ML16250A433

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Forward Final Report on Independent Confirmatory Survey Summary and Results for the Discharge Canal and Annex Building 6 at the Humboldt Bay Power Plant, Eureka, California
ML16250A433
Person / Time
Site: Humboldt Bay
Issue date: 10/28/2015
From: Harpenau E
Oak Ridge Associated Universities
To: John Hickman
Division of Decommissioning, Uranium Recovery and Waste Programs
References
Download: ML16250A433 (38)
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Text

OAAU Further Together October 28, 2015 Mr.John Hickman U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards Division of Decommissioning, Uranium Recovery, and Waste Programs Reactor Decommissioning Branch Mail Stop: T8F5 11545 Rockville Pike Rockville, MD 20852

SUBJECT:

FINAL REPORT-INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE DISCHARGE CANAL AND ANNEX BUILDING 6 AT THE HUMBOLDT BAY POWER PLANT, EUREKA, CALIFORNIA (RFTA NO.15-006); DCN 5272-SR-01-0

Dear Mr. Hickman:

ORAU is pleased to provide the enclosed final report detailing the independent confirmatory survey activities of the Discharge Canal and Annex Building 6 at the Humboldt Bay Power Plant in Eureka, California.

This report provides the summary and results of activities performed by ORAU, under the Oak Ridge Institute for Science and Education (ORISE) contract, during the period of July 20-23, 2015. Comments for additional clarification on the September 2015 draft version of this report have been incorporated.

You may contact me at 865.241.8793 or Erika Bailey at 865. 576.6659 if you have any questions.

Sincerely, Evan M. Harpenau Health Physicist ORAU EMH:fs electronic distribution:

G. Schlapper, NRC S. Roberts , ORAU E. Bailey, ORAU D. Stearns, NRC T. Vitkus, ORAU File/5272 P.O. Box 117 Oak Ridge, TN 37831

  • www.orau.org INDEPENDENT CONFIRMATORY SURVEY

SUMMARY

AND RESULTS FOR THE DISCHARGE CANAL AND ANNEX B U ILDING 6 AT T HE H U MBOLDT BAY POWER PLAN T , E U REKA, CALIFORNIA FINAL REPORT OR i\U Further. Together.

Prepared by Evan M. Harpenau OCTOBER 2015 Prepared for the U.S. Nuclear Regulatory Commission Prepared by ORAU under the Oak Ridge Institute for Science and Education contract, nwnber DE-AC05-060R23100, with the U.S. Department of Energy under interagenc y agreement (NRC FIN No. F-1244) between the U.S. uclear Regulatory Commission and the U.S. Department of Energy. Humboldt Bay Confirmatory Surve y Rep o rt 5272-SR-01

-0 ORAU INDEPENDENT CONFIRMATOR Y S U RVE Y S U MMAR Y AND RES U L T S FOR T HE DISCHARGE CANAL AND ANNE X B U ILDING 6 AT THE HUMBOLDT BAY POWER PLAN T , E U REKA , CALIFORNIA EXEC U TIVE S U MMAR Y The U.S. uclear Regulatory Commission (NRC) requested that ORAU, working under the Oak Ridge Institute for Science and Education (ORISE) contract, perform an independent confirmatory s urve y at the Humboldt Ba y Power Plant (HBPP) in Eureka, California.

Pacific Gas and Electric Company (PG&E), who owns and operates the site, is currentl y engaged in the decontamination and decommissioning of the U nit 3 boiling water nuclear reactor, along with the impacted areas associated with its operation.

This report focuses on confirmatory survey activities performed in support of decommissioning the Discharge Canal and Annex Bui l ding 6. ORAU performed independent assessment activities including gamma, beta , and alpha radiation surveys and soil sampling during the period of July 20-23, 2015. Confirmatory survey activities included surveys of two structural survey units with 29 s i de-by-side direct measurements; 20 independent direct measurements and smears; and one land area unit with two split and s i x random soil samples. The results of ORAU gamma, beta and alpha, radiation surveys, combined with laboratory analytical results from soil samples, support the conclusion that survey unit OOL 01-02 of the Discharge Canal and Annex Building 6 satisfies the NRC-approved soil and surface activity derived concentration guideline levels (DCGLs) described in PG&E's final status survey planning do c uments. Humboldt Ba y Co nfirmatory Survey Report 5272-SR-01-0 O IV\.U INDEPENDENT CONFIRMATORY SURVEY SUMMAR Y AND RESULTS FOR THE DISCHARGE CANAL AND ANNEX B U I L DING 6 AT THE HUMBOLDT BAY POWER PLANT, EUREKA, CALIFORNIA

1. INTROD U CTION The Pacific Gas & Electric Company (PG&E) operated the Humboldt Ba y Power Plant (HBPP) Unit 3 nuclear reactor near Eureka, California under Atomic Energy Commission (AEC) provisional license number DPR-7. HBPP Unit 3 achieved initial criticality in February 1963 and begun commer c ial operations in August 1963. Unit 3 was a natura l circu l ation boiling water reactor with a direct-cycle design. Stainless stee l fuel claddings were used from startup until cladding failures resulted in plant system contamination.

A number of spills and gaseous releases were reported during operations, resulting in a range of mitigation activities (ESI 2008). In Jul y 1973 , Unit 3 was shut down for annual refueling and se ismic modification

s. However, b y December 1980 it was concluded that completing the required upgrades and restarting Unit 3 would be cost prohibitive. PG&E decided in June 1983 to decommission U nit 3, received a possession-onl y licen se amendment , and placed the unit into cold shutdown and sa fety storage (SAFSTOR). The impacted areas associated with nit 3 are currentl y undergoing decommissioning.

As part of the Humboldt Ba y Repowering Project (HBRP), PG&E has built ten new fossil fuel units (16.3 MWe [megawatt electric]

each) on the site in the vicinity of Unit 3. Decommissioni n g activities have also been completed on the adjacent fossil fuel Units 1 and 2, with all materials being removed to ground level (ESI 2008). The U.S. Nuclear Regulatory Commission (NRC) is responsible for oversight of permitted license activities that are currentl y being conducted at Unit 3 of the HBPP. The NRC requested that ORAU, under the Oak Ridge Institute for Science and Education (ORISE) contract, perform confirmatory surveys of the Discharge Canal and Annex Building 6. Hereafter , Annex Building 6 will be referred to as the Annex. Humboldt Ba y Co nfirmator y Surve y Report 5272-SR-01

-0 L OAAU 2. SITE DESCRIPTION The HBPP site, owned by PG&E, consists of 143 acres on the southern edge of Humboldt Ba y four miles southwest of the town of Eureka, in Humboldt County, California (Figure A-1). PG&E maintains ten new operating electric generating units at the HBPP site (in the New Generation Footprint Area) that run on fossil fuels. The new fossil fuel electric-generating units ha ve replaced the former Fossil Units 1 and 2. The remaining property includes mostl y open areas and protected wetlands.

