ML040900350

From kanterella
Jump to navigation Jump to search
to Saxton Nuclear Experimental Corporation Facility License Termination Plan.
ML040900350
Person / Time
Site: Saxton File:GPU Nuclear icon.png
Issue date: 02/29/2004
From:
GPU Nuclear
To:
Office of Nuclear Reactor Regulation
References
Download: ML040900350 (29)
v • d • e


Text

Saxton Nuclear Experimental Corporation Facility License Termination Plan Rev. 3

\\ !\-\ t ) j '*. ' ' &'E-' ' -

February 2004 Prepared by GPU Nuclear, Inc.

List of Effective Pages Page No. Rev. No. Page No. Rev No.

TOC I ii 1 iii I iv 2 v 2 vi 1 vii 2 CHAPTERS 1-1 I 1-2 2 1-3 I 1-4 I 2-1 2-2 I 2-3 2-4 I 2-5 2-6 I 2-7 2-8 I 2-9 2-10 I 2-11 2-12 I 2-13 2-14 I 2-15 2-16 2 2-17 2-18 2 2-19 2-20 2 2-21 2-22 I 2-23 2-24 I 2-25 2-26 2 2-26a 2-27 2-28 2-29 2-30 2-31 2-32 2-33 2-34 2-35 2-36 2-37 2-38 2-39 2-40 2-41 2-42 243 2-44 2-45 2-46 2-47 2-48 2-49 2-50 2-51 2-52 2-53 2-54 2-55 2-56 2-57 2-58 2-59 2-60 2-61 2-62 2-63 2-64 2-65 2-66 2-67 2-68 2-69 2-70 2-71 2-72 2-73 2-74 2-75 2-76 2-77 2-78 l

List Of Effective Pages Page No. Rev. No. Page No. Rev. No.

2-79 1 2-80 I 2-81 2 2-82 I 2-83 1 2-84 1 2-85 1 2-86 2 2-87 2 2-88 1 2-89 1 2-90 1 2-91 1 2-92 1 2-93 1 2-94 1 2-95 3 2-96 1 2-97 1 2-98 1 2-99 I 2-100 1 2-101 I 2-102 1 2-103 I 2-104 1 2-105 I 2-106 1 2-107 I 2-108 1 2-109 1 2-110 1 2-111 1 2-112 I 2-113 1 2-114 I 2-115 1 2-116 I 2-117 1 2-118 I 2-119a-119i 2 2-119 1 2-120 (Fig. 2-1) I 2-121 (Fig. 2-2) 2 2-122 (Fig. 2-3) I 2-123 (Fig. 2-4) I 2-124 (Fig. 2-5) I 2-125 (Fig. 2-6) I 2-126 (Fig. 2-7) I 2-127 (Fig. 2-8) I 2-128 (Fig. 2-9) I 2-129 (Fig. 2-10) I 2-130 (Fig. 2-11) I 2-131 (Fig. 2-12) I 2-132 (Fig. 2-13) I 2-133 (Fig. 2-14) I 2-134 (Fig. 2-15) 1 2-135 (Fig. 2-16) I 2-136 (Fig. 2-17) 1 2-137 (Fig. 2-18) I 2-138 (Fig. 2-19) 0 2-139 (Fig. 2-20) 0 2-140 (Fig. 2-21) 0 2-141 (Fig. 2-22) 0 2-142 (Fig. 2-23) 1 2-143 (Fig. 2-24) I 2-144 (Fig. 2-25) 0 2-145 (Fig. 2-26) 0 2-146 (Fig. 2-27) 0 2-147 (Fig. 2-28) 2 2-148 (Fig. 2-29) I 2-149 (Fig. 2-30) 1 2-150 (Fig. 2-31) 0 2-151 (Fig. 2-32) 0 2-152 (Fig. 2-33) I 2-153 (Fig. 2-34) 2 2-154 (Fig. 2-35) 2 3-1 3 3-2 3 3-3 3 3-4 3 3-5 I 4-1 I 4-2 I 4-3 I 4-4 I 5-1 2 5-2 I 5-3 1 5-4 I 5-5 3 5-6 I 5-7 3 5-8 I 2

List Of Effective Pages Page No. Rev. No. Page No. Rev. No.

5-9 3 5-10 2 5-11 3 5-12 3 5-13 3 5-14 3 5-15 3 5-16 2 5-17 1 5-18 1 5-19 1 5-20 1 5-21 1 5-22 1 5-23 2 5-24 2 5-25 1 5-26 3 5-27 1 5-28 1 5-29 1 5-30 1 5-31 1 5-32 2 5-33 1 5-34 1 5-35 I 5-36 1 5-37 I 5-38 2 5-39 I 5-40 1 5-41 I 5-42 1 5-43 I 5-44 1 5-45 I 5-46 2 5-47 I 5-48 1 5-49 I 5-50 1 5-51 2 5-52 I 5-53 1 5-54 2 5-55 2 5-56 I 5-57 1 5-58 I 5-59 1 5-60 3 5-61 2 5-62 I 5-63 3 5-64 I 5-65 1 5-66 2 5-67 I 5-68 2 5-69 3 5-70 1 5-71 I 5-72 2 5-73 I 5-74 I 5-75 2 5-76 I 5-77 I 5-78 I 5-79 I 5-80 I 5-81 I 5-82 I Figure 5-1 I Figure 5-4 0 Figure 5-5 0 6-1 2 6-2 I 6-3 2 6-4 2 6-5 2 6-6 2 6-7 2 6-8 2 6-9 2 6-10 I 6-11 I 6-12 I 6-13 I 6-14 1 3

List Of Effective Pages Page No. Rev. No. Page No. Rev. No.

6-15a 1 7-1 2 8-1 1

  • Appendix 6.1 contains information on DandD DCGL Calculations for Building Occupancy Surface Area Model executed on 9/28/99 for Am-241 (3 pages), C-14 (2 pages), Co-60 (2 pages), Cs-137 (2 pages), Eu-152 (2 pages), H-3 (2 pages), Ni-63 (2 pages), Pu-238 (3 pages), Pu-239 (3 pages), Pu-241 (3 pages) and Sr-90 (2 pages) 4

SNE.-ACILITY LICENSE TERMINATION PLAN C REVISION 3 C Table 2-29 (Contd.)

