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 DRP s.

Upon further evaluation, ZionSolutions does not find the previous re s ponse provided in February 2021 regardin g w ind entrainment, or tran s po rt, to be s ufficientl y clear. Neither that response, nor the previou s RAI response, ha s 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 tran sport 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.

1 Prot o cols were put in plac e for safety and radi o logical co ntrol purp ose s. In accordance with OSHA sta nd ards, if wi nd speeds exceeded 23 miles per hour (mph) the movement of material was suspen d ed. At Zion, the Waste Manager designated adm ini strative limit s ( 15-18 mph) when lifting material from the HLRS was su spended. If the s e wind speeds were encountered, the load wa s taken back into the Containment Building. Additional ly, if medium to hard rain was enc o untered, th e load wa s taken back int o the Containm e nt Buildin g.

Regardless of the safety and radio logical c o ntrol protoc o ls, it was still possible for a DRP to be di s lo cated from the material during the rem o va l proce ss. The DRP s co uld hav e been di s located due to rain, wind, or per s onnel interacti o n ( e.g., during rigging it is pos s ible for the rigging cables and strap s to have rubbed on the equipment or component).

If a DRP was identified on the ground, it was usuall y found directly beneat h t h e HLRS or near the equipment hatch. The DRP was captured a nd removed from the area and a gamma scan survey using a Nal detector and s low scan speeds was performed in the immediate surro undin 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 lit erature suggest that the deposition rate of sim il arly s ized DRPs is on the order of 1 m /second.

ZionSolution s can confirm by isotopic compositio n t h at other particle s found on the site were from other event s a nd sources.

2 3

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 DRP s (DRP zo ne) to areas that have not been surveyed for DRPs. RESRAD OFFSlT E was used, in conjunction with other inputs, to estimate the mass of soil that is tran s ported. Giv e n the ma ss o f so il tran s po rted and th e number of DRP s per gram of so il, the number of DRP can be est im ated. The conclu s ion of the eva lu ation is that 2 DRP s could potentially be transported, via the airborne pathway, from the DRP s zo ne to s ite areas that ha ve not undergone DRP s urv ey. The ca lc ul ation method s are de sc ribed bel ow.

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 unlikel y. The model se tup is conceptual and not intend ed to represent actual s ite configuration. It is conservative in that the clean areas are immediately adjacent to the areas surveyed for DRP s and a se n s itivit y analysis was u sed to determine the offsite area size that result s in the highe st number of DRP s transported, which is 2.

There is no simple way to se t up a suspension-deposition mode l for large particle s. ZionSolution s is not aware of references that discuss large partic le airborne transport-the literature that is available, which is extensive, pertains to respirab le particles, i.e., < IO µm. Nonetheless, we belie ve 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 "offs ite dwelling area " ). The site areas are deri ved from TSD 22-001,

Discrete Radioa ctive Particl e Survey Report, Revision 0. The site areas provided in Rev is ion 1 were checked to confirm the expectation that the maximum number of DRP does not change with the s l ightl y 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

• atmosph eric transport - meteorolo gica l star file for Chicago O ' Hare airport

• r e leas e height- 0.1 m

• deposition velocity of all particulates - 0.01 m i s 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 l ied to provide a conservative estimate of airborne transport due to long ha lf-life

• distribution coeffic ient - 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 Th e nearest temporary offsite dwelling would be a campsit e at th e Illinois Beach S tat e Park to th e south of the si te.

5

• thicknes s of prima, y contamination - 0.3048 m

• soi l mixing depth in dwelling a r ea - 0.23 m. Se n s iti v ity a na lys i s indi cates t h at a 0.23 m mi x in g dept h m ax imi zes th e ma ss of so il t ran sporte d to d we llin g a rea. A 0.23 m mi x in g d epth was appli e d in t he Z io n D CGL ca lcul at io n s and is co n s idere d th e max imum va lu e for t h e o rder of m agnitu d e t ra ns p ort ca lc ul at io n s

