ML22293A432
| ML22293A432 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 10/17/2022 |
| From: | ZionSolutions |
| To: | Office of Nuclear Material Safety and Safeguards |
| References | |
| ZS-2022-044 | |
| Download: ML22293A432 (18) | |
Text
ZionSolutions ZS-2020-044 Zion Nuclear Power Station, Units 1 and 2 Response to Audit Questions
Explain the difference between the responses to the two RAis (2021 and May 2022) regarding the likelihood of wind transport.
The March 25, 2022 response to RAls, as further supplemented by the information herein, takes precedence over the February 10, 2021 response regarding the likelihood of wind entrainment of DRPs.
Upon further evaluation, ZionSolutions does not find the previous response provided in February 2021 regarding wind entrainment, or transport, to be sufficiently clear. Neither that response, nor the previous RAI response, has placed significance on this transport mechanism. While the February 2021 response cites wind entrainment as the "... most likely cause for the DRPs identified within the Security Restricted Area... ", it also notes that "... the DRPs were not highly mobile and were not easily dispersed throughout the site."
Wind entrainment is an explanation that is often given regarding particle transport. While it is a known mechanism that has been widely studied, the reference to it in the February 2021 RAI response is more apocryphal than evidentiary. For the DRPs in question, ZionSolutions does not believe that there is evidence that wind transport is viable over a significant distance. By significant we mean, from the survey unit where the DRP was initially deposited to an adjacent survey unit.
The following is meant to augment the response provided in March 2022 (NRC RAI-lc, Response no. 2) regarding the origin of how pa1ticles were introduced to the environment via the movement of potentially contaminated equipment/components (hereinafter "material") through the equipment hatch openings of each Containment Building prior to the erection of the waste loadout tents.
The equipment hatches for each Containment Building were positioned approximately 20 feet off the ground. A Heavy Lift Rail System (HLRS) was installed at each equipment hatch that was comprised of a cart and pulley system used to bring equipment and components into and out of the Containment Buildings. A set of protocols was established for use of moving material out of the Containment Building via the HLRS. Radiation Protection had overall control of material entering or being removed from the Containment Building.
Material that was slated to be removed from the Containment Building was remediated (as necessary), wrapped (as necessary), and surveyed prior to being loaded on the HLRS cart for removal. Note, the radiological surveys performed in the Containment Building would primarily have consisted of swipes for loose surface contamination with a normal limit of <l 000 dpm/100 cm2. Due to high background radiation levels inside of the Containment Building, surveys for DRPs would have not been possible. Once the item was removed from the Containment Building it was outfitted with rigging to enable movement to the ground or a transport vehicle. In some cases, the transfer would take hours or days and the items would remain outside. The photos below depict a liner, a steam generator, and a reactor head being removed from a Containment Building using the HLRS.
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Protocols were put in place for safety and radiological control purposes. In accordance with OSHA standards, if wind speeds exceeded 23 miles per hour (mph) the movement of material was suspended. At Zion, the Waste Manager designated adm inistrative limits (15-18 mph) when lifting material from the HLRS was suspended. If these wind speeds were encountered, the load was taken back into the Containment Building. Additional ly, if medium to hard rain was encountered, the load was taken back into the Containment Building.
Regardless of the safety and radiological control protocols, it was still possible for a DRP to be dislocated from the material during the removal process. The DRPs could have been dislocated due to rain, wind, or personnel interaction ( e.g., during rigging it is possible for the rigging cables and straps to have rubbed on the equipment or component).
If a DRP was identified on the ground, it was usually found directly beneath the HLRS or near the equipment hatch. The DRP was captured and removed from the area and a gamma scan survey using a Nal detector and slow scan speeds was performed in the immediate surrounding area to bound the area of potential contamination. These surveys verified that the DRP was deposited in the immediate vicinity once it was dislodged from the material. Data in the literature suggest that the deposition rate of sim ilarly sized DRPs is on the order of 1 m/second.