Listing of all "Hard to Detect Nuclide"/Transuranic Analysis Samle o. naysi AnalyteoalnlecitoLoatin/escipton Sample location See Table 5-2 H-3 Sr-90 Co-60 Cs-137 Am-241 Pu-238 Pu-239 Pu-241 C-14 Ni-63 Eu-152 SX-SL.2655 6/11/02 AnnulusWellA-4,5-10' OL1 (1) <1.89 <0.0497 0.0135 <0.03 <0.0123 <2.28 <0.187 <1.94 SX-SL-2374 6/11/02 SE CV Yard ABH 40, 30-40' OL1 (1) <1.87 <0.0268 <0.0585 <0.019 <0.00134 <2.37 <0.186 <1.64 SX-SL-2484 6/11/02 N CV Yard ABH 22, 50-60' OL1 (1) <2 <0.038 <0.0273 <0.0269 <0.0155 <2.45 <0.178 <1.84 SX-SL-2662 6/11/02 W CV Yard ABH 10, 20-30' OL1 (1) <1.98 <0.0338 <0.041 <0.0307 0.0293 <1.96 <0.189 <1.34 SX-SL-2960 6/11/02 NRC CV Angle Well, Initial Sample (3) <0.299 <0.327 <0.276 <0.16 <24.9 SX-SL-2634 6/11/02 W CV YardABH 9, 10-15 OL1 (1) <2.06 <0.0287 0.0324 <0.0295 <0.0104 <1.68 <0.191 <2.3 SX-SL-2649 6/11/02 Annulus Well A-2, 5-10' OL1 (1) <2.0 <0.0314 <0.00978 <0.0133 <0.00938 <1.87 <0.183 <1.75 SX-SL-2664 6/11/02 W CV Yard ABH 10, 40-50' OL1 (1) 1.99 <0.0303 <0.0216 <0.0105 <0.0256 <2.28 <0.182 <1.84 SX-SL-2660 6/11/02 Annulus Well A-3,15-20' OL1 (1) <1.9 <0.0268 <0.0372 <0.00949 <0.0134 <1.65 <0.189 <2.0 SX-SL-2425 6/11/02 E CV Yard ABH 33,60-90' OL1 (1) <1.99 <0.0339 <0.0261 <0.0264 <0.0118 <1.71 <0.188 <1.57 SX-ST-3077 8/30/02 CV Shell Steel, Internal CV1 <1.17 <1.17 <8.54 SX-ST-3087 8/30/02 CV Shell Steel, Internal CV1 <2.12 <1.45 <36.21 SX-ST-3069 8/30/02 CV Shell Steel, Internal CV1 <1.49 2.28 <10.83 SX-ST-3086 8/30/02 CV Shell Steel, Internal CV1 <1.22 <1.13 <5.9 SX-ST-3085 8/30/02 CV Sheln Steel, Internal CV1 <1.48 <1.38 <9.46 SX-ST-3084 8/30/02 CV Shell Steel, Internal CV1 <2.29 2.55 <5.11 SX-ST-3083 8/30/02 CV Shell Steel, Internal CV1 <1.56 <1.69 <4.33 SX-ST-3082 8/30/02 CV Shell Steel, Internal CV1 <1.66 3 <4.02 SX-ST-3081 8/30/02 CV Shell Steel, Internal CV1 <1.91 3.44 <4.99 SX-ST-3067 8/30/02 CV Shell Steel, Internal CV1 <3.39 16.9 <3.89 SX-ST-3070 8/30/02 CV Shell Steel, Internal i CV1 2.64 180 <10.96 SX-ST-3078 8/30/02 CV Shen Steel, Internal CV1 <1.11 2.3 <4.84 SX-ST-3076 8/30/02 CV Shell Steel, Internal CV1 _ 2.78 24.8 . <4.46 SX-ST-3079 8/30/02 CV Shell Steel, Internal CV1 <2.13 5.08 <20.46 SX-ST-3080 8/30/02 CV Shell Steel, Internal CV1 <1.02 1.9 <4.29 SX-ST-3074 8/30102 CV Shell Steel, Internal CV1 <1.96 3.36 . <3.8 SX-ST-3075 8/30/02 CV Shell Steel, Internal CV1 2.06 1.37 <5.25 SX-ST-3073 8/30/02 CV Shell Steel, Internal CV1 <1.79 3.03 <12.8 SX-ST-3072 8/30/02 CV Shell Steel, Internal CV1 <0.96 1.12 <27.11 _

SX-ST-3071 8/30/02 CV Shell Steel, Internal CV1 3.43 9.26 20.23 _

SX-ST-3068 8/30/02 CV Shell Steel, Internal CV1 <2.85 59 <30.35 DA-SX4PC990114 10/25/01 CV Dome Paint Chips (see (2) < 1.5 < 1.7

_ __ __ _ _ __ _ _ _ _1 1061 1) (PSO-5) __ _ _ _ I__ __ _ _I__ _ I____

Footnotes: (1) Subsurface Sample (2) Component to be Removed (3) Groundwater (4) Outside Impacted Area 2-95

SNEC SNE FACILITY FAILT LICENSE LIES TERMINATION TEMNTO PLAN PLA REVISION REISO II Table 2-30 CV Backfill & Subsurface Sample Results (see Figs. 2-31 & 2-32)

TYPE No. Cs-137 Co-60 11-3 SAMPLE SAMPLE DESCRIPTIONTION

_ _ o. _ AT I SL 1627 <0.I <0.12 - 10/30/2001 0- 'N. CV YARD GC HOLE 800'#1 (ELEVATION 800')

2 SL 1628 <0.07 <0.08 - 10/30/2001 10-20' CV YARD GC HOLE #1 3 SL 1631 <0.08 <0.11 - 10/30/2001 30'-40'N. CV YARD G.C. HOLE #1 4 SL 1632 <0.07 <0.09 - 10/30/2001 20'-30'N. CV YARD G.C.I IOLE #1 5 SL 1633 <0.1 <0.11 - 10/30/2001 40-50' N. CV YARD G.C. HOLE #l 6 SL 1634 <0.1 <0.12 - 10/30/2001 50-60'N. CV YARD G.C. HOLE #I 7 SL 1635 <0.1 <0.12 - 10/30/2001 60-70'N. CV YARD G.C. HOLE#l 8 SL 1636 <0.09 <0.09 - 10/30/2001 70-75' N. CV YARD G.C. HOLE #1 9 SL 1637 <0.11 <0. 13 - 10/30/2001 0-IV N. CV YARD G.C. IIOLE #3 800 10 SL 1638 <0.08 <0.09 - 10/30/2001 10-20 N. CV YARD G.C. HOLE #3 11 SL 1639 <0.09 <0.08 - 10/30/2001 20-30' N. CV YARD G.C. HOLE #3 12 SL 1640 <0.06 <0.07 - 10/30/2001 30-40' N. CV YARD G.C. HOLE #3 13 SL 1641 <0.13 <0.147 - 10/30/2001 40-50'N. CV YARD G.C. IIOLE #3 14 SL 1642 <0.09 <0.12 - 10/30/2001 50-60' N. CV YARD G.C. HOLE #3 15 SL 1643 <0.11 <0. 13 - 10/30/2001 60-70'N. CV YARD G.C. HOLE #3 16 SL 1648 <0.05 <0.07 - 10/30/2001 70-75' N. CV YARD G.C. HOLE #3 17 SL 1649 <0.07 <0.07 - 10/30/2001 0-10' N. CV YARD G.C. HOLE #5 (ELEVATION 800')

18 SL 1650 <0.07 <0.07 - 10/30/2001 10-20' N. CV YARD G.C. 11OLE #5 19 SL 1651 <0.12 <0.13 - 10/30/2001 20-30'N. CV YARD G.C. HOLE #5 20 SL 1652 <0.1 <0.11 - 10/30/2001 30-40' N. CV YARD G.C. HOLE #5 21 SL 1653 <0.12 <0.13 - 10/30/2001 40-50'N. CV YARD G.C. HOLE #5 22 SL 1654 <0.1 <0.11 - 10/30/2001 50-60' N. CV YARD G.C. HOLE #5 23 SL 1655 <0.09 <0.11 - 10/30/2001 60-75' N. CV YARD G.C. IOLE #5 24 SL 1656 <0.09 <0.09 - 10/30/2001 0-10' N. CV YARD G.C. HOLE #7 (ELEVATION K>

800')

25 SL 1657 <0.12 <0.15 - 10/30/2001 10-20' N. CV YARD G.C. HOLE #7 26 SL 1658 <0.11 <0.13 - 10/30/2001 20-30' N. CV YARD G.C. HOLE #7 27 SL 1659 <0.08 <0.08 - 10/30/2001 30-40' N. CV YARD G.C. HOLE #7 28 SL 1660 <0.13 <0.13 - 10/30/2001 40-50'N. CV YARD G.C. IIOLE #7 29 SL 1661 <0.11 <0.12 - 10/30/2001 50-60'N. CV YARD G.C. HOLE #7 30 SL 1662 <0.09 <0.1 - 10/30/2001 60-75' N. CV YARD G.C. HOLE #7 31 SL 1680 <0.12 <0.14 - 10/30/2001 0-10 N. CV YARD G.C. HOLE #9 (ELEVATION 800')

