ML17089A750

Safstor/Decommissioning Offsite Dose Calculation Manual, Vol.4, Rev. 4. Part 2 of 2

ML17089A750
Person / Time
Site:	Humboldt Bay
Issue date:	03/30/2017
From:	Pacific Gas & Electric Co
To:	Office of Nuclear Material Safety and Safeguards
Shared Package
ML17090A130	List: HBL-17-005, Safstor/Decommissioning Offsite Dose Calculation Manual, Vol.4, Rev. 4. Part 2 of 2 ML17089A747
References
HBL-17-005
Download: ML17089A750 (58)

Text

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 1-27 2.13 RADIOACTIVE WASTE INVENTORY LIMITING CONDITIONS 2.13.1 Liquid Radioactive Waste In Outdoor Tanks The radiological inventory of wastes in outdoor tanks that are not capable of retaining or treating tank overflows shall not exceed 0.25 Ci. APPLICABILITY:

At all times. ACTION: When the inventory exceeds the conditions as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Monitoring Program Report. 2.13.2 Deleted SURVEILLANCE REQUIREMENTS 2.13.3 An inventory of the estimated liquid radioactive waste in outdoor tanks inventory shall be maintained to verify the 0.25 Ci limit is not exceeded.

OR Provide overflow protection.

OR Use process lmowledge of typical concentration and tank volume to verify that the 0.25 Ci is not exceeded.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 1-28 3.0 SPECIFICATION BASES 3.1 Radioactive Gaseous Effluent Monitoring Instrumentation Basis Deleted -The plant stack ceased operation in 2015. Monitoring gaseous effluent is limited to sampling and analysis of Modular HEPA Units. 3.2 Liquid Effluent Concentration Basis Deleted-Liquid effluents are no longer discharged to Humboldt Bay. Effective December 31, 2013, discharge of processed radioactive liquid effluents to Humboldt Bay was terminated.

Any remaining or incidental radioactive liquid in concentrations exceeding 1 0 times 10 CFR 20, Appendix B, Table 2 Column 2 are manifested for disposal at a licensed disposal site. Sampling and manifesting requirements are consistent with the requirements of the receiving facility not subject to ODCM methodology.

3.3 Liquid Effluent Dose Basis Deleted -Liquid effluents are no longer discharged to Humboldt Bay. 3.4 Liquid Effluent Treatment Basis Deleted -Liquid effluents are no longer discharged to Humboldt Bay. 3.5 Gaseous Effluents Dose Rate Basis This specification provides reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA either within or outside the SITE BOUNDARY in excess of the design objectives of Appendix I to 10 CFR 50. The annual dose rate limits are the doses associated with the annual average concentrations of "old" 10 CFR 20, Appendix B, Table II, Column 1. The specification provides operational flexibility for releasing gaseous effluents to satisfy the Section II.A and II.C design objectives of Appendix I to 10 CFR 50. For a MEMBER OF THE PUBLIC who may at times be within the SITE BOUNDARY, the period of occupancy (which is bounded by the maximum occupational period while working in Units 1 or 2) will be sufficiently low to compensate for the reduced atmospheric dispersion of gaseous effluents relative to that for the SITE BOUNDARY.

The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. This specification does not affect the requirement to comply with the annual limitations of 10 CFR 20.130l(a).

NUCLEAR POWER GENERATION DEPARTMENT TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL SECTION ODCM VOLUME 4 REVISION 27 PAGE 1-29 Stack operation and monitoring ceased operation in 2015, so the reporting period for 2015 includes the dose contribution from the plant stack prior to ceasing operation.

Modular HEPA Ventilation Units continue to be sampled as a gaseous effluent pathway. Noble gas monitoring is not required because the spent fuel (noble gas source term) has been transferred to the ISFSI. Tritium monitoring is not required in gaseous effluents because the tritium source term was the spent fuel pool water which is now empty. Residual water in various plant drains and sumps contain low levels of tritium (generally at or below the drinking water standard (2E-5 uCi/ml or 20,000 pCi/L) and does not require monitoring as a gaseous plant effluent.

3.6 Deleted Gaseous effluent monitoring is not required for noble gases because the spent fuel (noble gas source term) has been transferred to the ISFSI. 3.7 Deleted 3.8 Gaseous Effluents:

Tritium and Radionuclides in Particulate Form Dose Basis This specification is provided to implement the requirements of Sections II.C , liLA , and IV.A of Appendix I , 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluent will be kept "as low as is reasonably achievable" (ALARA). The calculational methods specified in the Surveillance Requirements implement the requirements in Section liLA of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data , such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

The basis for the dose calculation threshold of 0.1 uCi alpha e1nission or Sr-90 in a week assumes a continuous ground level release of 1.65E-13 uCi/sec and an X/Q of 6.59E-3 sec/m 3. The limiting inhalation dose is to a teen age member of the public at the site boundary at appro x imately 0.3 mrem/wk (15 mrem/yr) to the bone from alpha emitters.

Compliance with this Specification has been established on a licensing basis by the SAFSTOR Environmental Report and NUREG-1166, "Final Environmental Statement for Decommissioning Humboldt Bay Power Plant." These reports have demonstrated that routine release of Tritium and radioactive materials in particulate form (with half-lives greater than 8 days) in gaseous effluents during decommissioning will not cause the Specification to be exceeded.

As long as routine releases do not exceed the baseline quantities evaluated in these reports , no further dose calculation is necessary.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE I-30 The previously evaluated tritium source term was the spent fuel pool water, which is now empty. Residual water in various plant drains and sumps contain low levels of tritium (at or below the drinking water standard (2E-5 uCi/ml or 20,000 pCi/L) and does not require monitoring as a gaseous plant effluent.

3.9 Solid Radioactive Waste Basis This Specification ensures that radioactive wastes that are transported from the site shall meet the disposal site(s) licensee and/or waste acceptance criteria for free standing liquids of the respective states to which the radioactive material will be shipped. It also implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. 3.10Total Dose Basis This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. The Special Repmi will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered.

If the dose to any MEMBER OF THE PUBLIC is estiinated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR part 190.11 and 10 CFR Part 20.2203a4, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed.

The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 2.3, 2.4, 2.6, 2.7 and 2.8. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is pmi of the nuclear fuel cycle. 3.11 REMP Monitoring Program Basis The quality-related portion of the REMP satisfies the requirements in 10 CFR Parts 20 and 50 that radiological environmental monitoring programs be established to provide data on measurable levels of radiation and radioactive materials in the site environs.

It is required to provide assurance that the baseline conditions established by the Environmental Repmi are not deteriorating and it supplements the SAFSTOR Environmental Report baseline NUCLEAR POWER GENERATION DEPARTMENT TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL SECTION ODCM VOLUME 4 REVISION 27 PAGE 1-31 environmental conditions by conducting onsite and offsite environmental monitoring to evaluate routine conditions during decommissioning and to document any increased nuclide concentrations and/or radiation levels resulting from accidents during decommissioning.

The SAFSTOR Environmental Report, submitted to the NRC as Attachment 6 to the SAFSTOR license amendment request, established baseline conditions for soil, biota and sediments.

The LLD's required by Table 2-9 are considered optimum for routine environmental measurements in industrial laboratories.

HBPP no longer includes water, milk, fish, food products, or sediment in its routine REMP sampling program. Sampling and analysis in support of the License Tennination Plan is independent of the ODCM requirements.

3.12 REMP Interlaboratory Comparison Program Basis The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid. 3 .13 Radioactive Waste Inventory Basis ) The requirements for limits on the accumulation of liquid radioactive waste in outdoor tanks were transferred from the license Technical Specifications.

4.0 ADMINISTRATIVE CONTROLS 4.1 Arinual Radiological Environmental Monitoring Report A report on the Decommissioning Radiological Environmental Monitoring Program shall be prepared annually in accordance with the NRC Branch Technical Position and submitted to the NRC by May 1 of each year. The Annual Radiological Environmental Monitoring Report shall include: a. Summaries, interpretations, and an analysis of trends of the results of the quality related Radiological Environmental Monitoring Program activities for the report period. The material provided shall be consistent with the objectives outlined in the ODCM, and in 10CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 1-32 b. A cmnparison with the baseline environmental conditions established in the Decommissioning Environmental Report. c. The results of analysis of quality related environmental samples and of quality related environmental radiation measurements taken during the period pursuant to the locations specified in Table 2-7 smnmarized and tabulated in the format of Table 4-1, Radiological Environmental Monito.ring Program Report Annual Summary, or equivalent.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the tnissing results. The missing data shall be submitted in the next annual report. d. A summary description of the Decommissioning Radiological Environmental Monitoring Program. e. Legible maps covering all sampling locations keyed to a table giving distances and directions from Unit 3. f. The results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required in accordance with Specification 2.12. g. The reason for not conducting the quality related portion of the Radiological Environmental Monitoring Program as required, and discussion of all deviations from the quality related sampling schedule of Table 2-7, including plans for preventing a recurrence in accordance with Specification 2.11. h. Deleted -water samples are not collected as a part of the REMP. 1. A discussion of all analyses in which the LLD required by Table 2-9 was not achievable.