• ma ss Lo a din g of all p ar tic ulat es - 3. 75E-04 g /m 3 is no m inal es timate o f to ta l part ic ul ate m ass loadi n g in co n struc ti o n zo n e. (Refere n ce: Mon itor ing Study o n Dust Dispersion Prop e rti es durin g Earthwork Co ns tru ction, Sc hoo l o f C ivi l E ngi neeri ng, C h o ng qin g U ni ver s it y, C hin a)

• s oil de nsity - 1.8 g /c m 3 (s ite s p ec ific va lu e)

• x and y dimensions of primary contam in ation area - 32 0 m

• lower and uppe r va lues for x coord inates of dwell ing ar ea - 0 m a nd 320 m

• lower and uppe r va lues for y coordinates of dwelling ar ea - 320 m and 1420 m

T h e numb e r of DRP that co uld b e tran sp orte d v ia th e a irb o rn e p a th way is es tim ate d u s in g th e fo ll ow in g e qu a ti o n :

CAm,OD mo v C DRP,P C DRPAT =

w he re:

DRP AT = num be r o fDRP tr a n sp o rt e d to a djace nt la nd a rea v ia a ir bo rn e path way C Am,OD = co n ce ntra ti o n of A m -2 41 in offs ite d we llin g a rea fro m RES RAD OFFSITE a n a lys is (p C i/g) m o o= m ass of so il in offs ite d we llin g area (g)

CoRP,PC = co n ce ntrati o n of DRP s in prim a ry co nt a min a ti o n ass umin g 2 071 DRP s prese nt ( DRP /g)

C Am,PC = ass um e d c o nc e ntrat io n of A m-24 1 in p r im ary co nta min at io n ( i.e., 1.0 pCi/g)

T h e mass of so il tra nsfe rr e d t o th e hy p o th eti ca l dw e llin g s ite (a nd th e co rr esp o ndin g numb e r of DRP s) i s d e p e n de n t o n bo th t he mi x in g d ep th and d we ll in g s ite ar ea. A se n si ti v ity a n a lys i s of th e two para m ete r s se p ara te ly co nc lud es th a t th e mi x in g d e p th and d we llin g sit e a r ea are bot h in verse ly p ro p ort io n a l t o t h e radi o nu c lid e co n ce ntr a ti o n in th e d we lli ng a rea. H oweve r, u s in g th e minimum mi x in g d epth and minimum d we llin g a rea s ize d oes no t res ult in th e m ax imum tran sfe r of so il m ass (an d th e co rr es p o ndin g numb er of DRP ). Severa l c o mbin at io n s of mi x in g d e pth and d we llin g a rea sizes we re eva lu ate d. As see n in th e t a bl e, t h e numb er of DRP s tra n sferre d ran ges fro m 0.22 to 2. 16. T h e m ax imum oc cur s w h e n th e mi x in g d ep t h a nd dwe llin g s ite area is m ax imi ze d. See ta bl e b e low.

6 Order of magnitude projection of the number of DRP transferred via the airborne pathway for a range of mixing depths and dwelling site areas

Mixing Dwelling concentration in (g) and inventory Number of Maximum soil Total soil mass 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. 00 E+ 0 3 4.80 E -0 2 4.4 2E+06 1. 6 1E -0l 2.56 E -0 2 1.14 E+ 0 5 5.60 E -0 3 2.9 3E+ 07 1.0 6E+ 00 2.56 E -0 2 3.51 E+ 05 2.3 0E -0 3 3. 72E+ 07 1.3 5E+ O0 7.67 E -0 2 2.00 E+ 0 3 2.2 0E -0 2 6.07 E+ 0 6 2.2 0E -01 7.67 E -0 2 1.14 E+ 05 2.60 E -0 3 4.08 E+ 07 l.48 E+ 00 7.67 E -0 2 3.51 E+ 05 l. l0 E -0 3 5.34E + 07 1.94E+ O0 2.3 0E -0l 2.00E+ 0 3 8.40 E -0 3 6. 96E + 0 6 2.52E -0l 2.3 0E -0I 1.14 E+ 0 5 l.00 E -0 3 4.70 E+ 0 7 l.71 E+ 00 2.3 0E -0l 3. 51 E+ 05 4. 1 0E -04 5.96E + 07 2. 16E+ 00

