ML24292A030
| ML24292A030 | |
| Person / Time | |
|---|---|
| Site: | NS Savannah |
| Issue date: | 10/16/2024 |
| From: | Koehler E US Dept of Transportation, Maritime Admin |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LAR 2023-01 | |
| Download: ML24292A030 (1) | |
Text
0 U.S. Department of Transportation Maritime Administration Office of Ship Operations 1200 New Jersey Ave., SE Washington, DC 20590 Ref: 10 CFR 50.82 and 50.90 October 16, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Docket No. 50-238; License No. NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01 Response to the Second Request for Additional Infonnation
References:
(a)
Email from Tanya Hood (NRC) to Koehler, Erhard (MARAD), dated September 16, 2024, NUCLEAR SHIP SAVANNAH-Request/or Additional Information Re:
Submittal and Request for Approval of the License Termination Plan (EPID L-2023-LLA-0151)
(b)
Letter from Mr. Erhard W. Koehler (MARAD) to U.S. Nuclear Regulatory Commission (NRC), dated October 23, 2023, License Amendment Request No.
LAR 2023-01, Submittal and Request for Approval of the License Termination Plan (c)
Letter from Mr. Erhard W. Koehler (MARAD) to U.S. Nuclear Regulatory Commission (NRC), dated June 27, 2024, Response to Requests for Additional Information Pursuant to 10 CFR 50.90, the United States Maritime Administration (MARAD) hereby submits its responses to the September 16, 2024, Request for Additional Information (RAJ) (Reference a). In Reference (b), MARAD submitted a license amendment for the Nuclear Ship SAVANNAH (NSS). The proposed amendment would modify the license to add License Condition 2.C.( 4) to include License Tennination Plan (LTP) requirements to the Nuclear Regulatory Commission (NRC) NS-I license for the NSS.
In Reference (c), MARAD responded to the NRC's first round of RAls that resulted from reviewing Reference (b) After the NRC reviewed the infonnation submitted in Reference (c), the NRC determined that a second round of additional information is required to complete its review. MARAD's response to the September 16, 2024, RAls is provided in Enclosure 1 to this letter. Attachment I of Enclosure 1 clarifies how the Radionuclides of Concern (ROCs) were detennined.
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 If there are any questions or concerns with any issue discussed in this submittal, please contact me at:
0 : (202) 366-2631, M: ( 410) 776-8268, and/or e-mail me at erhard.koehler@dot.gov.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on October 16, 2024.
Enclosure Respectfully,
~ -------
Erhard W. Koehler Senior Technical Advisor, N.S. SAVANNAH Office of Ship Operations 2
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024
Enclosure:
- 1.
Response to the Second Request for Additional Information 3
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 cc:
Electronic copy NSS ESC NSS SRC MAR 610, 612, 615 Hardcopy, cover letter only MAR-600, 640, 640.2 Hardcopy w/ all enclosures MAR-I 00, 640.2 (rt)
EK/jmo USNRC (Tanya Hood, Andrew Taverna)
USN RC Regional Administrator - NRC Region I MD Department of the Environment (Eva Nair) 4
0 U.S. Department of Transportation Maritime Administration Office of Ship Operations 1200 New Jersey Ave., SE Washington, DC 20590 Docket No. 50-238; License No. NS-1; N.S. SAVANNAH ENCLOSURE 1 RESPONSE TO THE SECOND REQUEST FOR ADDITIONAL INFORMATION 1.0 Introduction.................................................................................................................................... 6 2.0 RAI 5-1: Minimum Detectable Concentration (MDC) Basis:.................................................... 6 3.0 RAI 5-2: Quality Assurance and Quality Control Measures for ISOCS Basis:....................... 7 4.0 RAJ 5-3: DCGL-elevated measurement comparison (DCGLEMc) and Area Factors Basis:... 8 5.0 RAI 5-4: Sensitivity Index Basis:................................................................................................ 11 6.0 RAJ 5-5: Scan and Fixed MDC Measurements Basis:.............................................................. 11 7.0 RAJ 5-6: Water and Sludge Remaining at License Termination Basis:................................. 17 8.0 RAJ 5-7: Samples of and Compliance Demonstration for Pipes and Systems Basis:............ 17 9.0 RAI 5-8: Remaining Dismantlement Activities Basis:.............................................................. 21 10.0 RAJ 5-9: Operational Checks Basis:.......................................................................................... 21 11.0 RAJ 5-10: Survey Unit Failure Basis:......................................................................................... 22 12.0 RAJ 5-11: Ambient Background Subtraction Method Basis:.................................................. 23 13.0 RAI 5-12: ALARA Analysis Basis:............................................................................................. 24 14.0 RAI 10-1: NRC Approval ofLTP Changes Basis:.................................................................... 27 15.0 References..................................................................................................................................... 27.
Clarification of how the Radionuclides of Concern (ROCs) were determined.... 29 I
5
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024
1.0 INTRODUCTION
On October 23, 2023, the United States Maritime Administration (MARAD) submitted a license amendment request (LAR) for the Nuclear Ship SAVANNAH (NSS) (Reference a). The purpose of the LAR was to modify the NS-1 License to add License Condition 2.C.(4) to include License Termination Plan (LTP) requirements.
On June 27, 2024, MARAD provided responses to eighteen requests for additional information (RAis)
(Reference b). The NRC reviewed the information submitted in References (a) and (b) and determined that additional information is required to complete its review. The response to those RAls is detailed below. Each RAI includes the basis, issue and question, which is followed by our response.
2.0 RAI 5-1: MINIMUM DETECTABLE CONCENTRATION (MDC) BASIS:
The regulations in 10 CFR 50.82 (a)(9)(ii)(D) requires detailed plans for the final radiation survey to be included in the L TP. The regulations in IO CFR 20.150 I requires adequate survey be performed to understand and know the site's radiological condition. Applicable guidance in NUREG-1575, Revision I, states that instrument sensitivities should be I 0% to 50% of the Derived Concentration Guideline Levels (DCGLs). The description of survey methods in the final status survey (FSS) plan should provide sufficient detail so that NRC staff can verify the MDCs are accurate and that the methods to be employed will be consistent with the stated MDCs.
Issue:
Section 5.7.1.2, "Fixed Measurements," of the LTP states "Fixed measurements are taken by placing the instrument at the appropriate distance above the surface, taking a discrete measurement for a pre-determined time interval, and recording the reading." Section 5.7.2.4.2, "MDCs for Beta-Gamma Scan Surveys for Structure Surfaces," of the L TP states "[i]n the case of a scan measurement, the counting interval is the time the probe is over the source ofradioactivity. This time depends on scan speed, the size of the source, and the fraction of the detector's sensitive area that passes over the source; with the latter depending on the direction of probe travel." The NRC staff notes that methods for scanning and fixed measurements described in the L TP do not include sufficient information to show how the MDCs in Table 5-4, "Available Instruments and Associated MDCs," would be met. Without an adequate understanding on the specifics of the distance of the detector to the surface, scanning speed (in terms of detector width per second), or a minimal time interval to which the operator will set an instrument scaler for fixed measurements, the NRC staff cannot verify methods employed will be consistent with the stated MDCs.
Request:
Clarify the methodology used for scanning measurements specific to the detector distance from surfaces and scanning speed to achieve the MDCs in Table 5-4 of the LTP. Provide details of detector distance and scaler settings for fixed measurements to achieve the MDCs in Table 5-4.
Response
RAJ 5-5 is similar in nature to RAI 5-1. See notes 1-6 at the end of the Table 5-4 in the response to RAJ 5-5. Those notes describe the detector distances from the surfaces and the scanning speeds to achieve the MDCs.