NUCLEAR POWER GENERATION DEPARTMENT TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL SECTION ODCM VOLUME 4 REVISION 27 PAGE 1-33 Table 4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

-EXAMPLE Name of Facility Humboldt Bay Power Plant Unit 3 Docket No. 50-133 , OL-DPR-7 Location of Facility Humboldt County , California Reporting Period January 1-December 31, 1997 (County , State) Medium or I Type and Total I I All Indicator Location with Highest Annual Control Locations Mean Locations Number of Pathway Sampled I Number of Lower Limit Mean, [Unit ofMeasurement]

Analyses of Detection a (Fraction)

Performed (LLD) & rRange] b Name, Mean, Mean, (Fraction)

Non routine Distance and (Fraction)

& [Range] b Reported Direction

& [Range] b Measurements AIRBORNE Particulates Not Required N/A N I A N/A N I A Not Required N/A DIRECT RADIATION

[ mR/ quatier] Direct radiation I 3 I 13.6 +/- 0.1 (64) (64/64) Station T7 15.4 +/- 0.2 12.7 +/- 0.3 I 0 (4/4) (4/4) I I I [ll.8-17.5] r13.s-17.51 D2.5 -12.91 WATERBORNE INGESTION Milk Fish and invertebrates TERRESTRIAL Soil _ _

__ Not Required Not Required --------I----1----N I A N/A N I A N/A -------------

---

N I A N I A N/A NIA __

__ Not Required Not Required N/A ------------

N/A N/A NUCLEAR POWER GENERATION DEPARTMENT SECTION VOLUME REVISION PAGE ODCM 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL TABLE 4-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

27 1-34 a The LLD is defined as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal. LLD is defined as the a priori lower limit of detection (as pCi per unit mass or volume) representing the capability of a measurement system and not as the EL posteriori (after the fact) limit for a particular measurement. (Current literature defines the LLD as the detection capability for the instrumentation only, and the MDA, minimum detectable concentration, as the detection capability for a given instrument, procedure and type of sample.) The actual MDA for these analyses was at or below the LLD. b The mean and the range are based on detectable measurements only. The fraction of detectable measurements at specified locations is indicated in parentheses; e.g., (10/12) means that 10 out of 12 samples contained detectable activity.

The range of detected results is indicated in brackets; e.g., [23-34]. I !2116 I Not Required -not required by the HBPP Offsite Dose Calculation Manual. Baseline environmental conditions for this parameter were established in the Environmental Report as referenced by the SAFSTOR Decommissioning Plan. N/ A -Not applicable Note: The example data are based on the 1997 monitoring results and are provided for illustrative purposes only.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL 4.2 Annual Radioactive Effluent Release Repmi VOLUME 4 REVISION 27 PAGE 1-35 This report shall be submitted prior to April 1 of each year. The following information shall be included:

a. A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant as outlined in Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, (Rev. 1, 1974) with data summarized on a quarterly basis following the format of Appendix B thereof. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 1 OCFR 50.36a and 1 OCFR Part 50, Appendix I,Section IV.B.I. Beginning in the reporting year 2014, liquid effluents shipped for processing or disposal at a regulated disposal site are included in the annual report. b. For each type of solid waste shipped off-site:

1. Container Volume 2. Total Curie Quantity (specified as measured or estimated)

3. Principal Radionuclides (specified as measured or estimated)

4. Type of Waste (e.g., spent resin, compacted dry waste) 5. Solidification Agent (e.g., cement) c. A list and description of unplanned releases beyond the SITE BOUNDARY.

d. Information on the reasons for inoperability and lack of timely corrective action for any radioactive gaseous monitoring instrumentation inoperable for greater than 30 days in accordance with Specification 2.2. Beginning the reporting year 2015, following cessation of the plant stack operation, the effluent monitoring instrumentation associated with Specification 2.2 ceased operation.

Inoperability and lack of timely corrective action is only applicable to the period of plant stack operation.

Anomalies associated with monitoring effluent from Modular HEP A Ventilation systems will be reported.

e. A summary description of changes made to: 1. Process Control Program (PCP) 2. Radioactive Waste Treatment Systems NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 1-36 f. A complete, legible copy of the entire ODCM if any change to the ODCM was made during the reporting period. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE 112116 I SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL 4.3 Special Reports VOLUME 4 REVISION 27 PAGE 1-37 The originals of Special Reports shall be submitted to the Document Control Desk with a copy sent to the Regional Administrator, NRC Region IV, within the time period specified for each report. These reports shall be submitted covering the activities identified below to the requirements of the applicable Specification.

a. Radioactive Effluents-Specifications 2.8 and 2.1 0. b. Radiological Environmental Monitoring

-Specification 2.11. 4.4 Major Changes to Radioactive Waste Treatment Systems a. Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid) shall be reported to the NRC in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed.

The changes shall be approved by the HB Director.

b. The following information shall be available for review: 1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CPR 50.59, 2. Sufficient information to totally support the reason for the change, 3. A description of the equipment, components and processes involved and the interfaces with other plant systems, 4. A evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously estimated in the Environmental Report submitted to the NRC as Attachment 6 to the SAPS TOR license amendment request, 5. An evaluation of the change which shows the expected 1naximum exposures to an individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the Environmental Report, 6. An estimate of the exposure to plant personnel as a result of the change, and 7. Documentation of the fact that the change was reviewed and approved in accordance with plant procedures.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE 112/16 I SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL 4.5 Process Control Program Changes VOLUME 4 REVISION 27 PAGE 1-38 a. Changes to the Process Control Program (PCP) shall be documented and records of reviews performed shall be retained as required for the duration of Decommissioning.

b. The following information shall be available for review: 1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and, 2. A determination that the change willtnaintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

3. A description of the equipment, components and processes involved and the interfaces with other plant systems. c. The change shall become effective after approval of the HB Director.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-1 PART II-CALCULATIONAL METHODS AND PARAMETERS 1.0 UNRESTRICTED AREA EFFLUENT CONCENTRATIONS 1.1 LIQUID EFFLUENT UNRESTRICTED AREA CONCENTRATIONS Specification 2.3 .1 requires that the Radioactive Liquid Effluent Sample concentrations (RLES) are calculated to ensure that the limits of Specification 2.3 are not exceeded (the instantaneous concentration of radioactive material released to UNRESTRICTED AREAS shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2). This requirement is defined by the following relationship.

L ci , Canal ::::; 1 i lOx ECLi where: (1-1) Ci-Canal = The concentration of isotope " i " in the canal discharge point to Humboldt Bay. ECLi = Effluent Concentration Limit for radionuclide " i " from 10 CFR 20, Appendix B, Table 2, Column, 2 1.1.1 If the outfall location is not at the furthermost pmiion of the canal from the entrance to the Bay the concentration of the isotope Ci-Canal is equal to the concentration being discharged at the outfall.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-2 1.2 UNRESTRICTED AREA GASEOUS EFFLUENT CONCENTRATIONS 1.2.1 Equation C-4 of Regulatory Guide 1.109 demonstrates how to calculate dose from inhalation:

The annual dose associated with inhalation of all radio nuclides}

to organ j of an individual in age group aJ is then: Dja(r,e) = Ra I:xi(r,8)DF Aija where Dja is the annual dose rate to organ j of an individual in age group a Ra is the breathing rate for age group a Xi(r,8) is the annual average ground-level concentration of nuclide i in air in sector eat distance r, in pCi/m 3 DF Aija is the dose factor for nuclide i to organ j of age group a To calculate xi(r,e) the annual average ground-level concentration of nuclide i in air in sector eat distance r, in pCi/m 3 the equation must be rearranged to: Dja(r,8)/( DF Aija Ra) = Xi(r,e) Assuming that: Americium-241 is the primary nuclide The maximally exposed group is the Teen based on breathing rates and DF Aija The DF Aija to the bone of a Teen from Am-241 is 1. 77 mrem/pCi The DFAija are taken from: NRC NUREG/CR-4013, "LADTAP-11 Technical Reference and User Guide" The Teen breathing rate is 8000 m 3 /year NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE VOLUME 4 SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL REVISION 27 PAGE 11-3 Therefore the ground-level concentration of Am-241 in air in sector 8 at distance r, in pCilm 3 that will produce a dose rate of 1500 mremlyear to the bone of a Teen is: (1500 mremlyear)

I (1.77 mremlpCi)

I (8000 m 3 1year) = 1.06E-1 pCil m 3 1.06E-1pCi/

m 3 = (1.06E-1 pCilm 3) I (1E6 pCi I (1E6 mllm 3) = 1.06E-13 1.2.2 Quantity of radioactive material released Equation C-3 of Regulatory Guide 1.1 09 demonstrates how to calculate the quantity of material that must be released to produce a given airborne concentration:

The annual average airborne concentration of radionuclide i at the location (r, B) with respect to the release point may be determined as where Xi(r,8) 3.17 X 10 4 is the annual average ground-level concentration of nuclide i in air in sector 8 at distance r, in pCilm 3 is the number of pCi/Ci divided by the number of sec/yr is the annual average atmosphere dispersion factor, in sec!n/ is the release rate of nuclide I to the atmosphere, in Ci/yr A value of 6.59E-3 seclm 3 was used for the incidental release path atmosphere dispersion factor at the site boundary (x/Q)D(r,8) for releases from Modular HEPA Units. This is based on a release rate of 2000 cfm. (Ref: Safstor ODCM, Appendix B, 2.0) This factor is based on the atlnospheric models of Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. To determine the release rate that will result in an average ground-level concentration the above equation must be rearranged to: Qi = Xi(r,8) I (3.17 x 10\x/Q)D(r,B))

Therefore the Modular HEPA Unit release rate of Am-241 required to equal the incidental ground-level concentration at the site boundary calculated above is:

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-4 1.06E-1 pCihn 3 I ((3.17E4 (pCi/Ci)/ (sec/yr))

• (6.59E-3 sec/m 3)) = 5.07E-4 Ci/yr or 5.07E2 uCi/yr 1.2.3 Transmission Fraction Deleted -no on line monitoring provided.