This report focuses on the Discharge Canal and the interior and exterior surfaces of the A nnex, pictured in Figure 2.1 courtes y of Google Earth 2015. The Discharge Canal is located in the northern sec tion of the HBPP site. The middle portion of the Discharge Canal, survey unit (SU) OOL 01-02, contains approximately 1,018 square meters of surface area, primaril y comprised of silt and soil (PGE 2015a and 2015b). Following remediation, groundwater seepage was regularl y pumped from the SU to maintain suitable conditions for performing surveys. The Annex, SUs OFA 01-01 and OFA 01-02, is a concrete block structure with a footprint of approximately 272 m 2 (to tal area of 745 m 2 including exterior walls and roof) located across from the Main Office Building and is bounded b y SU OOL10 on the north, west, and south sides and S OOL08 on the east. It was constructed in the 1980s an d primarily used for administrative office space (PGE 2015c). Figure 2.1. Discharge Canal and Office Annex Building Location Humboldt Bay Co nfirmato ry Survey Report 2 5272-SR-01-0 ORAU 3. OBJECTIVES The objectives of the confirmatory survey activities were to provide independent contractor field data reviews and to generate independent radiologica l data for use by the NRC in evaluating the accuracy and adequacy of the licensee's procedures and results. 4. APPLICABLE SITE GUIDELINES The primary radionuclides of concern (ROCs) identified for the Discharge Cana l and Annex are beta-gamma emitters-fission and activation products-resulting from reactor operation.

Licensee documentation states that two specific ROCs, cobalt-60 and cesium-137, acco unt for over 95% of the total activity observed in the canal and Class 3 SUs. Though all site-re l ated RO Cs are not listed for the canal soils in Table 4.1, the dose contributions from radionuclides determined to be insignificant have been subtracte d from the overall 25 mrem/yr release criteria.

NUREG-1757 guidance describes radionuclides and exposure pathways that contribute no greater than 10% of the dose criteria to be insignificant contributors.

It also details that dose contributions from all radionuclides and pathways must be accounted for in demonstrating compliance with the release criteria (NRC 2006). PG&E determined the insignificant radionuclide dose contribution to be 1.1 millirem per year (mrem/yr).

The values presented in Table 4.1 are the derived concentration guideline l evels that have been adjusted down accordingly to account for insignificant radionuclide 1.1 mrem/y dose and therefore correspond with a total dose of 23.9 mrem/yr (PGE 2015b). Table 4.1. Discharge Canal Soil DCGL"s Scaled to 23.9 mrem/yr ROC Inventory Limit (pCi/g) Co-60 3.6 Nb-94 6.8 Am-241 23.9 Cs-137 7.6 Eu-152 9.6 Eu-154 9.0 Np-237 1.1 Humboldt Ba y Confirmatory Surve y Report 3 5272-SR-01-0 ORAu Eac h sca l ed radionuclide-s pecific represents the concentrati o n above back gro und of a re si du a l radionuclide that wo uld re s ult in a radiological d ose o f 23.9 millirem per year (mr e m/y r) to the average member of the critical group. For consi s tenc y with the licen see's application of DCGLs to gross soil and surface activity concentrations suc h t h at data were not corrected for background contributi o n s, ORAU also reported d ata results without background correction

s. Bec a u se each of the indi vi dual r e presents 23.9 mrem/y r, the s um-o f-fraction s (SOF) approach was used to demon s trate compliance with the d ose limit. SOF calculations were performed as follows: n SOFTOTAL = i SOFi = . *-o L w.1 J-j=O Where Ci i s the concentration of ROC " j ," a nd is the for ROC "j." N o te that gross concentrations were con s idered here for con serva tism. PG&E's characterization data indicated th a t s urface activity le v els were near background for the A nne x. As s uch, final status survey (FSS) and confirmatory re s ults were expected to represent a small fr a ction of the s ite's overa ll surface activity DCGLs, listed in Table 4.2. Table 4.2. Surface Activity DCGLs ROC" DCGL ROC DCGL ROC (dpm/100cm 2)b (dpm/100cm
2) A m-241 3.00E+0 3 Eu-152 2.70E+04 Pu-238 C-14 7.00E+06 Eu-154 2.50E+04 Pu-239 Cm-243 4.30E+03 H-3 1.80E+08 Pu-240 Cm-244 5.50 E+03 I-129 , 4.90E+04 Pu-2 41 Cm-245 2.20E+03 b-9 4 1.90E+04 Sr-90 Cm-245 2.70E+03 Ni-59 6.30E+07 Tc-99 Co-60 1.30E+04 Ni-63 2.40E+07 Cs-137 4.60E+04 Np-237 2.40E+03 *From the Office A nn e x Final Status Survey Planning Worksheet (PGE 20 1 Sc) h dpm/100c m 2 = disintegrations per minute p er 100 s qu are centimeters Humb o ldt Ba y Co nfirm a t ory Surve y R e p ort 4 DCGL (dpm/100cm
2) 3.40E+03 3.10E+03 3.10E+03 1.40E+05 9.70E+04 9.60E+06 5272-SR-01

-0 OR i\U The surface activity mea s urement s that were above in s trument minimum detection capabilities were prim a ril y attributed to Cs-137 a nd Co-60, with 94% of the activity coming from Cs-13 7 and 6% from Co-60 (PGE 2015c). The fractional distribution information was used to calculate a gross area DCGL (DCGL G J for both beta/g amma and alpha surface activities (PGE 2015c). The DCGL GA s are presented after Table 4.2.

  • DCGL GA beta/ gamma
  • DCGL GA alpha 4.06E+04 dpm/100cm 2 3.0E+03 dpm/100cm 2 5. PROCED U RES The confirmatory survey activities were conducted during the period of Jul y 20-23, 2015, in accordance with the project-specific confirmatory survey plan, the ORAU Radiological and Environmental Survry Pro cedure Manual, and the ORAU Environmental Services and Radiation Training Quality Program Manual (ORAU 2015a, 2015b, and 2015c). 5.1 S URFACE S CANS ORAU performed high-density gamma radiation scans of the accessible soil surfaces in the Discharge Canal. Gamma scans were performed using a Ludlum Model 44-10 sodium iodide (Nal) detector coupled to Ludlum Model 2221 ratemeter-scaler

\vith audible indicator.