6
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 3.0 RAI 5-2: QUALITY ASSURANCE AND QUALITY CONTROL MEASURES FOR ISOCS BASIS:
The regulations in IO CFR 50.82 (a)(9)(ii)(D) requires detailed plans for the final radiation survey to be included in the LTP. The regulations in 10 CFR 20.150 I requires adequate surveys be performed to understand and know the site's radiological condition. Applicable guidance in NUREG-1507 discusses in situ gamma spectroscopy and potential areas of significant measurement uncertainty that could be addressed through a measurement program's quality assurance (QA) and quality control (QC) process.
Issue:
Section 5.11, "Quality Assurance and Quality Control Measures," of the LTP discusses QA/QC activities for the FSS effort to ensure that surveys are performed by trained individuals using approved written procedures and properly calibrated instruments that are sensitive to the suspected contaminant. There is no discussion of QA for in situ object counting system (!SOCS) measurements including training and experience requirements for the ]SOCS operator and verification of results to confirm radionuclides/depth profiles of contamination match the modeled configuration.
Request:
Clarify the QA process for use of ISOCS measurements. Describe the training and experience requirements of individuals using the ISOCS system. As well as, how the !SOCS results are verified including the number of duplicates measurements and how the radionuclides/depth profiles of contamination are confinned to match the modeled configurations.
Response
The operation of the !SOCS system is described by SAVANNAH Technical Staff (STS) procedure O1-CANBERRA ISOC-01, Operation of Canberra in Situ Object Counting Systems (Reference c). That procedure specifies that quality control checks be performed prior to collecting measurements and following the completion of work at the end of the day. The results of these checks are recorded in the Final Status Survey Package paperwork.
The individuals using the I SOCS system are provided by MARAD's joint venture decommissioning integrated services provider, Nuclear Ship Support Services, LLC (NSSSN).
NSSSN competes for these individuals within the wider domestic active decommissioning market. Given the large number of such projects ongoing, readily qualified personnel are in short supply; not just for !SOCS operation, but for many FSS activities. To address this shortfall, NSSSJV sourced FSS technicians from the population of highly trained Radiation Protection Technicians (RPTs) that service operating plants. To prepare these individuals to perform FSSs, a specialized three-day training course for Final Status Survey technicians, identified as the FSS Academy, was developed and deployed by NSSSN partner Radiation Safety and Control Services, lnc. (RSCS) in early 2024. The three instructors for this course were health physicists certified by the American Board of Health Physics (ABHP), and each have over thirty years of professional health physics experience including ISOCS modeling. The three instructors are Subject Matter Experts. The two day classroom portion was designed for managers, supervisors and technicians. It discussed an overview of MARSSIM, the NSS LTP and implementation of Final Status Survey packages..
On the third day, Final Status Survey technicians received specialized training on the use of instrumentation, including the !SOCS. The technicians were taught how to perfonn a quality control check as well as how to collect a measurement. The technicians only perform these two 7
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 tasks. The collected spectrum, along with a picture of the measurement surface and the distance from the detector to the measurement smface are sent to a qualified subject matter expert to generate the efficiency model in the Geometry Composer software and perform the analysis.
Academy RPT participants are required to pass a written exam and practical factors demonstration before qualifying as FSS Technicians.
Duplicate !SOCS measurements would only be perfonned when the entire data set in a survey unit was from !SOCS measurements. Currently, the !SOCS is being used to investigate or verify radioactivity concentrations causing an elevated gross activity measurement. There are no current plans to perform an entire I SOCS based survey in a survey unit. If that becomes the case, then a duplicate measurement will be prescribed in the Final Status Survey Package per STS-005-030, Preparation of FSS Packages (Referenced). If the JSOCS measurement results are grossly different than the fixed measurements, the Health Physicist performing the comparison would adjust the model parameters to reach better agreement or to reassess or re-perform the fixed measurements and/or scans as part of the data assessment process.
The depth profile analysis is a common technique that is applied to volumetric residual radioactivity in concrete for land-based nuclear power plants. For the NSS, residual radioactivity is limited to metal surfaces and is not distributed to depths within the metal surface to an extent where modeling differences would be manifested.
4.0 RAI 5-3: DCG L-ELEV A TED MEASUREMENT COMPARISON (DCGLEMc) AND AREA FACTORS BASIS:
The regulations in IO CFR 20.1402 requires radiological license termination criteria not to exceed 25 mrem/year and that the residual radioactivity be reduced to levels that are as low as reasonably achievable (ALARA). NUREG-1757, Volume 2, and NUREG-1575, Revision I, provide guidance on methods for demonstrating compliance with these criteria including what can be considered appropriate investigation levels.
Issue:
In its letter dated May 30, 2024 (ML24 l 57 A I 03), the NRC staff asked in RAJ 5-3 for the licensee to confirm plans to remediate materials with measurements above the DCGLw until residual radioactivity is less than the DCGL-wide (DCGLw) or provide instructions on how any elevated measurements will be addressed in demonstrating compliance with IO CFR 20.1402.
In its RAI response dated June 27, 2024 (ML24 I 83A27 I), MA RAD stated, "All external surfaces above the DCGLw have been or will be remediated to less than the DCGLw. Any internal surface in a system above the DCGLw will be remediated or removed and properly disposed. Our investigation level is set at 75% of the DCGLw which gives us confidence that measurement locations greater than the DCGLw will either be remediated or removed." Section 5.5.3.3, "Remediation," of the LTP, states "If during the performance of an FSS, any areas of residual activity are found to be in excess of the DCGLw and an outlier, those areas will be remediated with the goal to reduce the activity to less than or equal to the DCGLw.
Remediation actions are discussed in Section 4 and documented as described in Section 5.1 0." In addition, Table 5-2, "Investigation Levels," of the L TP identifies investigation levels for fixed and scanned measurements for the three survey unit classifications and systems and components. Table 5-2 states that for a Class 3 survey unit, an investigation will be performed for fixed and scanned 8
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 measurements if the results are greater than 25% of the operational DCGLw. The NRC staff notes that the licensee response to RAJ 5-3, is inconsistent with Table 5-2 and Section 5.5.3.3 of the LTP.
Request:
Clarify your June 27, 2024, response for RAI 5-3, with regards to investigation levels as compared to that stated in Table 5-2 and described in Section 5.5.3.3, of the LTP.
Response
The clarified response for RAJ 5-3 in Reference bis presented below.
All external surfaces above the DCGLw have been or will be remediated to less than the DCGLw.
Any internal surface in a system above the DCGLw will be remediated, or the component will be removed and properly disposed. Our investigation level is set at 75% of the DCGLw, which is identified as the Operational DCGLw in LTP Table 5-2. The Investigation Level for a Class 3 survey unit has been further reduced to 25% of the Operational DCGLw These Investigation Levels give us confidence that measurement locations greater than the DCGLw will either be remediated or removed.
Additional changes to the L TP are needed to complete the original response. The additional changes for RAI 5-3 are presented below.
Table 5-2 Investigation Levels (Marked up)
Survey Unit For fixed measurements, perform For scan measurements, perform Classification investigation if:
investigation if:
Class 1
> Operational DCGLw aHe-afl
> Operational DCGLw etttttef:
Class 2
> Operational DCGLw
> Operational DCGLw Class 3
> 0.25 x Operational DCGLw
>0.25 x Operational DCGLw Systems and
> Operational DCGLw
> Operational DCGLw Components Paragraph immediately following Table 5-2 (Marked up):
For Class l survey units, measurements above the DCGLw are not necessarily unexpected.