1.2.4 Effluent Concentration The Modular HEPA Unit concentration that would result in a release rate of 5.07E-4 Ci/yr is equal to: Total release (Curies/year)

I Release rate (cc/year)

The average annual Modular HEPA Unit flow rate is 2,000 cfm This results in a total volume of2.98E13 cc/yr This is based on (2000 ft 3/min

• 525,600 minutes/yr

• 28,317 cc/ft 3). (5.07E-4 Ci

• 1E6 I (2.98E13 cc/yr) = 1.70E-11 Therefore an indicated Modular HEPA concentration ofl.70E-11 at 2000 cfm for one calendar year would result in a dose of 1500 mrem to a member of the public at the site boundary.

Two times the indicated release rate is equal to3.4E-11 Two hundred times the indicated release rate is equal to 3.4E-9 1.2.5 Relationship to EPA PAG To compare the release rates calculated above the following assumptions were made: Am-241 dose conversion factor in rem I cm-3 hr, from EPA 400 = 5 .3E8 Since no credit is taken for an elevated release point or an annual average x/Q the same atmospheric dispersion factor is used in the calculations below. Assuming that an unplanned release occurs at two times the ODCM release rate for one hour the total activity released is equal to:

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-5 3 .4E-11

• 2000 ft 3 /min

• 28,317 cc/ft 3

• 60 min = 1.16E-1 (1.16E-1

• (5.3E8 rem I cm-3 uCi hr) * (6.59E-3 sec/m 3) I (IE6 cm 3/m 3) I (3600 sec/hour)=

1.13E-4 rem This is much less than the EPA PAG of 1 Rem Assuming that an unplanned release occurs at two hundred times the ODCM release rate for 15 minutes the total activity released is equal to: 3.4E-9

• 2000 ft 3/min

• 28,317 cc/ft 3

• 15 min= 2.89EO This results in a dose of: (2.89EO

• (5.3E8 rem I cm-3 uCi hr) * (6.59E-3 sec/m 3) I (1E6 cm 3/m 3) I (3600 sec/hour)=

2.80E-3 rem This is much less than the EPA PAG of 1 Rem.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-6 1.2.6 Relationship to 1 OCFR20 Appendix B Table 2 Effluent Concentration limits The 1 OCFR20 Appendix B Table 2 Effluent Concentration limit for Am-241 is 2E-14 The average annual ground-level concentration in air (xi) in pCi/m 3 is equal to: xi= (3.17E4 (pCi/Ci)/ (sec/year))*

Q * (A/Q) Where Q is equal to the quantity of radioactive material released in a year in Curies/year ODCM Modular HEPA Unit incidental release .x/Q = 6.59E-3 sec/ m 3 If xi= 2E-14 then: Q = (2E-14

• 1E6 ml/m 3

• 1E6 I ((3.17E4 (pCi!Ci)/ (sec/yr)*(6.59E-3 sec/ m 3)) Q = 9.57E-5 Ci/yr The average annual Modular HEPA Unit volume based on the ODCM is 2.98E13 cc/yr. This is based on (2000 cfm

• 525,600 minutes/yr

• 28,317 cc/cfm). Therefore, the Modular HEPA Unit effluent concentration required to result in a fence-line concentration of 2E-14 is: (9.57E-5 Ci/yr

• 1E6 I (2.98E13 cc/yr

• 1 cc/ml) = 3.2E-12 1.2.7 Conversion Factor from Effluent Concentration to The release rate in

=Modular HEPA Unit concentration in

• 2000 ft 3 /min

• 1440 minutes/day

• 28317 cc/ ft 3 The release rate in

=Modular HEPA Unit concentration in

• 8.16E10 cc/day 1.2.8 Conversion Factor from to % ofNUE An NUE is equal to a release rate of 3000 mrem/year

%NUE = (Offsite dose rate I N*UE threshold)

• 100 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-7 %NUE =((Conversion Factor* Release Rate) /NUE threshold)*

100 %NUE =((Conversion Factor

• 1 00) I NUE threshold)

• Release Rate The Conversion Factor is equal to (1.77E6 mrem/J.tCi)

• (6.59E-3 sec/ m 3) * (8000 m 3/year) I (8.64E4 sec/day) This is equal to 1.08E3 mrem/year per J.!Ci/day 1.2.9 Results The 1 OCFR20 Appendix B Table 2 Effluent Concentration limit for Am-241 is 2E-14 J.!Ci/ml.

The Modular HEP A Unit effluent concentration that would result in a fence-line concentration of2E-14 J.!Ci/ml is 3.2E-12 J.!Ci/ml.

3.2E-12 uCi/ml

• 8.16E10 cc/day

• 1ml/cc

• 1.08E3 mrem-day/uCi-yr

= 4.70E2 mrem/yr. 470 mrem/yr I 8760 hr/yr = 5.365E-2 mrem/hr Assuming that an unplanned release occurs at two times the ODCM release rate for one hour the offsite dose corresponding to an NUE would be 1.07E-4 rem (0.107 mrem) which is much less than the EPA PAG. Assuming that an unplanned release occurs at two hundred times the ODCM release rate for fifteen minutes the offsite dose corresponding to an Alert would be 2.675E-3 rem (2.7 mretn) which is much less than the EPA PAG. Note that Am-241 is used in the example calculations and is expected to be limiting.

Other alpha emitting isotopes such as Pu-238, Pu-239/240 and Cm-243/244 are evident in the contamination at HBPP. Since the Effluent Concentration Limits (ECLs ), Derived Air Concentration (DAC) values and organ Dose Conversion Factors (DCFs) are similar, the Am-241 values may be assumed to be gross alpha with appropriate compensation for naturally occurring isotopes.

Other radionuclides (Co-60, Sr-90, Cs-137, etc.) are impmiant in determining actual offsite dose and in demonstrating compliance with the ECL using the sum of the fractions rule. The example calculations are used similarly for each isotope in the mix with their respective ECL, DCF and exposure pathway (inhalation, ingestion, and submersion).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-8 Although not relevant to the hypothetical offsite dose calculation in the ECL and NUE analysis above, assumed effluent concentrations are approximately 1 DAC, 2 DAC,' and 200 DAC for Am-241 at the point of release. Airborne radioactivity control measures to control worker dose, also limits the potential offsite dose.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-9 2.0 LIQUID EFFLUENT DOSE CALCULATIONS 2.1 MONTH (31 DAY PERIOD) Deleted 2.2 CALENDAR QUARTER-Deleted 2.3 CALENDAR YEAR-Deleted 2.4 LIQUID EFFLUENT DOSE CALCULATION METHODOLOGY As of December 31, 2013, HBPP has ceased liquid radioactive effluent discharges via the discharge canal to Humboldt Bay. Any remaining processed liquid radioactive waste is transported offsite for land disposal at an authorized disposal facility.

The following calculation methodology is preserved as a part of the ODCM for ease of reference to site specific parameters in the event of an accidental release of liquid radioactive effluent.

No recurring liquid effluent dose calculations are expected for the remainder of decommissioning.

The equations specified in this section for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. Equation (2) of Regulatory Guide 1.109 provides for the use of a site specific mixing ratio (i.e. reciprocal of the dilution factor) that describes the near term and near field mixing of the tidal flow from the Discharge Canal into Humboldt Bay. A two-dimensional numerical analysis, depth-averaged, finite element hydrodynamic model (reference 12.1) was developed by CH2MHILL and used to estimate the dispersion of the canal discharge in the Bay. The analysis indicated that an additional dilution factor of 80 for batch release applications or a dilution factor of 20 for continuous release applications can conservatively be used to describe the Bay dilution.

A factor of 20 will be applied in this calculation to address any combination of release modes. Since the intake canal contains a larger volume of water, use of the above dilution factors for effluent releases to the intake canal provides a simplified, conservative methodology for calculating annual dose from effluent releases to the intake canal. The dose contribution to the total body and each individual organ (bone, liver, kidney, lung and GI-LLI) ofthe maximum and average exposed individual (adult, teen, child, and infant) will be calculated for the nuclides detected in effluents.

The dose to an organ of an individual from the release of a mixture of radionuclides will be calculated as follows:

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE VOLUME 4 SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL REVISION 27 PAGE 11-10 n D = L[Ci-BaydilutedX DFx {(BFish,i X UFish)+ (Bmv , i X Umv)}] i=l (2-1) where: D Ci-Bay diluted DF BFish,i = The dose commitment, mrem per year, to an organ (or to the whole body) due to consumption of aquatic foods. = The average diluted Bay concentration, pico-Curie/liter, for radionuclide, i. If the outfall to the canal is at the furthest most portion of the canal from the entrance to the Bay, this will be estimated by calculating the total activity released (e.g. effluent concentration Ci effluent in pCi/L times the discharge volume V D in Liters) then dividing the total activity of the nuclide discharged during the period, pi co-Curies, by the dilution volume (e.g. total discharged volume V D plus total tidal flow V TD during the period in liters), and by the Bay dilution factor of20. The total annual tidal flow for the outfall canal is 2.47E+9 Liters/year (e.g., 6.77E+6 Liters/day).

If Gross Alpha radioactivity is determined to be in the effluent, Pu-241 will be considered to be present at 3.25 times the amount of detected Gross Alpha radioactivity.

Note that the resulting dose commitment is the annual dose rate (mrem per year) for a time frame with this average concentration.

Doses (NOT dose rates) for periods shorter than a year must be proportionately reduced. C Ci-EffluentX VD i -Baydiluted (VD +VrD)x20 (2-2) If the outfall is not located in the furthest most portion of the canal from the entrance to the Bay, no credit for tidal dilution of the canal will be taken and the diluted Bay concentration will be calculated using the following equation.