A dditionall y, the gamma detector/ratemeter pair was coupled to a global positioning system (GPS) that enabled real-time gamma count rate and spatial data capture (Figure A-2). Although the area was pumped overnight, groundwater see page in the bottom of the Discharge Canal pre ve nted scan coverage in portions of that area. Low-to medium-density beta scans were performed on the interior and exterior of the A nnex that included the floor, walls, and roof. Beta s cans were performed using Ludlum Model 44-142 plastic scintillation detectors coupled to Ludlum Model 2221 ratemeter-scalers with audible indicators.

Additionally, data loggers were coupled to the detector/ ratemeter pair to allow for electronic data capture for dataset evaluation and presentation (Figures A-3 and A-4). Beta radiation surface scans concentrated on areas with the highe s t likelihood of contamination potential (i.e., high-traffic pathways, accumulation points, and drainage paths). Humb o ldt Ba y Co nfirmat ory Surve y Rep o rt 5 5272-SR-01

-0 O AA U 5.2 S U R FA C E A CTIVITY M EASUREMENTS Independent direct surface activity measurements were collected on the Annex's interior and exterior surfaces to assess total residual alpha and beta activity levels. ORAU collected surface measurements from randomly-generated locations using a Ludlum Model 43-92 and Model 44-142 plastic scintillation detectors.

Visual Sample Plan (VSP) software, Version 7.4 was used to generate random measurement locations.

There were no judgmental measurement locations judgmentall y as no elevated direct radiation, with respect to the alpha and beta DCGL GA s, was observed during scan surveys (Figures A-5 and A-6). Additionall y , as PG&E elected to report PSS surface activity results based on gross rather than net measurement data, ORAU did not subtract material-specific backgrounds from measurement data, such that the results would be directl y comparable with the licensee's.

Smear samples to determine removable gross alpha and gross beta activity levels were also collected at each random measurement location. In addition, ORA U collected side-b y-side direct measurements at PG&E's PSS locations in the two SUs associated with the Annex to determine if the PG&E and ORAU instrumentation exhibited good correlation as well as review the results for any systematic bias. 5.3 SOIL S AM P LI NG A ranked set sampling (RSS) design was used to estimate the mean radionuclide concentration in the Discharge Canal. The number of locations to evaluate and sample within each S U were calculated b y using the contractor's PSS planning data and VSP (PGE 2015b). As a result of the sample planning inputs, 18 ranking locations were evaluated in SU OOL 01-02 of the canal. Following completion of walkover surve y s, the RSS locations were laid out as illustrated in Figure A-7. A one-minute static gamma measurement was made with the Nal detector at eac h ranking location.

The surface measurements were then ranke d , which resulted in the selection of six locations for sampling.

The six sample locations are presented in Figure A-8. At the RC's request, ORAU collected two split samples from locations that PG&E had selected judgmentall y from the water-saturated portion of the canal. Due to standing water in the canal and abundant water content in the samples themselve s , ORAU was not able to collect direct Humboldt Ba y C o nfirmatory Surve y Report 6 5272-SR-01

-0 OR i\U measurement data. The PG&E sample IDs corresponding with the ORAU split samples are listed in Table 5.1 below. Table 5.1. Corresponding Split Sample IDs for the Discharge Canal PG&E O RAU OOL 01-02-023-F-B 5272S0007 OOL 01-02-025-F-B 5272S0008

6. SAMPLE ANAL Y SIS AND D A TA INTERPRETATI O N Samples and data collected on site were delivered to the Radiological and Environmental A nalytical Laboratory (REAL) for analysis and interpretation.

Sample custody was transferred to the REAL in Oak Ridge, Tennessee.

Sample analyses were performed iri accordance with the ORAU Radiological and En v ironmentalAna!Jtical Laboratory Procedures Manual (ORAU 2015d). Soil sample s were analyzed b y gamma spectroscopy for gamma-emitting ROCs and results were reported in units of picocuries per gram (pCi/ g). Smear samples were analyzed for gross alpha/beta activity using a low-background proportional counter. Smear sample and direct measurement results are reported in units of disintegrations per minute per one hundred square centimeters (dpm/100 cm 2). 7. F INDINGS AND RES U LTS The results of the confirmatory survey are discussed in the subsections below. 7 .1 DO CUME N T R EVIEW The FSSP worksheet for OOL 01-02 uses the terms "hard-to-detect (HTD) nuclides" and "deselected nuclides" when discussing fractional dose contributions to the overall release criteria (PGE 2015b). The discussion aligns with the term "in s ignificant radionuclides" used in NUREG-1757 which describes radionuclides and exposure pathways that contribute no greater than 10% of the dose criteria to be insignificant contributors, and states that dose contributions from all radionuclides and pathways must be accounted for to demonstrate compliance with the release criteria (NRC 2006). \V'hile both terms, HTD and insignificant radionuclides, could be considered synonymous when discussing the ROCs in the soils of the Discharge Canal, the continued use of Humboldt Ba y C o nfirmatory Surve y Report 7 5272-SR-01-0 HT D s could resu l t in improper radionuclide associations in future documents.

ORA is of the opinion that the licensee should make the distinction between radionuclides that are in fact hard to detect and those considered to be insignificant contributors to dose. The distinction should then be clearly communicated in future documents to mitigate potential confusion.

7.2 O B SE R VAT IO NS During survey activities of the Annex, SU OFA 01-01, ORAU observed the PG&E technicians collecting removable activity smears after collecting the beta measurement but prior to the alpha direct measurements.

Had removable activity been present, this sequence could have led to biased low total alpha surface activity resu l ts. The smear sample should be collected after both the beta and alpha direct measurements are completed.

In addition, ORA noted that FSS measurement locations placed on the exterior walls of the Annex, S OFA 01-02, had been in v erted when compared to the survey map. Each concern was raised with the NRC and the contractor committed to correcting the survey sequence for collecting smears, and to maintaining consistency when l aying out wall and ceiling locations for future FSS activities.