32 SL 1681 <0.11 <0.11 - 10/30/2001 10-20'N. CV YARD G.C. IIOLE #9 33 SL 1682 <0.1 <0.11 - 10/30/2001 20-30' N. CV YARD G.C. HOLE #9 34 SL 1683 <0.07 <0.08 - 10/30/2001 3040' N. CV YARD G.C. HOLE #9 35 SL 1684 <0.11 <0.12 _ 10/30/2001 40-50'N. CV YARD G.C. IIOLE #9 36 SL 1685 <0.07 <0.05 - 10/30/2001 50-60' N. CV YARD G.C. HOLE #9 37 SL 1686 <0.11 <0.12 - 10/30/2001 60-75' N. CV YARD G.C. HOLE #9 38 SL 1687 <0.11 <0.12 - 1O/30/2001 0-10 N. CV YARD G.C. HOLE #11 (ELEVATION 800')

39 SL 1688 <0.12 <0.147 - 10/30/2001 10-20'N. CV YARD G.C.IIOLE #1I1 40 SL 1689 <0.13 <0.13 - 10/30/2001 20-30'N. CV YARD G.C. HOLE # 11 41 SL 1690 <0.12 <0.13 - 10/30/2001 30-40'N. CV YARD G.C. IIOLE #1I1 42 SL 1691 <0.11 <0.13 - 10/30/2001 40-50'N. CV YARD G.C.HOLE # 11 43 SL 1692 <0.1 <0.11 - 10/30/2001 50-60'N CV YARI) G.C. HOLE #11 _

44 SL 1693 <0.08 <0.09 - 10/30/2001 60-75' N. CV YARD G.C. HOLE #I 1 45 SL 1695 <0.07 <0.08 _ 10/31/2001 0-10 N. CV YARD G.C. HOLE #13 (ELEVATION I 800')

NOTE: All water samples are reported in uCi/ml except for 11-3 i hich is reported in pCi/L Soil samples are in pCilg.

2-96

s iI SNEC FACILITY LICENSE TERMINATION PLAN REVISION 3 3.0 IDENTIFICATION OF REMAINING DISMANTLEMENT ACTIVITIES

3.1 INTRODUCTION

As discussed previously, this final phase of the SNEC Facility Decommissioning Project commenced in April 1998 following NRC approval of a License Amendment authorizing decommissioning. Since that time, the greater part of 1998 was devoted to removing and shipping the SNEC Facility Large Components (i.e. Reactor Vessel, Pressurizer and Steam Generator). Following removal of the large components, the latter part of 1998 and the beginning of 1999 were devoted to removing the bulk of the remainder of the systems from the SNEC Facility. Following that, concrete decontamination commenced and continued into 2000 when it was determined that complete removal of the concrete from the Containment Vessel (CV) would be required in order to release the site in accordance with NRC Regulations (10 CFR Part 20.1402, Radiological Criteria for Unrestricted Use). Concrete removal was completed in October 2002. The interior of the CV below the -804 ft. elevation was then decontaminated and a final status survey was performed. The NRC has accepted the results of the survey and the CV has been backfilled to the -804ft. elevation in preparation of removal of the upper CV shell.

As described in Section 2.2.4 the site is divided into eight areas in describing radiological conditions. These same area designations will be used in describing remaining site dismantlement activities.

FIGURE 3-1 Photograph of SNEC Facility CV & DSF Buildings ASNE CV' A I SNEC DSF (Removed) 3-1

SNEC FACILITY LICENSE TERMINATION PLAN REVISION 3 REVISION 3 SNEC FACILITY LICENSE TERMINATION PLAN 3.2 REMAINING TASKS 3.2.1 Area 1 (CV Basement) Remaining Site Dismantlement Activities Area 1 was completely removed to the steel CV liner.

3.2.2 Area 2 (Primary Compartment) Remaining Site Dismantlement Tasks Area 2 was completely removed to the steel CV liner.

3.2.3 Area 3 (Auxiliary Compartment) Remaining Site Dismantlement Activities Area 3 was completely removed to the steel CV liner.

3.2.4 Area 4 (Operating Floor) Remaining Site Dismantlement Activities Remaining tasks in Area 4 include removal of miscellaneous steel and removal of the polar crane and its girders. The polar crane is the last large component to be removed from the SNEC Facility. It will be removed following removal of the upper dome of the CVThe Containment Vessel is a special case. The SNEC Facility License requires the interior of the Containment Vessel to be maintained as an exclusion area Thus, prior to removal of the upper dome of the Containment Vessel the SNEC Technical Specifications will need to be revised. A change request to the SNEC Facility Technical Specifications (Reference 3-1) has been submitted to permit this activity (Note: this request was approved on 1/9/04). During removal of the upper dome, the backfilled portion of the lower Containment Vessel will be protected concrete cap.

This area will be resurveyed after completion of the removal of the upper dome.

3.2.5 Area 5 (Pipe Tunnel) Remaining Site Dismantlement Activities As described in Section 2.2.2, Area 5 consisted of a concrete tunnel, which extended 235 degrees around the CV. The tunnel has been removed except for the portion under the Material Handling Bay (MHB) section of the Decommissioning Support Facility.

The portion of the tunnel under the MHB will be removed and the area surveyed and remediated as appropriate to meet site release levels.

3.2.6 Area 6 (Reactor Compartment and Storage Well) Remaining Site Dismantlement Activities Area 6 was completely removed to the steel CV liner.

3.2.7 Area 7 (Exterior of Containment Vessel Dome) Remaining Site Dismantlement Activities Remaining tasks in Area 7 consist of removing the CV Dome to at least three feet below grade and completing the requisite Final Status Survey and backfilling the remaining below grade portions of the CV. Remediation, if needed, will be performed as appropriate.

K>,

3-2

SNEC FACILITY LICENSE TERMINATION PLAN REVSION . 3 SNEC FACILITY LICENSE TERMINATION PLAN . RFVISION

. ._, ._._. 3_

3.2.8 Area 8 (SNEC Facility Yard Areas) Remaining Site Dismantlement Activities As described in Section 2.2.4.1.2, the Decommissioning Support Facility (DSF) building walls and roof have been dismantled and released from the site. Concrete floors for the remaining DSF subdivision areas [i.e. Decommissioning Support Building (DSB), Material Handling Building (MHB) and the Personnel Access Facility PAF)] still remain. Remaining tasks in Area 8, with the exception of the Saxton Steam Generating Station (SSGS) area will consist of some soil removal and removal of utilities located outside the CV.

The SSGS was used by the SNEC Facility to support its operation, generate electricity and to provide dilution flow for radioactive effluent liquid discharges. Surveys in this area indicate there was some residual contamination that needed to be removed. This included water and sediment removal, some concrete decontamination, and extensive embedded piping surveying and removal. These activities were completed..

3.3 ACCESS CONTROL MEASURES Since all decommissioning activities were not completed prior to the start of final status survey work, measures as described in Section 5.3 have been implemented to protect survey areas from recontamination during and after the final status survey. In all cases, decommissioning activities creating a potential for the spread of contamination will be completed within each survey area prior to starting a final status survey in the area. Additionally, decommissioning activities that create a potential for the spread of contamination to adjacent survey areas will be evaluated and controlled using barriers, covers, or restricting or rescheduling activities.

3.4 10 CFR 50.59 REVIEW 10 CFR 50.59 'Changes, Tests and Experiments" permits licensees to make changes to the facility as described in the Safety Analysis Report (as updated), make changes in procedures as described in the Safety Analysis Report (as updated), and conduct tests or experiments not described in the Safety Analysis Report (as updated) without obtaining a license amendment under certain conditions.

With respect to the SNEC Facility this License Termination Plan serves to support a License Amendment to perform the Final Status Survey and terminate the license. In addition as indicated in Section 3.2.4 above a change request to the SNEC Facility Technical Specifications (Reference 3-1) has been submitted to allow removal of the upper Containment Vessel dome.