7

Pro vide estimated radionuclide ratios for a representative particl e for e a c h of th e three type s 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 re sponse to the NRC's Request for Additio n a l In format ion (RAI-10, Spec ific Cons id era ti on 3b), NRC d ated Ma rch 28, 2022, ZionSolutions provided a radionuclide mi x tur e assumed for dose ana lys is regarding activated m e tal (Tab le 14). This mi xt ur e is base d on the activity of the hi ghest Co-60 DRP fo un d by ORI SE. The remainder of the radionuclides n ot contai ned in the gam m a spectrosco p y d ata reported by O RISE we re sca led to Co -60 from th e act i vat io n analysis performed for the reactor inte rnal s (hig h es t leve l of act ivat e d m eta l believ ed to be contained in the reactors). T hi s mi x tur e is pro v id ed 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 -5 5 7.15E + 03 2. 12 5%

Co-60 l.00 E+ 0S 29.721%

Ni-59 1.66E+03 0.493%

Ni-63 2.2 7E+05 67.468 %

Nb-94 5.54E+ 00 0.002%

Tc-99 l.18E+00 0.000%

As can be see n, Co -60 and N i-6 3 comprise over 97% of the total activity for th ese ty pe s of particle s.

Activated Concrete DRPs A lso in it s r es p o nse to RAI-10, Specific Co n s ideration 3b, ZionSolutions pro v ided a radionuclid e mi xture assumed fo r dose analysis re gardin g activated concrete (Table 17). This mixture is deri ve d from sca lin g the activated concrete nuclide s (Eu-152, Eu-154, and Ba-13 3) to Co-60 and th e n adding th e act iva te d m etal nuclides from the pri or a n a lys i s to account for the activated rebar within the co ncr ete. This acco unt e d for the potential presence of 12 radionuclides.

9 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%

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 hav e compared this mixture against po st-remed iation characterization data of activated concrete detected during FSS. 2 This data was co ll ected from the activated concrete region below the reactor vessel from Un it 2 representing 19 sa mples with reported quantities from 8 radionuc lid es 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 se lected the rad ionuc lid es with average abundances that exceed 1 %

(excludi n g 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 li zed each data set t o the act ivities for these four radionuclides as summarized in the table below.

2 Zion Station Restoration Projec t, Final Status Survey Final Report - Phase 2, Appendix 4, FSS Release Record, Survey Units 02100 and 02110 (Unit 2 Con tainm ent above 565 foot and Unit 2 Containment Under Vessel Areas).

10 Radionuclid e Abund an ce 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 %

E u-1 54 3.5% 2.5 %

The above Table shows a rea so nable comparison b etwee n the radionuclide mixtur es from our RAJ response a nd the activated co ncr e te c ha racterization data.

Irradiated Fuel DRPs For irradiated fu e l 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. l s this lower than-expected qu a ntity of the fis s ion product Cs-137 due to chemical activity in the environment or to some other process ?

To eva luate th e radionuclide mixture for irradiated fuel DRPs, ZionSolutions has compared the only particle of this type that ha s been identified (by ORJSE) to a generic PWR fuel mixture from Table B. l O ofNUREG -7227 3

• This table provide s for radionuclide activities present in unit s of Ci/MTU for nine cooling time s 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 (U nit 1) was in February 1997. During the operational period, the plant experienced some challenging fuel performance issue s (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 degree s of failure ranging from pinholes to severed pin s.

Therefore, the production of irradiated fuel DRP s could have been fr o m approximately 24 to 48 yea rs prior to the identification of the particle found by ORISE in Ap ril 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 ma y not apply s ince a failure could occur at any time during thi s period, followed immediatel y by a particle 's escape fr o m the core region.

De spite the se p ote ntial sources of discrepancies, we have conducted a comparison using the s ignificant 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 ORJS E fuel particle activity for six of the significant radionuclid es.

3 NUREG /CR-7227, US Commercial Spent Nuclear Fuel Assembly Chara cteristics: 1968-2013, U.S. NRC,

Sep tem b er 2016.