HoweYer, sueh a result 1nay still indieate a seed fer further ienstigatioA if it is sigAifieaetly different than the other measuremeets made within the same s\\:lrvey unit. Th\\:ls, some additi01rnl e¥aluation eriterioe is Reeded to assess if res\\:llts from fo.ea measuremeets ie a Class 1 Sl1/4FVe)'
ueit that m.eeed the DCGLw warraet further atteetioe. Measurements in Class 1 survey units that exceed the DCGLw aea differ from the mean of the remaieieg measuremeets l=>y more thae three
(> 3) standard deYiatioes will therefore be investigated. Measurements iA Class 1 \\:!nits that e~.eeed the DCGLw, l=>ut do not differ from the meae l=>y less thaH or equal to three (:'.S 3) staAdard deYiatioes may still l=>e ievestigated l=>ased OR professioeal judgment, as may aA)' measuremeets that differ sigeifieantly from the rest of the measuremeets made *Nithie a givee survey ueit.
Paragraph immediately following Table 5-2 (Clean):
9
Docket No. 50-238; License NS-1; N.S. S AVANNA H License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 For Class I survey units, measurements above the DCGLw are not necessarily unexpected.
Measurements in Class I survey units that exceed the DCGLw will be investigated.
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01,,Response to the Second Request for Additional Information October 16, 2024 5.5.3.3 Remediation (Marked up)
If during the performance of an FSS, any areas of residual activity are found to be in excess of the DCGLw tmd an outlier, those areas will be remediated with the goal to reduce the activity to less than the DCGLw. Remediation actions are discussed in Section 4 and documented as described in Section 5.10.
5.0 RAI 5-4: SENSITIVITY INDEX BASIS:
The regulations in 10 CFR 50.82 (a)(9)(ii)(D) requires detailed plans for the final radiation survey to be included in the L TP. The regulations in 10 CFR 20.1501 requires adequate surveys be performed to understand and know the site's radiological condition. NUREG-1575, Revision 1, provides guidance on determining scan MDC which incorporates use of a sensitivity index (d') as a factor in the scan MDC equation.
Issue:
Section 5.7.2.4.2, "MDCs for Beta-Gamma Scan Surveys for Structure Surfaces," of the LTP, states that the sensitivity index of 1.38 was chosen for use in the scan MDC Equation 5-8. The NRC staff notes that significant training is needed to ensure appropriate surveyor techniques and appropriate sensitivity to elevations in count rate to ensure the value of 1.3 8 is appropriate.
For informational purposes, NRC's contractor, who helped author MARSSIM methods in this technical area, typically utilizes 2.32 as the d' value in their scan MDC determinations.
Request:
Provide justification for using a sensitivity index of 1.38 in the scan MDC equation.
Response
MARAD reviewed MARSSIM and determined that a sensitivity index (d') factor of 1.38 is appropriate for our use. This is a conservative administrative decision and the training we conducted with the FSS technicians included field scanning using simulation technology which stressed that the scanning should be reversed and slowed for areas where an elevated response is perceived. MARAD's use of 1.38 for the sensitivity index is consistent with our FSS training ln addition, use of 1.38 has been an industry standard in nuclear power plant License Termination Plans such as Humboldt Bay, Zion, LaCrosse, Fort Calhoun, Yankee Rowe, Connecticut Yankee, and Maine Yankee (i.e., it is a precedent) and is used as the performance goal in an example calculation of the minimum detectable count rate in MARSSIM.
6.0 RAI 5-5: SCAN AND FIXED MDC MEASUREMENTS BASIS:
The regulations in 10 CFR 50.82 (a)(9)(ii)(D) requires detailed plans for the final radiation survey to be included in the LTP. The regulations in 10 CFR 20.1501 requires adequate surveys be performed to understand and know the site's radiological condition. NUREG-1575, Revision 1, provides guidance on determining both the scan MDC and fixed measurement MDC which incorporates use of both an instrument efficiency and a source efficiency as a denominator in the equations.
Issue:
Table 5-4 of the LTP, "Available Instruments and Associated MDCs," includes information on instrumentation, application, nominal efficiency, background, scan and fixed MDC measurements. The 11
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 NRC staff notes that Table 5-4 does not indicate if source efficiency corrections were included in the nominal fixed and scan MDC values. In addition, source efficiency values for maximum beta energies between 150-400 keV are not indicated. When utilizing the default source efficiency for low energy beta radiation and a sensitivity index of 2.32, it appears that the Geiger-Mueller pancake scanner may be unsuitable for use for FSS scanning.
The NRC staff notes that both the MDC values presented for the High Purity Germanium (HPGe) detector is in units of pCi/g (which is inconsistent with the proposed DCGL units) and that certain instruments mentioned in the LTP which are suitable for scanning or other use in this situation are not listed.
Request:
a) Clarify if the nominal MDC and nominal Scan MDC values in Table 5-4 include source efficiency correction. If Table 5-4 nominal MDC and Scan MDC values do not include a source efficiency, recalculate the values to include the source efficiencies. ln addition, recalculate the nominal MDC and Scan MDC in Table 5-4 with the appropriate sensitivity index value from your response to RAJ 5-4.
Response
See the response to RAI 5-5(b).
b) Provide the correction/correlation of the nominal MDC values for HPGe radiation detectors in dpm/ 100cm2.
Response
The static and scan MDCs in the updated Table 5-4 were calculated with the total weighted efficiencies as documented in STS report CR-164, Calculation of Weighted Efficiencies and MDCs of Ludlum Detectors for Final Status Surveys on the NS SA VANNAH (Reference e ). The efficiencies used to calculate the MDCs in Table 5-4 are based upon the total efficiency, not with the source and instrument efficiencies. MARSSIM states that the use of a total efficiency derived from measurements made on certified 4n activity traceable sources "... is not a problem, provided that the calibration source exhibits characteristics similar to the surface contamination (i.e.,
radiation energy, backscatter effects, source geometry, self-absorption)." Those four parameters were evaluated in CR-164 with the conclusion that the calibration sources exhibit characteristics similar to the surface contamination on the ship. Table 5-4 has been revised to list the MDC for the ISOCS system in dpm/100 cm2. The JSOCS system is a 2"x 2" NaJ detector connected to on Osprey MCA with a 90-degree collimator. The evaluation of the I SOCS for use in Final Status Surveys has been documented in STS report CR-167, Use of In-Situ Gamma Spectroscopy for Final Status Survey (FSS) at the NS SA VANNAH (Reference t).
c) If pipe crawlers are planned to be used for FSS surveys, update Table 5-4 with the appropriate information.
Response
A pipe crawler, per se, will not be used for Final Status Surveys. The detector chosen for internal pipe and duct measurements for FSS is a Ludlum 44-159-1, which is a 0. 7 inch diameter Csl scintillator. The MDCs listed in Table 5-4 were taken from STS report CR-165, Efficiency and MDC of Ludlum Pipe Detectors for Final Status Survey (FSS) on the NS SAVANNAH (Reference g).
12
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 d) Explain how/when Sodium Iodide (Nal) detectors may be used for scanning purposes and update Table 5-4, if appropriate.
Response
Sodium Iodide (Nal) detectors are used for scanning during Remedial Action Support Surveys (RASS) to identify elevated locatidns in a survey unit. The only NaJ detector used in FSS is the
!SOCS.
The following table presents the updated Table 5-4.