ZionSolutions can confirm by isotopic composition that other particles found on the site were from other events and sources.
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Provide information on why resuspension is not a viable method for particle transport.
ZionSolutions evaluated the airborne transport of DRPs from areas of the site that have been surveyed for DRPs (DRP zone) to areas that have not been surveyed for DRPs. RESRAD OFFSlTE was used, in conjunction with other inputs, to estimate the mass of soil that is transported. Given the mass of soil transported and the number of DRPs per gram of so il, the number of DRP can be estimated. The conclusion of the evaluation is that 2 DRPs could potentially be transported, via the airborne pathway, from the DRPs zone to site areas that have not undergone DRP survey. The calculation methods are described below.
The RESRAD OFF SITE analysis is a semi-quantitative, order of magnitude projection of DRP wind transport. It supports the position that wind transport of large particles ( ~ 100 µm DRP) is unlikely. The model setup is conceptual and not intended to represent actual site configuration. It is conservative in that the clean areas are immediately adjacent to the areas surveyed for DRPs and a sensitivity analysis was used to determine the offsite area size that results in the highest number of DRPs transported, which is 2.
There is no simple way to set up a suspension-deposition model for large particles. ZionSolutions is not aware of references that discuss large particle airborne transport-the literature that is available, which is extensive, pertains to respirable particles, i.e., <IO µm. Nonetheless, we believe this estimation to support the position that wind transport is not a significant means for dispersing DRPs across the site.
The following changes were made to the RESRAD OFFSITE default parameter set to provide a rough estimate of the radionuclide concentration in the areas that have not been DRP surveyed (represented by the "offsite dwelling area"). The site areas are derived from TSD 22-001,
Discrete Radioactive Particle Survey Report, Revision 0. The site areas provided in Revision 1 were checked to confirm the expectation that the maximum number of DRP does not change with the slightly modified areas.
site layout - area of primary contamination is 104,000 m2 which is the total area of the potential DRP zone, shown on the figure below site layout - hypothetical offsite dwelling I located immediately adjacent to the primary contamination with a total area of ~349,000 m2 atmospheric transport - meteorological star file for Chicago O'Hare airport release height- 0.1 m deposition velocity of all particulates - 0.01 mis which is maximum allowed by RESRAD OFFSITE. The maximum is used because the DRP are relatively large (mean of ~ 100 µm diameter). The deposition velocity increases with increasing particle size.
radionuclide -Am-241 was generically applied to provide a conservative estimate of airborne transport due to long half-life distribution coefficient - nominal value of 5,000 cm3/g assigned to Am-241 to ensure that the source term available for transport is not reduced by leaching 1 The nearest temporary offsite dwelling would be a campsite at the Illinois Beach State Park to the south of the site.
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thickness of prima,y contamination - 0.3048 m soil mixing depth in dwelling area - 0.23 m. Sensitivity analysis indicates that a 0.23 m mixing depth maximizes the mass of soil transported to dwelling area. A 0.23 m mixing depth was applied in the Zion DCGL calculations and is considered the maximum value for the order of magnitude transport calculations mass Loading of all particulates - 3.75E-04 g/m3 is nominal estimate of total particulate mass loading in construction zone. (
Reference:
Monitoring Study on Dust Dispersion Properties during Earthwork Construction, School of Civi l Engi neering, Chongqing University, China) soil density - 1.8 g/cm3 (site specific value) x and y dimensions of primary contamination area - 320 m lower and upper values for x coordinates of dwelling area - 0 m and 320 m lower and upper values for y coordinates of dwelling area - 320 m and 1420 m The number of DRP that could be transported via the airborne pathway is estimated using the fo llowing equation:
where:
DRPAT =
CAm,OD mov C DRP,PC DRPAT = number ofDRP transported to adjacent land area via airborne pathway CAm,OD = concentration of Am-241 in offsite dwelling area from RESRAD OFFSITE analysis (pCi/g) moo= mass of soil in offsite dwelling area (g)
CoRP,PC = concentration of DRPs in primary contamination assuming 2071 DRPs present (DRP/g)
CAm,PC = assumed concentration of Am-241 in primary contamination (i.e., 1.0 pCi/g)
The mass of soil transferred to the hypothetical dwelling site (and the corresponding number of DRPs) is dependent on both the mixing depth and dwelling site area. A sensitivity analysis of the two parameters separately concludes that the mixing depth and dwelling site area are both inversely proportional to the radionuclide concentration in the dwelling area. However, using the minimum mixing depth and minimum dwelling area size does not result in the maximum transfer of soil mass (and the corresponding number of DRP). Several combinations of mixing depth and dwelling area sizes were evaluated. As seen in the table, the number of DRPs transferred ranges from 0.22 to 2. 16. The maximum occurs when the mixing depth and dwelling site area is maximized. See table below.