11 Ra d io nu c li d e ac tivitie s (Ci/MTU)

Coo li ng T im e fro m Am - 24 1 P u-23 8 Pu-239 Cs -1 37 Sr-90 Cm -244 S hu tdow n IO years 2460 4720 397 11 8000 814 0 0 4050 50 y ear s 5300 3440 397 46800 3 11 00 876 O RI S E Fu e l Pa rt ic le 79900 26188 7450 98900 157043 14800 (pCi)

F rom the a b ove table th e re lative act ivit y fract io n s are calc ul ate d and show n be low.

Activity Fract ion s Coo lin g Ti m e fr o m Am-2 41 Pu -238 Pu-239 C s -1 37 Sr-90 C m-244 Shutdow n l O y ears 1. 2% 2.2 % 0.2 % 55.9 % 3 8.6 % 1.9 %

50 y ears 6. 0 % 3.9% 0.5 % 53.2 % 35.4 % 1.0 %

ORI SE Fue l P ar ti c le 2 0.8% 6.8% 1.9 % 25.7 % 4 0.9% 3.9%

T he d ata in th e a bove ta bl e sh ow so m e di ffere nc es betwee n th e o bserve d ac ti v ity frac ti o n s a nd th e exp ec t e d fract io n fo r full -b urnup fu e l fo r bo th th e 10- and 5 0- yea r coo lin g tim es. Seve ra l factors can lea d to th e o b se r ve d di ffe re n ce as di sc usse d be low.

• Burn-up time. Tf a fu e l da m age eve nt occ ur s ear ly in th e fu e l ir ra di ati o n hi sto r y a nd esca p es th e co r e r eg io n, t h e n t he ac ti v ity ge n erat io n wo uld be ge n e ra ll y favo r th e s h orter li ve d ra di o nu c lid es. T he d ata s uggests a re la ti ve ly co n s istent ra ti o b etween t he ORI SE DRP activity fract io n a nd th e 50- year Ta ble B. l O frac ti o n s for th e a c tin id e radi o nu c lid e a nd thi s ra ti o ran ges fro m 1.7 (Pu -23 8) to 4.29 (P u-23 9). In co ntra st, th e Cs - 13 7 ORI SE ac ti v ity rat io is app rox im a te ly 50 % of th e 5 0- yea r Ta bl e B. 10 frac ti o n. Thi s tr e nd su ggests th at s u ch a DRP m ay h ave bee n w ithin t he co re flu x reg io n for ma ny cyc les ra th er th a n for a sh ort p e ri o d.

In s u ch a case, th e U - 235 cont e nt of s uc h a p arti c le would be s ig nificantl y re duc ed,

th e re by cea s in g th e p ro du cti o n of Cs - 13 7 fr o m t he t h e rm a l ne utro n fi ss io n of U - 235 w hil e t he ac ti va ti o n of U - 23 8 th ro u g h fas t ne utro n a bso rp t io n co n t inu e d, resu ltin g in t h e ge n era ti on of t h e re m a inin g ac tinid es. T hi s p ot e n tia l h y p o th es is is n ot dir ect ly s upp o r ted by th e p rese n ce of S r-90 (w hi c h a lso wo uld have bee n ex p ecte d to cease ge n era ti o n in th e a bse nce ofU -235 fiss io n). H oweve r, th e p o te n t ia l c h emic a l be h av io r of th ese sp ec ies are n ot we ll un d erstoo d in a co mpl ex e nv iro nm e n t invo lv in g hi gh -te mp era tu re reac t o r coo la n t. For Am-24 I, t h e prin c ip a l pro du ct io n is fro m th e ul t imate d ecay of Pu -2 4 I w hi c h was n ot qu a nti fie d in th e O RI SE an a lys is a nd co ul d be attr ibut e d to its p ro du ct io n fo r a lo n g irra di a ti o n interva l.

• Dissolution in Reactor or SFP Water. T h e co mp a ri so n for Cs -1 3 7 s h ow s a lowe r-t h a n exp ecte d re la ti ve ac ti vity fo r th e O RI SE a n a lys is. T hi s co ul d a lso be att ribu te d t o a lo n g p er io d of expos ur e in reac t o r o r sp e nt fu e l p oo l wat e r w h e re so m e o f th e Ces ium

12 in ve ntory may ha ve dissolved. D espite these sma ll differe n ces, the comparison between these data shows reasonably good agreeme nt.

