Text

Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) (TAC No. MF6050) - Response to NRC Request for Additional Information - Radiation Protection and Consequence Branch
	ML16054A661
Person / Time
Site:	Watts Bar
Issue date:	12/22/2015
From:	James Shea; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CNL-15-216, TAC MF6050, WBN-TS-15-03
	Download: ML16054A661 (131)
	v • d • e

1101 Market Street, Chattanooga, Tennessee 37402 CNL-15-216 December 22, 2015 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License Nos. NFP-90 NRC Docket No. 50-390

Subject:

Application to Revise Technical Specification 4.2.1, "Fuel Assemblies" (WBN-TS-15-03) (TAC No. MFBOSO) - Response to NRC Request for Additional Information - Radiation Protection and Consequence Branch

Reference:

1. Letter From VA to NRC, CNL-1 5-001, "Application to Revise Technical Specification 4.2.1, 'Fuel Assemblies,' (WBN-TS-15-03),"

dated March 31, 2015 (ML15098A446)

2. Letter from WVA to NRC, CNL-1 5-077, "Correction to Application to Revise Technical Specification 4.2.1, 'Fuel Assemblies' (WBN-TS-15-03)," dated Aprill 28, 2015 (ML15124A334)

3. Letter From NRC to TVA, 'W/atts Bar Nuclear Plant, Unit 1.'

Supplemental Information Needed for Acceptance of Requested Licensing Action Regarding Application to Increase Tritium Producing Absorbing Rods (TAC NO. MFOO5O)," dated May 14, 2015 (ML15127A250)

4. Letter from WVAto NRC, CNL-1 5-092, "Response to NRC Request to Supplement the Application to Revise Technical Specification 4.2.1, 'Fuel Assemblies' (WBN-TS-15-03)," dated May 27, 2015 (MISIt547A6I1t)

U. S. Nuclear Regulatory Commission CNL-1 5-216 Page 2 December 22, 2015

5. Letter from TVA to NRC, CNL-1 5-093, "Response to NRC Request to Supplement Application to Revise Technical Specification 4.2.1,

'Fuel Assemblies' (WBN-TS-15-03) - Radiological Protection and Radiological Consequences," dated June 15, 2015 (ML15167A359)

6. Letter from TVA to NRC, CNL-15-172, "Application to Revise Technical Specification 4.2.1, "Fuel Assemblies" (WBN-TS-15-03)

(TAC No. MF6050) - Response to NRC Request for Additional Information - Reactor Systems Branch," dated September 14, 2015 (ML15258A204)

7. Electronic Mail from Jeanne Dion (NRC) to Thomas A. Hess (TVA) and Clinton Szab0 (TVA), 'T'PBAR RAls Part 3b- ARCB," dated October 2, 2015 By letter dated March 31, 2015 (Reference 1), Tennessee Valley Authority (TIVA) submitted a license amendment request (LAR) to revise Watts Bar Nuclear Plant (WBN), Unit 1 Technical Specification (TS) 4.2.1, "Fuel Assemblies," to increase the maximum number of Tritium Producing Burnable Absorber Rods (TPBARs) that can be irradiated per cycle from 704 to 1,792. The proposed change also revises TS 3.5.1, "Accumulators," Surveillance Requirement (SR) 3.5.1.4 and TS 3.5.4, "Refueling Water Storage Tank (RWST)," SR 3.5.4.3 to delete outdated information related to the Tritium Production Program. TIVA provided a correction letter on April 28, 2015 (Reference 2).

By letter dated May 14, 2015 (Reference 3), the Nuclear Regulatory Commission (NRC) requested that TIVA provide additional information to supplement the LAR. TVA provided the requested supplemental information in TVA letters dated May 27, 2015, and June 15, 2015 (References 4 and 5, respectively).

By electronic mail dated October 2, 2015 (Reference 7), the NRC requested that TVA provide additional information to support the NRC review of the LAR. The response to the request for additional information (RAI) was due November 2, 2015. The due date was subsequently extended to December 24, 2015. Enclosure 1 to this letter provides "IVA's RAl response.

There is one new regulatory commitment associated with this submittal. New regulatory commitment 3 is associated with the response to Radiation Protection and Consequences Branch (ARCB) RAl l .b. Enclosure 2 provides a complete updated commitment list that supersedes the previous commitment lists provided in the Reference 1 and Reference 6 letters.

Consistent with the standards set forth in Title 10 of the Code of FederalRegulations (10 CFR), Part 50:92(c), TVA has determined that the additional information, as

U. S. Nuclear Regulatory Commission CNL-15-216 Page 3 December 22, 2015 provided in this letter, does not affect the no significant hazards consideration associated with the proposed application previously provided in Reference 1.

Additionally, in accordance with 10 CFR 50.91 (b)(1), TVA is sending a copy of this letter and the enclosures to the Tennessee Department of Environment and Conservation.

Please address any questions regarding this request to Mr. Edward D. Schrull at (423) 751-3850.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 2 2 nd day of December 2015.

Re *ectfully, J.V Sheba Vice President, Nuclear Licensing

Enclosures:

1 Tennessee Valley Authority, Watts Bar Nuclear Plant, Unit 1, Response to NRC Request for Additional Information

2. Watts Bar Nuclear Plant, Unit 1 Tritium Producing Burnable Absorber Rods License Amendment Request Updated Regulatory Commitment List Enclosures cc (Enclosures):

NRC Regional Administrator - Region II NRC Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

ENCLOSURE I TENNESSEE VALLEY AUTHORITY WA'l-S BAR NUCLEAR PLANT UNIT I Tennessee Valley Authority, Watts Bar Nuclear Plant, Unit 1, Response to NRC Request for Additional Information Contents:

NRC Request for Additional Information Attachments:

I DOE-HDBK-1079-94, "Primer on Tritium Safe Handling Practices"

2. WBNNAL3003, Revision 5, "Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-1 8.1-1984"

3. WBNTSR1 00, Revision 12, "Design Releases to Show Compliance with I10CFR20" CNL-15-216 CNL-15-216Enclosure 1, Page 1 of 14

NRC Request for Additional Information By letter dated March 31, 2015, the Tennessee Valley Authority (TVA) submitted an application for license amendment to revise the Technical Specifications to increase the maximum number of tritium producing burnable absorber rods and to delete outdated information related to the tritium production program at Watts Bar Nuclear Plant (WBN) Unit 1 (ADAMS Accession No. ML15098A446). These changes would revise TS 4.2.1 ," Fuel Assemblies," TS 3.5.1 Accumulators," Surveillance Requirement (SIR) 3.5.1.4, TS 3.5.4, "Refueling Water Storage Tank," and SR 3.5.4[.3]. T'VA supplemented this request with letters dated in May and June 2015 (ADAMS Accession Nos. ML15147A611I and ML15167A359). In response to previous requests for information from NRC staff, TVA submitted letters dated September 14 and 25, 2015 (ADAMS Accession Nos. ML15258A204 and ML15268A568).

The staff of the Radiation Protection and Consequence Branch (ARCB) of the office of Nuclear Reactor Regulation (NRR) has reviewed the application and supplement dated June 15, 2015 (ADAMS Accession No. ML15167A359) and has determined that additional information is required to complete the review.

ARCB Request for Additional Information (RAI) 1 Supplement dated June 2015, Enclosure 1, Page 2 of 9, indicates that the radiation protection tritium control program is based in part on Regulatory guide 8.32, "Criteria for Establishing a Tritium Bioassay Program," and DOE-HDBK-1 079-94, "Primer on Tritium Safe Handling Practices."

a. Provide a copy of DOE-HDBK-1079-94, and describe how its guidance is incorporated into the actions outlined in the table on page 3 of 9.

TVA Response A copy of DOE-HDBK-1079-94, "Primer on Tritium Safe Handling Practices," is provided as Attachment 1Ito this enclosure. TVA has utilized DOE-HDBK-1079-94 since the start of the Tritium Production Program at WI'A as an information resource for personnel who manage and supervise the tritium production function at WVA. Although this information was used as background information in the development of the tritium control procedures, DOE-HDBK-1079-94 was not used to establish any of the action levels in Radiological Control Instruction (RCI)-137, "Radiation Protection Tritium Control Program," as provided in the table on page 3 of 9 of WVA Supplement dated June 2015, Enclosure 1. Please note that RC1-1 37, Table 3.1, is being revised in response to ARCB RAl l .b. Because this supplemental information merely clarifies the use of DOE-HDBK-1079-94 in developing WVA RC1-137, no change to the WVA letter dated June 15, 2015, is warranted.

b. Regulatory Guide 8.32 stipulates that bioassay for tritium (H-3) be provided for individuals that work around 10 kg or more of open reactor coolant with H-3 concentrations above .01 Ci/Kg.

Provide a basis for why this is not an action statement in the table on page 3 of 9.

WVA Response W-A will revise RC1-137, "Radiation Protection Tritium Control Program," Table 3.1, 'Tritium Action Levels," to incorporate the 0.01 Curies/kilogram (Ci/kg) (i.e., 10 pCi/g) criteria from NRC Regulatory Guide (RG) 8.32, "Criteria for Establishing a Tritium Bioassay Program," prior CNL-15-216 CNL-1-216Enclosure 1, Page 2 of 14

to increasing the number of TPBARs loaded in the reactor core above the currently allowed 704 TPBARs. The associated 0.01 pCi/mi (i.e., 10 pCilg) action level will be specified by RC1-I137, Table 3.1. In addition, RC1-I137, Table 3.1 will be simplified as follows.

TRITIUM ACTION LEVELS Basis for Bioassay Process Tritium (Regulatory guidance Concentration DAC, DAC- Mode of Tritium Survey and TVA Procedure (piJ~lml) hrs Exposure Requirements Recommended Action Requirements)

NIA direct measurement of Urinalysis foliowing skin US NRC Regulatory

> 0.01 contact process water contact, ingestion, or Guide 8.32 absorption through cuts or abrasions. Divng RCDP-7 requires routine bioassays as specified

___________in Note1..

~ 00inhalation measurement of Udinalysis foliowing US NRC Reguiatory

> 00process water exposure to air in a Guide 8.32 and tditium air room whenever samples empioyees are exposed to greater than 10 kg of water containing 0.01 Ci/kg or when water containing a total of more than 0.1 Ci of tritiumn is in contact with

___________ ___________air (such as a fuel pool).

> 0.3 DAC inhalation tritium air Urinalysis RCDP-7 samples recommended, see Note

______________2

> 4 DAC-hrs inhalation tritium air Urinalysis shall be RCDP-7, basis is 10 in 7 sampies with requested for any mremlweek, which is consecutive DAC-hr tracking employee who exceeds easily detected and days this limit verified by bioassay.

N**'For underwater diving operations in trdtiated water exceeding 0.01 pCi/mI, RCDP-7 Bioassay and Internal Dose Program specifies that collection and analysis of urine samples is recommended for each dive. (a) prior to the first on-site dive, (b) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the completion of the initial dive, (c) once each week while diving operations are in progress, (d) upon completion of diving operations, and (e) whenever diving suit leakage results in skin contact with tritlated water.

NoFor work activities where workers are known or may be exposed to tritium atmospheres exceeding 0.3 DAC or other site project criteria, the collection and analysis of urine is recommended as detailed: (a) pre-job should be performed to establish a baseline value, (b) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the completion of the first exposure, (c) weekly to ten days for the duration of the work involvng tritium exposure, and (d) upon completion of the work involving tritium exposure.

The proposed RC1-1 37, Table 3.1 simplifies the information provided in response to the NRC Request to Supplement E2-1 in TVA letter dated June 15, 2015.

TVA procedure RC1-137 along with TVA procedures NPG-SPP-05.1, "Radiological Controls,"

and RCDP-7, "Bioassay and Internal Dose Program," provide a graded approach for bioassay based on risk, work, and airborne conditions. TIVA's program establishes criteria for performing in vitro bioassay: (1) based on process water concentrations using the guidance in RC1-137, Table 3.1, (2) when an individual worker's exposure is greater than or equal to four Derived Air Concentration (DAC)-hours (hrs) in seven consecutive days, and (3) for skin contamination with tritiated water concentrations exceeding 0.01 pCi/mI. Tritium DAC-hr tracking is initiated whenever a worker is in an airborne area, exposed to concentrations of 0.3 DAC or greater. TVA procedures cover types of bioassay, selection of individuals for bioassay, bioassay collection, sample volume, sample storage, packaging and shipping, detection limits, and internal dose calculation methods. TVA requires a Minimum Detectable Activity (MDA) of less than I E+04 pCi/L (i.e., 0.01 pdi/L) tritium for urine bioassays. This MDA is less than the detection limit of 0.3 pdi/L specified in ANSI N13.14 and for most CNL-15-216 CNL-1-216Enclosure 1, Page 3 of 14

conditions (e.g., bioassay obtained within two weeks of exposure) will achieve detection of an intake resulting in a committed effective dose equivalent less than I mrem. The frequency of bioassays is determined based on the work, the exposure scenario, and trigger levels as described in RCDP-7. Baseline tritium bioassays are obtained for all divers prior to the first on-site dive. Baseline (pre-job) tritium bioassays are also required for work activities where workers are known or may be exposed to tritium atmospheres exceeding the trigger levels of greater than 0.3 DAC. All data (i.e., bioassay, air samples with DAC-hr tracking, and whole body counts) are reviewed and evaluated to arrive at the best estimate of the worker's intake and Committed Effective Dose Equivalent (CEDE) for each radionuclide. In some cases, a single bioassay is enough to evaluate the exposure; in other cases, multiple bioassays are obtained for work in tritium airborne conditions that may continue for days or weeks.

Sample collection guidance in RCDP-7 for tritium includes collecting urine samples no sooner than two hours following the tritium exposure event to allow the activity to equilibrate, and no later than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following an acute exposure if possible. The first voiding of the bladder following the exposure is not used for the urinalysis. Sample collection instructions are provided to the worker with the bioassay container. This guidance in RCDP-7 is consistent with the guidance from RG 8.32.

ARCB RAI 2 Supplement dated June 2015, Enclosure 2, page 33, provides estimated doses to the maximally exposed member of the public in Table 10. Verify that these doses were calculated using the methods, assumptions, and input parameters consistent with the latest version of the Watts Bar Offsite Dose Calculation Manual (ODCM). If this is not the case provide the calculations and calculational basis for these estimated doses.

TVA Response The calculated doses provided in Enclosure 2 of the June 15, 2015, TIVA supplement were calculated using the methods and assumptions in Revision 24 of the WBN 0DCM using the realistic source term data as input. Revision 25 of the WBN 0DCM has since been issued. The changes made to the document did not affect the calculation results. Both revisions of the WBN ODCM were provided to the NRC in a letter dated May 1, 2015, 'Watts Bar Nuclear Plant Unit 1, Annual Radioactive Effluent Release Report - 2014" (ML15121A826).

CNL-15-216 CNL-1-216Enclosure 1, Page 4 of 14

ARCB RAI 3 Supplement dated June 2015, Enclosure 2, pages 9, 23 and 24, indicate that the amounts of H-3 released from Watts Bar in radwaste discharges are estimated to increase by a factor of about 14 (or up to 26,889 Ci per year). Since this license amendment request is to allow loading a maximum of 1792 TPBARs into the core, and the 26,889 Ci/yr value is based on a TPC of 2500 TPBARs, verify that these statements are incorrect and that there is no intentions of releasing 26,889 Ci/yr of H-3. Clarify and provide corrected text.

TVA Response The 26,889 Ci/yr of tritium value was used to demonstrate the adequacy of the Radioactive (Radwaste) System. TWA does not intend to release 26,889 Ci/yr of tritium.

The expected tritium release for operation at the licensed TPBAR loading limit would be 6,093 Ci/year based on 1,792 TPBARs; the expected permeation rate would be 3.4 Ci/yr for the Mark 9.2 TPBAR. In no case is the tritium release from WBN, Unit 1 projected to exceed the established regulatory limits.

The information provided in Enclosure 2 of the June 2015 supplement on page 9 is revised as follows with deleted text lined out:

CONCLUSION Upon review of the documents and the analyses described above, this review determined that there were no substantial changes, that is, WBN Unit I will continue to demonstrate effluent release performance well within the regulatory As Low as Reasonably Achievable (ALARA) public dose guidelines of 10 CFR Part 50 Appendix I and occupational radiation exposure continues to be bounded by the station dose assessment of record10 , associated with the radiological impact analyses that were relevant to environmental concerns, nor were there any significant new circumstances or information relevant to environmental concerns which bore on the radiological impacts associated with the tritium production program. The impact of WBN Unit 1 operation with a TPC containing up to 2,500 TPBARs (Design Basis) will have a minimal impact on the Radwaste System Design Basis and realistic fission and corrosion product sources and the treatment of these isotopes in liquid and gaseous waste". Tho.m., Radact.. So,,m*.. Doin. Bacic *. ...... uro t-...iu- o..octimat.d to.

As indicated in Table 10, "Annual Projected Impact of TPC (1,900 TPBARs) on Effluent Dose to Maximally Exposed Members of the Public and Total Public Dose," the differences noted in the source terms for TPC operation would not affect the ability of the plant to remain within the applicable regulatory requirements relative to radioactivity in effluents to unrestricted areas (i.e., 10 CFR Part 20), the "as low as is reasonably achievable" criterion (i.e., 10 CFR Part 50 - Appendix I). Ho '-o or, RDwat, Syct,.om,.=, Docign.. B,...i= tr,,iu, CNL-15-216 1, Page 5 of 14 CNL-15-216Enclosure

The information provided in Enclosure 2 of the June 2015 supplement on page 23 is revised as follows with deleted text lined out and inserted text underlined:

Normal Operation Because of weepage through valve stems and pump shaft seals, some coolant escapes into the containment and the auxiliary buildings. A portion of the RCS leakage flashes to steam/evaporates, thus contributing to the tritiated water vapor source term, and a fraction remains as liquid, becoming part of the liquid source term. The relative amount of leakage entering the gaseous and liquid phases is dependent upon the temperature and pressure at the point where the leakage occurs. 10% due to flashing and Spent Fuel Pool (SFP) evaporative losses is the assumed gaseous effluent fraction for dose impact modeling (NUREG-0017, Revision 1), whereas WBN Unit 1 effluent history indicates an average of =5.0%.

As Tritiated water vapor is not removed by filtration or ion exchange, it will be released as gaseous effluent to the environment. A breaker-to-breaker run will potentially produce the maximum RCS tritium concentration. Cycles 11 and 12 with 544 TPBARs were estimated to peak atme* just less than 7.0 pCi/gm. With the assumption of routine boron cotoand 2,500 TPBARs at 10 Ci/TPBAR/year,.

the estimated average Design Basis RCS tritium concentration is calculatedestimated to be 29.8 pCi/gm. With the assumption of 1,900 TPBARs at 5 Ci/TPBAR/year, the estimated average Realistic Basis RCS tritium concentration is calculatedestimated to be 12 pCi/gm 32 .

and The calculatedpeeee tritium release to the RCS with a TPC containing TPBARs releasing tritium at the radwaste system design basis maximum rate (i.e., Table 3)I will result in about a factor of fourteen increase over the Non-TPC tritium production rate, that is, Ratio = (TPC) 26,889 Ci/year / (Nominal Core) 1,889 Ci/year = 14.2 The expected tritium release for operation at the licensed TPBAR loading limit would be 6.093 Ci/vear based on 1,792 TPBARs: the expected permeation rate would be 3.4 Ci/yr for the Mark 9.2 TPBAR.

The information provided in Enclosure 2 of the June 2015 supplement on page 24 is revised as follows with deleted text lined out and inserted text underlined:

Radwaste System Design Basis Operation The radwaste system design basis source term was updated from the License Amendment 40 source term to reflect a new permeation rate of 10 Ci/TPBAR/year, which bounds the realistic permeation rate with 100% margin. This source term also has increased margin by assuming -40% more TPBARs are loaded. The source term was also updated to delete the contribution of two failed TPBARs, because such failures are not the design basis case for radwaste system operation. It should be noted that the two failed TPBAR assumption previously used reflected an unrealistic and excessively conservative case, because the CNL-15-216 CNL-1-216Enclosure 1, Page 6 of 14

maximum TPBAR tritium Ci loading occurs at end of cyclecycle. life and Thethemid maximum cycle RCS tritium concentration due to permeation occurs mid case is the proper design basis case for radwaste management to support continued operation for the remainder of the cycle. The modified source term provides sufficient margin to bound reasonable off-normal operational cases.

The effect of WBN Unit 1 operation with a TPC containing up to 2,500 TPBARs (Design Basis) will have a minimal effect on the Radwaste System Design Basis and realistic fission and corrosion product sources and the treatment of these isotopes in liquid and gaseous wastes. The Radwaste System Design Basis tritium sources are assumedeetiemated to increase tho amo'-nt of t;rit'-m .h**. ic di.-eharg.. , :*nn.u'll by a factor of about 14. The analyzed gaseous releases are based on the radwaste system design basis tritium source term. For liquid releases, _athe maximum allowable liquid concentration of tritium released to the environment was determined.

ARCB RAI 4 Supplement dated June 2015, Enclosure 2, page 23, indicates that the assumed "design basis source term" (from a tritium production core (TIPC) of 2500 TPBARs, with a H-3 permeability of 10 CIITPBAR/yr) and "realistic source term" (from a TPC of 1900 TPBARs, with a H-3 permeability of 5 CI/TPBAR/yr), result in equilibrium H-3 reactor coolant concentrations of 29.8 Ci/gm and 7.0 Ci/gm respectively. Provide calculations, including all assumptions and parameters, demonstrating how these coolant concentrations were derived from their respective assumed source terms.

