ML22293A432

Response to Audit Questions

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 Attachment 2 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 regardin g wind entrainment, or tran sport, 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 Bui lding 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 cm 2 . 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 enab le 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 radi o logical control purposes. In accordance with OSHA sta nd ards, if wind speeds exceeded 23 miles per hour (mph) the movement of material was suspended . At Zion, the Waste Manager designated adm ini strative 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, th e load was taken back into the Containment Building.

Regardless of the safety and radio logical control protoco ls, it was still possible for a DRP to be dislocated from the material during the remova l process. The DRPs co uld have been di slocated 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 surroundin g area to bound the area of potential contamination. T hese 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 1m/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 tran sported . Given the mass o f so il transported and th e number of DRPs per gram of so il , the number of DRP can be estimated. The conclusion of the eva luation 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 ca lcul ation 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 partic le 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 Rev ision 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 m 2 which is the total area of the potential DRP zone, shown on the figure be low

• site layout - hypothetical offsite dwelling I located immediately adjacent to the primary contamination with a total area of ~349,000 m 2

• 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 ve locity increases with increasing particle size.

• radionuclide -Am-241 was generically app lied to provide a conservative estimate of airborne transport due to long ha lf-life

• distribution coefficient - nominal value of 5,000 cm3/g assigned to Am-241 to ensure that the so urce 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 ofprima, y contamination - 0.3048 m

• soil mixing depth in dwelling area - 0.23 m. Sensiti vity ana lysis indicates that a 0.23 m mix ing depth max imizes the mass of so il transported to dwe lling area. A 0.23 m mi x ing depth was appli ed in the Z ion DCGL calculations and is co nsidered the max imum valu e for the order of magnitu de tra nsport ca lcul at ions

• mass Loading of all p articulates - 3.75E-04 g/m 3 is nominal estimate of tota l partic ul ate mass loadi ng in co nstructi on zone. (

Reference:

Mon itoring Study on Dust Dispersion Properties during Earthwork Construction , Schoo l of C ivi l Engi neeri ng, Chongqin g Uni versity, China)

• soil density - 1.8 g/cm 3 (s ite specific value)

• x and y dimensions ofprimary 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 th e a irborn e pathway is estim ated using th e fo ll owing equ ati on:

CAm,OD mov C DRP,PC DRPAT =

where:

DRPAT = num ber ofDRP transported to adjacent land area v ia a irbo rne pathway CAm,OD = concentrati on of Am-241 in offs ite dwelling area fro m RESRAD OFFSITE ana lys is (pC i/g) moo= mass of so il in offsite dwe lling area (g)

CoRP,PC = concentration of DRPs in primary contamination ass uming 2071 DRPs present (DRP/g)

CAm,PC = assum ed concentration of Am-24 1 in primary contamination (i.e., 1.0 pCi/g)

The mass of so il transferred to th e hypoth eti cal dwelling site (and th e corresponding number of DRPs) is dependent on both the mixing depth and dwe ll ing site area. A sensi ti vity analysis of the two param eters separately co nc ludes that the mixing depth and dwe lling site area are both inversely proportiona l to the radi onuclide concentrati on in the dwe lli ng area . However, using the minimum mi xing depth and minimum dwe lling area size does not result in the max imum transfer of so il mass (and the correspondin g number of DRP). Several combinations of mix ing depth and dwe lling area sizes were eva luated. As seen in th e tabl e, the number of DRPs transferred ranges from 0.22 to 2. 16. The maximum occurs when the mix ing depth and dwelling site area is max imized. See tabl e 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 Total soil mass Maximum soil Mixing (g) and inventory Dwelling concentration in Number of depth in (pCi) transported site area dwelling area due DRPs dwelling to dwelling area (m2) to airborne transported area (m) via airborne transport (pCi/g) transport 2.56E-02 2.00E+03 4.80E-02 4.42E+06 1.6 1E-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 1.3 5E+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 2.3 0E-0l 2.00E+03 8.40E-03 6. 96E+06 2.52E-0l 2.3 0E-0I 1.14E+05 l.00E-03 4.70E+07 l.71 E+00 2.3 0E-0l 3.51 E+05 4.10E-04 5.96E+07 2. 16E+00 7

Pro vide estimated radionuclide ratios for a representative particle for each of th e 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 Cons ideration 3b), NRC dated March 28, 2022, ZionSolutions provided a radionuclide mi xture assumed for dose ana lys is regarding activated metal (Table 14). This mi xture is based on the activity of the highest Co-60 DRP fo und by ORI SE. The remainder of the radionuclides not contai ned in the gamma spectroscopy data reported by ORISE were sca led to Co-60 from the activation analysis performed for the reactor internal s (highest level of activated meta l believed to be contained in the reactors). Thi s mixture is provided in the table below.

Activity Radionuclide 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 A lso in its response to RAI-10, Specific Consideration 3b, ZionSolutions provided a radionuclide mixture assumed fo r dose analysis regardin g activated concrete (Table 17). This mixture is derived from scaling the activated concrete nuclides (Eu-152, Eu-154, and Ba-13 3) to Co-60 and th en adding th e activated metal nuclides from the prior analysis to account for the activated rebar within the concrete. This acco unted for the potential presence of 12 radionuclides.