• Environmental Degradation. Z ionSol utions h as no bas is to be li eve that the reduced leve l of Cs-13 7 is the resu lt of enviro nm e ntal degradation once the particle was released to the environ m e nt. There are no aspects of the env ironm e nt at the s ite that wo uld s upp ort s uc h a s upp os iti o n. Even if rainwater is mildl y acidic, th e exposure of a partic le would be limi ted to a brief, episodic period. This wou ld co ntinue t o be true over th e 1, 000 yea r comp li ance period.

In s ummar y, Z ionSolu tion s be li eves that the most lik ely so urce of irradiated fuel partic le de gra dation was in the reactor pri m ary coolant system or t he spent fue l pool. The leaching of fission product s is m ore li kely durin g dec a de s of imm e rs io n in tho se mi ldl y acidic env iro nm ents than in the natural env iro nm ent.

13 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 st ructure s uch that the lowe st point of the openings were level with the Charging Floor inside each containment. Ventilation sys tems with HEPA filter s were in sta lled to keep air pressure negat ive in s id e each co nta inm e nt.

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 railcar s. 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 u sed 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 the y were grouted in place. When the interior of containment was being prepped for open-air demo,

the grouted sect ion of the transfer canal was removed as a monolith and placed in a railcar for disposal.

14 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 proce ss ing 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 co ncrete was locat ed.

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 sett led 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.

15 Each conta inm ent exter ior und e rwent Uncond iti ona l Release S urvey s prior to demolition. A few areas w ith e levated readin gs were id e ntifi ed. T he se areas we re remediated a nd resurveyed to e n s ure n o co nt am in a tion re m a in e d.

Once containment d e moliti o n was complete, the sac rifi c ia l so il layer was remov ed an d di sp osed of a s rad waste.

Auxiliary Building (2013-2017). Prior t o ope n-a ir demo, su r gica l remova l of systems,

str uctures, a nd co mp o ne nt s took place w ith radioactive materials be in g load e d int o s up er sacks or direct ly loaded in to b ot h high-a nd low-s id ed go nd ola cars. The picture below shows a grey

s up ersac k b e in g loa de d in a hi g h s ided gon d o la. The h ar d cover for the rail car is o n the ground next to th e r ai l car. The app roac h was to remove th e int e ri or of th e A ux Buildin g s uch that only a

" bathtub" ex isted w hen o p en-air d e m o was complete and FSS was performed on the base m e nt prior to backfill.

Spent Fuel Pool (2015-2017). Once th e dry fu e l s tora ge pool-to-pad ca mpai g n was co mplet e,

the Spe nt F ue l Pool (SF P ) was c lean e d a nd remaining GTCC inv entory was p laced in HIC s fo r later transfer to GTCC I in ers ge nera t e d during the R eactor Vessel Int erna ls seg m e ntat io n project.

Water leve l was lowe red, a nd the racks were lifted a nd hydrolased above t h e pool before they were size-r e du ce d u si n g a di amo nd wire saw. Rack pi eces we re placed in bags then loa ded into

16 high-sided go nd olas. The SFP was power-washed a nd locked down with la g coat. Only a couple of areas had equipment that had to be removed before the start of open-air demo.

Starting from the sw itch yar d s ide 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 in g of the power block area started in June of 2019 and was completed by August of 2019. An agronomist dev e lop ed the spec for soils that wou ld s upport natural plant growth.

The picture above s hows the placement of so ils 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 pil e 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 Conta inment, the A uxiliary Building, Turbine Buildin g, and the SFP.

Final Site Grading (2020). Final site grading and scarification of the rest of the site commenced on A ugu st 31, 2020, and was completed on September 23, 2020. A detailed timeline of final site grading and s upp ort ing map s are included in the enclosure to the March 2022 RAI response

("Final Site Grading and Seeding Timeline w ith Maps ").

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