TVA Response A copy of WBN calculation WBNNAL3003, Revision 5, "Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-1 8.1-1 984," is provided as Attachment 2 to this enclosure. The calculation describes the methodology and assumptions used to calculate the tritium concentrations in the RCS for the radwaste system design basis and routine effluent realistic source terms.

ARCB RAI 5 Supplement dated June 2015, Enclosure 2, page 25 implies that the radwaste mobile demineralizer is effective for processing H-3 from liquid radwaste. Provide the decontamination factor expected (and basis), or clarify this text.

TVA Response The discussion of the radwaste system operation provided in Enclosure 2 of the June 15, 2015, TVA supplement describes the overall system operation with an effluent stream containing all of the expected radionuclides. It does not suggest that the system is capable of processing tritium.

Tables 5, 6, and 7 provided in the Enclosure 2 of the June 15, 2015, TVA supplement are updates to the information presented in Updated Final Safety Analysis Report (UFSAR) Tables 11 .2-4a, 11 .2-4b, and 11 .2-4c to reflect the requested TPBAR loading increase. Information on the radwaste system processing capabilities is provided in UFSAR Table 11.2-4. The decontamination factor provided for tritium is 1, which means no decontamination.

CNL-15-216 CNL-1-216Enclosure 1, Page 7 of 14

ARCB RAI 6 Supplement dated June 2015, Enclosure 2, pages 24 and 25, indicate that Enclosure 2 Tables 5 through 7 demonstrate that the Effluent Concentration Limits (ECL) in 10 CFR 20 Appendix B are met assuming a TPC of 2500 TPBARs, with a H-3 permeability of 10 CI/T'PBAR/yr (i.e., the design source term). However, the TPC H-3 concentration entries in all three of these tables is set at the "maximum allowable tritium concentration" with no explanation of how this value was derived or any nexus to the assumed source term.

a. Provide the basis for this maximum allowable tritium concentration, and how is derived from the design basis source term.

TVA Response A copy of WBN calculation WBNTSR100, Revision 12, "Design Releases to Show Compliance with 10CFR20," is provided as Attachment 3 to this enclosure. The purpose of this calculation is to determine if 1% failed fuel (design fuel damage) would result in effluent releases exceeding 10 CFR Part 20 Appendix B Table 2 limits. This calculation takes the expected liquid releases and scales these releases to design levels (i.e., 1% failed fuel) taken from UFSAR data. The 10 CFR Part 20 limit requires this sum to not exceed unity. For liquid releases of tritium, a maximum allowable liquid concentration of tritium is back-calculated based on the available margin for the liquid release scenarios.

b. How does TVA propose to administer this maximum allowable tritium concentration? Will it be incorporated into the plant Technical Specifications or some other on-site document?

TV/A Response TWA administers the maximum allowable tritium concentration in accordance with WBN, Unit 1 TS 5.7.2.3, "Offsite Dose Calculation Manual," and 5.2.7.7, "Radioactive Effluent Controls Program." The ODCM specifies that a sum of the ratios calculation be performed to demonstrate compliance with TS 5.7.2.7.b. The sum of the ratios (R) for each release point is determined by the following relationship.

where:

R = the sum of the ratios for the release point.

C1 = concentration of radionuclide i, pCi per milliliter (ml).

ECLI = the ECL of radionuclide i, 1iCi/ml, from 10 CFR Part 20, Appendix B, Table 2, Column 2.

The sum of the ECL ratios must be < 10 following dilution due to the releases from any or all of the release points described above.

CNL-15-216 CNL-1-216Enclosure 1, Page 8 of 14

The permitted maximum concentration of any single nuclide (including tritium) is unique to each scheduled release, because it is a function of the sum total of radionuclide activity present in the release and the respective ECLs.

c. Demonstrate by calculation that the Watts Bar radwaste processing systems are capable of annually releasing the design basis source term, while maintaining H-3 concentrations below the maximum allowable concentration.

TVA Response WBN calculation WBNTSR1 00, provided as Attachment 3 to this enclosure, demonstrates that the WBN radwaste processing systems are capable of controlling releases in compliance with 10 CFR Part 20. Tables 6 and 7 in Enclosure 2 of the June 15, 2015, TVA supplement show aggregate concentration/ECLs that are _<1 ifthe tritium concentration after dilution is

< 3.26E-04 pCi/cc, which is equivalent to releasing approximately 13,145 Ci/yr of tritium. This tritium release represents approximately 55% of the design basis source term. Therefore, an amount representing approximately 45% of the design basis source term would need to be held up. The majority of the tritium comes from the Chemical Volume Control System (CVCS) letdown and primary coolant equipment drains, which accounts for approximately 2,512 gpd.

As stated above, approximately 45% of the release volume (i.e., approximately 413,000 gal) would be held up for later release. The Tritiated Water Storage Tank (TWST) has a capacity of 500,000 gal and therefore can hold this volume until more favorable release conditions exist.

Because most WBN liquid releases are performed in a batch mode, a more conservative evaluation is based on instantaneous discharge. As discussed in Enclosure 2 of the June 15, 2015, TIVA supplement, the peak RCS tritium concentration is expected to be approximately 2.5 times greater than the average. For the design basis source term, this would result in a peak concentration of approximately 74.5 pCilgm. The table below provides the quarterly waste volume, dilution volume, and radioactive waste volume reported in the annual radioactive effluent release reports (ARERRs) for calendar years 2008 through 2014.

The last column provides the average dilution factor for the quarter by taking the ratio of the total volume of water released (waste + dilution) and the radioactive waste volume released.

As shown in the table below, more than half of the quarters for calendar years 2008 through 2014 had an average dilution factor greater than 1 E+04, which results in a peak concentration after dilution of 7.5 E-02 IJCi/gm and concentration/ECL of 7.5. Consistent with the evaluation in the original DOE topical report and approval in NUREG-1672, the WBN ODCM allows the sum of the concentrations to be less than 10 times the concentration values specified in 10 CFR 20, Appendix B, Table 2, Column 2. For time periods where a dilution factor of 1 E+04 is not achieved, the TWST can be used to holdup those effluents until more favorable release conditions exist.

CNL-15-216 CNL-15-216Enclosure 1, Page 9 of 14

Total Waste Volume of Radioactive Volume Dilution Water Waste Volume ARERR Released Used Released Report Period (Liters) (Liters) (Liters) Dilution Factor 2008________

01 1.14E+08 1.12E+ 10 2.36E+06 4.79E+03 Q2 1.50E+08 1.10E+10 1.44E+06 7.74E+03 03 1__.56E+08 2.43E+10 1.84E+06 1.33E+04 Q4 1.64E+08 1.01E+10 1__.10E+06 9.33E+03 2009 _ _____

01 2.68E+08 6.46E+09 5.1 3E+05 1.31 E+04 Q2 1.82E+08 1.64E+10 1__.12E+06 1.48E+04 Q3 2.68E+08 1.57E+10 __1.58E+06 1.01E+04 04 1.82E+08 5.45E+09 1.03E+06 5.47E+03 2010__ _ _ _ _ _ ___ _ _ _ _ _ _

01 2.49E+08 8.64E+09 3.47E+05 2.56E+04 Q2 2.87E+08 1.48E+10 1.32E+06 1.14E+04 03 2.38E+08 1.77E+ 10 3.84E+06 4.67E+03 04 1.83E+08 1.27E+ 10 2.65E+06 4.86E+03 2011 _ _ _ _ _ _

01 1.52E+08 5.48E+08 1.16E+06 6.03E+02 02 1.87E+08 5.46E+09 1 .62E+06 3.49E+03 03 2.88E+08 1.90E+10 1.32E+06 1.46E+04 Q4 2.70E+08 9.28E+09 __1.44E+06 6.63E+03 2012__ _ _ _ _ _ _ __ _ _ _ _ _ _ __ _ _ _ _ _ _

01 1.13E+10 3.13E+08 5.95E+05 1__.95E+04 02 2.31E+09 9.20E+09 7.07E+05 1.63E+04 Q3 2.08E+08 9.55E+09 1 .93E+06 5.06E+03 Q4 2.70E+08 1.18E+10 8.62E+05 1.40E+04 2013_____

Q1 2.81E+08 2.42E+10 5.95E+05 4.11E+04 Q2 2.39E+08 1.81 E+10 8.69E+05 2.11E+04 03 1.47E+08 1.14E+ 10 1.22E+06 9.46E+03 Q4 1.69E+08 2.20E+10 4.06E+06 5.46E+03 2014 _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _

Q1 1.37E+08 1.19E+10 1.57E+06 7.67E+03 Q2 1.77E+08 1.56E+ 10 1 .49E+06 1 .06E+04 Q3 9.85E+07 4.93E+10 9.95E+05 4.96E+04 04 1.19E+08 1.37E+10 8.18E+05 1.69E+04

d. Tables 5 and 7 do not demonstrate that dual unit operation will result in annual liquid effluent discharges within the limits of 10 CFR 20, Appendix B. Correct the incorrect text on page 25 and provide a basis for why this is acceptable.

TVA Response Tables 6 and 7 in Enclosure 2 of the June 15, 2015, TVA supplement show aggregate ECLs that are < 1, demonstrating discharges do not exceed the limits of 10 CFR Part 20, CNL-15-216 CNL-15-216Enclosure 1, Page 10 of 14

Appendix B. Table 5 shows the result without radwaste system(same processing. WBN, Unit 1 UFSAR Section 11.2.6.5.2 notes that UFSAR Tables 11 .2-4a information as Table 5 except for revised tritium source term) and UFSAR Table 11 .2-4b (same information as Table 6 except for revised tritium source term) describe liquid releases for both untreated and treated waste relative to the requirements of 10 CFR Part 20. The current licensing basis UFSAR Table 11 .2-4a shows the sum of the concentrations/ECL for all isotopes is greater than unity for the case where all isotopes are at design values and the released liquid is not processed by the Mobile Demineralizers. The effects of the updated tritium source term shown in Table 5 leads to the same conclusion. Radwaste processing through the Mobile Demineralizers demonstrates results that do not exceed 10 CFR Part 20 requirements, as shown in UFSAR Table 11 .2-4b (current licensing basis) and Table 6 (for the revised tritium source term).

The results in Tables 6 and 7 represent a conservative (but not realistic) continuous release scenario. The radwaste system processing at WBN, Unit I is performed in batch mode. The releases are processed in accordance with the 00CM to ensure compliance with TS 5.7.2.7.b.

The information provided in Enclosure 2 of the June 2015 supplement on page 25 is revised as follows with deleted text lined out and inserted text underlined:

Tables 6 and5-h:eugh 7 demonstrate that the liquid releases do not exceedae-belew the 10 CFR Part 20 Appendix B Table 2 limits.

e. Provide a basis for operating at (Table 7) or near (Table 6) the maximum ECLs is ALARA or acceptable.

TV/A Response Tables 6 and 7 are based on the radwaste system design basis source term and are intended to demonstrate the adequacy of the radwaste system. These tables are not intended to demonstrate that as low as reasonably achievable (ALARA) objectives are met with respect to dose commitment to a member of the public. The ALARA objectives for radioactive effluent releases are met by implementing the 0DCM in accordance with WBN, Unit I TSs 5.7.2.3 and 5.2.7.7.

ARCB RAI 7 Supplement dated June 2015, Enclosure 2, page 7, indicates that the revised design basis source term no longer assumes two failed TPBARs. The source term used to design the reactor facility radwaste systems should include anticipated operational occurrence (AOOs) as well as normal operations. Provide a basis for not including this previously included AOO in the source term assumptions.

TI/A Response The calculations developed for the initial tritium program for the routine effluent realistic source term, the radwaste system design basis source term, and the non-loss of coolant accident CNL-15-216 CNL-1-216Enclosure 1, Page 11 of 14

(LOCA) source terms (realistic and design basis) used the conservative assumption of tritium from two TPBAR failures in addition to the tritium permeation."

TVA revaluated the source term assumptions for the March 31, 2015, LAR. TIVA noted that Department of Energy (DOE) document NDP-98-1 81, Revision 1, "Tritium Production Core (TPC)

Topical Report," Section 2.11.2.2 characterized the failure of a TPBAR to be beyond a reasonable design basis consideration and not expected to occur during normal operation or for any Anticipated Operational Occurrence (AOO). TVA has accumulated significant experience with TPBAR irradiation without experiencing a TPBAR failure.

TVA concluded that incorporating an assumption of two TPBAR failures for the routine effluent realistic source term is inconsistent with the regulatory intent of this source term. As such, TIVA removed the assumption of two failed TPBARs from this source term. The TPC realistic source term has a tritium component that is conservatively based on an assumed tritium permeation of 5 Ci/yr from 1,900 TPBARs. The source term is conservative in that it assumes: 1) a permeation rate that is -50% more than the expected permeation rate for the Mark 9.2 TPBAR and 2) a number of TPBARs that is -6% more than the 1,792 TPBARs requested by the March 31, 2015, LAR.

WI'A also concluded-that incorporating an assumption of two TPBAR failures for the radwaste system design basis source term is not appropriate, given the definition of an AOO.

10 CFR Part 50, Appendix A defines AOOs as those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power. Therefore, WVA removed the assumption of two failed TPBARs from this source term. The TPC radwaste system design basis source term has a tritium component that is conservatively based on an assumed tritium permeation of 10 Ci/yr from 2,500 TPBARs. The source term is conservative in that it assumes:

1) a permeation rate that is almost three times more than the expected permeation rate for the Mark 9.2 TPBAR and 2) a number of TPBARs that is -40% more than the 1,792 TPBARs requested by the March 31, 2015, LAR. These conservatisms represent a reasonable source term for evaluating the capability (i.e., sizing) of the radwaste system to handle off normal events.

WVA retained the assumption of two TPBAR failures for the non-LOCA accident source term.

This source term has a tritium component that is conservatively based on an assumed tritium permeation of 10 Ci/yr from 2,500 TPBARs. The source term is conservative in that it assumes:

1) a permeation rate that is almost three times more than the expected permeation rate for the Mark 9.2 TPBAR and 2) a number of TPBARs that is -40% more than the 1,792 TPBARs requested by the March 31, 2015, LAR. The assumption of two TPBAR failures is conservative given the DOE failure assessment in NDP-98-181, Revision 1 and the operational experience at WBN. The non-LOCA accident source term provides sufficient margin to bound the TPC for Chapter 15 AOO and Postulated Accident dose consequence evaluations.

ARCB RAI 8 Supplement dated June 2015, Describe the measures (including onsite surveillance and groundwater monitoring) that WVA will employ to insure that Watts Bar operations with this increased H-3 source term will be within the requirements of NRC Bulletin No. 80-10, "CONTAMINATION OF NONRADIOACTIVE *SYSTEM AND RESULTING POTENTIAL FOR CNL-15-216 CNL-15-216Enclosure 1, Page 12 of 14

UNMONITORED, UNCONTROLLED RELEASE OF RADIOACTIVITY TO ENVIRONMENT," and 10 CFR 20.1406.

TVA Response Based on a September 28, 2015, conference call with the NRC, TVA understands that ARCB RAI 8 does not request a response to NRC Bulletin 80-10 for WBN. Instead, the RAI is requesting an assessment of the effects of the LAR on previous assessments of the potential for contamination of groundwater or non-radioactive water systems.

TWA provided a summary of the groundwater monitoring program for WBN, Unit I in the letter to NRC dated August 4, 2006, "Groundwater Protection - Data Collection Questionnaire for Browns Ferry, Sequoyah, and Watts Bar Nuclear Plants" (ML062280598). The letter was submitted as a result of an industry initiative sponsored by the Nuclear Energy Institute (NEI) to compile baseline information about the status of site programs for monitoring and protecting groundwater and to share that information with NRC.

WBN, Unit i is committed to the industry initiative described in NEI 07-07, "Ground Water Protection Initiative," (ML072610036) as documented in UFSAR Section 1.1.3. NEI 07-07 identifies actions to improve utilitY management and response to instances where the inadvertent release of radioactive substances may result in low but detectible levels of plant-related materials in subsurface soils and water.

The most recent results of the WVA groundwater monitoring program were provided to the NRC in the letter dated May 1, 2015, 'W*/atts Bar Nuclear Plant Unit 1, Annual Radioactive Effluent Release Report - 2014" (MLL5121 A826).

UFSAR Section 9.2.4 describes the plant features for the Potable and Sanitary Water Systems.

UFSAR Section 9.2.4.1.2, "Safety Evaluation," states that the potable water system is not cross-connected with any radioactive system. UFSAR Section 9.2.4.2.3, "Safety Evaluation," states that the sanitary water system does not receive radioactive waste. The NRC review of the Potable and Sanitary Water Systems is documented in NUREG-0847, "Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units I and 2" (ML072060490), dated June 1982.

NDP-98-1 81, Revision I addresses the Potable and Sanitary Water Systems as follows:

SRP 9.2.4 Potable and Sanitary Water Systems: The acceptance criteria in this section are based on the relevant requirements of 10 CFR 50, Appendix A, GDC 60. They deal with the design provisions provided to control the release of liquid effluents containing radioactive material from contaminating the Potable and Sanitary Water System (PSWS). The design of the PSWS is not being modified, therefore, the design features which prevented contamination of the PSWS in the reference plant (i.e., no cross-connection between the PSWS and any potentially radioactive system and the use of backflow prevention devices where plumbing fixtures are located in areas susceptible to potential radiological hazard) are still present for the TPC plant. Therefore, there is no impact on this system.

CNL-15-216 CNL-1-216Enclosure 1, Page 13 of 14.

In NUREG 1672, "Safety Evaluation Report related to the Department of Energy's Topical Report on the Tritium Production Core," dated March 1999, the NRC made the .following decision regarding NDP-98-1 81:

Potable and' Sanitary Water Systems (SRP Section 9.2.4)

DOE evaluated the effect of TPBARs on the Potable and sanitary water systems for the reference plant against the guidance of SRP Section 9.2.4 and concludes that there is no effect on this system. The staff agrees With this evaluation.

These conclusions are consistent with the tritium program operational experience at WBN, Unit 1 and remain valid for the proposed operation with 1,792 TPBARs requested by the March 31, 2015, LAR.

For systems that are considered as nonradioactive, but could possibly become radioactive through interfaces with radioactive systems, WBN Chemistry Manual, Chapter 3.01, "System Chemistry Specifications," -establishes the following routine sampling/analysis or monitoring program.

Refueling Water Storage Tank Tritium checked per request (associated with Component Cooling Water System Tritium checked per request (associated with Spent Fuel Pool Tritium checked quarterly Feedwater Tritium checked weekly Demineralized Water Head Tank Tritium checked monthly Condensate Storage Tanks Tritium checked per request (associated with Auxiliary Boiler Tritium checked monthlydurnn operation Raw Cooling Water Tritium checked quarterly Yard Holding Pond Tritium checked weekly Low Volume Waste Treatment Pond Tritium checked weekly Potable Water System Tritium checked weekly The system features, monitoring program, and Commitments described in the response to ARCB RAI 8 are not changed by the proposed LAR.

CNL-15-216 CNL-15-216Enclosure 1, Page 14 of 14

ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 TVA Response to NRC Request for Additional Information Attachment 1 DOE-HDBK-1 079-94, "Primer on Tritium Safe Handling Practices" CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I I, Page 1 of 48 Attachment

DOE-H DBK-1 079-94 December 1994 DOE HANDBOOK PRIMER ON TRITIUM SAFE HANDLING PRACTICES U.S. Department of Energy FSC-691 0 Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 1, Page 2 of 48 Attachment

This document has been reproduced directly from the best available copy.

Available to DOE and DOE contractors from the Office of Scientific and Technical Information. P.O. Box 62, Oak Ridge, TIN 37831; (615) 576-8401.

Available to the public from the U.S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161.

(703)487-4650.

Order No. DE95003577 CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 1, Page 3 of 48 Attachment

DOE-HDBK-1079-94 TPJTIUMSAFE HANDLING PRACTICES FOREWORD The Primer on TritiUm Safe Handling Practices is approved for use by all DOE Components. It was developed as an educational supplement and reference for operations and maintenance personnel involved in tritium handling. This Primer is intended to help operations and maintenance personnel perform their duties safely by increasing their knowledge of tritium handling and control, thus reducing the likelihood of tritium releases and exposures, and by providing basic emergency responses in the case of a tritium release.

The Department of Energy (DOE) Primers are a set of fundamental handbooks on safety-related topics of interest in the DOE Complex. The Primers are written as an educational aid for operations and maintenance personnel. The Primers attempt to supply information in an easily understandable form which will help them perform their duties in a safe and reliable manner. Persons trained in other technical areas may also find the Primers useful as a guide or as a reference source for further investigation.

The DOE Primer series draws heavily upon the subject-specific Primers and training materials previously developed by DOE sites (Savannah River, Rocky Flats, and Mound) and is intended for distribution to all DOE contractors. Information is also drawn from the applicable volumes of the DOE FundamentalsHandbook series developed by the DOE Office of Nuclear Safety Policy and Standards. References to other material sources are indicated in the text where applicable and a bibliography is included.

Beneficial commtents in the form of recommendations and any pertinent data that may be of use in improving this document should be addressed to John A. Yoder EH-63/GTN U.S. Department of Energy Washington, D.C. 20585 Phone (301) 903-5650

Facsimile (301) 903-6172 by using the U.S. Department of Energy Standardization Document Improvement ProPosal Form (DOE F 1300.x) appearing at the end of this document or by letter.