The remaining tasks described in Section 3.2 are bounded by the Definition for Decommissioning Activities in the SNEC Facility Technical Specifications (Reference 3-3) and are thus permitted by the Technical Specifications. Additionally these tasks have been reviewed against the criteria of 10 CFR 50.59 and are bounded by the SNEC Facility USAR (Reference 3-4).

Thus, although some changes will be made to the facility as described in the SNEC Facility USAR, e.g. Removal of the Polar Crane, these activities were anticipated to be performed.

Further, greater than 99% of the curies present at the SNEC Facility at the commencement of decommissioning in 1998 have been safely removed from the site and shipped for proper disposition. Finally, each activity requiring a procedure as defined by the SNEC Facility Technical Specifications is reviewed to ensure it satisfies the criteria of 10 CFR 50.59. Thus, 3-3

SNEC FACILITY LICENSE TERMINATION PLAN REVISION REVISION 33 SNEC FACILITY LICENSE TERMINATION PLAN with the exception of the removal of the Containment Vessel upper dome, none of the remaining tasks identified in Section 3.2 requires prior NRC approval in accordance with 10 CFR 50.59.

3.5 DECOMMISSIONING TASKS REQUIRING COORDINATION WITH OTHER FEDERAL OR STATE REGULATORY AGENCY Prior to License Termination, tasks requiring coordination with any other Federal or State Regulatory Agency include disposal of the water in Area 8. Additionally, some water treatment will be necessary for water collecting in excavated areas e.g., -the yard area North of the CV.

Discharge of this water will be controlled radiologically in accordance with the requirements of the ODCM (Reference 3-2). Non-radioactive sampling and analysis will be in accordance with a program accepted by the Commonwealth of Pennsylvania. Additionally, the Army Corps of Engineers gave its consent for locating temporary on-site excavations and infiltration galleries.

3.6

SUMMARY

The tasks remaining for site dismantlement are permitted by the SNEC Facility License Technical Specifications and USAR. Based on current estimates, these tasks will be completed in the second quarter of 2003 for an estimated exposure of approximately 37 person-rem (Table 3.1). Of this total approximately 35 person-rem has been expended to date. Estimated remaining low level radioactive waste generation is provided in Table 3.2.

3.7 REFERENCES

3-1 SNEC Facility Technical Specification Change Request No. 62, April 2002.

3-2 SNEC Procedure E900-PLN-4542.08 Off-Site Dose Calculation Manual" 3-3 SNEC Facility Technical Specifications 3-4 Updated Safety Analysis Report for Decommissioning the SNEC Facility 3-4

SNEC FACILITY LICENSE TERMINATION PLAN REVISION 3 To obtain the best approximation of the actual ratio between radionuclides present in any area, a sufficient number of measurements or samples, spatially separated throughout the area must be collected. The number of measurements or samples needed to determine the ratio is based on the chemical, physical, and radiological characteristics of the radionuclides expected to be present in the area. The MARSSIM surrogate method will be applied when similar physical and geological characteristics are present in a survey unit (Reference 5-21). This method will be used to determine the most representative ratio for each positively identified radionuclide present in the data set with the surrogate radionuclide. From the sum of the ratios a mean ratio value for each radionuclide will be calculated and used to determine an effective surrogate DCGLw. In conclusion, selecting a surrogate ratio ensures that potential exposures from individual radionuclides are not underestimated. The surrogate DCGLw for volumes with multiple radionuclides is calculated using the following equation (NUREG-1575, Reference 5-5, Equation 1-14):

Dot= 1/ [11D,+R21D2+R31D3+... +R.IDnJ Where: Dtotai is the DCGLW for the surrogate radionuclide when the concentration of that radionuclide represents all radionuclides that are present in the survey unit. DI is the initial DCGLw for the surrogate radionuclide. D2, D3 ,.... Dn are DCGLw values for radionuclides 2, 3,... n.

R2, R3... Rn are the respective ratio values between the concentration of the surrogate radionuclide and radionuclide 1, 2, ....n.

Post-remediation sampling may be performed to re-establish surrogate ratios when there is reason to believe that the remediation process may have significantly changed the ratios of the original radionuclide concentration mix. However, the surrogate ratio will usually be established using data collected prior to remediation. Additional sample data may be collected or an evaluation performed when there is reason to believe a surrogate ratio is non-representative of the survey area. In all cases, SNEC will use the best representative data for the survey area. A radionuclide mix justification will be documented in each survey package.

5.2.3.2.4 Gross Activity DCGLs Where multiple radionuclides are present, a gross activity DCGL may be developed. The gross activity DCGL enables field measurements of gross activity rather than the determination of individual radionuclide activity, for comparison to the radionuclide-specific DCGL. The gross activity DCGL, or DCGLGA, for surfaces or volumes with multiple radionuclides is calculated using the following equation (NUREG-1 575, Reference 5-5, Equation 4-4):

Gross Activity DCGL = 1I((f 1 IDCGL 1 ) + (f 2 IDCGL 2) +...(f.IDCGL))

Where:

f, = relative fraction of the total activity contributed by radionuclide 1 f2 = relative fraction of the total activity contributed by radionuclide 2 fn = relative fraction of the total activity contributed by the nth radionuclide in the mix 5-7

I ItF=Kl.rF TF=DM1K1^TlfKl 0 AKI 4 QMf F:At'll 1TV QKNIf' AI IT

...... I t~. ~

ItMQ

--- iI~AIM I~IorL

,l.n r IA

..- n,IVL.

MIC1110ink VI OEJIV I In situations where an area has unknown or a highly variable concentration of radionuclides throughout, it is acceptable to select the most conservative mix of radionuclides present when "

developing a gross DCGL. Remediation efforts at the SNEC Facility will seek to remove contamination from concrete surfaces by removing significant near surface volumes of concrete.

On other structural surfaces (e.g., support steel), abrasive decontamination techniques may be used to remove surface deposited contamination. Additionally, by aggressively chasing cracks and other structural anomalies, little (if any) surface deposited radionuclides should remain.

Those that do remain will be at a significantly reduced fraction of their previous level.

Characterization data and/or post-remediation sampling data will be used to adjust gross DCGL values before performing the FSS. In either case, a radionuclide mix justification will be documented in each survey package.

Gross activity measurements will be performed in both reference areas (background areas) and areas to be surveyed.

Open land areas will be scanned by semi-automated contractor supplied measurement equipment and/or by hand-held Nal detector instrumentation. Samples will also be obtained.

When necessary, spectra of the components of a radiation field may be used to differentiate between site related radioactive materials and natural occurring radioactivity.

5.2.3.2.5 Elevated Measurement Comparison (EMC) DCGLs Elevated Measurement Comparison (EMC) DCGLs are DCGL values that have been modified by reducing the dose model area size in the dose modeling computer code. The effect is to allow higher levels of contamination to exist in smaller well-defined on-site areas. Assuming the residual radioactivity is concentrated in a much smaller area rather than uniformly over the entire survey unit is the basis for developing the DCGLEMC. The methodology used to calculate the DCGLEMC is given in Appendix 5-1. Area factors for individual radionuclides have been calculated and are presented in Tables 5-15 and 5-15A.

5.2.3.2.6 Unity Rule Typically, each radionuclide specific DCGL corresponds to the release criteria (e.g., regulatory limit in terms of dose or risk). However, in the presence of multiple radionuclides, the total of the DCGLs for all radionuclides would exceed the release criteria. In this case, the individual DCGLs need to be adjusted to account for the presence of multiple radionuclides contributing to the total dose. One method for adjusting the DCGLs is to modify the assumptions made during exposure pathway modeling to account for multiple radionuclides. The surrogate measurement method discussed previously, describes one method for adjusting the DCGL to account for multiple radionuclides. Additionally, another method includes the use of the unity rule.