7.3 S U R FA C E SCAN S During surve y activities in the Discharge Canal (SU OOL 01-02), the upper slopes of the walls proved unsafe to wa l k on, especially along the northwest side of the SU. As a result, ORAU was not ab l e to survey the northwestern-most portion of the SU. Additionally, continuous inflow of ground water into the bottom of the canal prevented access to approximate l y one-sixth of the overall SU. The confirmatory gamma scan results of SU OOL 01-02 exhibited radiation levels within the detector background range of 4,200-8,000 gross counts per minute (cpm) (Figure A-2). Scans did not identify any areas for further investigation. Scan results for the interior and exterior surfaces of SUs OFA 01-01 and OFA 01-02, respectively, were also indicative of background radiation levels. Though the interior scan ranged from 120 to 730 cpm, there is no identifiab l e step-up of the data when displayed in a Q-plot which would typically be indicative of residual contamination (Figure A-3). Additionall y, the data show a normal distribution with all data points within three standard deviations of the mean. Much like the interior Humboldt Ba y Confirmatory Surve y Report 8 5272-SR-01-0 OAAU survey, the scan range for the roof, SU OFA 01-02, was 130 to 960 cpm and s howed a normal distribution about the mean without the pre se nce o f any ste p s in the d ata (Figure A-4). 7.4 SURFACE ACTIVITY MEASUREMENTS Direct alpha and beta radiation activity results during the collection of the random confirmatory measurements were consistent with typical background le ve ls. Total gross surface activity for the confirmatory measurement locations of the A nnex interior ranged from 0 to 35 alpha dpm/100 cm 2 and 980 to 1 , 600 beta dpm/100 cm 2* The total surface activities observed on the buildin g's exterior ranged from 17 to 300 alpha dpm/100 cm 2 and 1,600 to 2,100 beta dpm/100 cm 2* Removable activity results for interior and exterior s urfaces exhibited a maximum of 1 dpm/100 cm 2 alpha and 5 dpm/100 cm 2 beta. Tables B-1 and B-2 provide a det a iled s umma ry of the independent confirmatory measurement data. The s ide-b y-si de measurement data show a general agreement between observed alpha surface activities.

Howe ve r, PG&E's beta results using Ludlum Model 43-68 gas-proportional detectors are consistentl y lower than ORAU's corresponding data for all but three of the exterior measurement locations on the Annex, and the relative percent difference is greater than 25% for 17 of the 30 mea s urement results. Though the percent difference could lead to potential issues should s urface activities approach the DCGL GA s, the surface activities calculated for the Annex are less than 10% of the release criteria.

ORAU will continue to monit o r compar a ti v e measurement re s ult s between instrumentation when additional data are collected and provide notification to the NRC s hould there be a c;oncern about whether a S U meets or exceeds the release criteria. The side-by-side mea s urement data are summarized in Tables B-3 and B-4. 7 .5 RADIONUCLIDE CONCENTRATIONS IN SOIL A nal y tical results for the independent confirmatory and s plits samples collected from the Discharge Canal are provided in Table B-5. A ll s ix RSS samples exhibited ROC concentrations below the respective anal y tical minimum detectable concentrations (MDCs) for the ROCs listed in Table 4.2. However, the split samples 5272S0007 and 5272S0008 exhibited ROC concentrations above the re s pective analytical MDC s for Co-60 and Cs-13 7, re s p e cti v ely (Table B-5). Even though Co-60 and Cs-137 were identified in the split sa mples, the concentrations represent a small fraction of the Humboldt Ba y Co nfirm atory Survey Rep ort 9 5272-SR-01

-0 O RAu respective Due to the extremely low radionuclide concentrations in the samp l es and only a couple values being above the analytical MDCs, ORAU did not perform SOF calculations.

ORAU has not received PG&E's corresponding data for the split samp le s, and thus, was unable to perform an inter-laboratory comparison of the results. 8.

SUMMARY

At the NRC's request, ORAU conducted confirmatory survey activities of SU OOL 01-02 in the Discharge Canal and the interior and exterior surfaces of Annex Building 6, SUs OFA 01-01 and OFA 01-02, at the HBPP during the period of July 20-23, 2015. The survey activities included visual inspections, gamma and beta radiation surface scans, gamma, beta and alpha radiation measurements, and soil and smear sampling.

The gamma walkover surface scans and total surface activity measurements were not distinguishable from background.

The two split samples collected from PG&E judgmental samples within the Discharge Cana l contained radionuclide concentrations above analytical MDCs for Co-60 and Cs-137, but a ll sample concentrations were well below the respective The minor issue regarding the sequence of performing direct surface activity measurements followed by smears for removable contamination was corrected during onsite confirmatory surveys. For the second issue of measurement locations on exterior walls being inverted for measurement collection, PG&E stated that the same map orientation implemented for splayed walls during the Annex survey would be maintained for future FSS surveys. Another potential issue identified is the systematic bias between PG&E and ORAU beta detector responses.

As discussed in Section 7.2, ORAU will continue to evaluate the detector results during upcoming confirmatory surveys. However, based on the results of the confirmatory survey activities, ORAU is of the opinion that A nne x Building 6 and SU OOL 01-02 of the Discharge Canal satisfies the RC-approved soil and surface activity DCGLs described in PG&E's final status survey planning documents.

Humboldt Ba y Confirmatory Surve y Report 10 5272-SR-01-0 O RA.u 9. REFERENCES ESI 2008. Histori ca l Site Assessme nt. Draft. Prepared for the Humboldt Ba y Power Plant Pacific Gas & E lectric Company. E ureka , California.

September.

NRC 2006. Consolidated Decommissioning G11ida11ce

-Characterizati on, Sun1ey, and Determination of Radiological Criteria. REG 1 7 57, Vo l. 2 R ev, 1. U.S. Nuclear Regulat ory Co mmi ss i o n. Washington, DC. September.

ORAU 2014. ORAU Radiation Prot ec tion Manual. Oak Ridge Asso ciated U ni ve rsities. Oak Ridge, Tennessee.

October. ORA U 201 Sa. Proj ect-specific Plan for the Confirmatory Survey Activities at the Humboldt B ery Pow er Plant , Dis charge Canal and Offi ce Annex Bu ilding, Eureka, California.

Oak Rid ge Asso ciated U ni ve rsitie s. Oak Rid ge, Tenne ssee. July 13. ORAU 2015b. ORAU Radiological and Environmental Survey Procedures Man11al Oak Ridge Asso ciated U ni v ersities.

Oak Ridge, Tennessee.