Key Words: Hydrogen, Tritium, Half-.life, Radiation, Beta particles, Contamination, Health Physics, Monitoring, Bioassay, Metabolism, Emergency Response Rev. 0 Tritiumi CNL-15-216, Enclosure 1 Attachment. 1, Page 4 of 48

DOE-HDBK-1079-94 TRITIUM SAFE HANDLING PRACTICES OVERVIEW Tritium Safe HandlingPracticeswas prepared as an information resource for personnel who perform tritium handling functions at DOE facilities. A basic understanding of the properties and hazards associated with tritium will help personnel understand the impact of their actions on the safe and reliable operation of facility systems.

Tritiumn Safe HandlingPracticescontains an introduction and sections on the following topics:

Radiological Fundamentals

Physical and Chemical Properties of Tritium

Biological Properties of Tritium

Tritium Monitoring

Radiological Control and Protection Practices

Emergency Response This Primer is provided as an information resource only, and is not intended to replace any radiation worker or hazardous materials training. The Primer presents the theoretical concepts and good practices that form the basis of safe tritium handling.

Specific references have been cited as footnotes, and a bibliography is available at the end of the Primer. These sources provide further reading and specific guidance. This document contains selected information from the Health PhysicsManual of Good Practicesfor Tritium Facilities,which was prepared by EG&G Mound Applied Technologies at the Mound tritium facility. The authors acknowledge their expertise and experience.

Tritium Rev. 0 CNL-i5-216, Enclosure I Attachment 1, Page 5 of 48

Department of Energy Handbook PRIMER ON TRITIUM SAFE HANDLING PRACTICES Enclosure I1 CNL-15-216, Enclosure CNL-15-216, Attachment 1, Page 6 of 48 Attachment 1, Page 6of 48

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Tritium Primer DOE-HDBK-1079-94 TABLE OF CONTENTS TABLE OF CONTENTS LIST OF FIGURES ...................................................... ii LIST OF TABLES........................................................ iv INTRODUCTION...........................1 RADIOLOGICAL FUNDAMENTALS ........................................ 3 Hydrogen and Its Isotopes.. ............ *............ ,.......... ........ 3 Sources of Tritium......................... ............. ............. 4 Stable and Unstable Nuclides .............................. ....... ..... 5 Ions and Ionization................... .... ........................... 5 Types of Radiation.................... .................. ......... ... 6 Radioactivity........................................... ...... ..... 9 PHYSICAL AND CHEMICAL PROPERTIES OF TRITJIUM.......................11 Nuclear and Radioactive PrOperties .................................... . 11 Penetration Depths of Beta Particles ............ .... ......... ............. 11 Chemical Properties........................13 Contamination..........................14 BIOLOGICAL PROPERTIES OF TRITTIUM.................15 Metabolism of Gaseous Tritium....................15 Metabolism of Tritiated Water ......................................... 15 Metabolism of Other Tritiated Species ............... .............. 16 Metallic Getters.........................16 Tritiated Liquids.........................16 Other Tritiated Gases ............................................... 17 Biological Half-Life of HTO......................17 Bioassay and Internal Dosimetry............ *........................... 18 Sampling Schedule~and Technique...................18 Dose Reduction.............. :......................,,............. .. 19 TRITIUM MONITORING........................21 Air Monitoring.................................................... 21 Differential Air Monitoring...................... ................ 22 Discrete Air Sampling ........................ ....... .......... 22 Process Monitoring .......................................... .. 23 Rev. 0 Page i Tritium CNL-15-216, Enclosure 1 Attachment 1, Page 8 of 48

Tritium Primer DOE-HDBK-1079-94 TABLE OF CONTENTS Surface Monitoring............. ................................... 23 Tritium Probes ...... *........,... ............................ 24 Off-Gassing Measurements.................... ................. 25 Liquid Monitoring................................................. 25 RADIOLOGICAL CONTROL AND PROTECTION PRACTICES ................... 27 Airborne Tritium ................................................. 27 Secondary Containment ..................... ...... ............ 27 Temporary Enclosures ....... ............... .............. ... 28 Protection by Local Ventilation ................................. 28 Supplied-Akr Respirators................................... ... 29 Supplied-Air Suits.................... 29 Protection from Surface Contamination ............. ,....... ............. 29 Protective Clothing ...... ................................... 30 Lab Coats and Coveralls . .............................. 30 Shoe Covers ........... *......... .......................... 30 Gloves................................... ................ 31 EMERGENCY RESPONSE...................... 33 Emergency Steps to Take......... :....... ............... ............. 33 Decontamination of Personnel .. ........... .......................... 33 Decontamination of Surfaces... ............... ........................ 34 BIBLIOGRAPHY......................... 35 Tritium Page ii Rev. 0 CNL-15-216, Enclosure 1 Attachment 1, Page 9 of 48

Tritium Primer DOE-HDBK-1079-94 LIST OFFIGURES LIST OF FIGURES Figure 1 Hydrogen isotopes .................................................. 3 Figure 2 Radioactive decay............... .................................. 5 Figure 3 Neutral and ionized atoms ..................................... ;...... 6 Figure 4 Alpha particle (or nucleus of a helium atom) .............................. 7 Figure 5 Alpha shielding................................................... 7 Figure 6 Neutron shielding ................................................. 9 Figure 7 Half-life example................................................. 10 Figure 8 Reducing HTO uptake by washing after exposure to HTO vapor ............... 32 Rev. 0 Page iii Tritium CNL-15-216, Enclosure 1 Attachment 1, Page 10 of 48

Tritium Primer DOE-HDBK-1079-94 LIST OF TABLES LIST OF TABLES Table 1 Hydrogen isotopes........................4 Table 2 Important nuclear properties of tritium.................................. 11 Table 3 Penetration depths of tritium betas..................12 Tritium Page iv Rev. 0 CNL-15-216, Enclosure 1 Attachment 1, Page 11 of 48

Tritium Primer DOE-HDBK-1079-94 INTRODUCTION INTRODUCTION This Primer is designed for use by operations and maintenance personnel to improve their knowledge of tritium safe handling practices. It is applicable to many job classifications and can

beused as a reference for classroom Work or for self-study. It is presented in general terms so that it can be used throughout the DOE Complex.

The information in this Primer should enable the jeader to do the following:

Describe methods of measuring airborne tritium concentration.

List the types of protective clothing that are effective against tritium uptake from surface and airborne contamination.

Name two methods of reducing thae body dose after a tritium uptake.

Describe the most common method for determining the amount of tritium uptake in the body.

Describe the steps to take following an accidental release of airborne tritium.

Describe the damage to metals that results from absorption of tritium.

Explain how washing hands or showering in cold water helps reduce tritium uptake.

Describe how tritium~ exchanges with normal hydrogen in water and hydrocarbons.

The organization of the Primer is as indicated in the Overview. The following section contains background information on "Radiological Fundamentals." Those familiar with these topics may elect to skip this section and begin reading at the section entitled "Physical and Chemical Properties of Tritiunt" AdditiOnal information about tritium is available from the sources listed in the "Bibliography" section.

Rev. 0 Page 1 Tritium CNL-!5-216, Enclosure 1 Attachment 1, Page 12 of 48

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Tritium Primer DOE-HDBK-1079-94 RADIOLOGICAL FUNDAMENTALS RADIOLOGICAL FUNDAMENTALS This section provides a review of radiological fundamentals. The reader is assumed to be famniliar with this information from radiological worker training. The section discusses hydrogen and its isotopes and describes basic radiological concepts.

Hydrogen and Its Isotopes Atomic nuclei of a particular element (such as hydrogen or oxygen) have the same number of protons (positively charged), but may have a different number of neutrons (no net charge).

Those that have a different number of neutrons are isotopes of that element. Most elements exist in nature in several isotopic forms. For example, hydrogen has one proton. The isotopes of hydrogen either have no neutrons (normal hydrogen, called protium), one neutron (deuterium),

or two neutrons (tritium) (Figure 1). Although isotopes of an element have almost the same chemical properties, the nuclear properties can be quite different.

1

1. Protium - 1H (1 proton, 1 electron)

-Stable, (not radioactive)

-Comprises 99. 985% of natural hydrogen.

2. Deuteriumn- *Nor D (1 proton, 1 neutron, 1 electron)

-Stale notradioactive)"

-Comprises 0.015% of natural hydrogen.

3

3. Tritium - 1 H or T (1 proton, 2 neutrons, 1 electron)

-Radioactive- 18

-Comprises 10 parts of natural hydrogen.

Figure 1 Hydrogen isotopes Tritium

'*L°-15-216, Enclosure 1 ~¶~L15-16,Encosue AttachmentI *apage 14 of 48

RADIOLOGICAL FUNDAMENTALS DOE-HDBK-1079-94 Tritium Primer Nuclear notation uses the chemical symbol (H for hydrogen) and an arrangement of subscripts and superscripts. The total number of protons and neutrons is shown as a superscript: 1H for protium, 211 for deuterium, and 3 H for tritium. The number of protons (which identifies the element) is shown as a subscript. However, the common practice of using H, D and T for these isotopes, respectively, will be followed in this document, except where nuclear reactions are illustrated.

The atomic masses, symbols, and natural abundances of the three isotopes of hYdrogen are given in Table 1.

Table 1 Hydrogen isotopes Symbol Mass Natural Abundance (mass Physical Common Name (%) units)

IHH Protium 99.985 1.007825

HD Deuterium 0.015 2.01400

HT Tritium 1 x 10-i 3.01605 Sources of Tritium Tritium occurs naturally in the environment. Reactions between cosmic radiation and gases in the upper atmosphere produce most of the world's natural tritium. For example,

' 4N 3 0 n-' 1 - H+/-+1 Tritium converts into water and reaches the earth's surface as rain. An estimnated production rate of 4 x 106 Cilyr results in a world steady-state natural inventory of -70 x 106 Ci.a In addition, commercial producers of radioluminescent and neutron generator devices release about 1 x 106 Cilyr. Atmospheric nuclear test explosions from 1945 to 1975 added about 8 x 109~ Ci of trtium to the environment, much of which has since decayed. However, about 5 x i08 Ci remain in the environment, mostly diluted in the oceans. Underground nuclear tests appear to add little tritium to the atmosphere. The nuclear power and defense industries now

a. The curie (Ci) is a unit of activity defined as 3.7x 1010 disintegrations per second (dps). A more basic unit is 1 dps, which is the defini'tion of the becquerel (Bq). Throughout this Primer, the curie will be used instead of the becquerel.

Tritium Rev. 0 CNL-5-21, AttachmentEnlosue I 1,'Page 15 of 48

Tritium Primer DOE-HDBK-1079-94 RADIOLOGICAL FUNDAMENTALS release 1-2 x 106 Ci/yr, a small fraction of which comes from light-water reactors. Tritium is also a by-product of light-water and heavy-water nuclear reactor operation. In their coolants, these reactors produce about 500 to 1,000 and 2 x 106 Ci/yr, respectively, for every 1,000 MW(e) of power. Tritium is a fission product within nuclear fuel, generated at a rate of 1-2 x 104 Ci per year/1000 MW(e). U.S. DOE reactors have produced tritium by the neutron bombardment of lithium (6Li).

Stable and Unstable Nuclides In the lighter elements, the ratio of neutrons to protons mn their stable nuclei is usually 1 (one neutron for every proton). For the heavier elements, this ratio gradually increases to about 1.5 (1.5 neutrons for every proton). Although one cannot always predict from its ratio whether an isotope is stable or unstable, the relationship between the number of protons and neutrons is extremely important.

When an isotope is unstable, its nucleus will emit rays or particles or it may split into two different nuclei. Some combinations of neutrons and protons lead to stable nuclei. If there are too many or too few neutrons, the resulting nucleus is not stable. This unstable nucleus tries to become more stable by releasing excess energy. Atoms with unstable nuclei are radioactive.

The process of nuclei releasing this energy is referred to as radioactivedecay or disintegration (Figure 2). If a nucleus is still unstable after radioactive decay, further decay will occur.

Unstale nuleusParticles emitted Unstblenuceusaccompanied by /

excess energyQ .

Figure 2 Radioactive decay Ions and Ionization Atoms can combine chemically to form molecules. Atoms and molecules are surrounded by orbiting electrons (negatively charged). If the number of electrons equals the total number of protons (positively charged) in the nucleus, the atom or molecule is neutral (uncharged).

Rev. 0 Pace 5 Tritium CNL-15-216, Enclosure. 1 Attachment *l.'Page 16 of 48

RADIOLOGICAL FUNDAMENTALS DOE-HDBK-1079-94 Tritium Primer Electrically charged atoms or molecules are called ions. Ions are either positively or negatively charged, depending on the number of orbiting electrons relative to the number of protons in the nucleus. As shown in Figure 3, ions with more electrons than protons are negatively charged, while ions with more protons than electrons are positively charged. The process of breaking a neutral atom or molecule into electrically charged parts is called ionization. This process requires energy. Ionization removes electrons from the atom, or molecule, leaving an ion with a positive charge. The negatively charged electron (which can attach itself to a neutral atom or molecule) and the positively charged ion, are called an ion pair. Radiation that causes ionization is called ionizing radiation.

Figure 3 Neutral and ionized atoms Types of Radiation There are four basic types of ionizing radiation emitted from nuclei: alpha particles, beta particles, gamma rays, and neutrons.

alpha particle (a)--consists of two protons and two neutrons and is the same as the nucleus of a helium atom (4He) (Figure 4). Generally, oniy the heavy nuclides can emit alpha particles.

A typical example of an az-emitting nuclide is uranium-23 8"

+ 234Th + energy a38 The mass of an alpha particle is about four times the mass of a single neutron or proton, and has a positive charge of++/-2 (it has no electrons). This positive charge causes the alpha particle Tritium CNL15-16,Encosue AttachmentI 1 'Pge 17 of 48 Rev. 0

Tritium Primer DOE-HDBK-1079-94 RADIOLOGICAL FUNDAMIENTALS 2 neutrons, 2 protons no electrons /

Figure 4 Alpha particle (or nucleus of a helium atom) to ionize nearby atoms as it passes through body tissue. The strong positive charge and its relatively slow speed (resulting from its large mass) causes the alpha particle to interact strongly with orbiting electrons of atoms and molecules and to lose large amounts of energy in a short distance. This limits the penetrating ability of the alpha particle, making it easy to stop. A few centimeters of air, a sheet of paper, or the outer layer of skin stops alpha particles (Figure 5).

Figure 5 Alpha shielding Alpha particles are not an external radiation hazard because they are easily stopped by protective clothing or the outer layer of skin. However, if an alpha emitter is inhaled or ingested, it becomes an internal radiation hazard. Because the source is in close contact with body tissue, the alpha particle will dissipate its energy in a short distance of the tissue.

beta particle ([3)--is equivalent to an electron except for its source. Beta-emitting

nuclides have too many neutrons. A neutron emits a [3 particle, and the neutron is then converted to a proton. Tritium decay provides a good example of this process:

]H - [3 + He +energy.

0N-52 Rev. ,Enlsr1 CNL-5-21, Enlosue I 1I,Page 18 of 48 Attachment Tritium

RADIOLOGICAL FUNDAMENTALS DOE-HDBK-1079-94 Tritium Primer A [3 particle is identical to an electron, and its mass and charge are the same as those of an electron. As in the case of alpha particles, beta particles ionize atoms by removing electrons from their orbits. This reaction occurs from charged particle interactions or "collisions" with orbiting electrons.

Beta particles penetrate further than alpha particles of the same energy. A high-energy beta particle can penetrate a few centimeters of organic tissue. The higher the energy, the greater the penetrating ability. However, low-energy beta particles of tritium can be shielded by skin, paper, or only about 6 mm of air.

gamma ray (y)---is emitted when the nucleus of a nuclide releases stored energy without releasing a particle. Many gamma-emitters are found among the products of nuclear fission.

During pure y emission the nucleus does* not emit particles or change its nuclear structure or chemical characteristics. For instance, Gamma radiation is in the form of electromagnetic waves (or photons). Gamma rays are similar to x-rays, but they differ in their origin and energy. Gamma rays originate within the nucleus, and x-rays originate outside the nucleus.

Gamma rays have a very high penetrating power because they have no charge or mass.

Depending on their energy, a stream of gamma rays may penetrate with gradually diminishing intensity through several inches of concrete or similar material. They can be shielded effectively by very dense materials, such as lead and uranium. Gamma rays are a whole-body hazard. That is, because of their penetrating ability, the damage caused by gamma rays is not restricted to any particular body organ.

neutron (n)--may be emitted spontaneously by heavy nuclei during fission or may be emitted during radioactive decay. They are uncharged particles that have mass and a high penetrating ability.

A neutron has about 2,000 times the mass of an electron, but only one-fourth the mass of an alpha particle. Neutrons are difficult to stop because they lack a charge. Neutrons mainly interact with matter by stiig hydrogen nuclei or interacting with the nucleus of atoms. These collisions generally cause charged particles or other radiation to be emitted. These particles may then ionize other atoms. Collisions between neutronsand hydrogen nuclei (protons) are effective in stopping or slowing down high-energy neutrons. Neutrons are best shielded by materials with a high hydrogen content, such as water or plastic (see Figure 6).

Tritium CNL-5-21, AttachEnlment osue I 1e, age 19 of 48 Rev. 0

Thitium Primer DOE-HDBK-1079-94 RADIOLOGICAL FUNDAM'ENTALS 0

Paper Steel Wvater or Plastic Figure 6 Neutron Shielding Radio As radioactive isotopes decay, the number of radioactive nuclei decreases. The time required for half of the nuclei in a sample of a specific radioactive isotope to undergo decay is called its (physical) half-life (Figure 7). Each radioactive isotope has its own characteristic half-life.

Radioactive isotopes decay to less than 1% of their original quantity after about seven half-lives.

Half-lives vary widely with different radionuclides, as shown by the following examples:

' 6N - 7.35 seconds 3H - 12.43 years 2 -8 4.5 x 109 years (4.5 billion years)

The activity of a radioactive isotope sample is defined as the number of nuclei that decay per unit of time.

It has been shown that for a pure radioactive isotope the number of nuclei decaying per unit time (rate of decay) is proportional to the number of nuclei available to decay. If the substance is not being replenished, its activity will decrease accordingly. Therefore, in terms of half-life, the remaining activity after a period of time can be expressed as follows:

At A x (1/2)" or Ao/At 2" CNL15-16,Enclosure 1 Attachment 1,Page 20 of 48 Tritium

RADIOLOGICAL FUNDAMENVTALS DOE-HDBK-1079-94 Tritium Primer

'I)

(U O 50 0

12.5 8*.25

=*3.12 Time (half-lives)

Figure 7 Half-life example where:

Ao = the initial activity (Ci)

At = activity after elapsed time, t (Ci) n = number of half-lives during elapsed time, equal to t/Tla, where t = elapsed time (time units)

Tn = half-life (time units).

For example, assume we have stored 10,000 Ci of tritium. How many curies will be left after 50 years of storage?

A0o = 10,000OCi n =50/12.43 = 4.02 A, 10,000 (1/2)4.02,

- 10,000 x 0.062

- 620 Ci (approximately)

Tritium CNL-15-216, Enclosure 1 AttachmentfT, Page 21 of 48 Rev. 0

Tritium Primer DOE,-HDBK-1079-94 PHYSICAL AND CHEMTCAL PROPERTIES PHYSICAL AND CHEMICAL PROPERTIES OF TRITIUJM This section reviews the nuclear properties of tritium and discusses of some of the physical and chemical properties that are important in understanding tritium handling, containment, and contamination control.

Nuclear and Radioactive Properties Being an isotope of hydrogen, tritium has many of the properties of ordinary hydrogen (such as chemical reactions, permeability, and absorption). Differences may occur because the decaying tritium atoms can speed up (catalyze) reactions of undecayed tritium, or because atoms that have undergone decay have changed into helium atoms (3H-e). Additionally, small differences in chemical reaction rates may result from the relative masses of the isotopes.

Some of the useful properties of'tritium are listed in Table 2. Note that the properties listed are those of T2 . The specific activity and power density of HT and DT are approximately one-half those for T2. The activity density of HiT and DT is exactly one-half that of T2.

Table 2 Impgortant nuclear properties of tritium Half-Life 12.43 yrs Specific Activity 9,545 Ci/g Power Density 0.328 W/g Activity Density (Tz gas, 1 atm, 0°C) 2.589 Ci/cm3 (T2 gas, 1 atm, 25 °C) 2.372 Ci/cm 3 Penetration Depths of Beta Particles The penetration and absorption of beta particles in a material are important factors for detecting tritium and understanding the mechanisms by which tritium can degrade materials. A beta particle interacts with matter by colliding with electrons in the surrounding material. In each collision, the

.Rev. 0 Page 11 Tritium.

CNL-15-216, Enclosure I Attachment 1, Page 22 of 48

PHYSICAL AND CHEMICAL PROPERTIES DOE-HDBK-1079-94 Tritium Primer beta particle may lose several electron volts (keV)b of energy, and the electron is stripped from its atom (ionization) or promoted to an excited state. The beta particle has a finite penetration depth that depends on its energy.

Recall that tritium undergoes beta decay according to the following equation:

H-, +He

+3 + energy.

The helium daughter (*I-e) is stable, but lighter than common helium (*J-e). The decay energy is constant (18.6 keV), but is shared between the beta particle and an antineutrino (a tiny particle).

The result is that not all beta particles have the same energy. The average energy is 5.7 keV.

Consequently, not all tritium betas have the same penetration depth in a given material. Where beta ranges are given, it is customary to list both the highest energy and the average, most representative energy, as listed in Table 3.