The unity rule, represented in the expression below, is satisfied when radionuclide mixtures yield a combined fractional concentration limit that is less than or equal to one (1) (NUREG-1575, Equation 4-3):

(Cl1 DCGL1 ) + (C2 IDCGL 2 ) + ...(C I DCGL.) < 1 Where:

C = concentration of radionuclide (1, 2, ...,n)

DCGL = guideline value for each individual radionuclide (1, 2, ...,n) 5-8

C SN FACILITY LICENSE TERMINATION PLAN REVISI 3 Table 5-2 Initial Classifications of Site Areas Survey Unit Designations of the SNEC Facility and Surrounding Impacted Areas i Survey DescriptionUl Classification Survey Unit Area (mi) l Number of Tpo f l Number_________Desc_____p ____on________ 1 3WlsCing Ohr Survey UnitsAple Nub r1 2___________

3___ Floor I als C ilnIO h MISCELLANEOUS AREAS & ITEMS MA1 Airborne Monitoring Stations X c10 1 1 MA2 SSGS Discharge Tunnel Outfall (Land Area) X 600 1 2 MA3 Weir Outfall X 25 1 2 MA4 Weir Outfall Buffer X 200 1 2 MA5 Northeast Dump Site X 7000 1 2 MA6 Northwest Open Land Area X 4100 1 2 MA7 Northwest Open Land Area X 100 1 2 MAB Miscellaneous Concrete Slabs (Around Site) X I100 1 each I CONTAINMENT VESSEL (CV) -INTERIOR & EXTERIOR STEEL SHELL CV1-1 to 4 Interior Vertical Wall of CV Shell - Plates 0 to G X 392 4 1 "'

CV2-1 to 22 Intemal Support Rings - 774' El to 802' El X 410 22 I (

CV2-23 to 28 Internal Support Ring Weld Areas - 774 El to 802' El X 66 22 (a) (e)

CV3-1 to 3 Interior Curved Bottom of CV Shell - Plates A to C X 256 3 1 "'

CV4-1 Exterior Wall - 802.6' El up to Cut-off X 7 1 1" CV5 Exterior Wall 1 Meter Below Class 1Area (Down to 797.6' El) X 16 1 s CV6 & CV6-1 External Rock Anchor Support Ring Assembly Area X 66 1( I (

NOTES:

(a) XX designates a sequential number starting with 1, and defines a survey unit within a survey area.

(b) This data was estimated with best available information. No survey unit, regardless of its classification will exceed 10,000 square meters.

(c) NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs = 2 (d) Survey units were established as the ring areas became available to field personnel doing the survey work.

(e) Activation of CV steel liner to be addressed when region Is accessible.

5-11

SNFC FACILITY LICENSE TERMINATION PLAN REVISION 3-REVISION 3 SNEC FACILITY LICENSE TERMINATION PLAN Table 5-2 (continued)

Initial Classifications of Site Areas imrvrav IUnit flarinnfinna of the SNFI' FPnrilift nnri !Q;rrninelinn ImnartarI Aroac Survey Unit e Classification Survey Unit Area (M 2 ) 2 Number of T Type of DCGL Number (Iaj 1 2 l 3 Floor I Walls ICeilingI Other Survey Units ° Applied)

DECOMMISSIONING SUPPORT FACILITY (DSF - SNEC AREA DB1-1 & 2 DSB Floor Slab X 212 2 1 PF1 PAF Floor Slab X 36 = = = 1 1 DB5 DSB Carport Floor Slab X 62 1 1 SAXTON STEAM GENERATING STATION (SSGS), INTAKE, DISCHARGE & STEAM TUNNELS' SS1 Floor of Discharge Tunnel (first -150') X l l___ 120 l_ll_l 1 1 SS2 Floor of Discharge Tunnel (next -235') X 175 1 1 SS3 Floor of Discharge Tunnel (last -315') X 234 l 1 1 SS4 Ceiling of Discharge Tunnel (first -150') X l 120 1 1 SS5 Ceiling of Discharge Tunnel (last -550') 7 l l X l_7 _l_ 400 1 1 SS6-1 & 2 Walls of Discharge Tunnel (first -1 50') X l_l _l 290 3 1 SS7-1 & 2 Walls of Discharge Tunnel (last -550') X l___ 600 1 1 SS8-X In DT -Seal Chambers (1 2, & 3) X 230 3 1 NOTES:

(a) 'X" designates a sequential number starting with 1, and defines a survey unit within a survey area.

(b) This data was estimated with best available information. No survey unit, regardless of its classification will exceed 10,000 square meters.

(c) NRC Default Surface DCGLs = 1,Site Specific Volumetric DCGLs = 2: SNEC plans to use surface area DCGLs as noted in SSGS section. However, if geometry of surface is not appropriate for a surface area measurement then guidance as specified in LTP Chapter 6, Section 6.2.1 may need to be implemented.

5 et

( {7

SNQ e-ACILITY LICENSE TERMINATION PLAN C. or-VI0Ic J.

Table 5-2 (continued)

Initial Classifications of Site Areas Survev Unit Deslanations of the SNEC Facility and Surroundina Imoacted Areas Survey Unit e Classification Survey Unit Area (m 2 ) lb' Number of Type of DCGL Number Descnption

"' 1 l 2l 3 Floor I Walls I Ceiling I Other Survey Units ( Applied (c)

SAXTON STEAM GENERATING STATION (SS S), INTAKE, DISCHARGE & STEAM TUNNELS (cont'd)

SS9 Spray Pump Pit Floor X l_ l__ 120 1 1 SS10 Spray Pump Pit Walls Below 795' El. X 20 1 1 SS1 1 Spray Pump Pit Walls Above 795' El. X 100 1 1 SS12 SSGS Boiler Pad (811 El.) X 1800 - 1 1 SS13-1 & 2 SSGS Firing Aisle (806' El.) X 560 80 1 1 SS14-1 to 6 SSGS Basement Area Floor (790' El.) X _ 420 6 1 SS15 SSGS Basement Walls - East End X 100 1 1 SS16 SSGS Basement Walls Up to 2 Meters X 240 1 1 SS17 SSGS Basement Walls > 2 Meters X 350 1 1 SS18- X Top of Seat Chambers X 70 1 1 SS19-X Section of SSGS Intake Tunnel Floor X 493 = 3 1 SS20-X Section of Intake Tunnel Walls X 2150 3 1 SS21 Section of Intake Tunnel Ceiling X 493 3 1 SS22-1 to 4 CV Steam Tunnel Remnants X - 20 45 15 c30 4 1 SS23 Discharge Tunnel Access Area X = 97 1 1 SS24 Miscellaneous SSGS Area Concrete Slabs X < 400 1 1 NOTES:

(a) *X-designates a sequential number starting with 1, and defines a survey unit within a survey area.

(b) This data was estimated with best available Information. No survey unit, regardless of its classification will exceed 10,000 square meters.

(C) NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs = 2: SNEC plans to use surface area DCGLs as noted in SSGS section. However, if geometry of surface is not appropriate for a surface area measurement then guidance as specified In LTP Chapter 6, Section 6.2.1 may need to be Implemented.