August 6. ORAU 2015c. ORAU E n vironmental Services and Radiation TrainingQualiry Pro gram Manual Oak Rid ge Asso ciated U ni vers itie s. Oak Ridge, Tenn essee. A ugu st 7. ORA 2015d. ORAU Radiological and Enviro11me11talA11a!Jtical Laborato ry Pro cedures Manual Oak Rid ge Asso ciated U ni ve rsities. Oak Ridge, Tenne ssee. May 7. ORAU 2015e. ORAU Health and S cifery Manual Oak Rid ge Asso ciated U ni vers itie s. Oak Rid ge, Tennessee. June. PGE 2015a. Email correspondence between D. Randall (PG&E) to E. Harpen a u (O RA U) titled , " Draft Surv ey Plan s for the upcoming i nsp ection/ visit at Humboldt B ery Po1v er Plant. "Pacific Gas & E lectric Company. San Francisco, California. Jul y 8. PGE 201 Sb. Final Status Survey Plannin g (FSSP) Worksheet-Discharge Canal (PGE Middle Unit): OOLO 1-02, Re v. OD. RCP FSS-2, A tt. 9 .1. Pacific Gas & Ele ctric Company. San Francisco, Ca lif ornia. Jul y 7. PG E 20 1 Sc. Final S talus Survey Planning (FSSP) Worksheet-Office Annex: O F A O 1-01 and OF A O 1-02, R ev. OD. Pacific Gas & E lectric Company. San Francisco, California.

Jul y 7. Humb o l dt Ba y Co nfirmato ry Survey R e port 11 52 72-S R-0 1-0 Humboldt Ba y Confirmatory Surve y Report APPENDIX A FIG U RES 5272-SR-01

-0 c:J Discharge Canal Boundary c:J Annex Bu i ld i ng 0 3 5,()(Jll), 000 -=:::::i M ete r s ORAU Humboldt Bay Power Plant Eureka, California Ctta 1 cd by: A. K inhlink Date: Scp1cmbcr

22. 201.5 \':\IEA\T GIS il'l Humbokll Ba J Figure A-1. Location of the Humboldt B ay Power Plant, Eureka, California Hum b o ld t Bay Co n firmatory Survey R e p ort A-1 5272-SR-0 1-0

-5451 -6100 --Canal Boundary 7401 -7957 -4801 -5450 6751 -7400 -4151 -4800 -6101-6750

-<4150 0 1.5 5 272 Humboldt B ay C anal02 G amma Walk o ver 3 ORAU Coa t ed by:

th.lid£ D iue; Auci;nt 6, lO'lS Figure A-2. Discharge Can a l SU OOL 01-02-G a mma Walko v er Sc a n Humb o ldt Ba y Confirmatory Survey Rep ort A-2 5272-SR-01

-0 Annex Building Interior -.. --* ... / 11111.GD 118.0D -_ 110.00 j-oa : 450.00 I GOOD 390.00 1 *.oa 33000 i JClOOO 270.00 240.00 210.00 180.00 15000 120.00 ... 9000 ,. *' Theoretlclll Quantiles (Standard Normal) *I nterior Figure A-3. Q-Plot for Interior Scan Survey of the Annex Building 6 I l umbo l dt Bay o nfirmat o r y Sun-cy Report r\-3 5272-SR-0 1-0

--IOO.m --100.00 300.00 200.00 100.00 .. .. .. Annex Building Roof .... ... ,..---------

  • ' "' Theoretical Quantiles (Standard Normal) Figure A-4. Q-Plot for Roof Sc a n Surve y of the Anne x Building 6 H umbo l d t Ba y C o n fi rma t o ry Sun-c y R eport A-4 5272-S R-0 1-0 Figure A-5. Random Measurement Locations for Interior of Annex Building 6 Humbo l dt Ba y Confirmatory Survey Rep ort A-5 5272-SR-01-0 5272R0014 Figure A-6. Random Measurement Locations for Exterior of Annex Building 6 Humboldt Bay Confirmatory Survey R eport A-6 5272-SR-01-0

--Cana l Bounda ry 5 272 H umb ol dt B ay Canal02 R SS L o c a ti o n s ORAU F i gure A-7. R a nked Set Sample Loc a tions for the Discharge Can a l, SUOOL 01-02 H umb ol dt B a y Co nfirm a t ory S u rve y R e p o rt A-7 5 272-S R-0 1-0

  • Random Samples --Canal Boundary 5 272 HumholdtBay Cana102 Sample Locations ORAU Figure A-8. Sample Locations for the Discharge Can al, SUOOL 01-02 Humb o ldt B ay Co nfirm a t ory S u rvey R e p o rt -8 5 2 72-S R-0 1-0