Ci -Effluent Ci-Baydiluted

=----20 (2-3) The dose conversion factor, mrem/pico-Curie for the nuclide, organ, and age group being calculated.

This factor is taken from Tables 2-1, 2-2, and 2-3. The bioaccumulation factor, pica-Curie/kilogram per pico-Curie/liter, in fish for the radionuclide in question.

This value is taken from Table 2-4.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-11 Binv,i = UFish = Uinv = The bioaccumulation factor, pico-Curie/kilogram per pico-Curie/liter, in invetiebrates for the radionuclide in question.

This value is taken from Table 2-4. Usage factor (consumption) offish, kilogram/year, for the age group and individual (average or maximum) in question.

This factor is derived from Table 2-5 or 2-6. Usage factor of invertebrates, kilogram/year, for the applicable age group and individual (average or maximum).

This factor is from Table 2-5 or 2-6. The total exposure to an organ (or whole body) is found from the summation of the contributions of each of the individual nuclides calculated.

Note that the infant age group is not considered to consume either fish or other seafood, and exposure to this age group need therefore not be calculated.

Dose calculations can be performed using the above methodology for the current month, quarter, or year.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-12 Table 2-1 Ingestion Dose Factors for Adult Age Group (mrem/pico-Curie ingested)

Selected Nuclides from (LADTAP II input values) Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 5.99x1o-8 5.99 x 1o-8 5.99 x 1o-8 5.99 x 1o-8 5.99x1o-8 Co-60 No Data 2.14 x 1o-6 4.72 x 1o-6 No Data No Data 4.02 x 1o-5 Ni-63 1.30 x 1 o-4 9.01 x 1o-6 4.36 x 1o-6 No Data No Data 1.88 x 1o-6 Sr-90 8.71 x 1o-3 No Data 1.75 x 1o-4 No Data No Data 2.19 x 1o-4 Cs-137 7.97 x 1o-5 1.09 x 1o-4 7.14 x 1o-5 3.7o x 1o-5 1.23 x 1 o-5 2.11 x 1o-6 Y-90 9.62 x 1o-9 No Data 2.58 x 1o-1o No Data No Data 1.02 x 1o-4 Pu-241 1.57 x 1o-5 7.45 x 1o-1 3.32 x 1o-1 1.53 x 1 o-6 No Data 1.40 x 1o-6 Am-241 7.55 x 1o-4 7.05 x 1o-4 5.41 x 10-5 4.07 x 1o-4 No Data 7.42 x 1o-5 Gross a 7.55 x 1o-4 7.05 x 1 o-4 5.41 x 1o-5 4.07 x 1o-4 No Data 7.42 x 1o-5 Table 2-2 Ingestion Dose Factors for Teen Age Group (mrem/pico-Curie ingested)

Selected Nuclides frmn NUREG/CR-4013 (LADTAP II input values) Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 6.04 x 1o-8 6.04 x 1o-8 6.04 x 1o-8 6.04 x 1o-8

• 6.04 x 1o-8 Co-60 No Data 2.81 x 1o-6 6.33 x 1o-6 No Data No Data 3.66 x 1o-5 Ni-63 1.77 x 1o-4 1.25 x 1 o-5 6.oo x 1o-6 No Data No Data 1.99 x 1 o-6 Sr-90 1.02 x 1o-2 No Data 2.04 x 10-4 No Data No Data 2.33 x 1o-4 Cs-137 1.12 x 1o-4 1.49 x 1o-4 5.19 x 1o-5 5.07 x 1o-5 1.97 x 1o-5 2.12 x 1o-6 Y-90 1.37x1o-8 No Data 3.69 x 1o-1o No Data No Data 1.13 x 1o-4 Pu-241 1.75 x 1o-5 8.40 x 1o-1 3.69x1o-7 1.11 x 1o-6 No Data 1.48 x 1o-6 Am-241 7.98 x 1o-4 7.53 x 1o-4 5.75 x 10-5 4.31x1o-4 No Data 7.87 x 1o-5 Gross a 7.98 x 1o-4 7.53 x 1o-4 5.75 x 1o-5 4.31x1o-4 No Data 7.87 x 1o-5 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-13 Table 2-3 Ingestion Dose Factors for Child Age Group (mrem/pico-Curie ingested)

Selected Nuclides from NUREG/CR-4013 (ladTAP II in_put values) Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.16 x 1o-1 1.16 x 1o-1 1.16 x 1o-1 1.16 x 1o-1 1.16 x 1o-1 Co-60 No Data 5.29x1o-6 1.56 x 1o-5 No Data No Data 2.93 x 1o-5 Ni-63 5.38 x 1o-4 2.88 x 1o-5 1.83 x 1o-5 No Data No Data 1.94 x 1o-6 Sr-90 2.56 x 1o-2 No Data 5.15 x 1o-4 No Data No Data 2.29 x 1o-4 Cs-137 3.27 x 1o-4 3.13 x 1o-4 4.62 x 1o-5 1.02 x 1o-4 3.67x1o-5 1.96 x 1 o-6 Y-90 4.11 x 1o-8 No Data 1.10 x 1o-9 No Data No Data 1.11 x 1o-4 Pu-241 3.87 x 1o-5 1.58 x 1 o-6 8.04x1o-7 2.96 x 1o-6 No Data 1.44 x 1 o-6 Am-241 1.36 x 1o-3 1.17x1o-3 1.02 x 1o-4 6.23 x 1o-4 No Data 7.64 x 1o-5 Gross a 1.36 x 1o-3 1.17x1o-3 1.02 x 1o-4 6.23 x 1o-4 No Data 7.64 x 1o-5 Table 2-4 Bioaccumulation Factors for Saltwater Environment (pCi/kg per pCi/liter)

Selected Nuclides from Regulatory Guide 1.109, Table A-1 and from NUREG/CR-4013 Element Fish Invertebrate H 9.o x 1o-1 9.3 x 1o-1 Co 1.0 X 102 1.0 X 103 Ni 1.0 X 102 2.5 X 102 Sr 2.0 2.0 x 101 Cs 4.o x 101 2.5 x 101 y 2.5 x 101 l.Ox1o3 Pu 3.0 2.0 X 102 Am 2.5 x 101 l.Ox1o3 Gross a 2.5x1o1 1.0 X 103 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM SAFSTORIDECOMMISSIONING OFFSITE VOLUME 4 TITLE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-14 Table 2-5 Average Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

From Regulatory Guide 1.109, Table E-4 Other Seafood Fruits and Age Group Fish (Invertebrates)

Vegetables Milk Meat Adult 6.9 1.0 190 110 95 Teen 5.2 0.75 240 200 59 Child 2.2 0.33 200 170 37 Infant 0 0 0 0 0 Table 2-6 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

From Regulatory Guide 1.109, Table E-5 Other Seafood Fruits and Age Group Fish (Invertebrates)

Vegetables Milk Meat Adult 21 5.0 520 310 110 Teen 16 3.8 630 400 65 Child 6.9 1.7 520 330 41 Infant 0 0 0 330 0 3.0 LIQUID EFFLUENT TREATMENT 3.1 TREATMENT REQUIREMENTS 3.1.1 Deleted 3.1.2 Deleted 3.2 Deleted NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-15 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS 4.1 DOSE RATE 4.1.1 Deleted As explained in Specification Bases 3.7, Noble Gases are not required to be monitored, and the corresponding dose rate need not be calculated.

4.1.2 Tritium and Radioactive Particulates There are no short-lived radioactive particulates in the effluent, so radioactive decay can be neglected.

Meteorological parameters are assumed to be constant, and applied for the most conservative location.

Therefore, the radioactive particulates dose rate calculation methodology is the same as the radioactive particulates dose calculation methodology.

Refer to sections 4.3.3 through 4.3.8 for the appropriate equations.

As explained in Specification Bases 3.5, Tritium is not required to be monitored, and the corresponding dose rate need not be calculated.

Nevertheless, if such a calculation is required, refer to sections 4.3.9 through 4.3.13 for the appropriate equations.

4.2 Deleted 4.3 DOSE-TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM 4.3.1 Calendar Quarter The methodology for calendar quarter calculations is the same as for the calendar year calculations provided by section 4.3.3, and discussed in section 4.3.2, with the exception that the resulting values forD (annual dose commitment, mrem/year) must be divided by 4 to convert them to quarterly dose commitment, mrem/quarter.

4.3.2 Calendar Year The methods for calculating the dose due to release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Pati 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Water-Cooled Reactors," Revision 1, July 1977.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-16 The equations provided for determining the doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

4.3.3 Organ Dose Calculation Summation Methodology The release rate specifications for radioactive particulates with half-life greater than eight days are dependent on the existing radionuclide pathways to man, in areas at and beyond the SITE BOUNDARY.

The pathways which were examined in the development of these calculations were: 1) Individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leaf vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. The releases of radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents will be essentially limited to Cs-137, Co-60, and Sr-90. Radioactive decay may result in the dose from Transuranic radionuclides becoming significant.

If Gross Alpha radioactivity is determined to be released, Pu-241 will be considered to be present at 3.25 times the amount of detected Gross Alpha radioactivity.

The annual dose commitment will be calculated for any organ of an individual age group as follows: n D = L[Qi X (Rrnll,i + RGP , i + RMeat,i + RMilk , i + Rveg,i)] i=l (4-3) where: D Qi Rlnh,i == RGP,i = RM:eat,i = RM:ilk,i = Annual dose commitment, mrem/year.

The average release rate of the nuclide in question, Curies/second.