Table 3 Penetration depths of tritium betas Penetrati Material on Depth T2 gas, STpa 5.7 0.26 cm T2 gas, STP 18.6 3.2 cm Air, STP 5.7 0.036 cm Air, STP 18.6 0.45 cm Water, soft tissue 5.7 0.42 jtm (and oils/polymers of density = 1)

Water, soft tissue 18.6 18.6 5.2 jim (and oils/polymers of density 1)

Stainless steel 5.7 0.06 jim

a. STP = Standard temperature (0°C) and pressure (760 Torr).

b. An electron volt is a small unit of energy used in descriptions of nuclear and chemical reactions.

it equals the energy gained by an electron when it moves across a potential of 1 volt.

Tritium *Page 12 Rev. 0 CNL-15-216, Enclosure I Attachment 1, Page 23 of 48

Tritium Primer DOE-HDBK-1079-94 PHYSICAL AND CHEMICAL PROPERTIES With one unimportant exception, tritium is the weakest beta emitter known. The range of the most energetic tritium beta particles is only about 5 mm in air or 0.005 mm in Water or soft tissue. This range makes it a nonhazard outside the body, but presents a detection problem.

Where other radioisotopes can be detected by virtue of their penetrating radiation, tritium has to be introduced directly inside the detector or counter to be measured.

Chemical Properties Laboratories that have large quantities of tritium usually handle it in the form of HiT. However, at any time the tritium may be stored on metal getter beds (such as titanium, zirconium, or uranium). These beds form weak chemical compounds with hydrogen. Some of the beds are stable in air; and others are not and can only be used in certain atmospheres. The tritium is released (or delivered) by heating the beds to the required temperature.

Laboratories may also handle tritiated gases (such as ammonia and methane) and other compounds. By far the most common of these is HITO, which is formed from LIT whenever it is exposed to oxygen or water vapor. The conversion reactions are oxidation and exchange:

oxidation 2HT +0O2 -" 2HTO 2T2 + O2 -. 2T 20 exchange HT +H 2 0 -' H2 + HTO T + H2 0 -'HT +HTO These reaction rates are increased by radiation (from nearby tritium at high concentrations),

heat, or the presence of metal catalysts (especially palladium or platinum). All chemical reactions involving hydrogen can also be performed with tritium, sometimes at a higher rate if the tritium concentration is high enough to catalyze the reaction. one of the most important reactions occurs when a tritium atom exchanges with a loosely bonded hydrogen atom of an organic molecule. However, where LIT is dissolved in water (1120), the exchange process is

fairly Slow because the hydrogen in is tightly bonded and the reaction is not catalyzed.

Rev,. 0 Page 13 Tritium CNL-15-216, Enclosure 1 Attachment 1, Page 24 of 48

PHYSICAL AND CHEMICAL PROPERTIES DOE-HDBK-1079-94 Tritium Primer PHYSICAL AND CHEMICAL PROPERTIES DOE-HDBK-1079-94 Tritium Primer Contamination Tritium as HT or UTO will readily adsorb onto the surface of most metals (such as stainless steel, copper, or aluminum), plastics, and rubbers. The tritium will remain fairly close to the surface unless the metal is heated to a high temperature. At room temperature, permeation into these metals is usually extremely slow.

In the case of metal contamination, the tritium remains on or very close to the surface. The contamination can be removed with water or water vapor if the surface is contaminated with HTO or with hydrogen (H2 or D2 ) if the contamination is HT. Heating also speeds up the decontamination process. The initial application of heat to surfaces can also be used to prevent or lessen the contamination by HT or HTO. Metal surfaces exposed to high pressures of HiT or HTO for extended periods, especially at high temperatures, may allow enough penetration to cause structural damage to the metal. This is especially true if the decaying tritium causes a buildup of helium within the structure of the metal.

If adsorbed onto hydrogenous material, the tritium will easily permeate into the material. The HTO will move much more rapidly into the bulk material than will HT. The permeation rate varies with the type of material and is accelerated by increasing the temperature. As a result of this movement, plastics and rubbers exposed to tritium (especially as HTO) are readily contaminated deep into the bulk material and are impossible to decontaminate completely. After a period of time, the tritium exchanges with bulk hydrogen and presents little biological risk.

Highly contaminated metal or plastic surfaces may release some of the loosely-bound tritium immediately after exposure to the contaminating tritiated atmosphere or liquid. This is referred to as outgassing. The personnel risk from outgassing tritium is generally much less than that from making unprotected skin contact with the outgassing surface.

Tritium Page 14 Rev. 0 CNL-15-216, Enclosure I Attachment I, Page 25 of 48

Tritium Primer DOE-H-DBK-1079-94. BIOLOGICAL PROPERTIES BIOLOGICAL PROPERTIES OF TRITIUJM At most tritium facilities, the most commonly encountered forms of tritium are tritium gas (-IT and tritium oxide (HTO). Other forms of tritium may be present, such as metal tritides, tritiated pump oil, and tritiated gases such as methane and ammonia. As noted earlier, deuterated and tritiated compounds generally have the same chemical properties as their protium counterparts, although some minor isotopic differences in reaction rates exist. These various tritiated compounds have a wide range of metabolic properties in humans under similar exposure conditions. For example, inhaled tritium gas is only slightly incorporated into the body during exposure, and the remainder is rapidly removed (by exhalation) following the exposure. On the other hand, tritiated water vapor is readily taken up and retained in the body water. In this Primer, we will address only those compounds likely to be found at DOE laboratories: gaseous tritium, tritiated water, other tritiated species, metallic getters, and other tritiated liquids and gases.

Metabolism of Gaseous Tritium During a brief exposure to tritium gas, the gas is inhaled and a small amount is dissolved in the bloodstream. The dissolved gas circulates in the bloodstream before being exhaled along with the gaseous waste products (carbon dioxide) and normal watervapor. If the exposure persists, the gas will reach other body fluids. A small percentage of the gaseous tritium is converted to the oxide (HTO), most likely by oxidation in the gastrointestinal tract. Early experiments involving human exposure to a concentration of 9 jtCi/mL resulted in an increase in the HTO concentration in urine of 7.7 x 10-3 pxCi/mL per hour of exposure. Although independent of the breathing rate, this conversion can be expressed as the ratio of the [ITO buildup to the tritium inhaled as HT at a nominal breathing rate (20 L/main). In this context, the conversion is 0.003% of the total gaseous tritium inhaled. More recent~experiments with six volunteers resulted in a conversion of 0.005%. For gaseous tritium exposures, there are two doses: (a) a lung dose from the tritium in the air inside the lung and (b) a whole body dose from the tritium gas that has been converted to [HTO. The tr'itiated water converted from the gas in the body behaves as an exposure to tritiated water.

Intake of gaseous tritium through the skin has been found to be negligible compared with that from inhalation. Small amounts of tritium can enter the skin through unprotected contact with contaminated metal surfaces, which results in organically bound tritium in skin and in urine.

Ordinarily this is not a serious problem because surfaces highly contaminated with t--itium gas are inaccessible to skin contact. Also, most tritium exposed to air will be converted to the oxide form (water vapor) before the internal surfaces of equipment are handled during maintenance or repair operations.

Metabolism of Tritiated Water The biological incorporation (uptake) of airborne HTO can be extremely efficient: up to 99%

of inhaled [HTO is taken into the body by the circulating blood. Ingested liquid HTO is also Rev. 0 Page 15 Tritiumn CNL-15-216, Enclosure 1 Attachment 1, Page 26 of 48

BIOLOGICAL PROPERTIES DOE-HDBK-1079-94 Tritium Primer almost completely absorbed by the gastrointestinal tract and quickly appears in the blood stream.

Within minutes, it can be found in varying concentrations in the organs, fluids, and tissues of the body. Skin absorption of airborne HTO is also important, especially during hot weather, because of the normal movement of water through the skin. For skin temperatures between 30 and 40 0 C, the absorption of HTO is about 50% of that for HTO by inhalation (assuming an average breathing rate associated with light work, 20 Llmin). No matter how it is absorbed, the HTO will be uniformly distributed in all biological fluids within one to two hours. This tritium has a retention that is characteristic of water. In addition, a small fraction of the tritium is incorporated into easily exchanged hydrogen sites in organic molecules. Hence, retention of tritiated water can be described as the sum of several terms: one characteristic of body water, and one or more longer-term components that represent tritium incorporated into organic hydrogen sites.

Metabolism of Other Tritiated Species Most tritium handled in laboratories is in the form of tritiated gas or tritium oxide. However, tritium handling operations may form other compounds, such as tritiated hydrocarbons and metal tritides. Tritium may also contaminate Surfaces and liquids such as pump oil. These materials may present special safe handling problems.

Metallic Getters Although many metals are commonly used for gettering (chemically combining with) tritium, little information on their metabolic properties is available. Some of these compounds (such as uranium tritide and lithium tritide) are unstable in air. For these, exposure to air produces different results. Uranium tritide, being pyrophoric, releases large quantities of tritiated water; lithium tritide, a hydroxYl scavenger, releases mostly tritium gas.

Tritides of metals (such as titanium, niobium, and zirconium) are stable in air. For particles of these tritides, the primary organ of concern is the lungs. Some of the tritium may leach out in the lung fluids and then be incorporated into the body water. These particles may also produce organically bound tritium from contact with lung tissue, which would further complicate the metabolic process. However, in laboratories where such tritiated metals are handled, the possibility for exposure to airborne particulates of these metals is extremely remote except in accident situations.

Tritiated Liquids Next to HTO, the most common tritiated liquid is tritiated vacuum pump oil. Experience at DOE facilities has shown that the specific activities of pump oils can easily range from a few mCi/L to a few tens of Ci/L. The wide range in specific activities may result from variations in the tritium concentration and total throughput of tritium. Depending on the history of these pumps, the tritium may be found as HT, HTO, or tritiated hydrocarbons.

Tritium Page 16 Rev. 0 CNL-15-216, Enclosure 1 Attachment 1, Page 27 of 48

Tritium Primer DOE-HD.BK-1079-94 BIOLOGICAL PROPERTIES Next to pump oils, the next most common group is tritiated solvents. All solvents can be absorbed through the skin and are relatively volatile and toxic. The overall toxicity of tritiated solvents is usually dominated by the chemical nature of the solvent.

Other Tritiated Gases If tritium is released in a nitrogen- or air-filled glovebox, other tritiated gases may be. formed, such as ammonia and methane. The conversion of tritium to tritiated ammonia is small unless the tritium concentration is very high. The toxicity of these gases is not believed to be greater than that of tritium Oxide.

Biological Half-Life of HTO Studies of biological elimination rates of body water in humans date back to 1934 when the body water turnover rate was measured using HDO. Since that time, several additional studies have been conducted with HDO and HTTO. A simple average of the data suggests a value of 9.5 days for the measured biological half-life of water in the body with a deviation of +50%.

Calculations based on total fluid intake indicate a similar value. This is reasonable because the turnover rate of HTO should be identical to that of body water. In other words, the biological half-life of tritium is a function of the average daily throughput of water.

The biological half-life of HTO has been studied when outdoor temperatures varied at the time of tritium uptake. The data suggest that biological half-lives are shorter in warmer months. For example, the 7.5-day half-life measured in southern Nigeria is not surprising because the mean outdoor temperature there averages 27°C. In contrast, an average 9.5-day half-life was measured in North America, where the mean outdoor temperature averages 170 C. Such findings are consistent with metabolic pathways involving sensible and insensible perspiration. As such, the skin absorption and perspiration pathways can become an important part of body water exchange routes. It is important to note that personnel who are perspiring will have a greater absorption of tritium from contact with tritiated surfatces. For planning purposes, it is customary to use an average half-life of 10 days. However, it is not used to calculate doses from actual exposures.

Prolonged exposures can be expected to affect the biological half-life. Tritium's interaction with organic hydrogen can result in additional half-life components ranging from 21 to 30 days and 250 to 550 days. The shorter duration indicates that organic molecules in the body retain tritium relatively briefly. The longer duration indicates long-term retention by other compounds in the body that do not readily exchange hydrogen or that metabolize more slowly. However, the overall contribution from organically bound tritium is relatively small, that is, less than about 5% for acute exposures and about 10% for chronic exposures. Methods used to compute the annual limits on intake of air and water specify only the body water component and include the assumption of a 10-day biological half-life, as mentioned above.

Rev. 0 Page 17 Tritium CNL-15-216, Enclosure 1 Attachment 1, Page 28 of 48

BIOLOGICAL PROPERTIES DOE-HDBK-1079-94 Tritium Primer Bioassay and Internal Dosimetry Exposure to tritium oxide (HTO) is by far the most important type of tritium exposure. The HITO enters the body by inhalation or skin absorption. When immersed in tritiated water vapor, the body takes in approximately twice as much tritium through the lungs as through the skin.

Once in the body, it is circulated by the blood stream and finds its way into fluids both inside and outside the cells.

According to International Commission on Radiological Protection (ICRP) Publication 30, the derived air concentration (DAC)C for tritium gas (H-T) and HTO are 540,000 l+/-CI/m and 21.6 j*tCi/ms, respectively. The ratio of these DACs (25,000) is based on a lung exposure from the gas and a whole body exposure from the oxide. However, as was noted earlier, when a person is exposed to HiT in the air, an additional dose actually results: one to the whole body.

During exposure to HT, a small fraction of the tritium exchanges in the lung and is transferred by the blood to the gastrointestinal tract where it is oxidized by enzymes. This process results in a buildup of HTO until the HT is removed by exhalation at the end of the exposure. The resultant dose from exposure to this HTO is roughly comparable to the lung dose from exposure to HT. Thus, the total effective dose from an HT exposure is about 10,000 times less than the total effective dose from an equal exposure to airborne HTO. For both HTO and HT exposures, a bioassay program that samples body water for HTO is essential for personnel monitoring at tritium facilities.

Sampling Schedule and Technique After HTO enters the body, it is quickly distributed throughout the blood system and, within 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , throughout all water in the body. Once equilibrium is established, the tritium concentration is found to be the same in samples of blood, sputum, and urine. For bioassay purposes, urine is normally used for determining tritium concentrations in body water.

Workers who may be or who have been exposed to tritium are normally required to submit urine samples for bioassay periodically. The sampling period may be daily, biweekly, or longer, depending on the potential for significant exposure.

Special urine samples are normally required after an incident or a work assignment with a high potential for exposure. After a possible exposure, the worker should empty the bladder 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months later. A sample taken after the bladder is emptied should be reasonably representative of the body water concentration. A sample collected before equilibrium is established will not be representative because of dilution in the bladder, or because of initial high concentration in the blood. However, any early sample may still be useful as a sign of the potential seriousness of the exposure.

c. The DAC is defined as that concentration of a gas, which, if a worker were exposed to it for one working year (2,000 hours0 days 0 hours 0 weeks 0 months ), would result in an annual dose of 5 remn.

Tritium Page 18 Rev.O0 CNL-15-216, Enclosure 1 Attachment 1, Page 29 of 48

Tritium Primer DOE-HDBK-1079-94 BIOLOGICAL PROPERTIES A pure HT exposure is considered as a combination of a lung exposure from the HT and a whole body exposure from HTO. The HTO comes from the conversion of HiT dissolved in the blood. The whole body dose can be determined as outlined above by analysis for HTO in the urine. Because the effective dose equivalents from the lung and whole body exposures are about equal, the total effective dose can be obtained conservatively by multiplying the HTO whole body dose by 2. However, in general, this is too conservative because a release of pure tritium gas with less than 0.01% HTO is highly unlikely. With only a slight fraction (-0.1%)

of HTO in the air, the total effective dose is essentially the HTO whole body dose determined by bioassay.

As noted above, tritium-labeled molecules in the skin result from contact with metal surfaces contaminated with HT. This form is associated with a longer half-life. Lung exposure to airborne metal tritides may also cause unusual patterns of tritium concentrations in body water because of the slow release of tritium to the blood stream. If such exposures are possible at the facility, it is good practice to follow the elimination data carefully and to look for organically bound tritium in the urine.

The results of the bioassay measurements and their contribution to the worker's dose and general health must be shared with the worker in a timely fashion.

Dose Reduction The committed dose following an HTO exposure is directly proportional to the biological half-life, which in turn is inversely proportional to the turnover rate of body water. This rate varies from individual to individual. Such things as temperature, humidity, work, and drinking habits may cause rate variations. Although the average biological half-life is 10 days, it can be decreased by simply increasing fluid intake, especially diuretic liquids such as coffee, tea, beer, and wine. Even though the half-life may be easily reduced to 4 to 5 days in this way, a physician must be consulted before persons are placed on a regimen that might affect their health. Chemical diuretics require medical supervision because the resultant loss of potassium and other electrolytes can be very serious if they are not replaced. Such drastic measures can result in a decrease in half-life to 1 to 2 days. Even more drastic is the use of peritoneal dialysis or a kidney dialysis machine, which may reduce the half-life to 13 and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , respectively.

Such extreme techniques should be used only in life-threatening situations involving potential committed dose equivalents that would exceed about 100 rem without any treatment. Based on a 10-day half-life, the committed dose for an intake of 1 mCi of HTO is approximately 63 mrem.

Individuals whose urine concentrations exceed established limits should stop work that involves possible exposure to radiation, whether from tritium or other sources. Work restrictions are suggested or imposed to make certain that the annual dose limits for workers are not exceeded.

The operating group may impose stricter limits on their staff than those imposed by the health physics group. Depending on the number of workers available and the importance of the work to be done, doses can be managed to safe levels (from 5 to 100 *tCi/L in urine).

Rev. 0 Page 19 Tritium CNL-15-216, Enclosure 1 Attachment 1, Page 30.of 48

BIOLOGICAL PROPERTIES DOE-HDBK-1079-94 Tritium Primer Results of bioassay sampling should be given to workers who have submitted samples as soon as they are available. The results may be posted, or the workers may be notified personally.

Moreover, the results must be kept in the workers' radiation exposure records or medical files.

Like any other radiation exposure, any dose in excess of the limits specified by applicable regulations must be reported to DOE.

Tritium Page 20 Rev. 0 CNL-15-216, Enclosure 1 Attachment 1, Page 31 of 48

Tritium Primer DOE-HDBK-1079-94 TRITIUM MONITORING TRITIUM MONITORING The tritium monitoring system at a tritium handling facility is critically important to its safe operation. Operators and others at the facility need to be informed of the status of the processes, the development of any leaks in the primary or secondary containments, or of any releases to the room or environment so that protective measures and corrective action may be taken quickly.

The location arid degree of surface contamination are equally important to prevent accidental uptakes of tritium by personnel.

In this section, the various techniques used to monitor for tritium in gases (including air), in liquids, and on surfaces will be discussed.

Air Monitoring Fixed ionization chamber instruments are the most widely used instruments for measuring gaseous forms of tritium in laboratory and process monitoring applications. Portable ionization chamber instruments are also used to control contamination and to supplement fixed instrumaent measurements. Such simple devices require only an electrically polarized ionization chamber, suitable electronics, and a method for moving the gas sample through the chamber--usually a pump. Chamber Volumes typically range from a tenth to a few tens of liters, depending on the required sensitivity. The output is usually given in units of concentration (typically J4Ci/mn 3) or, ifra commercial electrometer or picoammeter is used, in current units that must be converted to those of tritium concentration. The following rule-of-thumb can be used to convert current to concentration: 1015 x current (amps)/chamber volume (liters) = concentration (*tCi/rn 3 ). For real-time tritium monitoring, the practical lower limits of sensitivity range from 0.1 to 10 i*tCi/im 3 .

External background radiation or the presence of radon can lower the sensitivity of the instrument.

For measurements of low concentrations, sensitive electrometers are needed. For higher concentrations (>1 mCi/in 3 for example), the requirements on the electronics can be relaxed, and smaller ion chambers may be used. Smaller chambers also need less applied voltage.

Because of a greater ratio of surface area to volume, residual contamination in the chamber is more likely and is called "memory."' This residual contamination elevates the background chamber current. Response times for higher level measurements can be made correspondingly shorter. However, because small chambers and chambers operated at low pressures may have significant wall effects, the above rule-of-thumb may not apply. Such instruments would have to be calibrated to determine their response.

Although most ionization chambers are the flow-through type that require a pump to provide the flow, a number of facilities Use "open window" or "perforated wall" chambers. These chambers, which employ a dust cover to protect the chamber from particulates, allow the air or gas to penetrate through the wall to the inside chamber without the need for a pump. These instruments are used as single point monitors to monitor rooms, hoods, gloveboxes, and ducts.

~L95-26, Attach nclsur 1 me~t* p*:age 32 of 48 Tritum

J'L°-I 5-216, Enclosure 1

TRITIUM MONITORING DOE-HDBK-1079-94 Tritium Primer Differential Air Monitoring Because HTO is more toxic than HT (10,000 to 25,000 times greater), it may be desirable to know the relative amounts of each species following a significant release into a room or to the environment. In the case of stack monitoring, discrete samples of the stack effluent should be taken using bubblers or desiccants with a catalyst for oxidizing the NT. Another technique for differential monitoring uses a desiccant cartridge in the sampling line of an ionization chamber monitor. The result is a measurement of the HT concentration. Without the cartridge, the total tritium concentration is measured.

Subtraction of NiT from the total produces the HTO concentration. The technique may be used with two instruments or one instrument in which the desiccant cartridge is automatically switched in and out of the sampling line.

Another technique uses a semipermeable membrane tube bundle in the sampling line to remove the HTO (preferentially over the HT), which is directed to an HT0 monitor.