Humb o ldt Ba y Co nfirmatory Survey Report APPENDIXB DATA TABLES 5272-S R-01-0 Table B-1. Surface Activity Levels for Confirmatory Measurements in Interior of Annex Building 6 Count Rate (cpm) Total Surface Activity Removable Activity Location ID Surface (dpm/100 cm 2) (dpm/100 cm2) Alpha Beta Alpha Beta* Alpha Beta 5272R0001 W all 2 339 17 1 ,400 1 5272R00 02 Wall 3 246 26 980 -1 2 52 72R000 3 W all 1 335 9 1 ,300 2 5272R0004 C eilin g 1 368 9 1,5 00 c l -1 5272R 000 5 W all 0 314 0 1,3 00 -1 0 5272R0006 Ce ilin g 2 305 1 7 1 ,200 -1 2 5272R 0007 Floor 1 311 9 1,200 -1 0 5272R 0008 C eilin g 3 333 26 1 ,300 -1 2 5272R 0009 Wal l 4 277 35 1 , 1 00 -1 2 5272R0010 W all 2 390 17 1 , 600 -1 0 *Det e ctor calibrated t o mula-s o urce beta energies for C o-60 and Cs-13 7 per fraca o naa o n p ro nd c d m PG E 20 1 Sc. Hwn bo ld t Bay Co nfirmat o ry Surycy R eport B-1 5272-SR-0 1-0 Table B-2. Surface Activity Levels for Confirmatory Measurements in Exterior of Annex Building 6 Count Rate (cpm) Total Surface Activity Removable Activity Location ID Surface (dpm/100 cm 2) (dpm/100 cm2) Alpha Beta Alpha Beta* Alpha Beta 5272 R 00 11 W all 30 4 0 5 260 1 ,600 1 4 52 72R00 1 2 Roof 6 4 56 52 1 , 800 -1 1 5272 R 00 1 3 R oo f 11 473 96 1, 900 -1 0 52 72R00 14 Wa ll 4 395 35 1 , 6 00 -1 0 5272 R 00 1 5 W all 2 423 17 1 ,700 2 52 72 R 00 1 6 R oo f 7 520 61 2, 1 00 4 5272R 00 1 7 R oo f 1 0 464 87 1 ,900 2 52 72 R 00 1 8 R oo f 9 509 78 2 , 000 -1 4 5 272 R 00 1 9 R oo f 8 466 70 1 , 900 -1 5 5272R0020 Roof 10 486 8 7 1 , 900 1 1 *D e tect o r c alibr a t ed t o mu l n-so ur ce beta co e r g>cs fo r Co-60 and Cs-1 3 7 per fra c o o n a o o n p r m ,d c d 111 P G E 20 1 5 c H um bo ldt Ba y Co n firm a t o r y Sun-c y R e port B-2 5272-SR-0 1-0 Table B-3. Surface Activity Levels for Side-By-Side Measurements in Interior of Annex Building 6 PG&E G ross Alph a Count Gro ss Bet a Count Rate Gross Alpha A c ti v i ty Gros s Beta Acti vity Loca t ion ID Surf a ce Rate (cpm) (cpm) (dpm/100 cm') PG&E O RA U PG&E ORA U PG&E OFAOl-01-00 1 Wall 1.5 1 179 330 23 OFAO l-01-002 Floor 0.5 0 270 357 8 OFA0 1-0 1-003 b W all -----OFAO l-0 1-004 Ceiling 2 1 267 329 31 OFAOl-01-005 F l oor 2 6 4 86 614 3 1 OFAO l-0 1-006 Wall 2 4 1 97 3 11 3 1 OFAO l-0 1-007 Door 0 2 244 346 0 OFAO l-0 1-008 Wall 0.5 0 185 287 8 OFAOl-01-009 Wall 0.5 2 181 246 8 OFAOl-01-0 1 0 Wall 1 2 212 257 15 OFAO l-0 1-0 11 Ce ilin g 0 2 378 415 0 OFAO l-0 1-0 12 Floor 1 0 277 346 15 OFAOl-01-013 Floor 0.5 1 272 357 8 OFAO l-0 1-0 14 F loor 0.5 2 285 367 8 OFAO l-0 1-0 15 Ceiling 0.5 1 240 332 8 , *D e te c t o r ca Libr at e d to muln-s o urce beta e n ergies fo r C o-60 and Cs-13 7 pe r fracn o nan o n provided m PGE 20 1 5c b Locatio n w a s inacceRsible, theref o r e not survey e d. Humboldt B ay Co nfirm a t o ry Sun-ey R epo rt B-3 ORA U 9 0 -9 52 35 17 0 17 1 7 1 7 0 9 1 7 9 (dpm/100 cm') PG&E ORA U" 669 1,30 0 1 , 009 1 ,400 --997 1, 300 1 ,8 15 2,5 00 736 1 , 200 9 11 1 ,400 691 1 , 1 00 676 980 792 1 ,000 1 , 41 2 1 , 700 1,035 1 ,400 1 ,0 16 1 , 4 00 1, 065 1 ,500 896 1 ,300 5272-SR-0 1-0 P G&E Gross Alpha Count G ross Beta Count Rate Surface Rate cm cm Location ID PG&E O RAU P G&E O RAU OFAO l-02-00 1 Wall 2 1 337 404 3 1 9 1 ,259 1,6 00 OFAO l-02-002 W all 6.5 3 355 234 1 00 26 1 ,326 940 OFAO l-02-003 Roof 4 5 502 457 62 43 1 ,875 1 ,800 OFAO l-02-004 Roof 8.5 6 455 487 13 1 52 1 ,700 1, 900 OFAO l-02-005 Wall 0.5 3 214 262 8 26 799 1 ,000 OFAO l-02-006 Roof 5 1 2 441 494 77 100 1 ,6 47 2,000 OFAO l-02-007 Roof 6.5 14 461 44 1 100 1 20 1,722 1 ,800 OFAO l-02-008 Wall 2.5 0 313 397 39 0 1 , 1 69 1,600 OFAO l-02-009 Wa ll 3 4 313 46 1 46 35 1 , 1 69 1 ,800 OFAO l-02-010 R oo f 1 9.5 34 423 440 30 1 300 1, 580 1 ,800 OFAO l-02-0 11 Wa ll 4 7 380 417 62 61 1 , 4 1 9 1 ,700 OFAO l-02-012 Wa ll 2.5 1 304 344 39 9 1 , 1 36 1 ,400 OFAO l-02-0 1 3 Roof 9.5 5 476 523 1 47 43 1 ,778 2,100 OFAO l-02-014 R oof 5.5 1 0 502 5 11 85 87 1 ,875 2,000 OFAO l-02-0 15 Roof 5.5 7 417 471 85 6 1 1 ,558 1 ,900 *Detect o r cali br ated t o multi-so urc e beta e n e rgi es for Co-60 and Cs-137 per fracti o nati o n pro,-id e d in PGE 2015 c Hum bo ldt Ba y Co nfirmat o r y SurYcy Report B-4 5272-SR-01-0 Table B-5. Radionuclide Concentrations in the Discharge Canal (SU OOL 01-02) Sample Sample ID' Am-241 Co-60 Cs-137 Data or Location ID Type or Statistic (pCi/g) (pCi/g) (pCi/g) Ran d o m 5272S000 1 R SS-1-1-2 -0.07 +/- 0.04' 0.0 1 +/- 0.03 0.44 +/- 0.05 527250002 R SS-1-2-2 -0.02 +/- 0.08 0.02 +/- 0.02 0.06 +/- 0.02 527250003 RSS-1-3-2 -0.04 +/- 0.04 0.02 +/- 0.0 4 0.56 +/- 0.06 527250004 RSS-2-1-1 -0.04 +/- 0.06 -0.0 1+/-0.02 0.02 +/- 0.0 1 527250005 RSS-2-2-1 0.00 +/- 0.05 0.03 +/- 0.03 0.03 +/- 0.02 527250006 R SS-2-3-1 0.09 +/- 0.11 O.D2 +/- 0.03 0.76 +/- 0.07 S plit 527250007 0.02 +/- 0.03 0.0 1 +/- 0.03 0.38 +/- 0.04 Sam pl es 527250008 O.Ql +/- 0.08 0.00 +/- 0.0 2 0.08 +/- 0.02 *Un ce rt a ma cs r e pr ese nt th e 95°1 0 co nfid e n ce l cY el , b ase d o n to tal pr o p aga t e d uncert ru na cs. hz cr o du e to ro und i n g Hum bo ldt Ba y Co nfirm ato r y S un-ey R epo rt B-5 Eu-152 Eu-154 Nb-94 (pCi/g) (pCi/g) (pCi/g) -0.07 +/- 0.06 -0.1 2 +/- 0.10 0.02 +/- O.D2 -0.0 1+/-0.04 -0.09 +/- 0.08 O.OO b +/- 0.0 1 0.0 1+/-0.06 -0.07 +/- 0.1 0 0.0 1+/-0.02 0.03 +/- 0.03 -0.13 +/- 0.08 0.00 +/- 0.0 1 0.0 1+/-0.06 -0.1 8 +/- 0.1 2 0.00 +/- 0.02 -0.0 1+/-0.04 -0.2 1 +/- 0.11 0.0 0 +/- 0.02 -0.02 +/- 0.05 -0.23 +/- 0.1 1 0.02 +/- 0.02 -0.02 +/- 0.0 4 -0.1 1 +/- 0.08 0.00 +/- 0.02 52 72-SR-01-0 Np-237 (pCi/g) 0.02 +/- 0.04 -0.0 1+/-0.03 -0.02 +/- 0.04 0.00 +/- 0.02 -0.0 1 +/- 0.04 0.00 +/- 0.0 4 0.02 +/- 0.02 -0.0 1+/-0.03 APPENDIXC S U RVEY AND ANALYTICAL PROCEDURES Humboldt Bay Confirmatory Survey Rep ort 5272-SR-01-0 C.1 PROJ E C T HEAL T H AND SAF ETY ORAU performed all survey activities in accordance with the ORAU Radiation Prot ection Manual, the ORAU Health and Sefery Manual, and the ORAU Radiological and Environmental Survey Proc edures Manual (ORAU 2014, ORAU 2015e, and ORAU 2015b). Prior to on-site activities, a work-specific hazard checklist was completed for the project and discussed with field personnel.