5-13

SqNEC FACILITY ILICENSE TEIRMINATION PI-AN REVISION 3 REVISION 3 NFC FACII ITY I IfFN5F TFRMINATIAN PLAN Table 5-2 (continued)

Initial Classifications of Site Areas Survey Unit Designations of the SNEC Facility and Surrounding Impacted Areas Survey Unit I Classification I Survey Unit Area (mi2 ) (b) Number of Type of DCGL Number Description 1 2 3 Floor I Walls ICeiling Other Survey Units b) Applied (

SAXTON STEAM GENERATING STATION (SSGS) SPRAY POND AREA SPi Open Land Area IX 1 6600 1 2 SNEC FACILITY SITE OPEN LAND AREA OL1-1 to X SNEC Facility Site & Near Site Area X l 11000 11 2 GPU ENERGY (PENELEC) SITE OPEN LAND AREA OL2-X Westinghouse and Adjacent Areas ( X 5700 l 6 l 2 OL3 Warehouse Burn Area X 200 1 2 OL4-X Buffer Zones X= 5600 1 2 REMAINING IMPACTED OPEN LAND AREA OL5-X Site Road Access Areas X 20500 9 2 OL6-X Stack Release Area (NNE) X _ 14600 3 2 OL7-X Stack Release Area (SSW) X 12700 2 2 OL8-X Buffer Zones X 47900 5 2 WAREHOUSE (LARGE GARAGE - SOUTH) - PENELEC AREA WA1-X Floor Slabs - X 450 l l l 2 1 WA2 Sump X 1 <10 1 1 WA3 Exterior Foundation Walls X 374 l I 1 1 NOTES:

(a) "X' designates a sequential number starting with 1, and defines a survey unit within a survey area.

(b) This data was estimated with best available information. No survey unit, regardless of its classification will exceed 10,000 square meters.

(c) NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs = 2 (d) Includes substation yard drainage area.

( K_ .

CL SNE%, r:ACILITY LICENSE TERMINATION PLAN C RFVIRIIC Table 5-2 (continued)

Initial Classifications of Site Areas Survey Unit Designations of the SNEC Facility and Surrounding Impacted Areas Survey Unit D p Classification Survey Unit Area (M2) (c) Number of Type of DCGL Number Descnpton 1 l2 l31 Floor I Walls Ceiling I Other Survey Units (b) Applied (c)

GARAGE (SMALL GARAGE - SOUTHWEST) - PENELEC AREA GA Floor Slab X 109 GA2 SXm < 10 1 1 LINE SHACK - PENELEC AREA LS1-X Floors & Walls Up to 2 Meters (Interior) X 290 177 2 1 LS2-X Upper Walls & Ceiling (Interior) X 191 412 1 1 LS3 Exterior Walls X 343 1 1 LS4 Roof X 324 1 1 LS5 Roof Drainage System X <10 1, 2 LS6 Line Shack Attic Area X < 50 1 1 PS i In ei r PENELEC SWITCHYARD BUILDING & YARD STRUCTURES

_ _ _ 5 89 I__ _ _ _ _ _ _ _ __I_ _ _

PSI0T Interlor X 55 89 55; 1T' PS2 Exterior Walls and Roof X 151 68 1 PS3 Switchyard Units - Base Pads = X <500 = 1 1 NOTES:

(a) *X' designates a sequential number starting with 1, and defines a survey unit within a survey area.

(b) This data was estimated with best available Information. No survey unit, regardless of its classification will exceed 10,000 square meters.

(c) NRC Default Surface DCGLs = 1, Site Specific Volumetric DCGLs = 2

  • 5-15

SNFC FACILITY LICENSE TERMINATION PLAN REVISION 22 REVISION SNEC FACILITY LICENSE TERMINATION PLAN 5.2.4.4 Changes in Classification Changes in classification are based on survey data and other relevant information that indicates a different area classification is more appropriate. Changes in area classifications which decrease an area classification will be in accordance with License Condition 2.E.(h).

5.2.5 Final Survey Process In general, FSS activities do not commence in the area to be surveyed until decontamination activities are believed to be complete and radioactive waste materials are removed. The FSS process begins with survey area preparation activities such as gridding and review of final remediation support survey information, as well as survey area walk-downs. Survey design calculations and the issuance of Survey Requests to field survey teams follow this phase. Field survey teams then collect the data and assemble the survey results in an organized and understandable format in accordance with site procedures. Data assessment and documentation concludes this process.

5.2.5.1 Survey Design Overview Survey design, as described in Section 5.4, identifies relevant components of the FSS process and establishes the assumptions, methods, and performance criteria to be used. Areas ready for FSS are classified as Class 1, Class 2 or Class 3 and are divided into survey units.

Systematic scan and static measurements are prescribed according to a pattern and frequency established for each classification. Investigation levels are established which, if exceeded, initiate an investigation of the survey data. A measurement from the survey unit that exceeds an investigation level may indicate a localized area of elevated residual radioactivity. Such locations are marked and investigated to determine the area and the level of the residual -

radioactivity present. Depending on the results of the investigation, the survey unit may require remediation, and/or re-survey or re-classification.

Quality Control (QC) measurements are prescribed to identify and control measurement error and uncertainty attributable to measurement methods or analytical procedures used in the data collection process. QC measurements provide qualitative and quantitative information to demonstrate that measurement results are sufficiently free of error and accurately represent the radiological condition of the SNEC Facility.

5.2.5.2 Survey Data Collection As deemed appropriate, a final post-remediation survey is performed using similar instrumentation, quality control and survey techniques to be used in the FSS process. The review of the final post-remediation survey data is then carried out to verify that residual radioactivity levels are acceptable and that no additional remediation will be needed in the survey unit. If an area of elevated residual radioactivity is identified, and remediation is determined to be ALARA, the area is remediated and re-surveyed to ensure meeting FSS requirements. The data collected during the final post-remediation survey (when performed),

5-16

SNEC FACILITY LICENSE TERMINATION PLAN REVISION 1

2. Planned dismantlement activities affecting or adjacent to the post remediation survey unit are completed, or are evaluated and determined to not have a reasonable potential to introduce radioactive material into the post remediation survey unit.
3. An operational radiation protection survey of the post remediation survey unit is completed and all outstanding items are addressed.
4. Planned physical work in, on, or around a post remediation survey unit, other than routine surveillance or maintenance, is complete.
5. Tools, non-permanent equipment, and material not needed for survey data collection are removed.
6. Housekeeping, clean up, and remediation of the survey unit are completed.
7. Scaffolding, temporary electrical and ventilation equipment and components, and other material or equipment needed for survey data collection is radiologically clean and left in place.
8. Transit paths to/through the post remediation survey unit are eliminated or re-routed.
9. Appropriate measures are instituted to prevent the re-introduction of radioactive material into isolated area from ventilation systems, drain lines, system vents, and other potential airborne and liquid contamination pathways.
10. Measures are instituted to control access and egress and otherwise restrict radioactive material from entering the survey unit.

5.3.3 Transfer of Control Once a walk-down has been performed and the isolation criteria are met, control of activities within the post remediation survey unit is transferred from the dismantlement organization to the FSS team. The need for localized remediation within the isolated area may Pe identified after transfer of control. Localized remediation may be performed under the control of the FSS organization.

However, if large areas require remediation, the isolated area may be transferred back to the dismantlement organization for further decontamination.

5.3.4 Isolation and Control Measures Prior to performing the FSS, the post remediation survey unit is isolated and controlled. Routine access, equipment removal, material storage, and worker and material transit through the area without proper controls are no longer allowed. One or more of the following administrative and physical controls will be established to minimize the possibility of introducing radioactive material from ongoing decommissioning activities in adjacent or nearby areas.

1. Personnel training
2. Installation of barriers to control access to the area(s)
3. Installation of postings with access/egress requirements
4. Locking or otherwise securing entrances to the area 5-25

SNEC FACILITY LICENSE TERMINATION PLAN REVISION 3

5. Installation of tamper-evident seals or labels Isolation and control measures are implemented through approved facility procedures and remain in place through the FSS data collection process until license termination.

5.4 SURVEY DESIGN The survey design identifies relevant components of the FSS process, and establishes the assumptions, methods, and performance criteria to be used. The methodology for planning a FSS, including a FSS in the subsurface region is identified in the applicable site procedure. Survey design is summarized in Table 5-5.