The dose factor for the inhalation pathway for the radionuclide, i, in units of mrem/year per pico-Curie/sec.

The dose factor for the ground plane (direct exposure from deposition) pathway for the radionuclide, i, in units of mrem/year per pico-Curie/sec.

The dose factor for the grass-cow-meat pathway for the radionuclide, i, in units of mrem/year per pico-Curie/sec.

The dose factor for the grass-cow-milk pathway for the radionuclide, i, in units of mrem/year per pico-Curie/sec.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-17 RVeg , i = The dose factor for the pathway of deposition on vegetation for the radionuclide, i, in units of mrem/year per pico-Curie/sec.

In general, the calculations for these pathways give results that represent trivial radiation exposure.

The values calculated for typical anticipated Decommissioning releases range from about 0.002 mretn/year (fruit/vegetable consumption pathway) to less than 1 x 1 o-6 mremfyear (for direct radiation exposure from material deposited on the ground). 4.3.4 Particulate Inhalation Pathway Dose Calculation Methodology Rlru1 ,i = (x/Q) x BRa x DFi ,a (4-3a) where: x/Q = BRa = DFi,a = The atmospheric dispersion parameter, seconds/cubic meter. 1.0 x 1 o-5 seconds/cubic meter for releases from the 50 foot stack. Refer to Appendix B, 1.2. 6.59 x 1 o-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B, 2.1. The breathing rate of the receptor age group (a), cubic meters per year. The values to be used are 1400, 3700, 8000, and 8000 cubic meters/year for the infant, child, teen and adult age groups, respectively.

The organ (or total body) inhalation dose factor, mrem/pico-Curie, for the receptor age group, a, for the radionuclide, i. The dose factors are given in Tables 4-1, 4-2, 4-3, and 4-4.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-18 4.3.5 Particulate Ground Plane Pathway Dose Calculation Methodology RGP,i = (D/Q) X SF X DFi X K X w (4-3b) where: K = DFi = SF = DjQ = unit conversion constant, 8760 hr/yr. The ground plane dose conversion factor for radionuclide, i, in mrem/hr per pCi/m 2 from Table 4-5. No values are provided for Transuranic radionuclides, as their dose contribution to this pathway is negligible.

The shielding factor (dimensionless).

Table E-15 of Regulatory Guide 1.109 suggests values of0.7 for the maximum individual.

The atmospheric deposition factor, with units of inverse square meters. = 3.0 x 10-8 inverse square meters for releases from the 50 foot stack. Refer to Appendix B, 1.3. = 5.39 x 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B, 2.2. w Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.1 09, for a 14 day removal half-life.

In this equation, W has the value of 1. 7 4 x 106 seconds.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-19 4.3.6 Particulate Grass-Cow-Milk Pathway Dose Calculation Methodology ( ,/ ) X (QF X Ua X FmyX DE,a X w) RMilk,i = D 1 Q (4-3c) where: QF = The cow's vegetation consumption rate. This is given as 50 kg/day per Regulatory Guide 1.109, Table E-3. Ua = The receptor's milk consumption rate, liters/year for the age group in question.

See Tables 4-6 and 4-7. y = The agricultural productivity by unit area of pasture. This parameter is 0.7 kg/m 2 per Regulatory Guide 1.109, Table E-15. DFi,a = The ingestion dose factor for radionuclide, i, for the receptor in age group (a), in units ofmrem/pico-Curie, from Tables 4-8, 4-9, 4-10, or 4-11. Fm = The fraction of the cow's intake of a nuclide which appears in a liter of milk, with units of days/liter.

This parameter is given by Table 4-12. DjQ = The atmospheric deposition factor, with units of inverse square meters. = 3.0 X 1 0-S inverse square meters for releases from the 50 foot stack. Refer Appendix B, 1.3. = 3.29 X 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B, 2.2. w = W eatheririg factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.1 09, for a 14 day removal half-life.

In this equation, W has the value of 1.74 x 106 seconds.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-20 4.3.7 Particulate Grass-Cow-Meat Pathway Dose Calculation Methodology RMe,,,i = (D/Q) x ( QF x u, x x Dfi,, x w) (4-3d) where: QF Ua y = Dfi,a = Ff = DjQ = = = w = The cow's vegetation consumption rate of 50 kg/day per Regulatory Guide 1.109, Table E-3. The receptor's meat consumption rate, kilogram/year.

Refer to Tables 4-5 and 4-7. The agricultural productivity by unit area of pasture. This parameter is 0.7 kg/m 2 per Regulatory Guide 1.109, Table E-15. The ingestion dose factor for radionuclide, i, for the receptor in age group (a), in mrem/pCi, from Tables 4-8, 4-9, or 4-10, as appropriate.

Note that this path is not considered to apply to the infant age group. The fraction of the animal's intake of a nuclide which finally appears in meat, days/kilogram.

This parameter is given in Table 4-13. The atmospheric deposition factor, with units of inverse square meters. 3 .Q X 10-8 inverse square meters for releases from the 50 foot stack. Refer to Appendix B, 1.3. 3.29 X 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B, 2.2. Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.109, for a 14 day removal half-life.

In this equation, W has the value of 1.74 x 106 seconds.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-21 4.3.8 Particulate Vegetation Pathway Dose Calculation Methodology ( 4-3e) where: UT = The total consumption rate of fruits and vegetables, kilogram/year.

This parameter is determined with the default values from Regulatory Guide 1.1 09, as reproduced in Tables 4-6 and 4-7. DjQ = The atmospheric deposition factor, with units of inverse square w y meters. = 3.0 x 10-8 inverse square meters for releases from the 50 foot stack. Refer to Appendix B, 1.3. = 3.29 X 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B, 2.2. Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.109, for a 14 day removal half-life.

In this equation, W has the value of 1.74 x 106 seconds. The agricultural productivity by unit area of pasture. This parameter is 0.7 kg/m 2 per Regulatory Guide 1.109, Table E-15. Note: this equation probably overestimates exposures, since it assumes that all of the deposition on a plant remains on the plant, while the Regulatory Guide allows a factor of0.25. Also, the quantities assumed consumed include grain (none is grown in the vicinity of the plant), as well as vegetables and fruit grown in other areas (imported to Humboldt county).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL 4.3.9 Tritium Organ Dose Calculation Methodology VOLUME 4 REVISION 27 PAGE 11-22 The annual dose commitment tnay be calculated for any organ of an individual age group as follows: D = Qm X ( Rmh, H3 + RaP, H3 + RMeat, H3 + RMi!k, H3 + Rveg, H3) (4-4) where: D Annual dose commitment, mrem/year.

QH3 The average release rate ofH-3, pico-Curies/second.

Rinh, H3 = The dose factor for the inhalation pathway for H-3, mrem/year per pico-Curie/sec.

RMeat,H3 = The dose factor for the grass-cow-meat pathway for H-3, mrem/year per pico-Curie/sec.

RMilk,H3 = The dose factor for the grass-cow-milk pathway for H-3, mrem/year per pico-Curie/sec.

Rveg,H3 = The dose factor for the vegetation consumption pathway, llli*em/year per pico-Curie/sec.

This pathway results in trivial offsite calculated radiation exposures.

A very conservative assumption of Tritium release is that Spent Fuel Pool water at 1 x 1 o-2 micro-Curies/ml H-3 is lost to the stack at a rate of 50 gallons/day.

With this assumption, the calculated maximum offsite exposure is 0.0013 mrem/year.

Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-23 4.3.10Tritium Inhalation Pathway Dose Calculation Methodology Rinh , H3 = (76) X BR. X DFH3 ,, where: (4-4a) x/Q = BRa = DFH3 , a = The atmospheric dispersion parmneter, seconds/cubic meter. 1.0 x 1 o-5 seconds/cubic meter for releases from the 50 foot stack. Refer to Appendix B, 1.2. 6.59 x 1 o-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B, 2.1. The breathing rate of the receptor age group (a), cubic meters per year. The values to be used are 1400, 3700, 8000 , and 8000 cubic meters/year for the infant, child, teen, and adult age groups, respectively.

The organ (or total body) inhalation dose factor for the receptor age group, a, for H-3. This is given in units ofmrem/pico-Curie by Tables 4-1, 4-2, 4-3, and 4-4. Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-24 4.3.11 Tritium Grass-Cow-Milk Pathway Dose Calculation Methodology The concentration of tritium in milk is based on the airborne concentration rather than the deposition:

n_ .. _ (x/l (0.75 X 0.5) -/Q) X H X QF X Ua X F m X DF a ( 4-4b) where: QF Ua DFa Fm 0.75 0.5 H = x/Q = The cow's vegetation consumption rate. This is 50 kg/day per Regulatory Guide 1.109, Table E-3. The receptor's milk consumption rate for age group, a, from Regulatory Guide 1.109. See Tables 4-6 or 4-7. The ingestion dose factor for H-3, for the reference group, mrem/pico-Curie, from Tables 4-8, 4-9, 4-10, and 4-11. The fraction of the cow's intake of a nuclide which appears in a liter of milk, with units of days/liter.