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Order of magnitude projection of the number of DRP transferred via the airborne pathway for a range of mixing depths and dwelling site areas Maximum soil Total soil mass Mixing Dwelling concentration in (g) and inventory Number of depth in (pCi) transported dwelling site area dwelling area due to dwelling area DRPs (m2) to airborne transported area (m) transport (pCi/g) via airborne transport 2.56E-02 2.00E+03 4.80E-02 4.42E+06 1.61E-0l 2.56E-02 1.14E+05 5.60E-03 2.93E+07 1.06E+00 2.56E-02 3.51 E+05 2.30E-03 3.72E+07 1.35E+O0 7.67E-02 2.00E+03 2.20E-02 6.07E+06 2.20E-01 7.67E-02 1.14E+05 2.60E-03 4.08E+07 l.48E+00 7.67E-02 3.51 E+05 l.l0E-03 5.34E+07 1.94E+O0 2.30E-0l 2.00E+03 8.40E-03 6.96E+06 2.52E-0l 2.30E-0I 1.14E+05 l.00E-03 4.70E+07 l.71 E+00 2.30E-0l 3.51 E+05 4.1 0E-04 5.96E+07
- 2. 16E+00 7
Provide estimated radionuclide ratios for a representative particle for each of the three types of particles to inform the risk assessment. Address the extent to which the fission products could separate from the actinides in the irradiated f ue/ particle.
Activated Metal DRPs In its response to the NRC's Request for Additional Information (RAI-10, Specific Consideration 3b), NRC dated March 28, 2022, ZionSolutions provided a radionuclide mixture assumed for dose analysis regarding activated metal (Table 14). This mixture is based on the activity of the highest Co-60 DRP found by ORI SE. The remainder of the radionuclides not contained in the gamma spectroscopy data reported by ORISE were scaled to Co-60 from the activation analysis performed for the reactor internals (highest level of activated metal believed to be contained in the reactors). This mixture is provided in the table below.
Radionuclide Activity Abundance (Ci)
H-3 2.53E+02 0.075%
C-1 4 3.59E+02 0.107%
Mn-54 2.85E+0l 0.008%
Fe-55 7.15E+03
- 2. 125%
Co-60 l.00E+0S 29.721%
Ni-59 1.66E+03 0.493%
Ni-63 2.27E+05 67.468%
Nb-94 5.54E+00 0.002%
Tc-99 l.18E+00 0.000%
As can be seen, Co-60 and N i-63 comprise over 97% of the total activity for these types of particles.
Activated Concrete DRPs Also in its response to RAI-10, Specific Consideration 3b, ZionSolutions provided a radionuclide mixture assumed for dose analysis regarding activated concrete (Table 17). This mixture is derived from scaling the activated concrete nuclides (Eu-152, Eu-154, and Ba-133) to Co-60 and then adding the activated metal nuclides from the prior analysis to account for the activated rebar within the concrete. This accounted for the potential presence of 12 radionuclides.