13
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 Table 5-4 Available Instruments and Associated MDCs (Marked up)
Instrument Application Nominal Nominal Nominal MDC (fixed Effieieee,*
Background
measurement)
Paneake GM Model beta gamma seans 10% (Go 60) 50 e13m 2393 dpm/1 00emi 44-9 or fi*ed (1 minute eount) measurements f+3/4Hi)
Zn8 sei-Atillator beta gamma seans 12% (Go 60) 150 e13m 4 00 d13m/l 00emi ( 1 minut1 Model 43 89 or fowd setffit) measurements (125emi)
ZnS scintillator beta-gamma scans 15% (Go 60) 150 cpm 394400 dpm/100cm 2--f}---HH Ludlum Model 43-or fixed setffit) 93 (100cm2) measurements Note 1 Gas Flow beta-gamma scans 15% (Go 60) 350 cpm NIA Proportional Counter Floor monitor Ludlum Model 43-37 (584 cm2)
HPGe in-situ gamma
- \\laries >,1,*ith
¥aries with 0.05 pGi/g Go 60 2" x 2" Nal detector spectroscopy energy and energy and 0.05 13Gi/g Gs 137 geometry geometr;*N/A connected to OsQrex (10 minute eounts)Co-60 MCA 2.20E3 dQm/100 cm2 Cs-137 2.15E3 dQm/100 c1 Note4
Docket No. 50-238; License NS-1 ; N.S. S AVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 Instrument Application
~
Effieieeey Nominal
Background
Nominal MDC (fixed measurem ent)
Nominal Scan MDC Pesitien sensiti,*e gas pfeper4ienel
~
esium Iodide {Cs])
scintillator Ludlum Model 44-159-1 SGIIH IIR0 f@G0 f0
~
gamma scans or fixed measurements inside pipes and ducts 18% (Ge liQ) 100 cpm 1310 dpm/ 100 cm2 Note 5 Typisel,*!!lye 1,92,§ Elpffl/ 1 QQsff!~,---+H6 MDC *,*Elfies willi Eleteeter speeEI !!REI ffi5laHee, 3 63 0 dpm/100~
Note 6 General notes: Efficiencies of the Ludlum 43-93 and 43-37 detectors have been determined with 2 large area NIST certified Tc-99 and Cs-137 calibration sources. Efficiencies of the ISOCS detector are determined with Geometry Composer software. Efficiencies of the Ludlum 44-159-1 detector have been determined by MicroShield and empirical testing.
Note I: detector in contact with surface, scalar setting equal to I-minute count time for both background and measurement, total weighted efficiency equal to 0.152 c/d.
Note 2: detector at 0.75 inches from the surface, circular hot spot area equal to 100 cm2, scan speed equal to I detector width per second (9. 14 cm/s), total weighted efficiency equal to 0. 104 c/d.
Note 3: detector at 0. 75 inches from the surface circular hot spot area equal to I 00 cm 2, scan speed equal to I detector width per second (15.9 cm/s) total weighted efficiency equal to 0.127 c/d.
Note 4: source to detector distance equal to I meter, detector with 90-degree collimator, circular plane geometry, source thickness equal to 2 mm, count time equal to I 000 seconds.
Noie 5: the Field of View (FOY) and source length is equal to I foot, detector in contact with bottom of pipe or duct, scalar setting equal to I-minute count time for both background and measurement, total weighted efficiency equal to 0.03 77 c/d for a 1.5 inch diameter pipe.
Note 6: the Field of View {FOY) and source length is equal to I foot, detector in contact with bottom of pipe or duct, scan speed equal to 0.25 mis, total weighted efficiency equal to 0.0377 c/d for a 1.5 inch diameter pipe.
15
Docket No. 50-238; License NS-I; N.S. S AVANNA H License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 Table 5-4 Available Instruments and Associated MDCs (Clean)
Instrument Application Nominal Nominal MDC (fixed
Background
measurement)
ZnS scintillator Ludlum Model beta-gamma scans or 150 cpm 394 dpm/100cm2 Note I 43-93 ( 100cm2) fixed measurements Gas Flow Proportional Ludlum beta-gamma scans 350 cpm N/A Model 43-37 (584 cm2) 2" x 2" Na! detector connected in-situ gamma N/A Co-60 2.20E3 d12m/l 00 cm2 to Os12rey MCA spectroscopy Cs-137 2. 15E3 dgm/ 100 cm2 Note4 Cesium Iodide (Csl} scintillator gamma scans or fixed JOO cgm 13 IO dgm/100 cm2 Note 5 Ludlum Model 44-1 59-1 measurements inside giges and ducts Nominal Scan MDC 1603 dpm/100cm2 Note2 452 dpm/100 cm2 Note3 N/A 3630 dgm/100 cm2 Note 6 General notes: Efficiencies of the Ludlum 43-93 and 43-37 detectors have been detem1ined with 2 large area NIST certified Tc-99 and Cs-137 calibration sources. Efficiencies of the !SOCS detector are detem1ined with Geometry Composer software. Efficiencies of the Ludlum 44-1 59-1 detector have been determined by MicroShield and empirical testing.
Note I: detector in contact with surface, scalar setting equal to I-minute count time for both background and measurement, total weighted efficiency equal to 0.152 c/d.
Note 2: detector at 0. 75 inches from the surface, circular hot sgot area equal to I 00 cm 2, scan sgeed equal to 1 detector width ger second (9.14 emfs}, total weighted efficiency equal to 0. 104 c/d.
Note 3: detector at 0.75 inches from the surface, circular hot sgot area equal to 100 cm2, scan sgeed equal to I detector width ger second
( 15.9 cm/s} total weighted efficiency equal to 0. 127 c/d.
Note 4: source to detector distance equal to I meter, detector with 90-degree collimator, circular glane geometry, source thickness equal to 2 mm, count time equal to I 000 seconds.
Note 5: the Field ofYiew (FOY} and source length is equal to I foot, detector in contact with bottom ofoige or duct, scalar setting equal to I-minute count time for both background and measurement, total weighted efficiency equal to 0.03 77 c/d for a 1.5 inch diameter gige.
Note 6: the Field ofYiew (FOY) and source length is equal to I foot detector in contact with bottom ofoige or duct, scan sgeed equal to 0.25 mis, total weighted efficiency equal to 0.0377 c/d for a 1.5 inch diameter gige.
16
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 7.0 RAI 5-6: WATER AND SLUDGE REMAINING AT LICENSE TERMINATION BASIS:
The regulations in 10 CFR 20.1402 requires radiological license termination criteria not to exceed 25 mrem/year and that the residual radioactivity be reduced to levels that are ALARA. NUREG-1757, Volume 2, and NUREG-1575, Revision 1, provide guidance on methods for demonstrating compliance with these criteria.
Issue:
Section 5.7.1.4 of the LTP, "Samples," states "If water or sludge is encountered in a system during FSS, sample results will be compared the Effluent Concentrations (ECs) listed in Table 2, Column 2 of Appendix B to IO CFR 20. If the sample results are greater than the ECs, the medium will be remediated or removed." The LTP states that the site contains no surface or ground water. The LTP does not discuss what MARAD plans to do if samples of water or sludge are below the ECs.
The NRC staff notes that there is no discussion in the LTP of how comparison of the effluent concentrations in Table 2, Column 2 of Appendix B to IO CFR 20 will be incorporated into the demonstration of compliance with the 25 mrem/yr TEDE dose limit. The LTP discusses dose from contaminated ship surfaces. If contained water or sludge is encountered in a system during FSS, the NRC staff are concerned that the total compliance dose equation may underestimate dose to an average member of the critical group because it does not currently account for dose from this potential additional source term. This supports insight regarding the risk-significant exposure scenarios related to both the preservation and shipbreaking end states.
Request:
Provide a discussion of how residual radioactivity from water and/or sludge remaining at license termination will demonstrate compliance with 25 mrem/yr.
Response
At license tennination, no reactor plant related water or sediment will remain on the ship.