The planned activities were thoroughly discussed with site personne l prior to implementation to identify ha za rds prese n t. A dditionall y, prior to performing work, a pre-jo b briefing and walkdown of the survey areas were completed with field personnel to identify ha za rds present and discu ss safety concerns.

Should ORAU have identified a hazard not co ve red in the ORAU Radiological and Environmental S11rvry Proc edures Manual or the project's work-specific hazard checklist for the planned survey and sa mpling procedures, work would not have been initiated or continued until it was addressed b y an appropriate job ha za rd analysis and hazard controls.

C.2 CALIBRATION AND QUALITY ASSURANCE Calibration of all field instrumentation was based on s tandards/ sources, traceable to ational Institute of Stan d ards and Tech n o l ogy (NIST). Field survey activities were conducted in accordance with procedures from the following ORAU document s:

  • ORAU Radiological and Environmental Surv ey Pro ced ur es Manual (O RAU 20 1 5b)
  • 0 RAU Radiologi cal and Environmental .AJJa/ytical Laboratory Pro cedttres Manual (O RA U 20 1 Sd)
  • ORAU Environmental Services and Radiation Trainin g Qttaliry Program Manual (ORAU 2015c) The procedures contained in these manuals were developed to meet the requirements of U.S. Department of Energy (DOE) Order 414.lD and the U.S. Nuclear Regu l atory Commissio n (N R C) Qttaliry Assttrance Manual for the Office of N ttclear Material S aftry and S efeg11ards a nd contain mea s ures to assess processes during their performance.

Humb o ldt B ay Co nfirmator y Surve y Rep o rt C-1 5272-SR-01-0 Quality control procedures include:

  • Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations
  • Participation in Mixed-Analyte Performance Evaluation Program, NIST Radiochemistry Intercomparison Testing Program , and Intercomparison Testing Program Laboratory Quality Assurance Programs
  • Training and certification of all individuals performing procedures
  • Periodic internal and external audits Detectors used for assessing surface activity were calibrated in accordance with IS0-7503 1 recommendations.

Total alpha and beta efficiencies (i:: wra U were determined for each instrument

/ detector combination and consisted of the product of the 27t instrument efficiency (i::;) and surface efficiency (i::,): E wra I = E; x i:: ,. ISO-7 503 recommends an s , of 0.25 for alpha emitters and also beta emitters with a maximum energy of less than 0.4 Me V and an s , of 0.5 for maximum beta energies greater than 0.4 MeV. Beta total efficiencies were determined based on a multi-point energy calibration using C-14, Tc-99, Tl-204, and Sr-90; development of instrument efficiency to beta energy calibration curves; and the selection of the E; and i:: , that represented the primary radionuclide of concern. Based on the data in PG&E's FSSP worksheet, a weighted efficiency for the fractional contributions of Co-60 and Cs-137 was calculated.

That total weighted efficiency was 0.25 for the plastic scintillators used to quantify beta surface activity.

Th-230 was selected as the alpha calibration source. The 2n alpha instrument efficiency (s;) factor was 0.46 for the plastic scintillation detectors, resulting in a total efficiency of 0.11 C.3 SURVE Y PROCED U RES C.3.1 S U R F AC E SC ANS Scans for elevated gamma radiation were performed by passing the detector slowl y over the surface. The distance between the detector and surface was maintained at a minimum. Specific scan 1 Int ernational Standard.

ISO 7 503-1, fa'aluation of Surface Co ntaminati o n -Part 1: Beta-e mitter s (maximum beta e n ergy greater than 0.1 5 M e V) and alpha-emitte r s. August l, 1988. Humboldt Ba y Confirmatory Survey Report C-2 5272-SR-01-0 minimum detectab l e concentration (MDCs) for the sodium iodide scintillation detectors (Nal) were not determined as the in s truments wer e u se d s olel y a s a qualitati v e mean s to identify ele v ated gamma radiation le v els in e x cess of background.

Identifications of elevated radiation le v els that could exceed the site criteria were determined based on an increase in the audible signal from the indicatin g instrument.

Beta scans were performed using small, hand-held scintillati o n detectors with a 1.2 mg cm 2 window. Identification of e l evated radiation levels was based on increases in the audible signal from the indicating in s trument. Beta surface scan MDCs were estimated using the approach described in NUREG-1507.