This parameter is given by Table 4-12. The fraction df total feed that is water. The ratio of specific activity of the feed grass to the atmospheric water. Absolute humidity of the atmosphere, 0.008 kilograms/cubic meter, according to Regulatory Guide 1.109. The atmospheric dispersion parameter, seconds/cubic meter. 1.0 x 10-5 seconds/cubic meter for releases from the 50 foot stack. Refer to Appendix B, 1.2. 6.59 x 1 o-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B, 2 .1. Once the spent fuel pool is emptied, this source tenn and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-25 4.3.12 Tritium Grass-Cow-Meat Pathway Dose Calculation Methodology (x/1 (0.75 x o.5] RMeat, H3 = / Q) X H ) X QF X Ua X FM X DF a (4-4 c) Equation (C-9) from Regulatory Guide 1.109 where: QF = Ua = DFa = FM = 0.75 = 0.5 = H = x/Q = The cow's vegetation consumption rate: 50 kg/day per Regulatory Guide 1.109, Table E-3. The receptor's meat consumption rate. See Table 4-6 and Table 4-7. The ingestion dose factor for H-3, for the receptor in age group (a), in mrem/pCi, from Tables 4-8 through 4-11. The fraction of the animal's intake ofH-3 which appears in a kilogram of meat, with units of days/kilogram.

This parameter is given by Table 4-13. The fraction of total feed that is water. The ratio of specific activity of the feed grass to the atmospheric water. Absolute humidity of the atmosphere, 0.008 kilograms/cubic meter, according to Regulatory Guide 1.1 09. The atmospheric dispersion parameter, seconds/cubic meter. 1.0 x 1 o-5 seconds/cubic meter for releases from the 50 foot stack. Refer to Appendix B, 1.2. 6.59 x 1 o-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B, 2.1. Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE VOLUME 4 SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL REVISION 27 PAGE 11-26 4.3.13 Tritium Vegetation Pathway Dose Calculation Methodology The concentration of tritium is based on the airborne concentration rather than the deposition:

Rvog,H3 = (:%) X ( 0*75 0*5) x Ur x DF, (4-4d) where: UT = H = 0.75 == 0.5 = DFa = x/Q = The total consumption rate of fruits and vegetables, kilogram/year.

This parameter is given in Tables 4-6 and 4-7. Absolute humidity of the atmosphere, 0.008 gm/m 3 per Regulatory Guide 1.109. The fraction of total feed that is water. The ratio of specific activity ofH-3 in the feed grass to the specific activity in atmospheric water. The ingestion dose factor for H-3, for the receptor in age group (a), in mrem/pCi, from Tables 4-8 through 4-11. The atmospheric dispersion parameter, seconds/cubic meter. 1.0 x 1 o-5 seconds/cubic meter for releases from the 50 foot stack. Refer to Appendix B, 1.2. 6.59 x 1 o-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B, 2.1. Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-27 Table 4-1 Inhalation Dose Factors for Adult Age Group (mrem/pico-Curie inhaled) Selected Nuclides from Regulatory Guide 1.109, Table E-7 and from NUREG/CR-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.58x1o-7 1.58x1o-7 1.58x1o-7 1.58x1o-7 1.58x1o-7 Co-60 No Data 1.44 x 1o-6 1.85 x 1 o-6 No Data 7.46 x 1o-4 3.56x 1o-5 Sr-90 1.24 x 1o-2 No Data 7.62 x 1o-4 No Data 1.20 x 1 o-3 9.02 x 1o-5 Cs-137 5.98 x 1o-5 7.76 x 1o-5 5.35 x 1o-5 2.78 x 1o-5 9.40 x 1o-6 1.05 x 1 o-6 Y-90 2.61 x *1o-7 No Data 1.01 x 1o-9 No Data 2.12 x Io-5 6.32 x 1o-5 Pu-241 3.42 x 1o-2 8.69 x 1o-3 1.29 x 1 o-3 5.93 x 1o-3 1.52 x 1o-4 8.65x1o-7 Gross a 1.68 1.13 7.75 x 1o-2 5.04 x 1o-1 1.82 x Io-1 4.84 x 1o-5 Table 4-2 Inhalation Dose Factors for Teen Age Group (mrem/pico-Curie inhaled) Selected Nuclides from Regulatory Guide 1.109, Table E-8 and from NUREG/CR-4013 Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.59xio-7 1.59x1o-7 1.59x1o-7 1.59x1o-7 1.59x1o-7 Co-60 No Data 1.89 x 1 o-6 2.48 x Io-6 No Data 1.09 x 1o-3 3.24 x 1o-5 Sr-90 1.35 x 1o-2 No Data 8.35 x 1o-4 No Data 2.o6 x 1o-3 9.56 x 1o-5 Cs-137 8.38 x 1o-5 1.06 x 1o-4 3.89x 1o-5 3.80 x 1o-5 1.51 x 1o-5 1.06 x 1o-6 Y-90 3.73 x 1o-1 No Data 1.00 x 1o-8 No Data 3.66 x 1o-5 6.99 x 1o-5 Pu-241 3.74 x 1o-2 9.56 x 1o-3 1.40 x 1 o-3 6.47 x 1o-3 2.60 x 1o-4 9.17 x 1o-1 Gross a 1.77 1.20 8.05 x 1o-2 5.32 x 1o-1 3.12 x 1o-1 5.13 x 1o-5 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-28 Table 4-3 Inhalation Dose Factors for Child Age Group (mrem/pico-Curie inhaled) Selected Nuclides from Regulatory Guide 1.109, Table E-9 and from NUREG/CR-40 13 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.04 x 1o-7 3.04 x 1o-7 3.04 x 1o-7 3.04 x 1o-7 3.04 x 1o-7 Co-60 No Data 3.55 x 1o-6 6.12 x 1o-6 No Data 1.91 x 1o-3 2.60 x 1o-5 Sr-90 2.73 x 1o-2 No Data 1.74 x 1o-3 No Data 3.99 x 1o-3 9.28x1o-5 Cs-137 2.45 x 1o-4 2.23 x 1o-4 3.47 x 1o-5 7.63 x 1o-5 2.81 x 1o-5 9.78 x 1o-1 Y-90 1.11 x 1o-6 No Data 2.99 x 1o-8 No Data 7.07 x 1o-5 7.24 x 1o-5 Pu-241 7.94 x 1o-2 1.75 x 1o-2 2.93 x 1o-3 1.10 x 1o-2 5.06 x 1o-4 8.90 x 1o-7 Gross a 2.97 1.84 1.28 x 1o-1 7.63 x 1o-1 6.08 x 1o-1 4.98 x 1o-5 Table 4-4 Inhalation Dose Factors for Infant Age Group (mrem/pico-Curie inhaled) Selected Nuclides from Regulatory Guide 1.109, Table E-10 and from NUREG/CR-4013 Oq an Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 4.62x1o-7 4.62x1o-7 4.62x1o-7 4.62x1o-7 4.62 x 1o-7 Co-60 No Data 5.73 x 1o-6 8.41 x 1o-6 No Data 3.22 x 1o-3 2.28 x 1o-5 Sr-90 2.92 x 1o-2 No Data 1.85 x 1o-3 No Data 8.03 x 1o-3 9.36 x 1o-5 Cs-137 3.92 x 1o-4 4.37 x 1o-4 3.25 x 1o-5 1.23 x 1o-4 5.09 x 1o-5 9.53 x 1o-7 Y-90 2.35 x 1o-6 No Data 6.30 x 1o-8 No Data 1.92 x 1o-4 7.43 x 1o-5 Pu-241 8.43 x 1o-2 1.85 x 1o-2 3.11 x 1o-3 1.15 x 1o-2 7.62 x 1o-4 8.97x1o-7 Gross a 3.15 1.95 1.34 x 1o-1 7.94 x 1o-1 9.03 x 1o-1 5.02 x 1o-5 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-29 -Table 4-5 External Dose Factors for Standing on Contaminated Ground (mrem/hour per pico-Curie/square meter) Selected Nuclides from Regulatory Guide 1.1 09, Table E-6 Total Nuclide Skin Body H-3 0 0 Co-60 2.oo x 1o-8 1.70x1o-8 Sr-90 2.60 x 1o-12 2.20 x 1o-12 Cs-137 4.90 x 1o-9 4.20 x 1o-9 Y-90 2.60 x 1o-12 2.20 x 1o-12 Values are not provided for Transuranic radionuclides, as their dose contribution to this pathway is negligible.

Table 4-6 Average Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

From Regulatory Guide 1.109, Table E-4 Other Seafood Fruits and Age Group Fish (Invertebrates)

Vegetables Milk Meat Adult 6.9 1.0 190 110 95 Teen 5.2 0.75 240 200 59 Child 2.2 0.33 200 170 37 Infant 0 0 0 0 0 Table 4-7 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

From Regulatory Guide 1.1 09, Table E-5 Other Seafood Fruits and Age Group Fish (Invertebrates)