The application of survey design criteria to structures and land areas will vary based on the type of survey media and the relative potential for elevated residual radioactivity. For facility systems, many of the survey design criteria applicable to structures and land areas do not apply or are dictated by the physical system layout and the accessibility to the system piping and components.

To accommodate these factors, the survey design integrates both non-systematic (random) and judgmental (biased) methods to data collection to achieve the overall objective of the final survey process. Survey design will be performed in accordance with SNEC procedures E900-ADM-4500.59, 'Final Site Survey Planning" and E900-ADM-4500.58, 'Treatment of Embedded Piping and Components".

Each survey design package will address the following areas of interest:

1. A brief overview describing the final status survey design;
2. A description and map or drawing of impacted areas of the site, area, or building classified by residual radioactivity levels (Class 1, Class 2, or Class 3) and divided into survey units, with an explanation of the basis for division into survey units and the boundaries for each survey unit or area indicated. Maps should have compass headings indicated;
3. A description of the background reference areas and materials, if they will be used, and a justification for their selection;
4. A summary of the statistical tests that will be used to evaluate the survey results, including the elevated measurement comparison, if Class 1 survey units are present, a justification for any test methods not included in MARSSIM, and the values for the decision errors ( and ) with a justification for values greater than 0.05;
5. A description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide;
6. For in-situ sample measurements made by field instruments, a description of the instruments, calibration, operational checks, sensitivity, and sampling methods, with a demonstration that the instruments, and methods, have adequate sensitivity;
7. A description of the analytical instruments for measuring samples in the laboratory, including the calibration, sensitivity, and methodology for evaluation, with a demonstration that the instruments and methods have adequate sensitivity; 5-26

SNEC FACILITY LICENSE TERMINATION PLAN I REVISION I1 REVISION SNEC FACILiTY LICENSE TERMINATION PLAN Table 5-12 ., .

Initial Survey Results and Conclusions When A Background Reference Area Is Not Used

Conclusion:

Survey unit meets Survey Result (Class I Areas) site release dose criteria Yes No All measurements less than or equal to DCGLw /

Mean greater than DCGLw /

Any measurement greater than DCGL with mean less than or equal to DCGL which passes Sign test Any measurement greater than DCGL with mean less than or equal to DCGL which fails Sign test Table 5-13 Initial Survey Results and Conclusions When a Background Reference Area is Used

Conclusion:

Survey unit meets Survey Result (Class 1 Areas) site release dose criteria Yes No Difference between maximum survey unit measurement and minimum background reference area measurement less than V or equal to DCGLw Difference between survey unit mean and - background /

reference area mean greater than DCGLW Difference between any survey unit measurement and any background reference area measurement greater than DCGLw, and difference between survey unit mean and background a reference area mean less than or equal to DCGLw which passes WRS test Difference between any survey unit measurement and any background reference area measurement greater than DCGL and difference between survey unit mean and background V reference area mean less than or. equal to DCGLW which fails WRS test 5-59

SNEC FACILITY LICENSE TERMINATION PLAN REVISION REVISION 3 3 SNEC FACILITY LICENSE TERMINATION PLAN 5.6.4.2 Sign Test The one-sample Sign statistical test is used if the radionuclide of concern is not present in background and radionuclide-specific measurements are made. The Sign test may also be used for gross activity measurements if one or more radionuclides are present in background at such small fractions of the DCGLw as to be considered insignificant. In this case, background concentrations of the radionuclides are included with the residual radioactivity (in other words, the entire amount is attributed to facility operations). Thus, the total concentration of the radionuclides is compared to the site release criteria. This option is only used if it is expected that ignoring the background concentration does not affect the outcome of the statistical test. The advantage of ignoring a small background concentration is that no background reference area is necessary.

The Sign test is applied as follows:

1. List the survey unit measurements, Xi, i = 1, 2, 3,. . .,n; where: n = the number of measurements.
2. Subtract Xi from the DCGLW to obtain the difference (DCGLw - Xi, i=1, 2, 3...n).
3. Discard differences where the value is exactly zero and reduce n by the number of zero measurements.
4. Count the number of positive differences. The result is the test statistic S+. Note that a positive difference corresponds to a measurement below the DCGLW and contributes evidence that the survey unit meets the site release criteria.

Compare the value of S+ to the critical values in Appendix I, Section 1.3 of NUREG-1575 (Reference 5-5). The Table columns equate to the false positive decision error rate, a. The value of a is the probability of passing a survey unit which actually fails to meet the site release criteria, which is obtained from the survey design (the initial value is 0.05 - see Appendix 5-2). If S+ is greater than the critical value for the false positive decision error rate given in the Table, the survey unit meets the site release criteria. If S+ is less than the critical value, the survey unit fails to meet the criterion.

5.6.4.3 Wilcoxon Rank Sum (WRS) Test The two-sample WRS statistical test is used when the radionuclide of concern appears in background, or when a measurement method is used that is not radionuclide-specific. Because gross activity measurements are not radionuclide-specific, they should be performed for both the survey unit being evaluated by the WRS test and for the corresponding background reference area.

The WRS test is applied as follows:

1. Adjust the background reference area measurements by adding the DCGLW to each background reference area measurement, Xi (Xi + DCGLW).
2. Sum the number of adjusted background reference area measurements, m, and the number of survey unit measurements, n, to obtain N (N = m + n).
3. Pool and rank the measurements in order of increasing size from 1 to N. If several 5-60

SNEC FACILITY-LICENSE -TERMINATION PLAN I REVISION 3 REVISION 3 SNEC FACILITY LICENSE TERMINATION PLAN

5. Survey measurement locations and corresponding survey data
6. Survey unit investigations performed with documented results, as applicable
7. Any survey unit data assessment results
8. Results of any special measurements performed for the survey unit 5.7.2 Final Survey Report A final survey report (Reference 5-22) will be prepared and submitted to the NRC. The report will provide a summary of any ALARA analysis, survey data results, and overall conclusions, which demonstrate that the SNEC Facility and site meet the radiological criteria for unrestricted use.

Information such as the number and type of measurements, basic statistical quantities, and statistical test results will be included in the report.

The following outline illustrates a general format that may be used for the final status survey report.

The outline below may be adjusted to provide a clearer presentation of the information. The level of detail will be sufficient to clearly describe the final status survey program and certify the results.

Information to be submitted (Reference 5-4, Section 14.5):

1. A summary of the results of the final status survey.
2. A discussion of any changes that were made in the final status survey from what was proposed in the LTP or other prior submittals.
3. A description of the method by which the number of samples were determined for each survey unit (see Reference 5-5, Section 5.5.2).
4. A summary of the values used to determine the numbers of samples and a justification for these values (see Reference 5-5, Section 5.5.2).
5. Survey results for each survey unit including:
  • Number of samples taken for the survey unit.
  • A map or drawing of the survey unit showing the reference system and random start systematic sample locations for Class 1 and 2 survey units, and random locations shown for Class 3 survey units and reference areas.
  • Measured sample concentrations.
  • Statistical evaluation of the measured concentrations (see Reference 5-5, Section 8.3, 8.4 and 8.5).
  • Judgmental and miscellaneous sample data sets reported separately from those samples collected for performing the statistical evaluation.

5-63

SNEC FACILITY LICENSE TERMINATION PLAN REVISION I

  • Discussion of anomalous data including any areas of elevated direct radiation detected during scanning that exceeded the investigation level or measurement locations in excess of the DCGLw.
  • A statement that a given survey unit satisfied the DCGLw and the elevated measurement comparison if any sample points exceeded the DCGLw.
6. A description of any changes in initial survey unit assumptions relative to the extent of residual radioactivity.
7. When a survey unit failed, a description of the investigation conducted to ascertain the reason for the failure and a discussion of the impact that the failure has on the conclusion that the facility was ready for final radiological surveys.
8. If a survey unit failed, a description of the impact that the reason for the failure has on other survey unit information.