After removing the remaining HTO with another membrane dryer, the sampled air is directed to the I-T monitor. Although this technique is slower than the one requiring a desiccant cartridge, it does not require a periodic cartridge replacement. Furthermore, it can be adapted to measure tritium in both species in the presence of noble gases or other radioactive gases by adding a catalyst after the HTO dryers, followed by additional membrane dryers for the HTO. However, because of its slow response, it is more suitable for effluent or stack monitoring than for room monitoring. Because significant releases into a room are quite rare, it is easier to treat any such release as one of HTO than use complicated techniques for continuous differential monitoring.

Discrete Air Sampling Discrete sampling differs from real-time monitoring in that the sampled gas (usually air) must be analyzed for tritium content (usually by liquid scintillation counting). The usual technique is to flow the sampled air through either a solid desiccant (molecular sieve, silica gel, or Drierite) or water or glycol bubblers. For low-flow rates (about 0.1 to 1 L/min), bubblers may be used. Bubblers are more convenient for sampling, but are less sensitive than the solid desiccant cartridges if the water in the desiccant is recovered by heating. Glycol or water may be used, but glycol is preferred for long-term sampling.

In any case, the collected water is then analyzed for HTO. For differential monitoring of NTO and NT, a heated catalyst (usually a palladiuma sponge) is used between the HTO desiccant cartridge or bubblers and the NT cartridge or bubblers. This is currently the preferred method for monitoring stacks for reporting purposes. In a different arrangement, palladium is coated on the molecular sieve in the NT cartridge to oxidize the NT into HTO, which is then absorbed by the molecular sieve. However, this technique is used primarily for environmental monitoring.

Another technique for sampling NTO in room air is to use a "cold finger" to freeze NTO out of the air. An alcohol and dry ice mixture in a stainless steel beaker works well. To Tritium Page 22 Rev.O0 CNL-15-216, Enclosure 1 Attachment 1, Page 33 of 48

Tritium Primer DOE-HDBK-1079-94 TRJTIUM MONITORING determine the concentration, the relative humidity must be known. Another sampling technique is to squeeze a soft plastic bottle several times to introduce the air (containing the HTO) into the bottle. A measured quantity of water is then introduced, and the bottle is capped and shaken. In a minute or less, essentially all the HTO is taken up by the water, which is then analyzed.

Other techniques involve placing a number of vials or other small specially designed containers of water, cocktail, or other liquid in selected locations in the area being monitored. After a period of time (usually a number of days), the liquid in the containers is analyzed. The result is qualitative (for open containers) to semiquantitative (for specially designed containers).

Process Monitoring Ionization chambers are typically used for monitoring stacks, rooms, hoods, glove boxes, and processes. The outputs can be used to sound alarms, activate ventilation valves, activate detritiation systems, and perform other functions. In general, it can be expected that stack, room, and hood monitors will require little nonelectronic maintenance (i.e.,

chamber replacement because of contamination). Under normal circumstances, the chambers are constantly flushed with clean air and are not exposed to high tritium concentrations. However, glove box monitors can be expected to eventually become contaminated, especially if exposed to high concentrations of HTO. Process HT monitor backgrounds can also be expected to present problems if a wide range of concentrations (4 to 5 orders of magnitude) are to be measured.

Mass spectrometers, gas chronmatograpbs, and calorimeters are the main instruments used for process monitoring. Because of their relative insensitivities, these instruments cannot detect tritium much below a few parts per million (Ci/m 3). For this reason, the analytical results and the related health physics concerns must be interpreted carefully.

It is not uncommon to find that samples showing no trace of tritium when analyzed on a mass spectrometer may actually have a concentration of several curies of tritium per cubic meter. In spite of their contamination problems, ionization chamber instruments are usefuil for measuring these lower concentrations and for providing instant indications of changing concentrations that are not possible with the more sophisticated instruments.

Surface Monitoring Any material exposed to tritium or a tritiated compound has the potential of being contaminated.

Although it is difficult to quantify tritium contamination levels, several methods are available to evaluate the extent of contamination, including smear surveys and off-gassing measurements.

Good housekeeping and work practices are essential in maintaining contamination at acceptable levels.

For health or safety implications, an indication of loose, removable tritium contamination is more valuable than a measurement of the total surface contamination. Loose tritium can be Rev. 0 *Page 23 Tritium CNL-15-216, Enclosure 1 Attachment 1, Page 34 of 48

TRITIUM MONITORING DO0E-HDBK-1079-94 Tritium Primer transferred to the body by skin contact or inhalation if it becomes airborne. As a result, loose contamination is routinely monitored by smears, which are wiped over a surface and then analyzed by liquid SCintillation or proportional counting.

The smears are typically small round filter papers used dry or wet (with water, glycol, or glycerol). Wet smears are more efficient in removing tritium, and the results are more reproducible, although the papers are usually more fragile when wet. However, results are only semiquantitative, and reproducibility within a factor of 2 agreement (for wet or dry smears) is considered satisfactory. Ordinarily, an area of 100 cm 2 of the suaface is wiped with the smear paper and quickly placed in a vial with about 10 mL of liquid scintillation cocktail, or 1 or 2 mL of water with the cocktail added later. The paper must be placed in liquid immediately after wiping because losses from evaporation can be considerable, especially if the paper is dry. The efficiency of the liquid scintillation cocktail is only slightly affected by the size of the swipe.

Foam smears are also available commercially. These smears dissolve in most cocktails and do not interfere significantly with the normal counting efficiency.

Smears may be counted by gas-flow proportional counting. However, because of the inherent counting delays, tritium losses before counting can be significant. Moreover, counting efficiencies may be difficult to determine and may vary greatly from one sample to the next.

Another drawback is potential contamination of the counting chamber when counting very "hot" smears. For all of these reasons, a liquid scintillation spectrometer is the preferred system.

An effective tritium health physics program must specify the frequency of routine smear surveys.

Each tacility should develop a routine surveillance program that may include daily smear surveys in laboratories, process areas, step-off pads, change rooms, and lunchrooms. In many locations within a facility, weekly or monthly routine smear surveys may be sufficient. The frequency should be dictated by operational experience and the potential for contamination. In addition to the routine survey program, special surveys should be made following spills or on potentially contaminated material being transferred to a less controlled area to prevent the spread of contamination from controlled areas.

The surface contamination levels acceptable for the release of materials from radiological areas may be found in the DOE RadiologicalControl Manual and DOE Order 5400.5.

Tritium Probes In general, the total tritium contamination on a surface can be measured only by destructive techniques. When tritium penetrates a surface even slightly, it becomes undetectable because of the weak energy of its beta particles. With open-window probes operated in the Geiger Mueller (GM) or ProPortional regions, it is possible to measure many of the betas emitted from the surface. Quantifying that measurement in terms of the total tritium present is difficult because the history of every exposure is different.

Consequently, the relative amounts of measurable and unmeasurable tritium are different.

Tritium Page 24 Rev. 0 CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I I, Page 35 of 48 Attachment

Tritium Primer DOE-HDBK-1079-94 TRITIUM MONITORING Such monitoring probes are used to survey areas quickly before more careful monitoring by smears, or to monitor the smears themselves whjle in the field.

The probe must be protected carefully from contamination. When monitoring a slightly contaminated surface after monitoring a highly contaminated one, contamination of the probe can be an immediate problem. Placing a disposable mask over the front face of the probe can reduce, but never eliminate this contamination, particularly if'the tritium is rapidly outgassing from the surface. Sensitivity of the instrument depends on many factors, but should be about 10o3 to 104~dpm/cm 2.

For highly contaminated surfaces (>1 mCi/100 cm2), a thin sodium iodide crystal or a thin-window GM tube can be used to measure the characteristic and continuous x-rays (Bremsstrahlung) emitted from the surface as a result of the interaction of the beta particle with the surface material.

Off-Gassing Measurements Off-gassing can be measured using one of two methods. The simplest method is to "sniff*' the surface for airborne tritium using a portable or fixed tritium monitor. The*

most reliable method, however, uses a closed-loop system of known volume and a flow-through ionization chamber monitor. By placing the sample inside the volume and measuring the change in concentration over time, tritium off-gassing rates can be determined accurately on virtually any material. The initial off-gassing rate is the required value because the equilibrium concentration may be reached quickly in a closed volume, especially if the volume is small because of recontamination by the airborne tritium.

The uptake of tritium from off-gassing materials is difficult to predict. Off-gassing tritium that is readily measured indicates contaminated equipment that should not be released for uncontrolled use.

Liquid Monitoring Liquid is almost universally monitored by liquid scintillation counting. Thae liquid must be compatible with the cocktail. Certain chemicals can degrade the cocktail. Others may retain much of the tritium; still others result in a high degree of quenching. In addition, samples that contain peroxide or that are alkaline may result in chemiluminescence that can interfere with measuring. Such samples should first be neutralized before counting. Chemiluminescence and phosphorescence both decay with time. Phosphorescence, activated by sunlight or fluorescent lighting, decays in the dark in a few minutes (fast component) to several days (slow component).

Chemiluminescence, the result of chemical interaction of sample components, may take days to decay at room temperatures, but t~akes only hours to decay at the cold temperatures of a refrigerated liquid scintillation spectrometer. Distillations may be necessary for some samples.

Rev. 0 Page 25 Trituium CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I I, Page 36 of 48 Attachment

TRITIUM MONITORING DOE-HDBK-1079-94 Tritium Primer For rather "hot" samples, as may be the case for vacuum pump oils, Bremsstrahlung counting may be useful. This technique may also be useful for active monitoring of "hot" liquids. Liquids may be monitored actively with scintillation flow cells, which are often made of plastic scintillator material or of glass tubing filled with anthracene crystals. However, both types are prone to memory effects that result from tritium contamination. In addition, flow cells are also prone to contamination by algae or other foreign material that can quickly degrade their counting efficiency.

Tritium Page 26 Rev. 0 CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I I, Page 37 of 48 Attachment

Tritium Primer DOE-HDBK-i079-94 RADIOLOGICAL CONTROL RADIOLOGICAL CONTROL AND PROTECTION PRACTICES Airborne Tritium Tritium released to room air moves readily with normal air current. The room or building ventilation system should be designed to prevent the air from being carried to uncontaminated areas, such as offices or other laboratories where tritium is not allowed. For that reason, differential pressure zoning is commonly used, and released tritium is directed outside through the building stack. In some newer facilities where the large quantities of tritium are being handled, room air cleanup systems are availaible for emergency use. Following a significant release, the room ventilation system is effectively shut down, the room is isolated, and cleanup of room air is begun.

Secondary Containment The most import~ant control for preventing a release of tritium to the room atmosphere is the use of containment around the source of tritium. This containment usually takes the form of a glove box, which is then a secondary containment if the tritium is already contained within the process plumbing, which is the primary containment. Even if the tritium is on the outside surfatce of a piece of equipment and located inside the glove box, through popular usage, the box is still referred to as the secondary containment.

Glove boxes used for tritium work typically are made of stainless steel or aluminum and use gloves made of butyl, neoprene, or Hypalon. Windows are made of glass or Lexan.

In order to reduce the amount of tritium released to the atmosphere, glove boxes where significant quantities of tritium are handled incorporate detritiation systems that process the glove box atmosphere and remove the tritium. These detritiation systems, including the room cleanup systems mentioned above, convert released HT to HTO and collect the HTO on a molecular sieve for later recovery or burial. Newer systems use metal getters that recover HT without resorting to oxidation. These getters, which can only be used in certain glove box atmospheres, can be heated to release and recover the HT easily.

The atmosphere in the glove box may be air, nitrogen, argon, or helium, depending on the type of activity in the box. Even in boxes with inert gas atmospheres, small amounts of moisture and oxygen exist. Any release of tritium gas in the box will eventually be converted to the oxide. As a result, the oxide will slowly diffuse through the gloves and contaminate their outside surfaces. For that reason, personnel using glove boxes that have had tritium releases are required to wear one or more additional pairs of disposable gloves when working in the glove box.

Glove box monitors are used to alert personnel of a release in the box and may be used to activate a cleanup system or to increase the rate of the cleanup process. With releases of tritium in the box, the monitor chamber will eventually develop a memory from 21521, Eclsur I ]*,~

~'N Attachment 38 of 48 Tritium

L0-15-216, Enclosure 1

RADIOLOGICAL CONTROL DOE-HDBK-1079-94 Tritium Primer contamination, mainly by HTO., Heated monitor chambers are useful in minimizing contamination by HTO.

The relative pressure of the glove box atmosphere is normally kept negative in order to prevent the gloves from hanging outside the box where passersby may brush against them and to prevent tritium from escaping into the room should a leak develop in the glove box. However, outward permeation of HTO through the gloves and inward permeation of room moisture are not affected by the pressure inside the glove box.

Temporary Enclosures At times, maintenance or repair work is done on equipment that cannot be moved into a glove box or fume hood and that has a high potential to release tritium. For these activities a temporary box ("tent"), may be constructed over the equipment, and an existing cleanup system installed to process the air. Alternatively, if the tritium at risk is not significant, the enclosing atmosphere may be purged to the stack. If the enclosure is small, gloves and glove ports may be fitted to the side of the enclosure. For larger enclosures entry may be required. In such cases,, personnel must work in air-supplied suits inside the enclosure.

Protection by Local Ventilation In spite of the greater protection afforded by glove boxes, fume hoods are commonly used at tritium facilities for handling or storing material with low quantities of tritium or with low-level contamination. Limits are generally imposed on the quantities used or stored in these hoods.

Fume hoods are also used to protect personnel at the outside door of glove-box pass boxes where materials are passed into and out of the boxes. Ideally, any tritium released in a hood from outgassing or a leaky container, for instance, is routed to the hood's exhaust duct. However, turbulence may occur at the hood entrance, resulting in backwash and possible contamination of personnel if the face velocity is not adequate for the design of the hood, the activities in the hood, or the local conditions (such as traffic in front of the hood). No hood should be used that has not been thoroughly surveyed and judged acceptable for tritium use.

For small operations local ventilation is commonly provided at the work site through a flexible ventilation duct ("elephant trunk") directed to the room exhaust system. The exhaust of these ducts is generally directed to the building ventilation exhaust system, which of itself may be adequate to supply the needed air flow for the duct without help from an additional in-line blower.

Attachment *a~ 39 of 48 R.ev. 0

uL'a15-216, Enclosure 1

Tritium Primer DOE-HDBK-1079-94 RADIOLOGICAL CONTROL Flexible ducts can provide adequate ventilation during maintenance in a glovebox with a panel removed. In this application, a flexible duct can be connected to a gloveport before the panel is removed, and then the work can proceed safely.

Supplied-Air Respirators In general, only supplied-air respirators are effective in preventing inhalation of airborne tritium. Two types of air-supplied respirators are available: self-contained breathing apparatus (SCBA) and full-face supplied air masks.

An SCBA, consisting of a full-face mask fed by a bottle of compressed air carried on the worker's back, provides excellent protection against HTO inhalation. Because the mask provides no protection against absorption by most of the skin, the SCBA is normally reserved for emergency use only. The protection factor of 3 or more afforded by the SCBA may be adequate for some applications. An SCBA can be used as an added precaution during certain maintenance or operations that experience has shown should not result in the release of significant amounts of HTO. Nevertheless, the potential for exposure is real, and the SCBA gives the worker time to leave the area if necessary before a skin exposure occurs.

Full-face supplied-air masks are also available. Because the air is normally supplied by a fixed-breathing-air system, they are not practical for many emergency situations and, consequently, are not as popular as SCBAs.

Supplied-Air Suits Because of the inherent disadvantages associated with respirators and other breathing apparatus, supplied-air plastic suits that completely enclose the body are often used by facilities that handle large quantities of tritium. Although they afford reasonably complete body protection, they are slow to don and cumbersome to wear. For these reasons, they are not favored for rescue work where time and mobility are important considerations. For certain maintenance operations outside of glove boxes with a high degree of risk, supplied-air suits may be quite useful.

For tritium work, supplied-air suits are constructed of materials that have acceptable permeation protection against HTO and provide good tear and abrasion resistance.

Because of the closed environment, and the additional background noise caused by the flow of air into the suits, communication between personnel may require special equipment or methods.

Protection from Surface Contamination Experience at tritium laboratories has shown that many tritium exposures to personnel occur as a result of contact with highly contaminated Surfaces. Sudden and significant releases of airborne tritium occur mostly as the less toxic form HT and are quickly detected by portable or Tritium

[L°-15-216, Enclosure 1 Attachment *,a*e 40 of 48

RADIOLOGICAL CONTROL DOE-HDBK-1079-94 Tritium Primer strategically placed, fixed tritium monitors. The result is: that the exposure and uptake of airborne tritium are minimiz ed. (Heavy-water reactors, of course, present a more significant risk of exposure to tritiated water vapor than to tritium gas.) The presence and degree of contamination may be unknown until measurements are made. Consequently, the importance of routine and special monitoring surveys for surfaces that personnel might contact cannot be overestimated.

Protective clothing worn by workers is one of the most important aspects of an effective health physics program. Because tritium cant be absorbed easily through the skin or by inhalation, personnel protective equipment must protect against both exposure routes. The following paragraphs describe protective measures and equipment.

Protective Clothing Lab Coats and Coveralls Lab coats and coveralls (fabric barriers) are worn in most tritium facilities. Lab coats are routinely worn to protect personal clothing. Coveralls are sometimes worn for added protection instead of a lab coat when the work is unusually dusty, dirty, or greasy. The

protection afforded by lab coats and coveralls is minimal (except for short exposures) when tritium is airborne, but they are more effective in preventing skin contact with contaminated surfaces.

Disposable water-proof and water-resistant lab coats and coveralls have been tested at various laboratories. They are not popular for everyday use because of the cost and excessive discomfort inflicted on the worker. Most facilities prefer using ordinary open-weave fabrics for lab coats and coveralls and using an approved laundry for contaminated clothing. Some facilities have chosen to use disposable paper lab coats and coveralls, exchanging the costs associated with a laundry for the costs associated with replacement and waste disposal.

Shoe Covers Although shoe covers provide protection against the spread of contamination and exposure, the routine use of shoe covers in a tritium facility is usually weighed against actual need.* Shoe covers can offer both a degree of personnel protection and control over the spread of contamination on floors. However, in modern facilities where tritium is largely controlled by the use of secondary containment, shoe covers may not be required. Such facilities can easily maintain a clean laboratory environment by the use of regular smear surveys and good housekeeping. Using liquid-proof shoe covers until spills are cleaned up should be considered following spills of tritium-contamninated liquids and solids to prevent the spread of local contamination.

Tritium Page 30 Rev. 0 CNL-15-216, Enclosure1I Attachment 1, Page 41 of 48

Tritium Primer DOE-HDBK-1079-94 "RADIOLOGICAL CONTROL Gloves*

In most operations, the hands and forearms of workers are vulnerable to contact with tritium surface contamination. The proper use and selection of gloves are essential.

Many fatctors should be considered in selecting the proper type of glove. These include chemical compatibility, permeation resistance, abrasion resistance, solvent resistance, glove thickness, glove toughness, glove color, shelf life, and unit cost. Gloves are commercially available in butyl rubber, neoprene, polyvinyl chloride (PVC) plastics, latex, etc.

The most common gloves found in tritium laboratories are the light-weight, disposable short glove (usually PVC or latex) used for handling lightly contaminated equipment.

Depending on the level of contamination, such gloves may be changed frequently (every 10-20 minutes), a second pair may be worn, or heavier gloves may be used instead.

When using gloves for this purpose, the work should be planned so that contaminated gloves doe not spread contamination to surfaces that are being kept free of contamination.

When working in a glove box using the box gloves, disposable gloves are worn to prevent uptake of HTO contaminating the outside of the box gloves. Again, depending on the level of contamination, more than one additional pair may be required, one of which may be a longer, surgeon's length, glove.

In spite of all the precautions normally taken, workers may occasionally be contaminated with tritium. The skin should be decontaminated as soon as possible after any potential skin exposure to minimize absorption into the body. Effective personal decontamination methods include rinsing the affected part of the body with cool water and soap. If the entire body is affected, the worker should shower with' soap and water that is as cool as can be tolerated. Cool water keeps the pores of the skin closed and reduces the transfer of HTO across the skin. The importance of washing the affected skin as soon as possible alter contamination cannot be over-emphasized. Figure 8d illustrates the effect of speed on reducing the uptake and the resultant dose. Even if gloveS are worn when handling contaminated equipment or when working in contaminated glove box gloves, it is good practice to wash the hands after removing the gloves.

d. W.R. Bush, Assessing and Controllingthe Hazardfrom Tritiated Water, AECL-4 150, Atomic Energy of Canada LTD., Chalk River, Ontario, 1972.

Rev. 0 Page 31 Tritium CNL-15-216, Enclosure I Attachment I, Page 42 of 48

RADIOLOGICAL CONTROL DOE-HDBK-1079-94 Tritium Primer Figure 8 Reducing HTO uptake by washing after exposure to HTO vapor.

Tritium Page 32 Recv. 0 CNL-15-216, Enclosure 1 Attachment 1, Page 43 of 48

Tritium Primer DOE-HDBK-1079-94 EAERGENTCYRESPONSE EMERGENCY RESPONSE It is important to examine the history of accidents that have occurred in tritium facilities and to consider foreseeable unplanned events in order to minimize or mitigate their effects or to prevent their taking place at all. When an accident does occur, requirements for reporting accidents must be followed.

Facilities that handle significant quantities of radioactive material must have a site-specific emergency plan. All radiological workers at the site must be familiar with certain aspects of this plan. In addition, job assignments involving radiological hazards are typically covered by procedures and work permits that include steps for emergency situations that may arise during the course of the work. Radiological workers must be familiar with these procedures or be accompanied by a radiological control technician (RCT) to provide guidance in case of an emergency.