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Radionuclide Ratio to Abundance Co-60 H-3 2.53E-03 0.016%
C-14 3.59E-03 0.023%
Mn-54 2.85E-04 0.002%
Fe-55 7.14E-02 0.462%
Co-60 1.00E+00 6.470%
Ni-59 l.66E-02 0.107%
Ni-63 2.27£+00 14.688%
Nb-94 5.53E-05 0.000%
- 1. l 7E-05 0.000%
Ba-133 5.07E-02 0.328%
Eu-152 l.15E+01 74.409%
Eu-154 5.40E-0l 3.494%
To provide a comparison of the above mixture to activated concrete characterization data, we have compared this mixture against post-remediation characterization data of activated concrete detected during FSS. 2 This data was collected from the activated concrete region below the reactor vessel from Unit 2 representing 19 samples with reported quantities from 8 radionuclides as shown below.
Radionuclide Average Abundance H-3 43.95%
Co-60 2.69%
Ni-63 5.97%
Sr-90 0.10%
Cs-134 0.00%
Cs-137 0.93%
Eu-152 44.98%
Eu-154 1.39%
For the comparison, we have selected the radionuclides with average abundances that exceed 1 %
(excluding tritium since it was not included in the ORISE analysis), which leaves 4 radionuclides: Co-60, Ni-63, Eu-154, and Eu-152.
We also have re-normalized each data set to the activities for these four radionuclides as summarized in the table below.
2 Zion Station Restoration Project, Final Status Survey Final Report - Phase 2, Appendix 4, FSS Release Record, Survey Units 02100 and 02110 (Unit 2 Containment above 565 foot and Unit 2 Containment Under Vessel Areas).
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Radionuclide Abundance from Abundance from RAI Table 17 Characterization data Co-60 6.5%
4.9%
Ni-63 14.8%
10.8%
Eu-152 75.1%
81.7%
Eu-1 54 3.5%
2.5%
The above Table shows a reasonable comparison between the radionuclide mixtures from our RAJ response and the activated concrete characterization data.
Irradiated Fuel DRPs For irradiated fuel DRPs, the question arises as to why the radionuclide ratios in the particle detected by ORISE differ from what would be expected based upon fuel burnup. ls this lower-than-expected quantity of the fission product Cs-137 due to chemical activity in the environment or to some other process?
To evaluate the radionuclide mixture for irradiated fuel DRPs, ZionSolutions has compared the only particle of this type that has been identified (by ORJSE) to a generic PWR fuel mixture from Table B. l O ofNUREG-7227 3* This table provides for radionuclide activities present in units of Ci/MTU for nine cooling times ranging from 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to 200 years.
For the Zion plant, the earliest start-up (Unit 1) was in December 1973 and the latest cessation of power operation (Unit 1) was in February 1997. During the operational period, the plant experienced some challenging fuel performance issues (fuel failures) that ranged across the operating life of the reactors. During the Dry Cask Storage Campaign, approximately 100 fuel assemblies were classified as failed and of these approximately 55 showed varying degrees of failure ranging from pinholes to severed pins.
Therefore, the production of irradiated fuel DRPs could have been from approximately 24 to 48 years prior to the identification of the particle found by ORISE in April 2021. The closest cooling times to this range in Table B.10 ofNUREG/CR-7227 are 10 and 50 years. However, the data in Table B.10 assume that the fuel was irradiated for a full burn-up period. However, in the case of failed fuel, this irradiation period may not apply since a failure could occur at any time during this period, followed immediately by a particle's escape from the core region.
Despite these potential sources of discrepancies, we have conducted a comparison using the significant radionuclides identified in the ORJSE analysis. The table below shows the activity values from Table B. l O from NUREG/CR-7227 (the "Table B. 10 values") along with the ORJSE fuel particle activity for six of the significant radionuclides.