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Ratio to Radionuclide 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%

Tc-99 1. l 7E-05 0.000%

Ba-133 5.07E-02 0.328%

Eu-152 l.15E+0 1 74.409%

Eu-154 5.40E-0l 3.494%

To provide a compariso n of the above mixture to activated concrete characterization data, we have compared this mixture against post-remed iation characterization data of activated concrete detected during FSS. 2 This data was coll ected from the activated concrete region below the reactor vessel from Unit 2 representing 19 sa mples with reported quantities from 8 radionuc lides as shown below.

Average Radionuclide 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 se lected the rad ionuc lides with average abundances that exceed 1%

(excludi ng tritium since it was not included in the ORISE ana lysis), which leaves 4 radionuclides: Co-60, Ni-63 , Eu-154, and Eu-152.

We also have re-norma lized 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 Con tainment above 565 foot and Unit 2 Containment Under Vessel Areas).

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Abundance from Abundance from Radionuclid e RAI Table 17 Characterization data Co-60 6.5% 4.9%

Ni-63 14.8% 10.8%

Eu-152 75 .1% 81 .7%

E u-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 fu el 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 qu antity of the fis sion 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 (fue l 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 Spen t Nuclear Fuel Assembly Characteristics: 1968-2013, U.S. NRC, September 2016.

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Radionucli de activities (Ci/MTU)

Coo li ng Time fro m Am-24 1 Pu-23 8 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 Part icle 79900 26188 7450 98900 157043 14800 (pCi)

From the above table the re lative activity fractions are calcul ated and shown be low.

Activity Fractions Coo ling 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%

ORI SE Fuel Parti cle 20.8% 6.8% 1.9% 25 .7% 40.9% 3.9%

The data in the above table show some di ffere nces between the observed activ ity fracti ons and the expected fract ion fo r full -burnup fu el fo r both th e 10- and 50-yea r coo ling times . Several factors can lead to the observed difference as discussed below.

• Burn-up time. Tf a fu el da mage event occurs early in th e fu el irradi ati on hi story and escapes the core reg ion, then the acti vity generation wo uld be generally favo r the shorter-lived radi onuclides. The data suggests a relati ve ly consistent rati o between the ORISE DRP activity fraction and the 50-year Table B . l O fracti ons for the actin ide radi onu clide and thi s rati o ranges fro m 1.7 (Pu-23 8) to 4.29 (Pu-23 9). In contrast, th e Cs-137 ORISE activity ratio is approx imately 50% of the 50-year Ta bl e B .10 fraction. Thi s trend suggests that such a DRP may have been w ithin the core flu x region for many cycles rath er th an for a short peri od.

In such a case, the U-235 content of such a parti cle would be significantly reduced, th ereby ceasing the production of Cs-13 7 from the thermal ne utron fi ss ion of U-235 w hile the acti vati on of U-23 8 th rough fast neutron absorption continued, resu lting in the generati on of the remaining actinides . Thi s potentia l hypothes is is not directly supported by the presence of Sr-90 (which also would have been expected to cease generati on in the absence ofU-235 fiss ion). However, the potential chemical behavior of these species are not we ll un derstood in a complex environment invo lving hi gh-temperatu re reactor coo lant. For Am-24 I , the principal produ ction is from the ultimate decay of Pu-24 I whi ch was not quantified in th e ORISE analys is and co ul d be attributed to its produ ction fo r a long irradiati on interval.

• Dissolution in Reactor or SFP Water. The compari son for Cs-1 37 shows a lower-than-expected relative activity for the ORISE analys is. This co ul d also be attribu ted to a long period of exposure in reactor or spent fu el poo l water where some of th e Ces ium 12

inventory may have dissolved. Despite these sma ll differe nces, the comparison between these data shows reasonably good agreement.

• Environmental Degradation. Z ionSolutions has no bas is to beli eve that the reduced leve l of Cs-13 7 is the resu lt of environmental degradation once the particle was released to the environ ment. There are no aspects of the environment at the site that wo uld support such a supposition. Even if rainwater is mildl y acidic, th e exposure of a partic le would be limited to a brief, episodic period. This wou ld co ntinue to be true over th e 1,000 yea r comp liance period .

In summary, Z ionSolutions believes that the most likely source of irradiated fuel particle degradation was in the reactor pri mary coolant system or the spent fue l pool. The leaching of fission products is more li kely durin g decades of immersion in those mi ldly acidic environm ents than in the natural env iro nment.

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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 insta lled 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 "pee l" the containment liner after each drop.

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Each conta inment exterior underwent Uncond iti ona l Release Surveys prior to demolition . A few areas with e levated readings were identified . These areas were remediated and resurveyed to ensure no contam ination rema ined.

Once containment demolition was complete, the sacrifi c ial soil layer was removed and disposed of as rad waste .

Auxiliary Building (2013-2017). Prior to open-air demo, surgica l remova l of systems, structures, and components took place with radioactive materials being loaded into super sacks or directly loaded in to both high- and low-s ided go ndola cars . The picture below shows a grey supersack being loaded in a hi gh sided gondola. The hard cover for the rail car is on the ground next to the rai l car. The approach was to remove the interi or 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 th e dry fu el storage pool-to-pad campaign was complete, the Spent Fue l Pool (SFP) was c leaned and remaining GTCC inventory was p laced in HICs fo r later transfer to GTCC Iiners generated during the Reactor Vessel Internals segm entation project.

Water leve l was lowered, and the racks were lifted and hydrolased above the pool before they were size-reduced usi ng a di amo nd wire saw. Rack pieces were placed in bags then loaded into 16

high-sided gond olas. 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 dev eloped the spec for soils that wou ld 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 rai l 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 w ith Maps").

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