8.0 RAI 5-7: SAMPLES OF AND COMPLIANCE DEMONSTRATION FOR PIPES AND SYSTEMS BASIS:
The regulations in IO CFR 20.1402 requires radiological license termination criteria not to exceed 25 mrem/year and that the residual radioactivity be reduced to levels that are ALARA. NUREG-1757, Volume 2, and NUREG-I 575, Revision 1, provide guidance on methods for demonstrating compliance with these criteria.
Issue:
Assuming functionality is not an issue, if residual radioactivity may remain in any pipe or system at license termination, then that pipe or system should be sealed and made generally inaccessible in the future by means such as grouting or use of similar materials to fill the void spaces. This will similarly make any future potential exposure much less probable as the pipe/system would then be more likely to be disposed of as waste instead of reused in the future.
Section 3.1.2 "Completed Dismantlement Activities" of the LTP states that the aft Reactor Coolant System piping may remain onsite at license temiination or may be one of the last low level radioactive waste shipments. In the event that this piping system remains onsite and contains residual radioactivity, 17
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 the NRC staff does not have a clear understanding of how the licensee plans to characterize the residual radioactivity nor the criteria MARAD will use to estimate dose to an average member of the critical group from the residual radioactivity to comply with the 25 mrem/yr TEDE dose limit.
Section 4.2.3 "High Pressure Water Blasting" of the LTP states that if the licensee discovers residual radioactivity in a pipe in excess of the release criteria and the system is going to remain, the licensee will perfonn in situ remediation. However, Section 5.6 "Survey Protocol for Non-structural Systems and Components" of the LTP states that "MARAD expects limited use of in-situ gamma spectroscopy or pipe crawlers." Section 5.6 "Survey Protocol for Non-structural Systems and Components" of the LTP states that "Evaluations as to whether I) material should be considered as a structure or a component will be made and 2) comparisons with the dose modeling scenarios used to develop the DCGLs that govern release of structures." The NRC staff are seeking clarification as to how material will be categorized.
The NRC staff are also seeking clarification on how the licensee will account for systems and components containing residual radioactivity at license tennination but not considered in the dose modeling scenarios.
At other sites, all non-structural components are typically removed prior to FSS. If components may be left (e.g., piping), those are considered separate survey units that, once surveyed, are combined with the structural or open land survey unit in which they reside so the SOFs are added when assessing compliance. Such pipes are also typically sealed and grouted to some extent if they can't be fully accessed and surveyed.
Request:
- 1. Provide any updates to the end state configuration plans of any pipe or systems to remain at license tennination, including plans to fill any pipe or system to make [it] inaccessible.
Response
The aft loop of the Reactor Coolant System pipe segments and valves were disposed in early 2024 as noted in STS-224, Annual Report for Calendar Year 2023 (Reference h). In all other respects, the end state at license termination will agree with the descriptions in LTP Sections 3.1, 3.1.1, and 3.1.2, with minor exceptions. Sections 3.1 and 3.1.1. of the L TP cross reference to the STS-100, Post Shutdown Decommissioning Activities Report (PSDAR), Revision 1 (Reference i).
The PSDAR lists systems slated for dismantlement in its Section 3.3. For clarity, Table 8-1,
Summary of System End States at License Termination, provides a description of the end state configuration for the systems covered by the PS DAR, and Sections 3.1 and 3. I. I of the L TP. The minor exceptions are described in the system status column where appropriate. There is no credible reuse scenario for the pipe in the preservation end state; the pipe would be disposed as scrap in the shipbreaking end state; and in the artificial reefing end state, the pipe is submerged and would be nearly impossible to access by divers based on its location deep within the engine room lower level. The tanks discussed in Table 8-1 will be surveyed.
Regarding the sentence, "This system has been removed in its entirety," note that bulkhead penetrations and deck penetrations for some of these systems remain. These systems were cut as close to the bulkhead or deck as possible, leaving the minimum amount of piping on each side of the bulkhead or deck. In numerous instances, the bulkhead was fitted with a sleeve through which the pipe passed. All of these penetrations have been surveyed.
18
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 TABLE 8-1
SUMMARY
OF SYSTEM END STATES AT LICENSE TERMINATION ID System Name Status CRD Control Rod This system has been removed in its entirety.
Drive PE Primary This system has been removed in its entirety with the exception of the Pressurizing retained portions of the pressurizer.
SL Buffer Seal This system has been removed in its entirety.
SC Shutdown This system has been removed in its entirety.
Circulation PR Primary This system has been removed in its entirety.
Relief WL Gaseous This system has been removed in its entirety.
Waste Collection and Disposal cc Containment This system has been removed in its entirety. All that remains is the Cooling support structure for Containment Air Coolers.
PD Equipment Piping for this system has been removed in its entirety. All that will Drain and remain are two tanks integral the ship's double bottom structure. These Waste tanks are PD-TS Contaminated Water Tank Starboard and PD-T6 Collection Contaminated Water Tank Port.
pp Primary Piping for this system has been removed in its entirety.
Loop Purification HA Hydrogen Piping for this system has been removed except as follows:
Addition Two orphan lengths of HA piping will be grouted In the Hydrogen Addition Room, several ancillary devices will remain and include components such as a pressure gauge, tubing, miscellaneous pressure control or monitoring devices, etc.
SP Soluble Piping for this system has been removed in its entirety. All that will Poison remain is the Soluble Poison Tank which has been drained and will be surveyed.
DK Emergency The DK system has been removed forward of the Engine Room (forward Cooling of Bulkhead 126). Heat exchangers and saltwater side of the system will remain intact and will be surveyed.
19
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 ID System Name Status cw Intermediate Piping and components in the engine room will remain. These include the Cooling sea water heat exchangers and connecting piping, and the supply and return Water fresh (intermediate cooling) water connections to the system outside of the engine room. All other piping and components forward of the engine room have been removed except for the Intem1ediate Cooling Water Surge Tank located in the Reactor Compartment Upper Level, which has been drained and will be surveyed. Six short lengths of pipe that looped through the ceiling in the port horseshoe area are cut at each end and orphaned. They will be surveyed.
RSV Reactor This system has been removed in its entirety with the exception of Space ductwork penetrations through various spaces. These penetrations are Ventilation accessible for survey.
WD Waste This system has been removed in its entirety.
Dilution and Disposal SA Primary This system has been removed in its entirety.
Sampling None Clean (salt LTP Chapter 3.1.1 identifies the Fresh Water Shield Tank (FWST) as water) subject to FSS. Although this integral double bottom hull structure tank is Ballast not part of a Reactor Auxiliary System, it was used to store the primary coolant drained from the primary system in the mid-1970s. Similarly, the Reactor Space Void Tank, another integral double bottom hull structure tank which is surrounded by the FWST is not part of a Reactor Auxiliary System. lt was impacted by a leak of the stored primary coolant from the FWST. The Void Tank is subject to FSS.
- 2.
Discuss how material will be considered as a structure or a component possibly including decision criteria.
Response
LTP Section 3.1.1 discusses, in part, the distinction between structures and components, and specifically includes two contaminated waste tanks (PD-TS and PD-T6) which are part of the PD system, as structures, because they are integral to the ship's double bottom hull structure. These contribute as structures in the dose model. LTP footnote 12 adds the Fresh Water Shield Tank (FWST), also an integral double bottom tank, as part of a system. Similar to these, the Soluble Poison Tank is integral to ship structure even though it is classified as a system component.
During operations, the Reactor Space Void Tank contained saltwater ballast. It is not a Reactor Auxiliary System component, but it has been impacted by in-leakage of stored primary coolant from the FWST. The sixth system tank is the surge tank of the Intermediate Cooling Water system. These six tanks are described above in Table 8-1. The retained steam generators, pressurizer shell, and neutron shield tank are all system components. The Containment Ves~el is considered a structure.