Emergency Steps to Take The initial steps to be undertaken following a serious accident must always include the following:

Warning others in the vicinity

Evacuating the laboratory if an airborne release has occurred

Requesting any necessary assistance

Giving urgent first aid in the event of serious injuries (This should take priority over problems that arise from contamination)

Starting personnel decontamination procedures

Submitting urine samples following the schedule outlined for nonroutine samples.

Decontamination of Personnel Personnel should be decontaminated by the following procedure:

Remove clothing thought to be contaminated

Wash hands with soap and cool water

Wash other parts of the body (such as face, hair, and arms) that may have been exposed to tritium, or immediately shower with cool water and soap Rev. 0 Page 33 "Tritium CNL-15-216, Enclosure 1 Attachment 1, Page 44 of 48

EMERGENCY RESPONSE DOE-HDBK-1079-94 Tritium Primer If mouth-to-mouth resuscitation must be given to a contaminated victim, the Victim's mouth should first be wiped with a damp cloth.

Decontamination of Surfaces Following a tritium spill involving a liquid with high specific activity, the area may have to be isolated and other protective measures taken before cleaning up the liquid. Monitoring for possible airborne tritium must be started to determine the need for respiratory protection or skin protection. After the spill has been cleaned up, residual contamination will remain. Depending on the level of contamination, any further steps needed to prevent the spread of contamination and reduce the level to an acceptable value should be determined.

Following a release of tritium gas, surfaces would not be expected to be heavily contaminated.

Iftritiated water. vapor is released, the contamination may be greater, depending on the amount and activity of the released vapor. In any case, smear and air surveys will be used to determine the course of action needed to control and reduce the contamination safely.

Tritium Rev. 0 CNL-15-216, Enclosure 1. Attachment Page 34 1, Page 45 of 48

Tritium Primer DOE-HDBK-1079-94 BIBLIOGRAPHY

~BIBLIOGRAPHY EG&G Mound Applied Technologies, Health Physics Manual of Good Practicesfor Tritium Facilities,MLM-37 19, Draft, Miamisburg, Ohio, December 1991.

International Atomic Emergency Agency, Safe Handlingof Tritium, IAEA-324, Vienna, Austria, 1991.

U.S. Department of Energy, RadiologicalControlManual, DOE/EHI-0256T, June 1992.

"Occupational Radiation Protection," 10 CFR 835, FederalRegister, 58, No. 238, December 1993.

T.B. Rhinehammer and P.H. Lamberger (eds.), Tritium Control Technology, WASH-1269, Monsanto Research Corporation, Miamisburg Ohio, 1973 U.S. Department of Energy, Radiation Protection for Occupational Workers, DOE Order 5480.11, Change 3, June 1992.

National Council on Radiation Protection and Measurements, Tritium Measurement Techniques, NCRP 47, 1976.

U.S. Department Of Energy, Occurrence Reporting and Processing of Operations Information, DOE Order 5000.3B, January 19, 1993.

U.S. Department of Energy, Radiation Protectionof the Public and the Environment, DOE Order 5400.5, Change 2, January 7, 1993.

Rev. 0 Page 35 Tritium CNL-15-216, Enclosure I Attachment 1, Page 46 of 48

CNL-15-216, Enclosure 1 Attachment 1, Page 47 of 48 CONCLUDING MA TERLIAL DOE-HDBK-1079-94 Tritium Primer CONCLUDING MATERIAL Review activities: Preparing activity:

DOE Facilities ANL-W, BNL, EG&G Idaho, DOE EH-63 EG&G Mound, EG&G Rocky Flats, Proj ect Number 6910-0036 LLNL, LANL, MMES, ORAU, REECo, WHC, WINCO, and WSRC.

DOE Program Offices AD, DP, EU, EM, ER, NP, NS, RW.

DOE Field Offices AL, CH, ID, NV, OR, RL, SR, OAK, RF.

Tritium Rev. 0 CNL-15-216, Enclosure 1 Attachment 1, Page 48 of 48

ENCLOSURE1I TENNESSEE VALLEY AUTHORITY WATFS BAR NUCLEAR PLANT UNIT I TVA Response to NRC Request for Additional Information Attachment 2 WBNNAL3003, Revision 5, "Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18,1 -1984" CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 1 of 49 Attachment

NPG CALCULATION coVERSHEETICTS UPDATE

Page 1 REV 0 EDMS/RIMS NO. CTS TYPE: EDMS TYPE: .EDMS ACCESSION NO (N/A for REV. O)

B45 860107 235 Calculation CALCULATIONS (NUCLEAR) 9,. I* 0 06 Cab¢

Title:

Reactor Coolant and Secondary Side Activities in Accordance With ANSIIANS-1 8.1-1 984 ORG PLANT BRANCH NUMBER CUR REV "NEW REV cALO ID NUC WBN NT8 WBNNAL3003 004 005 CTS UPDATE ONLYQ []No CTS Changes 0]

(Verifier and Approval Signatures Not Required) *. (For caic revision, CTS has been reviewed and no CTS changes required)

UNITS (check one) SYSTEMS UNIDS 0015*[,20--,30 N/A. N/A ___,__

DCNEDCNIA APPLICABLE DESIGN DOCUMENT(S) "CLASSIFICATION DCN 61599 N/A .E .

QUALITY. SAFETy RELATED? UNVERIFIED SPECIAL REQ*UIREMENTS DESIGN OUTPUT SAR/TS and/or ISFSI RELATED? .(if yes, QR = yes) ASSUMPTION AND/OR LIMITING CONDITIONS? ATTACHMENT? SAR/CoC AFFECTED*

Yes[] Nofl Yes[] Nofl Yes[] No[] " Yeas[ NoJ[] Yeas: No]* Yes[] No[]

.CALCULATION NUMBER, REQUESTER .PREPARING DISCIPLINE V/ERIFICATION METHOD.j NEW METHOD oF ANALYSIS Name:. PHONE: N .Design Review .] OYes O ]No PRPRRPIT AEA AE CHECKER (PRINT NAME SIG *) "DATE VREPRIFER (PRINT ANDAM I DATE AP R ENDSG)J) ,AT Barry C. Schwartz _"___________,___________l'tt*.*[1 '

STATEMENT OF PROBLEM/ABSTRACT

The methodology of ANSI/ANS-iB. 1-1984 was followed to calculate reactor coolant activities except for TPC tritium values which are discussed in the calculation. The calculations begin with the use of the base activities in Table 6 of ANSIIANS-l 8.1-1984. Each base nuclide activity (in microcuries per gram) is multiplied by an adjustment factor. The referenced standard provides the method for using plant-specific parameters to determine the adjustment factors..

The expected reactor coolant activities that are calculated are intended for possible use in environmental reports, in normal dose calculations for equipment qualification purposes, and for other applications where the use of expected average data over the life of the plant would be appropriate.

Revision .5is performed in support of increasing the TPC Iritium permeation rate to reflect a realistic and design basis source term.

The realistic source term is based on a permeation rate of 5 Ci/TPBAR/yr and is calculated for 704 TPBARs and 1900 .TPBARs.

The design basis source term is based on a permeation rate of 10 Ci/TPBAR/yr and is calculated for 2500 TPBARs.

MICROFICHE/EFICHE Yes 0] No 0] FICHE NUMBER(S)

TVA 40532 Page I of 2 NEDP-2-i E10-31-2011]1 CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 2, Page 2 of 49

NPG CALCULATION COVERSHEETICTS UPDATE Page 2 CALC D ORG PL~AN BRANCH NUMBER REV NUC WBN NTB WBNNAL3003 10051*

BUILDING .1 ROOM ELEVATION COORD/AZIM FIRM.-

N/A ] N/A I N/A N/A S&L CATEGORIES KEY NOUNS (A-add, D-delete)

ACTION KEYWORD A/D KEYWORD

_______ CROSS-REFERENCES (A-add, C-change, D-delete)

ACTION XREF XREF XREF XREF XREF A/ CODE PLANT TYPE NUMBER REV A P WBN DCN 61599 CTS ONLY UPDATES:

Following a~re required only when making keyword/cress reference OTS updates and pagle 1 of form NEDP-2-1 is not included:

PREPARER (PRINT NAME AND SIGN) DATE . CHECKER (PRINT NAME AND SIGN) DATE PREPARER PHONE NO. EDMS ACCESSION NO.

1 VJ*L4.U,*. rage z or z NI:UI-'-Z- 1 [1U-;*l-ZU11 J CNL-15-216, Enclosure 1 CNL-5-21, Attachment Enlosue I 2, Page 3 of 49

' Pa e 3 NPG CALCULATION RECORD OF REVISION.

CALCULATION IDENTIFIER WBNNAL3003 Title Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18.1-1 984 Revision DESCRIPTION OF REVISION No.

0 Initial Issue 1 Revision 1 was performed to add secondary side activities (water and steam) to the analysis. R0 reac'tor coolant activities were not changed. All pages were rewritten for legibility and to bring the calculation into conformance with NEP 3.1 R2

_____Pa es Chan..ed : all"

2 Revision 2 was performed to incorporate EDC E50629A which addresses the use of a Tritium Production Core. Only the amount of tritiumn produced in the core is changed. Since operations will not change all other isotopes remain the same. Updated file to new format. Revision bars will indicate changes from last revision: Classification forms were deleted as they are no longer needed.

Pages changed :1-18 Pages added :1, Computer file storage (2), Computer microfiche sheet (3)

Pages deleted : classification forms (2&3)

R2: total 43 pages 3 Revision 3 is in addition to R2., it includes the possibility of I or 2 rod failure in a TPC. The only change is to the tritium concentration in the event of these failures. It also changes the expected concentration of H3 with a TPC from a maximum value to an average (9 to 3.7ttCi/g). All margins in the body of the calculation were changed, so all pages are new, but actual text, changes are marked by revision bars.

Pages changed:lIc,2,3,7-8,1 1, 16-18 Pages added: I cover sheet Pages deleted : none

_______R3 :.total 44 pages 4 Revisionn4 is in support of DCN D51754, Steam Generator Replacement. With the new steam generators, the steam and water masses changed as well as the steam flow rate. These changes are incorporated in Appendix A and the results are shown to be less than that determined in P3. Therefore the user can choose which set of concentrations to use. The values in Appendix A will not be valid until DCN 5 1754.is in RTO status and the plant is out of the refueling outage it was installed. The current results are conservative for dose calculations, but are non-conservative for leak detection (WBNAPS3-052, -053) and minimum concentration calculations (WBNAPS3048). The FSA1R and Technical Specifications impacts, if any, are addressed in the screening review for DCN 51754.

Pages Replaced or Revised: 1,3, 4, 7-9, 16, 17 Pages Added: CCRIS Update Sheet (2), Appendix A (4.pages)

Pages Deleted: Design Verification Form, Computer Output Form

_______Total R4 pages -44 TVA 40709 [12-2000] Page 1 of1 NEDP-2-2 112-04-2000]

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue Attachment I 2, Page 4 of 49

Page 4 Revision DESCRIPTION OF REVISION No. .

5 Revision 5 is performed in support of increasing the TPC tritium permeation rate to reflect a realistic and design basis source term. The realistic source term is based on a permeation rate of 5 Ci/TPBAR'yr and is calculated for 704 TPBARs and 1900 TPBARs. The design basis source term is based on a permeation rate of 10 Ci/TPBAR/yr and is'calculated for 2500 TPBARs. This change impacts the TPC average tritium concentrations as well as the 1 and 2 TPBAR failure concentrations.

CTS .was reviewed for successor documents, and multiple successor documents were identified. The following documents are impacted by this revision: WBN Calculations T1534, T1535, WBNAPS3O44, WBNAPS307"7, WBNAPS3 118, WBNTSROO8, WBNTSR064, WBNTSR068, WBNTSR080, WBNTSRO84, WBNTSR088, WBNTSR093, and WBNTSR100. There are a significant number of additional successor calculations. However, those calculations do not use the TPC values for tritiumi, therefore they are not impacted by this revision.

See DCN 61599 for SAR/Tech Spec impact determination.

Pages Replaced or Revised: 1-5, 7-9, 13, 18, 19, 21, 22,'24 Pages Added: N'PG Calculation Verification Form (page 6).

Pages Deleted: none*

Total Revision 5 pages: 48.

The revision of all the pages has been changed to revision 5 and changes to the calculation are indicated by change bars.

TVA 40709 [12-2000] Page 1 of 1 NEDP-2-2 [12-04-2000]

Attachment 2, Page 5 of 49 CNL-15-216, Enclosure 1

Coversheet1 CTS update Sheet . 2 Revision Log .3 Table of Contents .*5 Calculation Verification 6 Computer Input File Storage Information Sheet 7 Purpose 8 Introduction 8 Assumptions. 8 Special RequlrementslLlmltlng Conditions "8 Calculations 9.*

Results *18 References 21 Appendix A - Steam Generator Replacement Results 22 Attachment A - Phone conversation with C.D. Thomas (1 page) 26 Attachment B - ANSI/ANS-18.1-1984 (22 pages) 27 TVA 40710 [12-2000] Page l ot I NEDP-2-3 f12-04-2000J CNL-1 5-21 6, Enclosure I Attachment 2, Page 6 of 49

" Page 6 NPG CALCULATION VERIFICATION FORM Calculation Identifier WBNNAL3003 Revision 005 Method of verification used: .

1, Design Review [

2. Alternate Calculation LI Verifier B..Sharz Dt 6t//r24 12t'Y

3. Qualification Test LI ..

Cornments:.

Ihave reviewed Calculation WBNNAL3003 Revision 5 and have found the portions of the calculation revised to be technically adequate. In conducting the verification of the portions of WBNNAL3003 revised as part of Revision 5, I have reviewed the inputs, computations, and results, which I have found to be complete and accurate. All comments have been resolved with the preparer.

TVA 40533 [10-2008] Page 1 of 1 NEDP-2-4 [10-20-2008]

CNL-15-216, Enclosure 1 CNL- I 2, Page 7 of 49 5-1 6Encosue Attachment

Page 7 NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document WBNL03Rv 05 Pat B

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18.1-1 984

-l Electronic storage Of the input files for this calculation is not required. Comments:

[]Sprovided Input filesbelow for thisforcalculation each. inputhave file. been (Any stored retrievedelectronically file requiresand sufficient identifying information is re-verification of its contents before use.)

The Microsoft Word file for R5 is permanently stored in FILEKEEPER # 321761.

(file Calcu/ation WBNNAL3003, Rev. 5.doc)-

El MicroficheleFiche TVA 40535 [12-2000] Page 1 of 1 NEDP-2-6 [12-04-2000]

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 8 of 49 Attachment

Calculation. No. WBNNAL3003 Rev: 005 Plant: WBN Page: 8 I

Subject:

Reactor Coolant and Secondary Side Activities' in Accordance with ANSIIANS-18.1'-1984I Purpose

The-purpose of this calculation is to calculate the expected primary coolant activities in accordance with ANSI/ANS-18.1-1984 methodology, except for the Tritium Production Core (TPC) tritium values which are discussed later. Revision 1 of this calculation added the expected secondary side (water and steam) activities to the analysis.

Introduction ANSIIANS-18.1-1984 (ref.2) (Attachment B) provides numerical values of coolant and steam activity based on available data from operating plants. For pressurized water reactors with U-tube steam generators, such as WBN, the numerical

values' are given in Table 6 of reference 2. These activities may be used without correction only if all plant parameters in Table 2 of reference 2 that affect primary coolant and steam activity have the same parameter nominal values presented in Table 2 of reference 2. Since some applicable parameter values for WBN are not identical to the nominal values in Table 2 of reference 2, adjustment factors need to be calculated. The adjustment factors are calculated and applied to the base activities in Table 6 of

reference 2 to obtain the normal coolant and secondary side (water and steam) activities for WBN.

Revision 5 is performed in support of increasing the TPC tritium permeation rate to reflect a realistic and design basis source term. The realistic source term isbased on a permeation rate of 5 Ci/TPBAR/yr and is calculated for 704 TPBARs and 1900 TPBARs. The design basis source term is based on a permeation rate ofl10 CiJTPBAR/yr and is calculated for.2500 TPBARs.

These TPC tritium permeation rates are based on assumption 6.

Assumptions

1. Several isotopes listed in ANS-18.l ANSI N237-1976 are not present in ANSI/ANS-l8.1-1984. The isotopes deleted were Kr-83m, Br-83, Br-85, Rb-S6, Te-125m, Te-127m, Te-127, and 1-130. These isotopes will not be inthe final list of this calculation.

Technical Justification: These isotopes were dropped from the list because they were deemed to be not as important as the listed isotopes. This was justified by one of the authors of ANSIIANS-18.1-1984 (see Attachment A for detailed justification).

2. Other isotopes were left off the ANSI/ANS-18.1-1984 listing because they are in secular equilibrium with the parent isotope.

These isotopes are Y-90, Rh-103m, Rh-106, Ba-137m, and Pr-l44. This report will include these isotopes in the final listing, however they will be marked as being in secular equilibrium with its parent..

Technical Justification: As in Assumption #1, the authors'ofANSI/ANS-18.1-1984 deemed these isotopes to be insignificant (see Attachment A). For conservatism and completeness, these isotopes will be included.

3. Pr-143 was left off the ANSI/ANS-l8.1-1984 list because it is a pure beta emitter (see Attachment A). This will be included in the f~mal listing of this report, with the appropriate notation. It is assumed that Pr-143 is in secular equilibrium with its parent Ce-143.

Technical Justification: This isotope is included for completeness. The assumption that it is in secular equilibrium (the activity is the same as the parent) is not true because Pr-143 has a longer half life (13.58 days, ref.5) than Ce-143 (33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months , ref.5).

However, this assumption is deemed conservative because it is in addition to those isotopes given in the ANSIIANS-l18.1-1984 standard.

4. A peak concentration of 2.5 ptCilg of tritium is assumed before dilution~occurs in a normal, non-TPC cycle.

Technical justification: Operational plant data for WBN and SQN show a peak of 2.5 pCi/g before dilution occurs in a normal cycle, i.e. no scrams or interruptions (ref.13).

5. A TPC will not change the normal operations of the plant, e.g no extra dilution, no recycle of primary coolant, and/or other changes that would impact coolant tritium concentrations, and thus the concentration of all other isotopes except tritium will remain the same.

Technical Justification: Tritium has a long half life and does not produce any daughters, therefore all other isotopes are not impacted by the increase in tritiumn.

6. A realistic and design basis TPC tritium permeation rate of 5 CifTPBAR/yr and 10 CiITPBARIyr, respectively, is assumed.

Technical justification: The realistic permeation rate of 5 Ci/TPBAR/yr is acceptable because it bounds the observed permeation rate. The design basis permeation rate of 10 CiiTPBALR/yr provides an additional factor of 2 margin and is therefore reasonable, but conservative and bounding.

Special Requirements/Limiting Conditions none" CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2,; Page 9 of 49 Attachment

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSIIANS-18.1-1984 Calculations Plant specific parameters used to obtain adjustment factors are Shown in the following table. The adjustment factors are calculated for all ref.2 Table 6 primary coolant and secondary side activities in classes 1, 2, 3, and 6. Note that the corrosion product activities are included in the Table 6 class 6 list. Calculation of the adjustment factors follows the procedures given in ref.2.

Operation with a TPC has a potential to incre'ase the amounts of tritium in the reactor coolant. To address the TPC impacts, the methodology in ref.15 is used in establishing a realistic and design basis average tritium concentration in the reactor coolant.

The realistic average tritium concentration is 5.1 ptCi/g for 704 TPBA.Rs and 12.0 *tiCi/g for 1900 rPBARs. This is roughly based on WBN having an average tritium concentration during normal operation of 1.0 j*tCi/g and projecting an approximate 5.1 and 12.0 times increase above the tritium production in a non-TPC core. This is based on assumption 5 that says the plant will operate the same as for a non-TPC, i.e. no extra dilution (feed and bleed) or other changes. The design basis average tritium concentration is

29.8 j*tCi/g for 2500 TPBARs and is also based on WEN having an average tritiumn concentration during normal operation of 1.0

tCi/g and projecting an approximate 29.8 times increase above the tritium production in a non-TPC core.

PARAMETERS USED TO DESCRIBE THE REACTOR SYSTEM. REALISTIC BASIS (values in parentheses are those given far the Replacement Steam Generators)

Nominal WBN Symbol Units ANS-18.l-1984 Value Reference 1984 Value (note)

P MWt 3400 3582 3,4 (a)

Thermal power Steam flow rate FS lb/hr 1. 5E+07 1. 5E+07 3,6 (1. 54E+07) 16, (j)

Weight of water in WP lb 5. 5E+05 5. 4E+05* 3,10 (b) reactor coolant system (5.78E+05) .16, (j)

Weight of water in WS lb 4.50OE+05 (c) 3.48E+05 3,9 all steam generators (6.79E+05) 16, (j)

Reactor coolant letdown FD lb/hr 3 .7E+04 3.7E+04 *3,8 (d]

flow rate (purification)

Reactor coolant letdown FB lb/hr 500 845 3 flow rate (yearly average f or boron control)

Steam Generator FBD lb/hr 7.50OE+04 3.00E+04 3,9 Ce)

Blowdown flow (total)

Fraction of NBD 1.0 1.0 (f) radioactivity in blowdown stream which is not returned to the secondary coolant system Flow through the FA lb/hr 3.7E+03 3.7E+03 3,8 (g) purification system cation demineral izer CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 10 of 49 Attachment

Calculation Reactor

Subject:

No. WBNNAL3003 Coolant and Secondary Side Activities Rev: 005 [Plant:

in Accordance withWBN [Page: 10 ANSiIANS-18.1-1984 t PARAMETERS USED TO DESCRIBE THE REACTOR SYSTEM REALISTIC BASIS .Cont'd Nominal WBN Symbol Units ANS-18.1-1984 Value Reference 1984 Value (note)

Ratio of condensate NC -0.0 0.55 3,9 (h) demineralizer flow rate to the total steam flow rate Fraction of the noble Y - 0.0 .0.0 3(i) gas activity in the letdown stream which is not returned to the reactor, coolant system (not including the boron recovery system).