The scan MDC is a function of many v ariables, including the background level. Additional parameters selected for the calculation of scan MDCs included a two-second observation interval , a specified level of performance at the first scanning stage of 95% true positive and 25% false positive rate, which yields ad' value of 2.32 U REG-1507 , Table 6.1), and a surve y or efficienc y of 0.5. The beta total weighted efficiency factoring in the fractional contributions of Co-60 and Cs-137 was 0.25. The detector used had a general background of 305 cpm. The minimum detectable count rate (MDCR) and scan MDC was calculated as: C.3.2 B i= (305)(2 s)(l min/60 s) = 10 counts MDCR = (2.32)(10 counts)1 1 2[(60 s/min)/2s) = 222 cpm MDCR.um ro r = 222/(0.5)1 1 2 = 314 cpm Scan MDC= (314)/(.25) = 1,257 dpm/100 cm 2 S U RF A C E AC TIV I TY ME ASUREME N TS Measurements of total beta and alpha surface activity le v el s were performed using hand-held scintillation detectors coupled to portable ratemeter-scalers.

Count rates (cpm), which were integrated over one minute with the detector held in a s tatic position, were converted to activity levels (dpm/100 c m 2) b y dividing the count rate b y the total static efficienc y (EiXE ,) and correcting for the ph y sical area of the detector , which for both detector s is 100 cm 2 ** ORAU did not determine construction material-specific background for each surface type encountered for determining net count rates. Instead, ORAU took the conservative approach followed by the licensee and reported gross activity values. However, should background subtraction be necessary, the ambient beta and alpha background (1 cpm) count rates for the area would be used (305 cpm used in the example below) when determining surface activity. An example a priori MDC for beta activity is given by: Humb o ldt B ay Co nfirmat o r y Surve y Rep o rt C-3 52 72-SR-0 1-0 MDC = 3 + ( 4.65v'B) G Etot \'{There:

B background

£t o t total efficiency G geometry correction factor (1.0) The a priori beta static MDC was approximately 335 dpm/100 cm 2 using the weighted efficiency calculated from the fractional contributions of Co-60 and Cs-137. C.3.3 SOIL SAMP LI NG Soil samples (approximatel y 0.5 kilogram each) were collected using a clean garden trowel, then transferred into a new sample container b y ORAU personnel.

In total, ORAU collected eight soil samples from the Discharge Canal during the Jul y 20-23, 2015 confirmatory survey. ORAU personnel labeled each sample in accordance with ORAU survey procedures and completed the required custody documentation. C.4 RADIOLOGICAL ANALYSIS C.4.1 GAMMA SP E CTROSCOP Y Samples were anal yz ed as received, mixed, crushed , and/ or homogeni ze d as necessary , and a portion sealed in a 0.5-liter Marinelli beaker. The quantity placed in the beaker was chosen to reproduce the ca li brated counting geometry.

Net materia l weights were determi n ed and the samp l es counted using intrinsic, high purity, germanium detectors coupled to a pulse height analyzer system. Background and Compton stripping, peak search, peak identification , and concentration calculation s were performed using the computer capabilities inherent in the analyzer system. All total absorption peaks (TA P s) associated wit h the ROCs were reviewed for consistency of activity. Spe c tra were also reviewed for other identifiable TAPs. TAPs used for determining the activities of ROC s and the typical associated MDCs for a one-hour count time were: Radionuclide" TAP (MeV) MDC (pCi/g) Am-2 41 0.0595 0.15 Co-60 1.1 7 3 0.06 Cs-137 0.662 0.05 Humboldt Ba y Co nfirmat ory Surve y Rep o rt C-4 5272-SR-01

-0 Eu-152 0.344 0.10 Eu-154 0.723 0.15 b-94 0.871 0.05 Np-2 3 7 0.312 0.08 as pectra were also reviewed for other identifiable TAPs. C.4.2 DETECTION LIMITS Detection limits, referred to as MDCs, were based on 95% confidence level. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides in samples, the detection limits differ from sample to sample and instrument to instrument.

Humb o ldt Ba y Confirmatory Survey Rep ort C-5 5272-SR-01-0 APPENDIXD MAJOR INSTRUMENTATION Humboldt Ba y Confirmatory Surve y Report 5272-SR-01

-0 The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the author or his employer.

D.1 SCANNING AND MEAS U REMENT INSTR U MENT /DETECTOR COMBINATIONS D.1.1 G AMMA Ludlum NaI Scintillation Detector Model 44-10, Crystal: 5.1 cm x 5.1 cm (Ludlum Measurements, Inc., Sweetwater, Texas) coupled to: Ludlum Ratemeter-scaler Model 2221 (Ludlum Measurements, Inc., Sweetwater, Texas) coupled to: Trimble Data Logger (Trimble Navigation Limited, Sunnyvale, California) D.1.2 B ETA Ludlum Plastic Scintillation Detector Model 44-142, 100 cm 2 physical area coupled to: Ludlum Ratemeter-scaler Model 2221 (Ludlum Measurements, Inc., Sweetwater, Texas) coupled to: Trimble Data Logger (Trimble avigation Limited, Sunnyvale, California) D.1.3 ALP HA Ludlum Plastic Scintillation Detector Model 43-92, 100 cm 2 physical area coupled to: Ludlum Ratemeter-scaler Model 2221 (L u dlum Measurements, Inc., Sweetwater, Texas) D.2 LABORATOR Y ANALYTICAL INSTR U M E NTATION High-Purity, Extended Range Intrinsic Detector CANBERRA/Tennelec Model o: ERVDS30-25195 (Canberra, Meriden, Connecticut)

Used in conjunction with: Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, Tennessee) and Multichannel Analyzer Canberra's Gamma Software Dell Workstation (Canberra, Meriden, Connecticut)

Humboldt Bay Confirmator y Surve y Report D-1 5272-SR-01-0 High-Puri ty, Intrinsic Detector Model No. GMX-45200-5 CANBERRA Model No: GC4020 (Canberra, Meriden, Connecticut) Used in conjunction with: Lead Shield Model G-11 Lead Shield Model SPG-16-K8 (Nuclear D a ta) Multichannel A nal yz er C anberra's Gamma Software Dell Work s tation (Canberra, Meriden, Connecticut) Low Background Gas Proportional Counter Model LB-5100-W (Tennelec/Canberra , Meriden, CT) Tri-Carb Liquid Scintillation A nal yz er Model 3100 (Packard Instrument Co., Meriden , CT) Humb o ldt B ay Co nfirmat ory Surve y Report D-2 52 72-SR-01-0

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