The DCGLs provided in Chapter 6 of the LTP apply equally to structures and components (also systems and portions of systems) since the dose model used to develop the DCGLs simply applies a conservative value to the time spent working with material containing residual radioactivity and 20
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 considers the combination of all exposure pathways and selects the lowest DCGL (highest dose) from each of the pathways and radionuclides of concern (ROCs). Therefore, the distinction between structure or components will have no impact on the selection or application of the DCGLs for final status surveys.
- 3. Explain how the residual radioactivity from systems and components not included in the dose modeling scenarios will be accounted for dose compliance with the 25 mrem/yr TEDE dose limit.
Response
See the response to RAJ 5-7, item 2 (above).
9.0 RAI 5-8: REMAINING DISMANTLEMENT ACTIVITIES BASIS:
The regulations in 10 CFR 50.82(a)(9)(ii)(B) requires identification of remaining dismantlement activities. Applicable guidance on identification ofremaining site dismantlement activities can be found in NUREG-1700, Revision 2, Section 2.3, and NUREG-1757, Volume 2, Revision 2, Appendix G.3.5, "Recontamination."
Issue:
Section 5.4.4, "Area Preparation: Isolation and Control," of the LTP, states, "Once the area meets the isolation and control criteria, isolation and control will be achieved through a combination of personnel training, physical barriers, postings, and site notices as appropriate, to prevent unauthorized access to an isolated survey unit. Isolation and control measures will be implemented through approved plant procedures. An administrative process will be used to evaluate, approve ( or deny), and document all activities conducted in these areas during and following FSSs." The NRC staff notes that Section 5.4.4 does not discuss periodic surveillance and surveys to verify that no new residual radioactivity has been reintroduced to a survey unit/area once FSS activities have been completed.
Request:
Provide criteria for surveillance and surveys that must be met and what corrective actions will be taken if criteria are not met. Discuss periodic surveillance and surveys to verify that no new residual radioactivity has been reintroduced to a survey unit/area once FSS activities have been completed.
Response
MARAD procedure STS-005-032, Survey Unit Turnover and Control (Reference j) was revised to require the performance of periodic surveillances on completed survey units.
10.0 RAI 5-9: OPERATIONAL CHECKS BASIS:
The regulations in 10 CFR 20.1501 require appropriate surveys appropriate to evaluate the concentrations and quantities ofresidual radioactivity. NUREG-1575, Revision 1, Sections 4.7.1, 6.2.2.2, and 6.5.4 provide guidance on operational and calibration checks on instruments.
Issue:
Section 5.7.2.3, "Operational Checks," of the LTP states "Instrumentation will be checked for proper operation in accordance with approved procedures. If the instrument operational check does not fall within the established range, the instrument will be removed from use until the reason for the deviation can be resolved and acceptable operation is again demonstrated. If the instrument fails a post-survey 21
Docket No. 50-238; License NS-1; N.S. SA VANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 source check, all data collected during that time period with the instrument will be carefully reviewed and possibly adjusted or discarded, depending on the cause of the failure." Page 8 of 10 Section 13 of the NS SAVANNAH Project STS-003-001, Revision 4 [Decommissioning Quality Assurance Plan] indicates that instrumentation will be properly controlled, calibrated, and adjusted during specified periods to maintain accuracy within necessary limits. NUREG-1575, Revision I, Section 6.2.2.2, "Bias," defines calibration checks as measurements performed to verify measurement performance each time an instrument is used.
While the L TP commitments are likely consistent with guidance in that the licensee commits to routine operational checks, typically operational checks are done daily or per shift, at a minimum, to assure response is within acceptable ranges and that any changes in background are not attributable to contamination of the detector.
Request:
Clarify that operational checks of survey instruments are being conducted daily or per shift when used.
Response
Performance of operational checks is specified in the Final Status Survey Packages per STS-005-030, Preparation of FSS Packages (Reference k). A special fonn for documenting operational checks is included in the survey packages. Operational checks are required on a daily basis and are conducted prior to performing the survey. Additional operational checks are performed during the shift and at the end of the day for the full duration of the survey.
11.0 RAI 5-10: SURVEY UNIT FAILURE BASIS:
Regulations at 10 CFR 20.1402 state a site to be acceptable for unrestricted use if the residual radioactivity results in a TEDE that does not exceed 25 mrem/y and be ALARA. NUREG-1575, Revision I, Section 8.5.3 provides guidance on what should be considered "If the Survey Unit Fails."
Issue:
Section 5.10, "Final Status Survey Release Records and Reports," of the LTP, states, "When a survey unit fails, the FSS report will include the following: a description of any changes in initial survey unit assumptions relative to the extent ofresidual radioactivity; a summary of the investigation conducted to ascertain the reason for the failure; a summary of the effect that the failure has on the conclusion that the facility is ready for final radiological surveys; and, a summary of the effect of the failure has on other survey unit information." The NRC staff note that if a survey unit "fails", then the survey unit cannot be determined to meet the license termination criteria, and thus license termination cannot be granted. If a survey unit fails, then the cause should be evaluated and corrections performed, to possibly include additional remediation and resurveying, in order to demonstrate compliance with the DCGLs and the license termination criteria. Only FSSRs which ultimately demonstrate meeting the release criteria should be submitted to the NRC.
Request:
Clarify actions that will be taken if a survey unit fails.
Response
MARAD understands that only FSSRs which ultimately demonstrate meeting the release criteria will be submitted to the NRC. If a survey unit fails, the cause will be evaluated, and corrections will be performed. This may also include additional remediation and resurveying. These actions are defined in MARAD procedure STS-005-033, Final Status Survey Data Assessment and 22
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 Investigation, Reference (I). After all actions have been taken and completion of a successful FSS, the Final Status Survey Release Record will include a discussion of the failure, any follow-up actions and.the results of the survey data and data analysis from the successful survey.
12.0 RAI 5-11: AMBIENT BACKGROUND SUBTRACTION METHOD BASIS:
The regulations in 10 CFR 50.82 (a)(9)(ii)(D) requires detailed plans for the final radiation survey be included in the LTP. The regulations in 10 CFR 20.150 I requires adequate surveys to be performed to understand and know the site's radiological condition. NUREG-1507, Revision 1, Section 4.5, "Ambient Background Count Rate," provides guidance on ambient background effects and Section 5.1,
"Background Count Rates for Various Materials," further discusses ambient background. In addition, Appendix A, Section A.2, "Methods," discusses some acceptable methods to determining ambient background.
Issue:
Section 5.8.1, "Sign Test," of the L TP, discusses the evaluation of gross activity measurements after subtracting the ambient background from each measurement. Note that NUREG-1575, Revision I, provides guidance that gross measurements should be evaluated using the Wilcoxon Rank Sum test which would likely alleviate ambient background subtraction as a possible issue if performed incorrectly.
Request:
Clarify how ambient background will be appropriately/conservatively determined for each gross activity measurement.
Response
MARAD is ignoring ambient background and is not subtracting ambient background from each gross activity measurement prior to performing the Sign Test. MARAD determined this is a conservative decision because the measurement is biased to a higher value. The following is the revised Section 5.8.1 of the LTP.
5.8.1 Sign Test Radionuclide specific measurements for which the radionuclide(s) of interest either does not exist in background or is not present in a concentration that is a relevant fraction of the DCGLw will be evaluated using the Sign test. In addition, the Sign test may be used to evaluate gross activity measurements from survey units containing multiple materials by s1:1btraetiAg the ambieAt baekgro1:1Ad from eael=t meas1:1re1neAt.