Notes:

a) The thermal power is 3411 MWt (ref. 4) . The value used in this calculation is 105% of 3411 = 3582 MWt.

b) Reference 10 gives the reactor coolant volume as 11375 cuft. Reference 4 gives the reactor inlet temperature as 558.1 degF and the outlet temperature as 618.2 degF. From ref. 11, the specific volume vr for compressed liquid at 500 degF is 0.0204 cult/lb, for 600 degF vf is 0.0236 cuft~lb, and for 620 degF v-f is 0.0247. Using linear interpolation, vffor 558.1 degF is then 0.0223 cuft/Ib and vf for 618.2 degF is 0.0246 cult/lb. These then give an estimate of the mass of the reactor coolant between 11375 cu~fl.I.0204 cuftllb = 5.576E5 lb and 11375 cuftl0.0246 cuft/ib = 4.624E5 lb. The use of 5.4E5 lb is therefore appropriate. Specific volume for the RSG was based on the reactor vessel inlet temperature of 557.3 (ref. 16) and was calculated to be 0.021986 ft 3/lb.

c) Reference 9 gives the weight of water in a steam generator as 4.745E7g which translates to 4"4,745E7 g /453.59 g/lb = 4.184E5 lb. The use of 3.48E5 will result in higher concentrations and can therefore be used.

d) Reference 8 gives the maximum letdown as 120 gpm =120 gal/rain

60 min/hr

62.4 lb/cuft/7.48 gal/cufi

-6E4 lb/br. The use of 3.7E41b/hr is therefore an average flow which wili result in larger radioisotope concentrations. Note that 3.7E4' is the design flow through the cation demineralizers (see note g below).

e) The blowdown is given as 28,900 lb/hr ref.9. The use of 30,0001b/hr is a rounded off value and is not inappropriate.

f) This means that all radioisotopes~are removed. This is not a bad assumption because virtually all noble gasses wili be removed, and reference 12 gives the condensate demineralizers removal efficiencies of a factor of 10. A NBD factor of 0.9 could be used, but to be consistent with previous calculations, the factor will remain at 1.0.

g)The design flow through the cation denmineralizers is 75 gpm = 75. gal/min

60 nin/hr

62.4 Ib/cuft /7.48 gal/cult = 3.75E4 lb/hr (ref. 8). The use of 3.7E4 lb/hr will increase the radioisotope inventory and is thus conservative to use.

h) Reference 9 gives a value for NC as 0.589. 0.55 is the minimum value from reference 3.

CNL-15-216, Enclosure 1 CNL-5-1 6Encosue Attachment I 2, Page 11 of 49

CalcuationNo. WB

Subject:

3 Rev: 005 Plant WBN Page: 11 Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18.1-1 984 i) The use ofo0.0 Will maximize the noble gas inventories by not stripping any noble gasses from the reactor coolant. Note that the Y values given in Table 11.6 of ref. 3 are not used in the ANS-18.1-1984 corrections methodology.

j) These values are those for the Replacement Steam Generators. They result in lower concentrations and thus the original values are conservative. See Appendix A for the results using these values.

Equations used: (ref.2)

Lambda=In (2) I (Half Life)

R*=FB+ (FD-FB) Y WP R*, 3,*(FD) (NBE*.) +/-(!-NB* [FB+FA (NA*.I.L R*=9.0E-4 (this value comes directly from ANSI/ANS-l8.l-l984)

R=2=6.7E-2 (this value comes directly from ALNSI/ANS-18.1-1984)

R.*=3.7E-2 (this value comes directly from ANSI/ANS-18.1-1984)

R,4=O (this valu~e comes directly from ANSi/ANS-18.l-1984)

R,,=6.6E-2 (this value comes directly from ANSI/ANS-18.1-l984)

NB2 =.99 (this value comes directly from ANSI/ANS-18.l-1984)

NB3 =O.5 Cthis value comes directly from ANSI/ANS-18.1-1984)

NB,=0.98 (this value comes directly from ANSI/ANS-18.l-1984)

NA2 =O (this value comes directly from ANSI/ANS-18.l-1984)

NA3 =O.9 (this value, comes directly from ANSI/ANS-18 .l-1984)

NA*=O.9 (this value comes directly from ANSI/ANS-18.l-1984) r 2.... =(FBD) (N'BD)+(NS) (FS) (NC) (NX)

WS NS 2==.O1 (this value comes directly from ANSI/ANS-18.l-1984)

N'S,,*=5E-3 (this value comes directly from ANSI/ANS-18.1-1984)

NS5 =1.O (this value comes directly from ANSI/ANS-18.l-1984)

NX1 ,,,=0 (this value comes. directly from ANSI/ANS-18.l-1984)

N~,=. (this value comes directly from ANSI/ANS-18..l-1984)

NX, ffi.5 (this value comes directly from ANSI/ANS-18.1-1984)

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 12 of 49 Attachment

Calculation No.

Subject:

WBNNAL3003 ]Rev: 005 Plant: WBN Reactor Coolant and.Secondary Side Activities in Accordance with ANSIIANS-18.1-1984 j Page: 12 f ....... =P (WP)((Rm~+/-um+AMBA; f 4 ,5 =l, 0; f' 23 *6 = (WSftL*+LAMBDA) (f WP(P,) (R1...... +LAMBDA) (WS) (r 2 , .+LAMBDA) r,.... =O.l7.(this 6 value comes directly from ANSI/ANS-18.l-l984) r,,3=0.15 (this value comes directly from A~NSI/ANS-18.l.-1984) f' 1 =0.0 (negligible noble gasses in secondary water) f' =_* ;f' 5=1.0; f"...=W* r~*LMD)!~

WS (WS) (r2.... +LAMBDA)

f" 1 =(FS)(f*j ; f" 4 =WLS ;f"5 =l.O; A=f*A.,; A'=f'*A', ;A"f*"

FS WS where Subscript n refers to the nominal value of the variabl~e subscript (number) refers to class number

Lambda=nuclide decay constant hr"'

R=removal rate (reactor coolant) hr"'

r=removal rate (steam/secondary water) hr"*

f=adjustment *factor, reactor coolant f'=adjustment factor, secondary water f"=adjustment factor, secondary steam NA=fraction of material removed by the. cation demineralizer.

NB=fraction of material removed by the purification demineralizer NS=Ratio of concentration in steam to that of the steam generator NX=Fraction of activity removed by the condensate demineralizers A=activity uCi/g, reactor coolant A'=activity uCi/g, Secondary water A"=activity uCi/g, secondary steam CNL-5-21, Enlosue I 2, Page 13 of 49 Attachment CNL-15-216, Enclosure 1

Calculation Reactor

Subject:

No. WBNNAL3003 Coolant and Secondary Side Activities Rev: 005 Plant:

in Accordance withWBN Page:8.1-1 ANSIIANS-1 13984 TPC Average Concentration and TPBAR Failure: methodology and values from ref. 15

~Tritium Sources = TPBAR releases + IFBA releases + Non-TPC sources Realistic Source Term

TPBAR releases = #TPBAR6

5 Ci/TPBAR/yr IFBA releases =40 Ci/yr

Non'-TPC sources.= 870 Ci/yr TPC coolant average concentration is the non-TPC average concentration (1 jiCi/g) multiplied by the ratio of the tritium sources for the TPC core to the normal tritium sources C*=(1 pCi/g) (#TPBARs

5 + 40 + 870)/(870)

(1 jiCi/g) (704

5 + 40 + 870)/870 = 5.2 pCi/g for 704 TPBARs (ilpCi./g) (1900

5 + 40 + 870) /870 = 12.0 pCi/g for 1900 TPBARs Increase in coolant average concentrations due to TPBAJR failure:

mass of RCS = 5.4E5 lb'* 453.9 g/ib.= 2.45E8 g I*= 11,600 Ci I*=RCS inventory prior to failure = average concentration x RCS mass 5.1 4Lci/g x 2.45E8 g = 1249.5 Ci for 704 TPBARs 12.0 pCi/g x 2.45E8 g = 2940.0 Ci for 1900 TPBARs 1 TPBAR Failure C,.= (11,600 + 1249.5) Ci! 2.45E8 g = 52.45 J.Ci/g for 704 TPBARs Cn*e (11,600 + 2940.0) Ci/ 2.45E8 g = 59.35 jiCi/g for 1900 TPBARs 2 TPBAR Failure C = (2*11,600 + 1249.5) Ci/ 2.45E8 g = 99.79 *1Ci/g for 704 TPBARs C = (2*11,600 + 2940.6) Ci/ 2.45E8 g = 106.69 *LCi/g for 1900 TPBARs where I is the inventory, and C is the average concentration Desion Basis Source Term TPBAR releases = #TPBARs

10 Ci/TPBAR/yr IFBA releases = 40 Ci/yr Non-TPC sources = 870.Ci/yr TPC coolant average concentration is the non-TPc average concentration (i jIci/g) multiplied by the ratio of the tritium sources for the TPC core to the normal tritium sources C* = (i jiCi/g) (#TPBARs

10 + 40 +. 870)/(870)

(l pCi/g) (2500

10 + 40 + 870)/870 = 29.8 pCi/g for 2500 TPBARs Increase in coolant average concentrations due to TPBAR failure:

mass of RCS = 5.4E5 lb

453.9 g/ib = 2.45E8 g ITPAR = 11,600 Ci I~s= RCS inventory prior to failure = average concentration x RCS mass 29.8 *iCi/g x 2.45E8 g = 7301.0 Ci for 2500 TPBARs 1 TPBAR Failure C = (11,600 + 7301.0) Ci/ 2.45E8 g = 77.15 *iCi/g for 2500 TPBARs 2 TPBAR Failure C = (2*11,600 + 7301.0) Ci/ 2.45EB g = 124.49 *iCi/g for 2500 TPBARs where I is the inventory, and C is the average concentration CNL-15-216, Enclosure I CNL-5-21, Enlosue I 2, Page 14 of 49 Attachment

CalculatiooNo. WB NNL3003 IRev: 00 Plant: WBN IPage: 14 [

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18.1-1984 REACTOR COOLAN~T ACTIVITIES nominal calculated half *removal adjust.* reactor reactor life Lambda rate factor coolant coolant Nu~clide .ref.5 1/hr R f ANSI-18.1 WBN uCi/gm uci/gsm Class 1 Kr- 85m 1. 55E-01 1.0 7E+00 1. 6E-01 1.71E-0l 4.481h 1.56E-03 Kr- 85 10. 7y 7. 39E-06 6.19E-01 4 .3E-01 2 .66E-0l 5.47E-0l 1.56E-03 1.5E-01 Kt- 87 7 6m 1. 07E+00 1. S1E-01 1.56E-03.

Kr- 88 2.84h. 2 .44E-01 1.07E+00 2 .8E-01 3 .00E-01 1.56E-03

Xe-131m 11. 77d 2. 45E-03 8. 96E-01 7.3E-01 6. 54E-01 1.56E-03 Xe- 133m 2 .19d 1. 32E-02 1. 02E+00 7.0OE-02 7. 17E-02*

1.56E-03 Xe-.133 5 .25d 5.50E-03 9. 72E-01 2. 6E+00 2. 53E+00 I. 56E-03 1. 3E-0i Xe- 135m 15 .6m 2.67E+00 1. 07E+00 1.39E-01

i. 56E- 03 Xe- 135 9.1lh 7.62E-02 1. 06E+00 8. 5E-01 9.04E-01

i. 56E- 03 Xe- 137 3. 82m 1. 09E+0! 1. 07E+00. 3 .4E-02 3 .65E-02.

1. 56E-03 1 .2E-01 Xe -138 14 .1m 2 .95E+00 1. 07E+00 1.29E-01 1..76E-02 Class 2 Br-84 31.8m l.31E+00 1. 07E+00 1. 6E-02 1. 72E-02 6.78E-02 1-131 8.04d 3 .59E-03 1. 06E+00 4. 5E-02 4 .77E-02 6 .78E-02 1-132 2.28h2 3. 04E-01 1. 07E+00 2.1lE-01 2 .25E-01 6 .78E-02 1-133 20.9h2 3.32E-02 1. 06E+00 1 .4E-O1 1 .49E-01 1-134 52.6m .6.78E-02

7. 91E-01 1. 07E+00 3 .4E-01 3. 64E-01 6.78E-02 1-135 6.61h2 1. 05E-01 1. 07E+O0 2. 6E-01 2. 78E-01 3.81E-02 Class 3 Rb-88 17.8m 2 .34E+00 1.07E+00 l.9E-01 2.04E-01l 3 .81E-02 Cs-134 2.062y 3. 84E-05 1. 04E+00 7 .1E-03 7.3 9E-03 3.81E-02 Cs-136 13.ld 2 .20E-03 1.04E+00 8.7E-04 9. 08E-04 3.8 1E-02 Cs-137 30.1l7y 2 .62E-06 1 .04E+00 9 .4E-03 9.79E-03 Class 4 N-16 7.13s 3. 50E+02 1 .00E+00 4.0E+01 4.00E+0l Note: N-16 will be 0.0 ,uCi/g outside the shield building, and up to 40 uCi/gm inside.

The exact concentration will be highly dependent on how long the water has been outside the core (due to the very short half life).

CNL-15-216, Enclosure I CNL-5-21, Enlosue I 2, Page 15 of 49 Attachment

Calculation No. WBNNAL3003 Rev: 005 Plant: WBN Page: 15

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-1 8.1-1 984 R~EACTOR COOLANT ACTIVITIES (continued) nominal calculated half removal adjust. reactor reactor life Lambda rate factor coolant coolant Nuclide ref.5 1/hr* R f ANSI-18.1 WBN uCi/gm uCi/gm Class 5 H-3 12.33y 6.42E-OG - 1.00E+00O l.0E+OO l.O0E+00 Class 6 Na-24 15.02h 4.G1E-02 G.73E-02 1.06E+0O 4.7E-02 4.99E-02 Cr-51. 27.7d l.04E-03 6.73E-02 1.05E+00O 3.lE-03 3.26E-03 Mn-54 312d 9.26E-05 6.73E-02 l.05E+00 1.6E-03 1.68E-03 Fe-55 2.7y 2.93E-05 6.73E-02 1.05E+00 1.2E-03 1.26E-03 Fe-59 44.6d 6.48E-04 6.73E-02 1.05E+00O 3.0E-04 3.16E-04 Co-58 70.8d 4.08E-04 6.73E-02 l.05E+OO 4.6E-03 4.84E-03.

Co-6O 5.2 7 1y 1.50E-05 6.73E-02 l.05E+00 5.3E-04 5.58E-04 Zn-G5 244.ld 1.18E-04 G.73E-02. 1.05E+00 5.1E-04 5.37E-04 Sr-89 50.5d 5.72E-04 6G.73E-02 1.O5E+OO 1.4E-04 1.47E-04 2

Sr-90 8.8y 2.75E-06 6.73E-02 1.05E+00O 1.2E-05 1.26E-05 Sr-91 9.5h 7.30E-02 G.73E-02 l.0GE+00 9.GE.-04 l.02E-03 Y-91m 49.7m 8.37E-L0 G.73E-02 1.07E+00 4.GE-04 4.93E-04 Y-91 58.5d 4.94E-04 6.73E-02 l.05E+OO 5.2E-06 5.47E-06 Y-93 10.2h 6.80E-02 6.73E-02 1.06E÷00 4.2E-03 4.46E-03 Zr-95 64d 4.51E-04 6.73E-02 l.05E+OO 3.9E-04 4.10E-04 Nb-95 35d 8.25E-04 6.73E-02 1.05E+OO 2.8E-04 2.95E-04 Mo-99 66.02h 1.05E-02 6.73E-02 l.06E+00O G.4E-03 6.75E-03 Tc-99m 6.02h 1.15E-01 6.73E-02 1.07E+O0 4.7E-03 5.0iE-03 Ru-103 39.4d 7.33E-04 6.73E-02 l.05E+OO 7.5E-03 7.89E-03 Ru-106 367d 7.87E-05 6.73E-02 1.05E+OO 9.0E-02 9.47E-02 Ag-l10m252d 1.15E-04 6.73E-02 1.05E+00 1.3E-03 l.37E-03 Te-12,9m33.5d 8.62E-04 6.73E-02 1.05E+O0 1.9E-04 2.00E-04 Te-129 67m 6.21E-O1 6.73E-02 1.07E+O0 2.4E-02 2.57E-02 Te-131m30h 2.31E-02 6.73E-02 1.06E+OO 1.5E-03 1.59E-03 Te-131 25m 1.66E+OO 6.73E-02 1.07E+O0 7.7E-03 8.26E-03 Te-132 78h 8.89E-03 6.73E-02 l.05E+00 1.7E-03 l.79E-03 Ba-140 12.79d 2.26E-O3 6.73E-02 l.05E+0O l.3E-02 1.37E-02 La-140 40.3h 1.72E-02 6.73E-02 1.O6E+OO 2.5E-02 2.64E-02 Ce-141 32.5d 8.89E-04 6.73E-02 l.05E+00 1.5E-04 1.58E-04 Ce-143 33h 2.10E-02 6.73E-02 1.O6E+OO 2.8E-03 2.96E-03 Ce-144 284d 1.02E-04 6.73E-02 1.05E+OO 4.OE-03 4.21E-03 W-187 23.9h 2.90E-02 6.73E-02 l.O6E+O0 2.5E-03 2.65E-03 Np-239 2.35d 1.23E-02 6.73E-02 l.06E+00 2.2E-03 2.32E-03 CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 2, Page 16 of 49

Calculation No. WBNNAL3003

Subject:

Reactor

Coolant and Secondary Side Activities. Rev: 005 Plant:

in Accordance withWBN jPage:8.1-116984.

ANSI/ANS-1 SECONDARY SIDE ACTIVITIES nominal nominal removal adjust. adjust, secondary secondary secondary secondary rate factor factor water water steam steam Nuclide r f1 f" ANSI-l8.1lWBN *ANSI-18.lWBN uCi/gm uCi/gm uci/gm uci/gm Class 1 Kr-85m O.OOE+OO O.OOE+OO 1.07E+OO 0.OE+OO O.OOE+OO 3.4E-O8 3.63E-08 Kr-85 O.OOE+OO 0.OOE+OO .6.19E-01 O.OE+OO O.OOE+00 8.9E-08 5.51E-08 Kr-87 O.O0E+00 O.OOE+OO 1.07E+O0 O.OE+OO O.00E+00 3.0E-08 3.22E-08 Kr-88 0.O0E+OO O.OOE+0O 1.07E+0O O.0E+00 O.OOE+OO 5.9E-08 6.31E-O8 Xe-131m0.O0E+00 0.OOE+OO 8.96E-01 0.0E+00 O.00E+00 l.5E-07 l.34E-07 Xe-133m0.OOE+00O O.OOE+OO 1.02E+OO O.OE+00O O.00E+OO l.5E-08 l.54E-08 Xe-l33 0.00E+OO O.OOE+O0 9.72E-01l 0.E+OO 0.O0E+00 5.4E-07 5.25E-07 Xe-135m0.00E+00 O.OOE+OO 1.07E+00 O.0E+00O O.00E+OO 2.7E-08 2.90E-08 Xe-135 0.OOE+00 0.0OE+0OO 1.06E+OO O.OE+OO O.00E+0O l.8E-07 l.91E-07 Xe-137 0.0OE+00 0.OOE+00 1.07E+OO O.OE+OO O.O0E+OO 7.1E-09 7.62E-09 Xe-138 0.OOE+OO O.OOE+OO l.07E+OO 0.0E+00 O.OOE+00 2.5E-08 :2.68E-08 Class 2 Br-84 3.O0E-O1 l.27E+OO0 1.27E-t00 7.5E-08 9.56E-08 7.5E-lO. 9.56E-lO 1-131 3.O0E-01 7.85E-O1 7.85E-Ol 1.8E-06 l.41E-06 1..8E-08* 1.41E-08 1-132 3.00E-01 l.09E+OO 1.09E+OO 3.1E-06 3.37E-O6 3.1E7 08 3.37E-08 1-133 3.00E-Ol 8.40E-O1 8.40E-Ol1 4.8E-06 4.03E-06 4.8E-08 4.03E-08 1-134 3.OOE-O1 1.22E+OO l.22E+OO 2.4E-06 2.93E-06 2.4E-O8 2.93E-08 1-135 3.OOE-O1 9.38E-O1 9.38E-Ol 6.6E-06 6.l9E-06 6.6E-08 6.19E-08 Class 3 Rb-B8 1.45E-Ol 1.39E+OO l.39E+00 5.3E-07 7.36E-07 2.GE-09 3.61E-09 Cs-134 1.45E-01 1.39E+OO l.39E+00O 3.3E-07 4.58E-07 1.7E-09 2.36E-09 Cs-136 1.45E-0l l.39E+00O l.39E+00O 4.OE-08 5.56E-08 2.0E-lO 2.78E-lO Cs-137 1.45E-O1 1.39E+OO 1.39E+OO 4.4E-07 6.llE-07 2.2E-09 3.05E-09 Class 4

N-16 O.OOE+00 l.29Ei-00 I.29E+OO 1.OOE-06 1.29E-06 1.OE-07 1.29E-07 Note: N-16 will be 0.0 uCi/g outside the shield building, and up to lE-6 uCi/gm inside.