Vegetables Milk Meat Adult 21 5.0 520 310 110 Teen 16 3.8 630 400 65 Child 6.9 1.7 520 330 41 Infant 0 0 0 330 0 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM SAFSTORIDECOMMISSIONING OFFSITE VOLUME 4 TITLE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-30 Table 4-8 Ingestion Dose Factors for Adult Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-ll and from NUREG/CR -4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.05xlo-7 1.05xlo-7 1.05xlo-7 1.05xlo-7 1.05xlo-7 Co-60 No Data 2.14 x 1o-6 4.72 x 1o-6 No Data No Data 4.02 x lo-5 Sr-90 7.58 x lo-3 No Data 1.86 x lo-3 No Data No Data 2.19 x lo-4 Cs-137 7.97 x lo-5 1.09 x lo-4 7.14 x lo-5 3.70 x lo-5 1.23 x 1 o-5 2.11 x 1o-6 Y-90 9.62 x lo-9 No Data 2.58 x 1o-10 No Data No Data 1.02 x lo-4 Pu-241 1.57 x 1 o-5 7.45xlo-7 3.32xlo-7 1.53 x 1o-6 No Data 1.40 x 1o-6 Gross a 7.55 x 1o-4 7.05 x 10-4 5.41 x 1o-5 4.07 x 1o-4 No Data 7.81 x lo-5 Table 4-9 Ingestion Dose Factors for Teen Age Group ( mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-12 and from NUREG/CR-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.06 x 1o-7 1.06 x lo-7 1.06 x 1o-7 1.06 x 1o-1 1.06 x 1o-7 Co-60 No Data 2.81 x 1o-6 6.33 x 1o-6 No Data No Data 3.66x1o-5 Sr-90 8.30 x 1o-3 No Data 2.05 x 1o-3 No Data No Data 2.33 x 1o-4 Cs-137 1.12 x 1o-4 1.49 x lo-4 5.19 x 10-5 5.07x 1o-5 1.97 x 1 o-5 2.12 x 1o-6 Y-90 1.37 x 1o-8 No Data 3.69 x 1o-1o No Data No Data 1.13 x lo-4 Pu-241 1.75 x 1o-5 8.40 x lo-7 3.69xlo-7 1.11 x 1o-6 No Data 1.48 x 1o-6 Gross a 7.98 x lo-4 7.53 x 1o-4 5.75 x lo-5 4.31 x lo-4 No Data 8.28 x 1o-5 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-31 Table 4-10 Ingestion Dose Factors for Child Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-13 and from NUREG/CR-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 2.03x1o-7 2.03x1o-7 2.03 x 1o-1 2.03x1o-7 2.03x1o-7 Co-60 No Data 5.29x1o-6 1.56 x 1 o-5 No Data No Data 2.93 x 1o-5 Sr-90 1.10 x 1o-2 No Data 4.31x1o-3 No Data No Data 2.29 x 1o-4 Cs-137 3.27 x 1o-4 3.13 x 1o-4 4.62 x 1o-5 1.02 x 1o-4 3.67 x 1o-5 1.96 x 1 o-6 Y-90 4.11 x 1o-8 No Data 1.10 x 10-9 No Data No Data 1.11 x 1o-4 Pu-241 3.87 x 1o-5 1.58 x 1o-6 8.04x1o-7 2.96 x 1o-6 No Data 1.44 x 1o-6 Gross a 1.36 x 1o-3 1.11 x 1o-3 1.02 x 1o-4 6.23 x 1o-4 No Data 8.03 x Io-5 Table 4-11 Ingestion Dose Factors for Infant Age Group ( mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-14 and from NUREG/CR-4013 Or! an Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.08x1o-7 3.08 x 1o-1 3.08 x 1o-1 3.08 x 1o-1 3.08 x 1o-1 Co-60 No Data 1.08 x 1o-5 2.55 x 1o-5 No Data No Data 2.57 x 1o-5 Sr-90 1.85 x 1o-2 No Data 4.71 x 1o-3 No Data No Data 2.31 x 1o-4 Cs-137 5.22 x 1o-4 6.11 x 1o-4 4.33 x 1o-5 1.64 x 1o-4 6.64 x 1o-5 1.91 x 1 o-6 Y-90 8.69 x 1o-8 No Data 2.33 x 1o-9 No Data No Data 1.20 x 1o-4 Pu-241 4.25 x 1o-5 1.76x 10-6 8.82x1o-7 3.17 x 1o-6 No Data 1.45 x 1o-6 Gross a 1.46 x 1o-3 1.21 x 1o-3 1.09 x 1o-4 6.55 x 1o-4 No Data 8.10 x 1o-5 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 27 DOSE CALCULATION MANUAL PAGE 11-32 Table 4-12 Stable Element Transfer Data For Cow-Milk Pathway (days/liter)

Selected Nuclides from Regulatory Guide 1.109, Table E-1 and from NUREG/CR-4013 Element Fm H 1.0 x 1o-2 Co 1.0 x 1o-3 Sr 8.o x 1o-4 Cs 1.2 x 1o-2 y 1.0 x 1o-5 Pu 5.o x 1o-6 Gross a 5.o x 1o-6 Table 4-13 Stable Element Transfer Data For Cow-Meat Pathway (days/kilo-gram)

Selected Nuclides from Regulatory Guide 1.109 , Table E-1 and from NUREG/CR-4013 Element Ff H 1.2 x Io-2 Co 1.3 x Io-2 Sr 6.0 x Io-4 Cs 4.0 x Io-3 y 4.6 x Io-3 Pu 2.0 x Io-4 Gross a 2.0 x Io-4 NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-33 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE 5.1 WHOLEBODYDOSE Specification 2.10 limits the whole body dose equivalent from the Uranium fuel to no more than 25 mrem/year.

The whole body dose is determined by summing the calculated doses from the following:

a. Deleted b. Modular HEPA Ventilation Particulate releases, using equation ( 4-3). c. Deleted. Tritium is no longer a gaseous effluent source term. d. Liquid releases, No longer applicable.

To this calculated exposure is added potential direct radiation exposure to an individual at the site boundary.

The only portion of the site boundary where there is significant direct radiation is near the radwaste facilities at the [PG&E] North edge of the site. Due to the possibility that an individual at the shoreline (fishing, bird watching, etc.) may use the path at the brow of the cliff for access, the TLD stations along the path are used to estimate an annual radiation exposure.

The time period used for this estimate is 67 hour7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br />s/year, given by Table E-5 of Regulatory Guide 1.109, as the maximum time for shoreline recreation for the Teen age group. 5.2 SKIN DOSE Specification 2.10 limits the dose to any organ (thyroid excepted) to less than or equal to 25 mrem/year.

The dose to the skin is determined by summing the calculated doses from the following:

a. Deleted b. Modular HEPA Ventilation releases, using equation ( 4-3). Tritium is no longer a gaseous effluent source term. c. Liquid releases, No longer applicable.

d. The potential direct radiation exposure to an individual at the site boundary based on TLD stations, as determined in Section 5.1 above.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL 5.3 DOSE TO OTHER ORGANS VOLUME 4 REVISION 27 PAGE 11-34 Specification 2.10 limits the dose to any organ (thyroid excepted) to less than or equal to 25 mrem/year.

The dose to any individual other than skin organ is determined by summing the calculated doses from the following:

a. Deleted b. Modular HEPA Ventilation releases, using equation (4-3). c. Liquid releases, No longer applicable.

d. The potential direct radiation exposure to an individual at the site boundary based on TLD stations, as determined in Section 5.1 above. 5.4 DOSE TO THE THYROID Specification limits the dose to the thyroid to less than or equal to 75 mrem/year.

Since Unit 3 has not operated since July 2, 1976, there is an insufficient radioactive iodine source term remaining onsite to approach this limit. Therefore, calculation of dose to the thyroid is not required.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-35 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING SOLIDIFICATION Deleted -Based on the status of decommissioning, HBPP no longer anticipates wastes exceeding a specific activity that is unacceptable to disposal site without solidification or exceeding Class A as defined in 1 0 CFR 61. 7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN HIGH INTEGRITY CONTAINERS 112/16 I Deleted -HBPP no longer anticipates wastes exceeding a specific activity that is unacceptable to disposal site without solidification or exceeding Class A as defmed in 10 CFR 61. HBPP no longer anticipates disposal of wastes requiring stabilization in a High Integrity Container (HI C). 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED RESINS AND OTHER WET WASTES 8.1 SCOPE This section pertains to bead-type spent radioactive demineralizer resin, filters and other wet wastes shipped for land burial which contain a total specific activity less than the disposal site(s) criteria for solidification, and which does not exceed the concentration limits for Class A waste as defined in 1 0 CFR 61. 8.2 PROGRAM ELEMENTS 8.2.1 The dewatered resin or wet wastes must meet the requirements of 10 CFR 61.56 or those of the disposal site( s) (whichever is more restrictive) for freestanding, noncorrosive liquid. 8.2.2 For bead resins, the preceding criterion will be met by following approved Plant Manual procedures for dewatering resin. 8.2.3 Liquid waste, that will not be thermal treated to remove freestanding liquid, must be solidified.

8.2.4 Contract vendor solidification or dewatering services are utilized in accordance with PG&E approved supplier list and procurement procedures.

8.2.5 Vendor services may be conducted off site in accordance with their facility license and procedures.

Vendor services include written confirmation of acceptable disposal waste form.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-36 8.2.6 Gross dewatering of resins and filters may be performed onsite to achieve transport requirements in preparation for additional processing to a final waste form by offsite vendor services.

8.2. 7 On site activities, such as managing wet soils from decommissioning excavations and process water shall be performed utilizing approved procedures or work instructions to ensure compliance with transportation regulations, disposal facility license requirements and/or waste acceptance criteria.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL VOLUME 4 REVISION 27 PAGE 11-37 9.0 PROGRAM CHANGES 112/16 I 9.1 PURPOSE OF THE OFFSITE DOSE CALCULATION MANUAL . The Offsite Dose Calculation Manual was developed to support the implementation of the Radiological Effluent Technical Specifications required by 10 CPR 50, Appendix I, and 10 CPR 50.36. The purpose of the manual is to provide the NRC with sufficient information relative to effluent monitor setpoint calculations, effluent related dose calculations, and environmental monitoring to demonstrate compliance with radiological effluent controls.

9.2 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL It is recognized that changes to the ODCM may be required during the Decommissioning period. All changes shall be reviewed and approved by the HB Director prior to implementation.

The NRC shall be informed of all changes to the ODCM by providing a description of the change(s) in the first Annual Radioactive Effluent Release Report following the date the change became effective.

Records of the reviews performed on change to the ODCM should be documented and retained for the duration of the possession only license. 9.3 HBPP is allowed to modify or reduce environmental requirements in the ODCM provided HBPP considers the modification or reduction from a technical and decommissioning 112/16 I perspective.