3 NUREG/CR-7227, US Commercial Spent Nuclear Fuel Assembly Characteristics: 1968-2013, U.S. NRC, September 2016.
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Radionuclide activities (Ci/MTU)
Cooling Time from Am-241 Pu-238 Pu-239 Cs-1 37 Sr-90 Cm-244 Shutdown IO years 2460 4720 397 11 8000 81400 4050 50 years 5300 3440 397 46800 311 00 876 ORI SE Fuel Particle 79900 26188 7450 98900 157043 14800 (pCi)
From the above table the relative activity fractions are calculated and shown below.
Activity Fractions Cooling Time from Am-241 Pu-238 Pu-239 Cs-1 37 Sr-90 Cm-244 Shutdown l O years 1.2%
2.2%
0.2%
55.9%
38.6%
1.9%
50 years 6.0%
3.9%
0.5%
53.2%
35.4%
1.0%
ORISE Fuel Particle 20.8%
6.8%
1.9%
25.7%
40.9%
3.9%
The data in the above table show some differences between the observed activity fractions and the expected fraction for full-burnup fuel for both the 10- and 50-year cooling times. Several factors can lead to the observed difference as discussed below.
Burn-up time. Tf a fuel damage event occurs early in the fuel irradiation history and escapes the core region, then the activity generation would be generally favor the shorter-lived radionuclides. The data suggests a relati vely consistent ratio between the ORISE DRP activity fraction and the 50-year Table B. l O fractions for the actinide radionuclide and this ratio ranges from 1.7 (Pu-238) to 4.29 (Pu-239). In contrast, the Cs-137 ORISE activity ratio is approximately 50% of the 50-year Table B.10 fraction. This trend suggests that such a DRP may have been within the core flux region for many cycles rather than for a short period.
In such a case, the U-235 content of such a particle would be significantly reduced, thereby ceasing the production of Cs-13 7 from the thermal neutron fission of U-235 while the activation of U-238 through fast neutron absorption continued, resulting in the generation of the remaining actinides. This potential hypothesis is not directly supported by the presence of Sr-90 (which also would have been expected to cease generation in the absence ofU-235 fission). However, the potential chemical behavior of these species are not well understood in a complex environment involving high-temperature reactor coolant. For Am-24 I, the principal production is from the ultimate decay of Pu-24 I which was not quantified in the ORISE analysis and could be attributed to its production for a long irradiation interval.
Dissolution in Reactor or SFP Water. The comparison for Cs-1 37 shows a lower-than-expected relative activity for the ORISE analysis. This could also be attributed to a long period of exposure in reactor or spent fuel pool water where some of the Cesium 12
inventory may have dissolved. Despite these small differences, the comparison between these data shows reasonably good agreement.
Environmental Degradation. ZionSolutions has no basis to believe that the reduced level of Cs-13 7 is the resu lt of environmental degradation once the particle was released to the environment. There are no aspects of the environment at the site that would support such a supposition. Even if rainwater is mildly acidic, the exposure of a particle would be limited to a brief, episodic period. This would continue to be true over the 1,000 year compliance period.
In summary, ZionSolutions believes that the most likely source of irradiated fuel particle degradation was in the reactor primary coolant system or the spent fuel pool. The leaching of fission products is more likely during decades of immersion in those mi ldly acidic environments than in the natural environment.
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Provide a narrative and approximate timeline for the Zion demolition.
Containment Structures (2011-2019). In October of 2011, an opening (with doors) was installed in each containment structure such that the lowest point of the openings were level with the Charging Floor inside each containment. Ventilation systems with HEPA filters were installed to keep air pressure negative inside each containment.
The picture above shows the Unit 1 Containment opening with the doors shut following Steam Generator removals using the Heavy Lift Rail System (HLRS). The Steam Generators were cut with diamond wire saws at the transition piece such that they could fit on railcars. They were then rinsed and "locked down" with blue fixative prior to leaving containment.
Once large components were removed from containment, the HLRS was removed and a "reach stacker" was used to place intermodals for direct loading on the charging floor.