5.7.3 Other Reports If requested by the NRC, computer-generated and/or summary data reports will be provided in hard copy or electronic form. Survey data include date, instrument, location, type of measurement, and mode of instrument operation. Other data, such as conversion factors, background reference areas, and the MDCs used, are available which will allow independent verification of the results.

Measurement results will also be presented graphically. The FSS report will be independently reviewed.

Any independent verification survey performed will be performed by an organization outside the SNEC Facility staff and management organization. Reports generated as a result of any independent verification survey process initiated by the SNEC Facility, will be available upon request.

5-64

SNEC FACILITY LICENSE TERMINATION PLAN REVISION II REVISION SNEC FACILITY LICENSE TERMINATION PLAN defined by existing facility physical features, such as a room, intersection of walls, column-and-row layout of a floor elevation, or structural I-beams.

24. Survey Location - In a structural or open land survey area, a survey location is usually represented by a single grid block. In a system survey area, a specified length of piping or a component such as a valve or tank is referred to as a survey location. A survey location can contain one or more survey points. Also referred to as measurement locations.
25. Survey Unit Release Record - A collection of information in a standardized format for controlling and documenting field measurements taken for the Final Status Survey. A survey unit release record is prepared for each survey area. The survey unit release record may include the survey instructions, a control form, grid map(s), survey measurement data sheets and survey maps. It may also be called a survey package.
26. Survey Point - A smaller subdivision within a survey location (grid block, system, component) where local measurements are taken. For structures and systems, a survey point generally refers to an area covered by a detector, or an area of 100 cm2 when a smear is taken. For open land areas, a survey point refers to the area covered by a detector (for paved surfaces), the point at which a dose rate measurement is taken, or the point at which a soil or pavement sample is collected.
27. Survey Unit - A geographical area consisting of structures or land areas of specified size and shape at a remediated site for which a separate decision will be made whether the unit attains the site-specific reference-based cleanup standard for the designated pollution parameter. Survey units are generally formed by grouping contiguous site areas with a similar use history and the same classification of contamination potential. Survey units are established to facilitate the survey process and the statistical analysis of survey data.
28. Total Effective Dose Equivalent (TEDE) - The sum of the deep dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).
29. Unity Rule - Where more than one radionuclide is present, the sum of the ratios of each radionuclide concentration to its respective DCGL should not exceed unity.

When this method is used, the effective DCGL is equal to one (1).

5-67

-qMFC FACILITY LICENSE TERMINATION PLAN REVISION 2 REVISION 2 SNEC FACILITY LICENSE TERMINATION PLAN

5.9 REFERENCES

5-1 Code of Federal Regulations, Title 10, Part 50.82, 'Termination of License" 5-2 Regulatory Guide 1.179, 'Standard Format and Content of License Termination Plans for Nuclear Power Reactors," January 1999 5-3 Code of Federal Regulations, Title 10, Part 20.1402, 'Radiological Criteria for Unrestricted Use" 5-4 NUREG-1727, "NMSS Decommissioning Standard Review Plan", September 2000.

5-5 NUREG-1575, Revision 1, 'Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)," August 2001 5-6 NUREG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys" 5-7 SNEC Facility Site Characterization Report, May, 1996 5-8 NUREG/CR-5512, 'Residual Radioactive Contamination From Decommissioning, Final Report," Volume 1, October 1992 5-9 Draft NUREG-1549, "Using Decision Methods for Dose Assessment to Comply With Radiological Criteria for License Termination," July 1998 5-10 Yu, C. F. et al., Manual for Implementing Residual Radioactivity Materials Guidelines Using RESRAD, Environmental Assessment Division, Argonne National Laboratory 5-11 Yu, C. F. et al., RESRAD-Build, A Computer Model for Analyzing the Radiological Doses Resulting from the Remediation and Occupancy of Buildings Contaminated with Radioactive Material. Environmental Assessment Division, Argonne National Laboratory 5-12 Regulatory Guide 4.15, "Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment" 5-13 SNEC Procedure, 1000-PLN-3000.05, "SNEC Facility Decommissioning Quality Assurance Plan" 5-14 SNEC Procedure, E900-PLN-4542.01, "SNEC Radiation Protection Plan" 5-15 SNEC Procedure, E900-ADM-4500.44, "SNEC Facility Calculations" 5-16 SNEC Procedure, E900-ADM-4500.04, "SNEC Records Retention Procedure" 5-17 SNEC Procedure, E900-ADM-4500.12, "Radiological Surveys: Requirements &

Documentation Procedure" 5-68

SNEC FACILITY LICENSE TERMINATION PLAN REVISION 3 5-18 NUREG-1507, 'Minimum Detectable Concentrations With Typical Radiation Survey Instruments for Various Contaminants and Field Conditions," June 1998 5-19 SNEC Facility Historical Site Assessment Report, January 2000 5-20 Deleted

- 5-21 SNEC Procedure, E900-ADM-4500.59, "Final Status Survey Planning and DQA" 5-22 SNEC Procedure, E900-ADM-4500.60, 'Final Status Survey Report' K2.

5-69

SNEC FACILITY LICENSE TERMINATION PLAN SNEC FACILITY LICENSE TERMINATION PLAN REVISION II REVISION APPENDIX 5-1 ELEVATED MEASUREMENT COMPARISON (EMC)

The EMC, sometimes called a 'hot spot test," is a simple comparison of measured values against a limit. There are two applications of this comparison in the final survey process. It is used when the sensitivity of the scanning technique is not sufficient to detect levels of residual radioactivity below the DCGL (i.e., where the MDCscn is greater than the DCGL). In this application, the number of static measurements may need to be adjusted. Appendix 5-2 describes how this is done. The second application in this appendix, is when one or more scan or static measurement data points exceed the DCGL. The use of the EMC for measurements above the DCGL provides assurance that unusually large measurements receive the proper attention and that any area having the potential for significant dose contributions is identified.

The EMC is intended to flag potential failures in the remediation process.

Locations, identified by scan or static measurements, with levels of residual radioactivity, which exceed the DCGL, are investigated (see Section 5.4.4). The size of the area where the elevated residual radioactivity exceeds the DCGL and the level of the residual radioactivity within the area are determined. The average level of residual radioactivity is then compared to the DCGLEMC. If a background reference area is to be applied to the survey unit, the mean of the background reference area measurements may be added to the DCGL or the DCGLEMC to which the average level of residual radioactivity is compared.

The DCGLEMC is calculated using the following equation (NUREG-1 575, Equation 8-1):

DCGLEMC = Area Factorx DCGL The area factor is the multiple of the DCGL that is permitted in the area of elevated residual radioactivity without requiring remediation. The area factor is related to the size of the area over which the elevated residual radioactivity is distributed. That area, denoted AEMC, is generally bordered by levels of residual radioactivity below the DCGL, and is determined by the investigation. The area factor is the ratio of dose per unit area or volume for the default surface area for the applicable dose modeling scenario to that generated using the area of elevated residual radioactivity, AEMC. It is calculated based on the methodology given in chapter 8 of NUREG-1505 (Reference 5-6).

If the average level of the elevated residual radioactivity is less than the DCGLEMC, there is reasonable assurance the site release criteria is still satisfied and the area does not require remediation. Radioactivity at the DCGLEMC distributed over the area AEMc delivers the same calculated dose as does residual radioactivity at the DCGL distributed over the default surface area. If the DCGLEMC is exceeded, the area is remediated and resurveyed. Area factors for open land areas at the SNEC Facility are provided in Table 5-15. Area factors for surface area DCGLs supplied by the NRC are provided in Table 5-15A.

5-70

Retrieved from "https://kanterella.com/w/index.php?title=ML040900350&oldid=2624457"