The Sign test is applied as described in the following steps:
I. For each survey unit measurement, subtract the gross activity measurement from the DCGLw and record the differences.
- 2. Discard any difference that is exactly zero and reduce the total number of measurements (N) by the number of zero differences.
- 3. Count the number of positive differences. This value is the test statistic S+.
- 4. Compare the number of positive differences (S+) to the critical values from Table 1.3 of the MARSSIM for the appropriate values ofN (total measurements) and a (decision error rate). (A positive difference corresponds to a measurement below the DCGLw and contributes evidence that the survey unit meets the release criterion.)
23
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 If S+ is greater than the critical value in MARSSIM Table 1.3, then the null hypothesis is rejected, and the survey unit meets the release criteria.
13.0 RAI 5-12: ALARA ANALYSIS BASIS:
The regulations in 10 CFR 20.1402 requires radiological license termination criteria not to exceed 25 mrem/year and that the residual radioactivity be reduced to levels that are ALARA. NUREG-1757, Volume 2, Revision 2, Appendix N, provides guidance on methods for demonstrating compliance with the ALARA portion of the regulation.
Issue:
Section 4.4, "ALARA Evaluation," of the LTP, states that the licensee will meet a pre-determined compliance measure of 15 mrem/y which is ALARA when the license termination criteria is 25 mrem/y.
The NRC staff notes that this is inconsistent with the guidance in NUREG-1757, Volume 2, Appendix N which should be used to evaluate the residual radioactivity remediation methods in order to determine a pre-determined compliance measure. If the ratio of ALARA concentration to 25 mrem/y DCGL is 3/5 or greater, then it is demonstrated that the 15 mrem/y licensee imposed limit will be ALARA. Otherwise, it is uncertain if ALARA would be an even smaller dose concentration. The NRC staff notes that the licensee could determine the ratio of the ALARA concentration to the 15 mrem/y DCGLs to demonstrate ALARA when the ratio is 1 or greater.
Request:
Perform an ALARA analysis consistent with guidance in NUREG-1757, Vol 2, Appendix N to demonstrate that 15 mrem/y DCGLs are ALARA or adjust the DCGLs to correspond to a dose limit that is considered ALARA as determined by the ALARA analysis.
Response
10 CFR 20.1402 Radiological criteria for unrestricted use states:
A site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a TEDE to an average member of the critical group that does not exceed 25 mrem (0.25 mSv) per year, including that from groundwater sources of drinking water, and the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA).
To ensure that the NSS meets these requirements, MARAD has established an administrative limit of less than or equal to 15 mrem (0.15 mSv) per year of residual radioactivity that is distinguishable from background radiation results in a TEDE to an average member of the critical group. This administrative limit includes as low as reasonably achievable (ALARA) considerations.
The initial decision to apply the 15 mrem/y criteria rather than the 25 mrem/y criteria stipulated in IO CFR 20.1402 was to avoid a potential regulatory conflict following license termination. This potential conflict could occur for future ship-breaking or preservation activities for the NSS which are governed by EPA regulations which endorse an annual radiation exposure limit to the public of 15 mrem/y. MARAD made this decision after it awarded its fixed price decommissioning services contract, such that the contractor's fixed price was based remediating to 25 mrem/y. When MARAD decided to remediate to 15 mrem/y, the contractor did not require a contract value increase to meet this objective. Therefore, the cost of the decision was zero despite minor possible increases in waste disposal, transportation and labor. MARAD and its 24
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 contractor believe those possible increases are too low to quantify since removal of surficial residual radioactivity is not controlled with enough precision to predict these costs. As such, the costs for remediation to levels even lower than 15 mrem/yr cannot be accurately predicted despite the fact that residual radioactivity levels on the NSS are well below the DCGLs associated with 15 mrem/y.
Because the cost for remediation to 15 mrem/y or below is known to be zero, the application of equations N-l and N-2 is artificially skewed. Therefore, we have prepared an evaluation of the benefits and costs for remediation in accordance with NUREG 1757, Volume 2, Rev. 1, Appendix N. In this case, the monetary equivalent for future dose savings at 15 mrem/y is calculated as a benefit and the cost is considered zero as described above. In accordance with NUREG 1757, Volume 2, Rev. l, Appendix N, if the benefit is greater than the cost then the action is considered ALARA and should be performed. In our case where the cost is zero, any positive benefit at 15 mrem/y would justify this lower dose limit.
This ALARA evaluation is based on a ship scrapping scenario. Our assumptions and estimates are based on MARAD's extensive experience in shipbreaking as discussed in Chapter 6.3.1 of the LTP. MARAD identified an exemplar shipbreaking project for this analysis. The project is the S.S. CAPE JOHN, completed in 2015 in Brownsville, TX. CAPE JOHN was a contemporary ofNSS and is of the basic hull and machinery design from which NSS was derived. It's hull form and power requirements were similar enough to NSS that the differences can be accounted for. Table 13-1, Comparison ofNSS to CAPE JOHN, summarizes these differences. The significant difference between the NSS and the CAPE JOHN is the machinery weight, which on NSS includes the weight of the nuclear power plant (3000 tons), including the Containment Vessel (CV) and Collision Boundary. With essentially all components removed from the CV, we estimate the remaining machinery is double that of the CAPE JOHN or 2000 tons. This value accounts for additional machinery weight of auxiliary systems associated with passenger operations (e.g., air conditioning), auxiliaries unique to the nuclear power plant (e.g., diesel generators, take-home motor), component differences based on saturated main steam (e.g., moisture separators, larger turbines) and the weight of the retained nuclear components.
TABLE 13-1 COMPARISON OF CAPE JOHN TO SAVANNAH Long tons (2240 pounds per ton)
CAPE JOHN SAVANNAH Lightship Weight 8,640 12,334 Steel Weight 5,210 5,300 Machinery Weight 1,040 1980
[4,420 total machinery weight -
(3000 + 2(250))]
Outfitting 2,390 2,480 CAPE JOHN was dismantled over seven months and required 16,313 man hours for on-board and ground cutting. For this evaluation, we estimated that approximately 11 workers would be required to scrap the NSS in a year. The DCGLs were derived with an exposure duration assumed to be 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s/day for 250 days or 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> per year for the ship scraping worker.
25
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 If each of the 11 workers were exposed to residual radioactivity represented by DCGLs set at 25 mrem/y for the working year of 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />, each worker would receive a dose of 25 mrem and would result in a collective dose of 275 p-mrem. If each of the workers were exposed, under the same conditions, at DCGLs equivalent to 15 mrem/y, each worker would receive a dose of 15 mrem and would result in a collective dose of 165 p-mrem. This is a collective dose savings of 100 p-mrem.
Therefore, under all exposure conditions, the individual and collective doses would be approximately three-fifths (3/5) of the doses if the DCGL was set to 15 mrem/y rather than at 25 mrem/y.
As described above, Equations N-1 and N-2 from NUREG 1757, Volume 2, Revision 2, Appendix N are used to determine the monetary benefit from MARAD's proposed action. In this analysis, the variables used in the calculation are shown in the table below. See the following table.
