ME 04333

Dear Pat:

Thanks for the information that you recently sent. I reviewed the calibration and use procedure for the VRM-1X large volume plastic scintillation detector that was used in much of the site survey work. I have some comments and questions that apply primarily to the calibration process.

Procedure RPPVRMO1 Rev. No. 0 of 9/30/97, supplied by IT Corp., states in paragraph 7.2.1 that the uncertainty in the background is determined by summing all count rates, taking the square root of the sum, then dividing the resultant value by n-I. This would be approximately correct if the counting time for each background count was one second (assuming count rates are expressed in counts per second).

In general for n counts, the standard error in the mean count rate would be obtained from a = (1R/t.)0 /n (1) where t, is the counting time for the i" observation. I could not find anything in the descriptive material you sent that indicated the time interval for which background counts were collected in making a determination of the background count rate. For times greater than one second uncertainty in the background rate would be less than the value calculated-by the recommended procedure.

In paragraph 7.2.4 of the same procedure the MDA is stated to be obtained from MDA = (2.71 + 4.65 Rbc3)/E (2) where Rb is the mean background count rate, and E is the efficiency determined from calibration. Again, it should be noted that this expression is valid only if the count integration time is one second and if the background and sample counting intervals are the same. In general, the MDA is given by

2 k 2/ts+b + k(Rjtb)r(I+tdt,,b)P M DA= ..................--------- (3)

E where tb is the background couting time, and tb is the sample counting time. For the case when the acceptance level for false positive and false negative conclusions as to activity presence are each set at 0.05, k has the value 1.645 and k2 = 2.7 1. Only when tb is one and tb = t,÷b does equation 3 reduce to /v,-,a.

equation 2.

The cited procedure describes the manner in which minimum detectable activities are to be determined for both point sources and for distributed activity. Paragraph 7.2.3 describes determination of an area calibration factor by summing the average count rates from all 50 grid areas that make up a 10'x4' grid, multiplying by the total grid area and dividing the result by the source activity multiplied by the number of grid areas (i.e., 50). This area calibration factor operates as an efficiency figure which is divided into the MDA of equation 2 to obtain the minimum detectable surface activity concentration. Aside from the problem with equation 3, described above, I do not have a problem with this approach. The Note that appears after Paragraph 7.2.4 is confusing. I believe they meant to say that the general area MDAs can 3 be converted to pCi/g (rather than pCi) by dividing by 11. Since they assume'a soil density of 1.44glcm 2 . 54 cnm") = 11 g/cm2 and and soil contaminated to a depth of 3 inches, the factor is (1.44g/cnm)(3")(

2 when this is divided into the area MDA, with units of pCi/cm , units of pCi/g are obtained. One difficulty with the latter conversion is that it assumes an arbitrary depth of dispersal of the activity in the soil, and no account is taken of attenuation of the gamma radiation in the soil. When the specific calibration was carried out using the Co-60 source at the Maine Yankee site, the equations used in the procedure cited above were used. Near the bottom of the handwritten first page of3 the IT2 calibration the 2 4 to minimum detectable surface area activity of 61.5 pCi/cm is divided by (1.44 g/cm )(3")( .5 cmn/")

obtain 5.6 pCi/g (handwritten sheet mistakenly says 5.6 pCi). I carried out calculations, using Microshield V. 5.01, to see how the response from a uniform dispersion of Co-60 in soil, 3" deep and 10' by 4' in area, would compare with a surface distribution of Co-60, with no soil attenuation; I assumed a height of 20" for the dose point above the ground (5,0 grid coordinate position). The end result was that a given amount of activity dispersed uniformly throughout the soil volume, 3" deep by 10' by 4', produced an exposure rate that was about 0.79 of that obtained when the same amount of activity was dispersed uniformly over a 10' by 4' area. This is simply to point out that soil attenuation may be a factor of concern if activity is distributed to any significant depth in the soil. Clearly, the greater the depth of dispersal of activity below the surface, the greater will be the degree of attenuation and the greater will be the disparity between the calculated minimum detectable activity per gram and the actual minimum detectable activity per gram. Thus, for activity distributed uniformly through a 12" depth of soil, the minimum detectable activity per gram would be about 1/2 of what would be predicted by the IT procedure.

It should be kept in mind that the calibration procedure discussed above applies to a stationary detector.

The IT procedure does not appear to take account of the fact that, in use, the detector is moving. If the soil is contaminated over large areas such that the detector integration time is short compared to the time required for the detector to pass over the contaminated area, then the calibration procedure, possibly modified to consider photon attenuation, is appropriate. For smaller areas of contamination, however, the motion of the detector may reducce sensitivity and yield higher MDAs than those calculated using the current procedure. An example may demonstrate this. If we assume that the vehicle carrying the detector is moving at the nominal maximum speed of 5 mph (7.333 fps), it is clear that the

,detector will not spend very much time over a small source. The detector is 3" wide and, traveling at a rate of 7.333 fps, will pass over a point source in 0.034 seconds. Naturally the detector responds even as it moves away from the source, although at a lower sensitivity. If we assume that the detector is on a track such that the geometric center of the detector will pass over the source, the maximum response of the detector likely will occur during an integration time, T, that includes the detector moving toward, over, and possibly past the source. Using the data (net count rates obtained by subtracting background rate of 650 cps from respective gross rates) taken at the MY site for (x,y) grid coordinates (5.0), (5,1),

(5,2), (5,3), and (5,4) (these coordinates track a line normal to the longitudinal axis of the detector and include the point directly below the geometric center of the detector), I fit the net count rate data vs. the y coordinate to obtain a reasonable expresssion for the expected count rate as a function of displacement away from the y =0 location. The resultant exponential fit was.

R = 2319 exp(-0.572 y), (4) where R is in cps and y is in feet. The y-value may be expressed in terms of velocity and time as y = vt. The total number of counts, C, expected as the detector moves towards the source during the detector integration time is obtained from T T 2

C = JfR dt = f2319 exp(-0.57 2 vt)dt = (2319/0.572v)(1 - exp(-0.57 vT) (5) 0 0 For this case I have assumed an integrating time of one second, consistent with the MDA equations used by IT; for v = 7.333 fps, we obtain C = 545 net counts. The count rate interpreted by the detector would then be 1195 cps (based on a background rate of 650 cps), less than 50% of the count rate obtained for the stationary detector centered above the source. The MDA determined from the gross count rate of 1195 cps and the determined background rate of 650 cps, assuming a one second integration time for each, would have been 2.2 rmcrocuries, about four times greater than the value of 0.57 microcuries obtained by IT. If the detector integrates for longer intervals than one second, the results would worsen. For example, for a five second integration time, the expected gross count would have been 3803, and the interpreted gross count rate would have been 761 cps, which would lead to an MDA of about 4.7 microcuries, about eight times larger than the IT value.

The discrepancies noted above may or may not be significant, depending on the actual extent of site contamination, how the survey data are being used and what additional surveys are planned. If you have any questions concerning the above, please contact me.

Sincerely, George Chabot