Reactor Vessel Internals were removed using a mechanical cutting process. Numerous liners of Class A, B, C, and Greater than Class C (GTCC) waste were generated during cutting operations.
Any chips that could not be collected were washed down into the Transfer Canal where they were grouted in place. When the interior of containment was being prepped for open-air demo, the grouted section of the transfer canal was removed as a monolith and placed in a railcar for disposal.
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Prior to the start of interior concrete demolition, tents were constructed and attached to each containment (2017). The tents had ventilation and HEPA systems as well as rail access such that contaminated materials from containment could be loaded under cover. The picture below is looking up from a lower level in containment where the inside of the waste processing tent is visible. Activated concrete from under the reactor vessel has not been removed at this point because more demolition was sti ll required to reach the sumps where the activated concrete was located.
Once all material was removed from containment, the liner and floor were decontaminated from 3' below grade down to the lowest level of containment. ISOCs were used as part of the FSS process to verify rad levels prior to lockdown and the start of open-air demo. Clean fill was placed in containment such that the level was brought up to about 4' below grade. A geotextile barrier was placed on top of the fill and then gravel was placed to bring the level up to 3' below grade.
Ohce conditions were established for open-air demolition, the waste processing tents (including the asphalt floors) were removed. The containment structure was dropped in 4' sections. The excavator and hammer attachment worked around the outside of containment, cutting wedges all the way around containment such that it settled by 4' after the last wedge is hammered. An excavator with a shear remained in containment (not occupied when wedges were being created) to "peel" the containment liner after each drop.
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Each containment exterior underwent Uncond itional Release Surveys prior to demolition. A few areas with elevated readings were identified. These areas were remediated and resurveyed to ensure no contamination remained.
Once containment demolition was complete, the sacrificial soil layer was removed and disposed of as rad waste.
Auxiliary Building (2013-2017). Prior to open-air demo, surgical removal of systems, structures, and components took place with radioactive materials being loaded into super sacks or directly loaded in to both high-and low-sided gondola cars. The picture below shows a grey supersack being loaded in a high sided gondola. The hard cover for the rail car is on the ground next to the rail car. The approach was to remove the interior of the Aux Building such that only a "bathtub" existed when open-air demo was complete and FSS was performed on the basement prior to backfill.
Spent Fuel Pool (2015-2017). Once the dry fuel storage pool-to-pad campaign was complete, the Spent Fuel Pool (SFP) was cleaned and remaining GTCC inventory was placed in HICs for later transfer to GTCC I iners generated during the Reactor Vessel Internals segmentation project.
Water level was lowered, and the racks were lifted and hydrolased above the pool before they were size-reduced using a diamond wire saw. Rack pieces were placed in bags then loaded into 16
high-sided gondolas. The SFP was power-washed and locked down with lag coat. Only a couple of areas had equipment that had to be removed before the start of open-air demo.
Starting from the switchyard side of the SFP, it was demolished up to the eastern wall (abuts the Auxiliary Building).
Final Grading of the Power Block Area (2019). Final grad ing of the power block area started in June of 2019 and was completed by August of 2019. An agronomist developed the spec for soils that would support natural plant growth.
The picture above shows the placement of soils that were seeded with fescue and durable grass seed mixtures. Two CCDD piles can be seen in the picture. The top right comer shows the pile west of the rail spur in the old employee parking lot. The other pile is center near the top just east of the rail and switchyard.
To be clear, the entire site did not have soil added to it. The picture above is about 60%
complete. The final area with new soi l is actually a square that covers the power block including the footprints of Containment, the Auxiliary Building, Turbine Building, and the SFP.
Final Site Grading (2020). Final site grading and scarification of the rest of the site commenced on August 31, 2020, and was completed on September 23, 2020. A detailed timeline of final site grading and supporting maps are included in the enclosure to the March 2022 RAI response
("Final Site Grading and Seeding Timeline with Maps").
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