TABLE 13-2 RESULTS OF EQUATIONS N-1 AND N-2 Parameter Parameter Parameter Parameter Description Value units Pd Population density 1
p/m2 A
Area being evaluated 10 m2 0.025 Annual dose limit for NSS 0.015 rem/y F
Fraction of the residual radioactivity 1
fraction removed Conc/DCGlw Ratio of cone to DCGLw I
fraction R
Monetary Discount Rate 0.07 y *I A
Decay Constant, Cs-13 7 0.0231 y-1 N
Number of years 1
y PW(ADcollective)
Present worth of future averted dose 0.1432 p-rem VAo BAD Value of averted dose 5,200
$Ip-rem Benefit from averted dose for remediation
$745 USD The analysis shows that the benefit of reducing the dose limit from 25 to 15 mrem has a monetary benefit of $745 in year I following license termination. If this analysis is performed for a period of 70 years (as presented in Chapter 6.1 4 of the L TP), the benefit would be well less than $5200 using the same parameter values for all other parameters. This $5200 benefit also ignores radioactive decay and is considered very small in comparison to most industrial activities including the decommissioning costs and ship breaking. Essentially, this low benefit value in comparison to the zero cost shows that using our lower dose limit meets the ALARA requirement from 10 CFR 20 Subpart E.
26
Docket No. 50-238; License NS-1; N.S. SAVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 14.0 RAI 10-1: NRC APPROVAL OF LTP CHANGES BASIS:
The regulations in 10 CFR 50.82(a)(I0) outlines the requirements for changes to the LTP. Applicable guidance on areas that cannot be changed without NRC approval can be found in NUREG-1700, Revision 2, "Appendix B, "LTP Areas That Cannot Be Changed Without NRC Approval." NUREG-1575, Revision 1, Section 4.7.1 provides guidance that a measurement system with an MDC between 10-50% of the DCGL should be selected.
Issue:
Chapter 10, "LTP Areas That Cannot be Changed Without NRC Approval," of the LTP, Bullet 5 states, "Increase the derived concentration guideline levels. Nominal values for the MDCs have been presented in Table 5-4 in the LTP. Using the methodology for calculating MDCs presented in Chapter 5 in the LTP, the actual MDCs will be calculated prior to performing the FSS.
Therefore, increasing the MDCs does not require NRC approval." The NRC staff notes that several terms for the MDC determination appear to have been omitted or non-conservatively selected. At a minimum, the NRC staff should be informed if the actual MDC values for instruments used to perform FSSs are potentially going to exceed 50% of the proposed DCGLs.
Request:
Make appropriate changes to Table 5-4 to present nominal values using appropriate terms for the scan MDC and fixed measurement MDC. Subsequently, ensure bullet number 5 is consistent with the guidance in NUREG-1700, Revision 2, Appendix B regarding contacting the NRC if increasing the MDC above what was approved.
Response
Table 5-4 has been updated. See response to RAJ 5-5.
See updated bullet 5 in Section 10 of the LTP below.
(Marked up) Increase the derived concentration guideline levels }-l01'AiHal ¥alues fer the minimum detectable cencentrations CMDCs) haYe been presented i:n Table 5 4 in the LTP. Usieg the metlledelegy fer calcHlatiHg I>.IDCs presented iH Cha-pter 5 i-B the LTP, the acrual I>.IDCs will be calcHlated prior to perferming the FSS. Therefore, increasing the MDCs does not reeiHire }-lRC appro11al and related minimum detectable concentrations (MDCs) for both scan and fixed measurement methods. lfMDCs are increased (relative to what was approved) the licensee should request NRC approval.
(Clean) Increase the derived concentration guideline levels and related minimum detectable concentrations (MDCs) for both scan and fixed measurement methods. IfMDCs are increased (relative to what was approved) the licensee should request NRC approval.
15.0 REFERENCES
- a.
Letter from Mr. Erhard W. Koehler (MARAD) to U.S. Nuclear Regulatory.Commission (NRC),
dated October 23, 2023, License Amendment Request No. LAR 2023-01, Submittal and Request for Approval of the License Termination Plan
- b.
Letter from Mr. Erhard W. Koehler(MARAD) to U.S. Nuclear Regulatory Commission (NRC),
dated June 27, 2024, Response to Requests for Additional Information
- c.
CANBERRA ISOC-0 I, Operation of Canberra in Situ Object Counting Systems, Revision 0
- d. STS-005-030, Preparation of FSS Packages, Revision 3 27
Docket No. 50-238; License NS-1; N.S. S AVANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024
- e. CR-164, Calculation of Weighted Efficiencies and MDCs of Ludlum Detectors for Final Status Surveys on the NS SAVANNAH, Revision 1
- f.
CR-167, Use of In-Situ Gamma Spectroscopy for Final Status Survey (FSS) at the NS SAVANNAH, Revision 0
- g.
CR-165, Efficiency and MDC of Ludlum Pipe Detectors for Final Status Survey (FSS) on the NS SAVANNAH, Revision 0
- h. STS-224, Annual Report for Calendar Year 2023, Revision 0
- 1.
STS-100, Post Shutdown Decommissioning Activities Report, Revision 1 J.
STS-005-032, Survey Unit Turnover and Control, Revision 3
- k.
STS-005-030, Preparation of FSS Packages, Revision 3 I.
STS-005-033, Final Status Survey Data Assessment and Investigation, Revision 1 28
Docket No. 50-238; License NS-1; N.S. SA VANNAH License Amendment Request No. LAR 2023-01, Response to the Second Request for Additional Information October 16, 2024 ATTACHMENT 1.
CLARIFICATION OF HOW THE RADIONUCLIDES OF CONCERN (ROCS) WERE DETERMINED As described in Section 2.1.3 of the L TP, the NSS MARS SIM characterization campaign was implemented after the first tranche of decommissioning funds was appropriated in FY 2017. MA RAD tasked the work to its integrated support contractor, who began planning for the characterization campaign to commence the following calendar year (note that funds were appropriated in the 3Q of FY 2017, and performance of the characterization effort began in 4Q FY 2018). Planning included reviews of the CR-003, Historical Site Assessment (HSA) (Reference 2-3 of the LTP) and all previous scoping and characterization surveys. An initial list of radionuclides of concern (ROC) was derived based on this review and is included in the L TP as Table 6-2. Among the factors considered when developing this initial list of ROCs was the effect of a fuel failure described in Section 3.1 of the HSA. The cladding failure was assessed to be minor and insignificant, as no alpha contamination was ever identified in the routine surveys conducted through the years since shutdown. The initial suite of radionuclides was used as the basis for the laboratory analyses of samples during the RC / CV characterization effort described in Section 2.3.5 of the LTP.
During the characterization of RC and CV areas of the ship, multiple smears were collected from 12 locations. The smears from each location were composited and analyzed by gamma spectrometry at an offsite lab. The composite smears from 5 systems were selected for additional analyses of the hard to detect radionuclides from LTP Table 6-2. One sludge sample was collected from the make-up storage tank (PD-T2). This sample was sent to an offsite laboratory for analysis for a full suite of radionuclides including actinides, hard-to detect activation products and gamma emitters (activation and fission products). The evaluation of these results from CR-139, Calculations to Support NS SAVANNAH Swface Contamination DCGLs (Reference 6-9 of the L TP) shows 11 radionuclides detected across the population of samples with some not detected in some of the samples. This data was summed across all samples and normalized into activity fractions for all eleven radionuclides. This list is consistent with the initial suite of ROCs with the addition of Ag-I 08111, Am-241, and Pu-239/240. This final list of ROCs is included in the LTP as Table 6-3.
In addition to developing the final ROCs, CR-139 evaluated the insignificant radionuclides once the DCGLs were derived. The final ROC list represents three (3) radionuclides (Co-60, Cs-137, Ni-63) that are shown to contribute 99.3% of the dose conh*ibution when considering all 11 radionuclides present.
The dose from the remaining eight (8) radionuclides are considered insignificant contributors and are removed from further consideration. From this list of three (3) radionuclides, a surrogate relationship between Co-60 and Ni-63 is established to account for the presence and dose contribution from Ni-63.
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