[CMT 10.1] 10.0 COMMITMENTS 12/16 I 10.1 HBPP does not intend to modify or reduce the environmental monitoring requirements as specified in the ODCM during the period of SAPS TOR and decommissioning activities.

This applies to those environmental samples and analysis identified as either quality or non-quality samples. This commitment is to be incorporated into the next revision of the ODCM. NOTE: HBPP is allowed to modify or reduce environtnental requirements in the ODCM provided HBPP considers the modification or reduction from a technical and decommissioning perspective.

11.0 RESPONSIBLE ORGANIZATION Radiation Protection Manager APPENDIX A SAFSTOR BASELINE CONDITIONS ODCM APPENDIX A Revision 27 Page A-1 ODCM APPENDIX A Revision 27 Page A-2 1.0 LIQUID AND GASEOUS EFFLUENTS 1.1 LIQUID EFFLUENTS Baseline levels of radioactive materials contained in liquid effluents during the SAPS TOR period were established in the Environmental Report submitted as Attachment 6 to the SAPS TOR license amendment request. These values are presented for cumulative annual release and average monthly discharge in Table A -1. As of December 31, 2013, HBPP ceased processed liquid effluent to the discharge canal and processed liquid effluent will be transpotied for disposal at a regulated disposal site. Storm water and groundwater associated with excavations and groundwater inleakage to structures during decommissioning will typically be treated and released using the Ground Water Treatment System in accordance with the Storm Water Pollution Prevention Plan (SWPP) and the associated NPDES permit. The GWTS is an Active Treatment System (ATS) is designed to remove suspended solids in order to meet release criteria of the SWPP. The system will be limited to treating water containing soluble radionuclides less than 10 times the "new" 10 CPR 20, Appendix B, Table 2, Column 2 effluent concentration limits (ECLs) in order to ensure concentrations at the Site Boundary are maintained less than limiting condition 2.3.1. 1.2 GASEOUS EFFLUENTS Tritium Baseline levels of radioactive materials contained in gaseous effluents established in the Environmental Report are presented for cumulative annual and average monthly release in Table A-2. Table A-1 Baseline Liquid Effluent Activity Type of Activity Annual Release Monthly Average Release (Curies) (Curies) 8.60E-2 7.17E-3 Principal Gamma Emitters (total) 1.85E-1 1.54E-2 Strontium-90 3.28E-4 2.73E-5 Table A-2 Baseline Gaseous Effluent Activity Type of Activity Annual Release (Curies) Tritium <4.0E-2 Particulate Gamma Emitters (total) 3.16E-4 Strontium-90 3.38E-6 ODCM APPENDIX A Revision 27 Page A-3 Monthly Average Release (Curies) <3.3E-3 2.63E-5 2.82E-7 Table A-3 below reflects the Gaseous Effluent Activity as a representation of the state of decommissioning during the calendar year 2013 relative to the Baseline above. Table A-3 2013 Gaseous Effluent Activity Type of Activity Annual Release Monthly Average Release (Curies) (Curies) Particulate Gamma Emitters (total) <1.5E-5 <1.3E-6 Strontium-90

<1E-6 <1E-7 Particulate Alpha Emitters (total) <1E-6 <1E-7 Table A-3 data is summarized from the 2013 Annual Effluent Release Report and are listed as less than values because sampling results were the composite of LLD values. Tritium is no longer monitored due to a lack of significant source term.

APPENDIXB ODCM APPENDIXB Revision 27 Page B-1 BASES FOR ATMOSPHERIC DISPERSION AND DEPOSITION VALUES 1.0 BASIS FOR DISPERSION/DEPOSITION VALUES -50' STACK ODCM APPENDIXB Revision 27 Page B-2 1.1 The instantaneous atmospheric dispersion factor (X/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0, titled "Determine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) ofthe calculation (N238C) provides "1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />" values for the instantaneous X/Q for the 50' stack for various stack flow rates, based on an EPA model named "IS CST". The instantaneous X/Q value used in the ODCM (6.52 x 10-4) is based on a stack flow of25,000 cfm. 1.2 The annual average atmospheric dispersion factor (X/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0, titled "Determine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) ofthe calculation (N238C) provides annual maximum values for X/Q for the 50' stack for various stack flow rates, based on an NRC model named "XOQDOQ".

The annual average X/Q value used in the ODCM (1.00 x 1 o-5) is based on a stack flow of 25,000 cfm. 1.3 The annual average atmospheric deposition factor (D/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0, titled "Detennine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) ofthe calculation (N238C) provides annual maximum values for D/Q for the 50' stack for various stack flow rates, based on an NRC model named "XOQDOQ".

The annual average D/Q value used in the ODCM (3.00 x 10-8) is based on a stack flow of25,000 cfm. 2.0 BASIS FOR DISPERSION/DEPOSITION VALUES -INCIDENTAL RELEASE PATHS 2.1 The atmospheric dispersion factor (X/Q) for incidental releases is 6.59 X 1 o-3 seconds/cubic meter, calculated as described below 2.1.1 This factor is based on the atmospheric models ofRegulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. These models are intended to estimate meteorological dispersion for "real tilne" conditions (i.e., hourly), rather than "annual average" conditions.

The applicable guidance is section 1.3 .1 (Releases Through Vents or Other Building Penetrations);

as it applies to all releases from points lower than 2.5 times the height of adjacent structures.

This calculation generally follows the guidance for the use of equations 1, 2 and 3 of Regulatory Guide 1.145.

ODCM APPENDIXB Revision 27 Page B-3 2.1.2 The assumed distance from the emission point to the potential receptor for this calculation is 150 meters. This is the approximate distance to publicly accessible areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the trail at the edge of the bluff). 2.1.3 The meteorological conditions assumed for this calculation are for stable "fumigation" conditions (Pasquill stability class G), with a wind speed of 1 meters/ second. 2.1.4 The applicable equations from Reg. Guide 1.145 are as follows: where: X/Q 1 (1) X/Q 1 (2) 1 X/Q = (3) ulO wind speed at 10 meters above grade, equal to 1 meter/second.

cry lateral plume spread, equal to 4.33 meters for Pasquill Class G at a distance of 150 meters. az vertical plume spread, equal to 1.86 meters for Pasquill Class G at a distance of 150 meters. A vertical cross-sectional area of structures, equal to 3 7 5 meters 2 , based on the Refueling Building dimensions (about 36 feet high, about 112 feet long). L:Y lateral plume spread (including meander and building wake), meters, equal to 6cry (for distances less than 800 meters, wind speeds below 2 meters/second, and stability class G). 2.1.5 With these values, the results for equations 1, 2, and 3 are as follows: X/Q = 4.70 x 10-3 seconds/meter 3 (1)

XIQ 1.32 X 10-2 seconds/meter 3 XIQ 6.59 X 10-3 seconds/meter 3 ODCM APPENDIXB Revision 27 Page B-4 (2) (3) Per the Reg. Guide, the higher value of equations 1 and 2 is to be compared with the value for equation 3, and the lower value of that comparison should be used, with this logic, the resulting value for X/Q is 6.59 x 1 o-3 seconds/meter

3. 2.2 The atmospheric deposition factor (D/Q) for incidental releases is 5.39 x 10-6 mete{2 for the Particulate Ground Plane Pathway, and is 3.29 x 10-6 mete{2 for all other deposition related pathways.

The factors are calculated as described below 2.2.1 These factors are based on the atmospheric models of Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-water-cooled Reactors.

The applicable guidance is section C.3.b (Dry Deposition), and Figure 6 (Relative Deposition for Ground-level Releases).

To determine the atmospheric deposition across a downwind sector, the value from Figure 6 is to be multiplied by the fraction of the release transpmied into the sector, and divided by the sector cross-wind arc length at the distance being considered.

For this calculation, the deposited contamination will be assumed to be evenly distributed across the width of the plume, rather than across an arbitrary angular sector. 2.2.2 Two factors are necessary because the nearest location (along the bay) is not a credible location for farming. For the purposes of estimating offsite doses from incidental releases, the nearest "farm" will be assumed to be beyond the railroad tracks, southeast of the plant. 2.2.3 For the Particulate Ground Plane Pathway, the assumed distance from the emission point to the potential receptor for this calculation is 150 meters. This is the approximate distance to publicly accessible areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the trail at the edge of the bluff). At this distance, Figure 6 provides a Relative Deposition Rate value of 1.4 X 1 o-4 meter-1. The plume width assumed for this calculation is the same as was used in equation 3 of section 2.1.4 (above), so that the plume width is approximately 6cry. For cry equal to 4.33 meters (Pasquill Class Gat a distance of 150 meters), D/Q is (1.4 x 10-4 mete{1)/ (6 x 4.33 meter)= 5.39 x 10-6 mete{2. 2.2.4 For the pathways involving farming or ranching, the assumed distance from the emission point to the potential receptor for this calculation is 220 meters. This is the approximate distance to publicly accessible "grazing" areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the other side of the railroad).

At this distance, ODCM APPENDIXB Revision 27 Page B-5 Figure 6 provides a Relative Deposition Rate value of 1.2 X 1 o-4 mete{1. The plume width assumed for this calculation is the same as was used in equation 3 of section 2.1.4 (above), with the plume width of approximately 6cry., but at a greater distance.

For cry equal to 6.07 meters (Pasquill Class Gat a distance of220 meters), D/Q is (1.2 x 10-4 meter-1)/ (6 x 6.07 meter)= 3.29 x 10-6 mete{2*

APPENDIXC Deleted ODCM APPENDIXC Revision 27 Page C-1