The exact concentration will be highly dependent on how long the water has been outside the core (due to the very short half life).

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue Attachment I 2, Page 17 of 49

Calculation No. WBNNAL3003 Rev: 005 Plant: WBN Page: 17

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSlIANS-1 8..1-984 SECONDARY SIDE ACTIVITIES (continued) nominal nominal removal* adjust, adjust. secondary secondary secondary secondary rate factor factor water water steam steam Nuclide r f, f" ANSI*-18. W~BN ANSI-18.1 WBN uCi/gm uCi/gm uCi/gm uCi/gm Class 5 R-3 0.O0E+OO 1.OOE+00 1.00E+00 1.0E-03 1.00E-03 l.0E-03. 1.00E-03 Class6 Na-24 1.93E-01 l.24E+00O 1.24E+00 1.5E-06 1.86E-06 7.5E-09 9.30E-09 Cr-51 1.93E-Ol 1.20E+OO 1.20E+OO 1.3E-07 1.56E-07 6.3E-I0 7.56E-1O.

Mn-54 1.93E-O1 l.20E+OO 1.2OE+00O 6.5E-08 7.80E-08 3.3E-I0 3.96E-I0 Fe-55 l.93E-Ol l.20E+OO 1.20E+00 4.9E-08 5.88E-O8 2.5E-lO .3.OOE-l0 Fe-59 l.93E-Ol l.20E+00 l.20E+OO l.2E-O8 l.44E-O8 6.lE-ll 7.32E-ll Co-58 l.93E-Ol 1.20E+00 l.20E+OO 1.9E-07 2.28E-07 9.4E-l0 1.13E-09 Co-60 1l.93E-O1 1.20E+O0 1.20E+OO 2.2E-08 2.64E-08 I.lE-l0 1.32E-lO

Zn-65 .1.93E-O1 1.20E+00 1.20E+00 2.1E-O8 2.52E-08 1.0E-10 1.20E-l0 Sr-89 l.93E-O1 1.20E+OO l.20E+OO 5.7E-D9 6.84E-09 2.9E-lI 3.48E-II Sr-90 1.93E-O1 1.20E+OO l.20E+OO 4.9E-I0 5.88E-I0 2.5E-12 3.00E-12

Sr-91 1.93E-Ol 1.26E+00O 1.26E+00 2.8E-08 3.52E-08 1.4E-10 l.76E-l0 Y-91m l.93E-O1 1.35E+OO 1.35E+O0 3.2E-O9 4.34E-09 l.6E-ll 2.17E-ll Y-91 1.93E-O1 1.20E+OO 1..20E+00O 2.1E-10 2.52E-lO. l IE-12 l.32E-12 Y-93 1.93E-01 l.25E+O0 1.25E+00O 1.2E-07 1.50E-07 6.lE-10 7.65E-10 Zr-95 1 93E-O1 1.20E+00 1.20E+00O 1.6E-08 1.92E-08 7.9E-If 9.48E-II Nb-95 l.93E-O1 1.20E+00 *l.20E+00 l.IE-O8 l.32E-08 5.7E-II 6.84E-ll Mo-99 1.93E-O1 l.21E+O0 1.21E+00 2.5E-07 3.03E-07 l.2E-09 l.45E-09 Tc-99m 1.93E-O1 1.28E+0O 1.28E+O0 l.lE-07* 1.40E-07 5.7E-10 7.27E-lO Ru-103 l.93E-01 l.20E+OO l.20E+00 3.lE-07 3.72E-07 l.6E-09 l.92E-09 Ru-106 1.93E-O1 1.20E+OO l.20E+O0 3.7E-06 4.44E-06 l..SE-08 2.16E-08 Ag-ll0ml.93E-0l 1.20E+00 l.20E+O0 5.3E-08 6.36E-08 2.7E-10 3.24E-l0 Te-129m1.93E-01 1.20E+00 l.20E+00 7.8E-O9 9.36E-09 3.9E-If 4.68E-ll Te-129* l.93E-O1 l.35E+O0 l.35E+OO 2.2E-07 2.96E-07 l IE-09 l.48E-09 Te-131m1.93E-O1 l.22E+00 1.22E+OO 5.4E-08 6.60E-08 2.7E-10 3.30E-I0 Te-131 1.93E-O1 1.37E+OO 1.37E+OO 2.9E-08 3.97E-08 l.5E-10 2.05E-l0 Te-132 1.93E-O1 1.21E+OO l.21E+00O 6.6E-08 7.98E-O8 3.3E-10 3.99E-lO Ba-140 1.93E-Ol 1.20E+OO l.20E+OO 5.2E-07 6.25E-07 2.6E-O9 3.12E-09 La-140 I.93E-01 l.22E+O0 .l.22E+00 9.3E-07 1.13E-O6 4.6E-09 5.60E-09 Ce-141 l.93E-O1 l.20E+00 l.20E+00 6.1E-09 .7.32E-09 3.lE-ll 3.72E-ll Ce-143 l.93E-Ol 1.22E+OO 1.22E+O0 1.OE-07 1.22E-07 5.IE-10 6.23E-10 Ce-144 l.93E-Ol l.20E+O0 l.20E+OO 1.6E-07 1.92E-07 8.2E-10 9.83E-1O W-187 1.93E-01 l.23E+00 1.23E+00 8.7E-08 1.07E-07 4.4E-10 5.40E-I0 Np-239 1..93E-01 1.21E+O0 l.21E+OO .8.4E-08 1.02E-07 4.2E-lO 5.09E-l0 CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2,. Page 18 of 49 Attachment

SCalculation No. WBNNAL3003 Rev: 005 Plant: WB ag:1

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSIIANS-1 8..1-1 984 Result The following table presents the adjusted expected reactor coolant and secondary side equilibrium nucdide activities. These activities are intended for possible use in environmental reports, in normal dose calculations for equipment qualification purposes,

and for other applications where the use of expected average data over the lfe* of the plant would be appropriae. It should be [

noted that the failure of a TPBAR is considered an abnormal event Reactor Coolant Secondary Water Secondary Steam Nuclide* WBN WBN WBN uCi/gm uCi/gm uCi/gm Class 1 Kr-85m 1.71E-Ol O.00E+00 3.63E-08 Kr-85 2.66E-01 0.OOE+O0 5.51E-08 Kr-87 1.61E-01 0.00E+00 3.22E-08 Kr*-88 3.Q0E-01 O.OOE+0O 6.31E-08 Xe-131m 6.54E-O1 0.00E+00 1.34E-07 SXe-133m 7.17E-02 0.0OE+00 1,54E-08 Xe-133 2.53E+00 0.00E+00O 5.25E-07 Xe-135m 1.39E-Ol 0.OOE+0O 2.90E-O8 Xe-135... 9.04E-01 Q.00E+00O 1.91E-07 Xe-137 3.65E-02 0.OOE+OO 7.62E-09 Xe-138 1.29E-O1. O.OOE+OO 2.68E-08 Class 2 Br-84 1. 72E-02 9.56E-08 9.56E-I0 1-131 4.77E-02 1.41E-06 1.41E-08 1-132 2.25E-O1 3.37E-06 3.37E-08 1-133 1.49E-O1 4.03E-06 4.03E-08 1-134 3.64E-O1 2.93E-06 2.93E-08 1-135 2.78E-O1 6.19E-06 6.19E-08 Class 3 Rb-88 2.04E-01 7.36E-07 3.61E-09 Cs-134 7.39E-03 4.58E-07 2.36E-Q9 Cs-136 9.08E-04 5.56E-O8 2.78E-1O Cs-137 9.79E-03 6.11E-07 3.O5E-O9 Class 4 N-16 4.OOE+Ol 1.29E-06 l.29E-07 CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 19 of 49 Attachment

Calculation No. WBNNAL3003 [Rev: 005 Plant: WBN Page: 19

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSllANS-18.1-1984 Reactor Secondary Secondary Nuclide Coolant Water Steam Class5 uCi/gm udi/gm uCi/gm H-3 1.00OE+00 1.00OE-03 1.00E-03 Realistic Source Term TPC 704 5.10E+00' 5.lOE-03 5.10E-03' 1 rod 5.25E+01 5.25E-02 5.25E-02

2 rod 9.98E+01 9.98E-02 9.98E-02

TPC 1900 1.20E+0l' l.20E-02 l.20E-02' 1 rod 5.94E+01. 5.94E-02 5.94E-02

2 rod 1.07E+02 1. 07E-01 l.07E-01' Design Basis Source Term TPC 2500 2 .98E+01' 2.98E-02 2.98E-02

1 rod 7. 72E+01 7.72E-02 7.72E-02

2 rod. 1.24E+02 l.24E-01 l.24E-01

Class 6 Na-24 4. 99E-02 1. 8GE-06 9. 30E-09 Cr-51 3 .26E-03 1. 56E.-07 7. 56E-I0 Mn-54 1.68E-03 7. 80E-08 3. 96E-10 Fe-55 1.26E-03 5. 88E-08 3.00GE-10 Fe- 59 3.16GE-04 1. 44E-08 7.32E-II Co- 58 4. 84E-03 2. 28E-07 1.13E-09 Co- 60 5. 58E-04 2. 64E-08 1. 32E-l0 Zn-65 5..37E-04 2. 52E-08 1. 20E-10 Sr-89 1. 47E-04 6. 84E-09 3 .48E-1I Sr-90 l.26E-05 5.88E-10 3. OOE-12 Sr-91 1.02E-03 3 .52E-O8 1. 76E-I0 Y-90

1.26GE-O5 5. 88E-l0 3.00OE-12 Y- 91m 4.93E-04 4.34E-09 2. 17E-ll Y- 91 5 .47E-O6 2 .52E-l0 1. 32E-12 Y-93. 4.46E-03 1. 50E-07 7.G5E-10 Zr-95 4.10OE-04 1. 92E-08 9 .48E-lI Nb- 95 2. 95E-04 1. 32E-08 6.,84E-II Mo -99 6. 75E-03 3.0 3E-07 1.45E-09 Tc -99m 5.O1E-03 1 .40E-07 7.27E-10 Ru- 103 7.89E-03 3. 72E-07 1 .92E-09 Ru- 106 9 .47E -02 4 .44E-0G 2.1l6E-O8 Rh-103m

7. 89E-03 3. 72E-07 1. 92E-09 Rh-106

9 .47E-02 4 .44E-O6 2.16E-08 Ag-ll~m l.37E-03 6. 36E-08 3 .24E-l0 Te-129m 2.00OE-04 9. 36E-09 4.6G8E-lI Te-129 2. 57E-02 2.96E-07 1 .48E-09 Te-131m l.59E-03 6 .60E-08 3 .30E-I0 Te-131 8.26GE-03 *3.97E-08 2. 05E-I0 Te-132 1. 79E-03 7.98E-08 3. 99E-I0 Ba-137m

9 .79E-03 6.11E-07 3.,05E-09 Ba-140 1.37E-02 6.25E-07 3. 12E-09 La-140 2.,64E-02 1. 13E-06 5.60E-09 Ce -141 1.58E-04 7. 32E-09 3.72E-II CNL-15-216, Enclosure 1 5-1 6Encosue I 2, Page 20 of 49 CNL-Attachment

CalculationNo.

Subject:

Reactor WBNNAL3003 fRev: 005 Plant: WBN fPage: 20 Coolant and Secondary Side Activities in Accordance with ANSIIANS-18.1-1984 Ce-1d3 2.96E-03 1.22E-07 6.23E-1O Ce-144 4.21E-03 1.92E-07 9.83E-I0 Pr-143 ** 2.96E-03 1.22E-07 6.23E-I0 Pr-144

4.21E-03 1.92E-07 9.83E-I0 W-187 2.65E-03 1.07E-07 5.40E-10 Np-239. 2.322-03 1.02E-07 5.09E-I0

Assumption 2 -daughter which is in secular equilibrium with parent see text for more detail

See Assumption 3

Maximum for TPC with no dilution under normal operation.

Secondary side is scaled to the primary side i.e. ratio is same for conventional core end TPC.

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 21 of 49 Attachment

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSIIANS-1 8.1-1 984.

References

1. ANS N237-1976 (ANS -18.1), "Radioactive Materials in Principal Fluid Streams of Light-Water-Cooled Nuclear Power Plants," May 11!, 1976

2. ANSL'ANS-18. -1984, Radioactive Source Term for Normal Operation of Light water Reactors -Tables 2,6,9,and 11 attached

3. WBN FSAR Table 11.!.6 Amendment 72

4. N3-68-4001 R2 "Reactor Coolant System" RJMS# B26 890403 004

5. Lederer, C.Micheal and Shirley, Virginia eds. "Table of Isotopes," seventh edition, John Wiley and Sons, Inc., New York, 1978

6. N3-1-4002 R3 "Main Steam System" RIMS# B26 880719 079

7. telephone conversation between M. Berg and C.D. Thomas, Jr. of Yankee Atomic Electric Company on 3/13/87 -attached

8. N3-62-4001 R2 "Chemical and Volume Control System" RIMS# B26 880726 058

9. WBNAPS3-052 RI "Minimum Detectable Leak Rate For the Steam Generator Blowdown System" RJMS# B45 880620 23 7

10. WBNNAL3-002 R3 "100-Day LOCA-DBA Source .Terms for the EGTS and ABGTS Filters, Containment, Sump, and Shield Building Annulus"

11. Kennan, Joseph H. and Keyes, Frederick G. "Thermodynamic Properties of Steam" 1st edition, John Wiley and Sons, Inc.

New York

12. NUiREG-00.17 R1 "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents From Pressurized Water Reactors" April 1976

13. Letter from Jim Chardos, TVA, to M.L. Travis, Wesdyne,* "TVA* Plant Specific Tritium Concentration Data," Sept. I11, 2000, RIMS# T35 000911 913

14. EDC E50629A i5. WAT-D- 10890, Westinghouse document NIDP-00-0326, "Transmittal of Waste Management Evaluation for the Watt's Bar Tritiujm Production Core," RIMS# T71 001204 807

16. WB1RSG-TR-02 R4, "RSG-OSG Comparison Document" transmitted via TVWES-0725

17. DCN 51754A

18. DCN 61599

This reference is for information only. It is not used for design input.

CNL-15-216, Enclosure 1 5-1 6 Encosue I 2, Attachment CNL- Page 22 of 49

Cl OculationNo.

Subject:

Reactor WBNNAL3003 Coolant and Secondary Side ActivitiesIRev: 00 Plant:

In Accordance withWBN IPage: 22 ANSI/ANS-18.1-1984 Appendix A Steam Generator Replacement Results This project revised the following parameters used in the body of t *his *calculation (OSG is the original steam generators):

OSG RSG (ref. li6)

Steam.Flow Rate - FS 1.5E+Q7 1.54E+07 Weight of Water in the RCS - WP 5.4E+05 S.78E+05 (based on density at RV +/-nlet temp)

Weight of Water in all SG's - WS 3.48E+C5 6.79E+05 The following are the results of following the same methodology in the body of this calculation using the above RSG parameters.

nominal calculated half removal *- reactor reactor adjust.

life Lambda rate factor coolant coolant Nuclide ref.S5 *1/hr R f ANSI -18 .1 WBN uCi/gm uCi/gm Class 1 Kr- 85m 4.48 1 .55E-01 1 .46E-03 9.9 95-01 1.605-01 1.60E-01 h

Kr- 85 10.7 y 7 .39E-06 1 .46E-03 6.19E-01 4.305-01 2.66E-01 Kr-87 76 m 5 .47E-01 1 .465-03 1.005+00 1.505-01 1. 50E-01 Kcr-88 2.84 h 2.44E-01 1.465-03 1. 00E+00 2.80E- 01 2.80E-01 Xe-131m 11 .77 d 2.45E-03 1.46E-03 8.S5E-01 7.30E-01 6.27E-01 xe- 13 3m 2.19 d 1 .32E-02 1.46E-03 9.64E-01 7.005-02 6 .75E-02 Xe- 133 5.25 d S5.50E-03 1.46E-03 9.22E-01 2.605+00 2.40E+00 Xe -135Sm 15.6 m 2. 67E+00 1.46E-03 1 1. 005+00 1.305-01 1.305-01 Xe -135 9.1 h 7.625-02 1 .46E-03 9.95E-01 S. 505-01 8.46E-01 Xe -137 3 .82 m 1. 09E+0! 1 .46E-03 1.00E+00 3.40E-02 3.415-02 Xe -138 14.1 m 2. 95E+00 1.465-03 1.005+00 l'.205-01 1.205-01 Class 2

Br- 84 31.8 m 1.31E+00 6.345-02 1.01E+00 1.605-02 1.61E-02 1-131 8.04 d 3.59E5-03 6.34E-02 1. 065+00 4.50E-02 4.75E-02 1-132 2.28 h 3.04E-01 6.34E- 02 1.015+00 2.10E-01 2.13E-01 1-133 20.9 h 3.325-02 6 .34E-02 1.04E+00 1.4 0E-01 1.46E-01 1-134 52.6 7. 91E-01 6.34E-02 1. 015+00 3.40E-01 3 .42E-01 1-135 6.61 h 1.05SE-01 6.345-02 1.025+00 2.60E-01 2.665-01 Class 3 Rb- 88 17.8 m 2.34E+00 3.565- 02 1.00E+00 1.905-01 1.91E- 01 Cs-134 2. 062 y 3.84E-05 3.565-02 1. 04E+00 7.105- 03 7.395-03 Cs -.13 6 13 .1 d 2.20E-03 3.56E-02 1.04E+00 8.70E-04 9.04E- 04 Cs -137 30.17 y 2.62E-06 3.;56E-02 1.045+00 9.4 05-03 9.79E-03 Class 4 N- 16 4.005+01 4. 00E+01 7.13 a 3.50E+02 0.OQE+00 1.005+00 Class S 5-3 6.42E-06 1.005+00 1.005+00 12.33 y 0.00E+00 1.005+00 Realistic Source Tern TPC 704/1900 5.10E+00' /I.20E+01' 1, rod 5.25E+01/5.94E+01 2 rod. 9.98E+01/1. 07E+02 Design Basis Source Term TPC 2500 2.985+01 '

1 rod 7.725+01 2 rod 1.24E+02 CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 2, Page 23 of 49

ICalculatino.o I

Subject:

- Reactor WB.NAL3003Coola'nt and Secondary Side Activities Rev: 005 jPlant:

in Accordance withWBN JPage: 23 ANSIIANS-18.1-1984*

Appendix A - Steam Generator Replacement Results (continued) nominal calculated half removal adjust. reactor *reactor life Lambda *rate factor coolant coolant Nuclide ref.5 1/hr R f ANSI-18.1 WBN uCi/gm uCi/gm

'Class 6 Na-24 15. 02 h 6.29.-02 1. 03E+00 4. 7oE-o2 4. 85E-02 4 .61E-02 Cr-51 27 .. 7 d 1. 04E-03 6. 29E-02 1. 05E+00 3.10OE-03 3 .26E-03 Mn-54 3 12 d 9 .26E-05 6 .29E-02 1. 05E+00 1. 60E-03 1.68E-03 Fe-55 2 .7. Y 2.93E-05 1.20E-03

.6.29E-02 1.05E+00 1.26E-03 Fe-59 44 .6 d 6.48E-04 6.29E-02 l.05E+00 3.O0E-04 3.16E-04 Co-58 70 .8 d 4.08E-04* 6.,29E-02 1.05E+00 4.60E-03 4.84E-03 Co-60 5.2 71 y 1..50E-05 6.29E-02 1.05E+O0 5.30E-04 5.58E-04 Zn-65 244 .i. d 1.18E-04 6.29E-02 1.05E+00 5.10E-04 5.37E-04 Sr-89 50 .5 d 5.72E-04 6.29E-02 1.O5E+-00 1 40E-04 1.47E-04 Sr-90 28 .8 Y 2.75E-06 6.29E-02 1.05E+00 l20E-05 1 1.26E-05 sr-91 9 .5 7.30E-02 6.29E-02 1.03E+-00 9.60E-04 9.85E-04 Y-9Q" 1.26E-05 Y-91m 49 .7 m 8.37E-01 . 6.29E-02 1.01E+00 4.60E-04 4.63E-04

Y-91 58 5" d 4.94E-04 6.29E-02 1.05E+00 5.20E-06 5.47E-06 Y-93 10 .2 h 6.80E-02 6.29E-02 1.03E+00 4.20E-03 4.31E-03 Zr-95 64 d 4.51E-04 6.29E-02 1.05E+00 3.90E-04 4.1OE-04 Nb -95 35 d 8.25E-04 6.29E-02 1.05E+00 2.80E-04 2.94E-04 Mo-99 66. 02 h 1.05E-02 6.29E-02 l.05E+00 6.40E-03 6.69E-03 Tc-99m 6. 02 h 1.15E-0'l 6.29E-02 l.02E+i0O 4.70E-03 4.79E-03 Ru-103 39 ).4 d 7.33E-04 6.29E-02 l.05E+-00 7.50E-03 7.89E-03 Ru-106 3 67 d 7.87E-05 6.29E-02 l.05E+00 9.00E-02 9.47E-02 Rh- 103m* 7.89E-03 Rh-106* -- 9.47E-02 Ag-llrn 2*52 d 1.15E-04 6.29E-02 l.05E+00 l.30E-03 1.37E-03 Te-129m 33 .5 d 8.62E-04 6.29E-02 1.05E+00 l.90E-04 2.OOE-04 Te-129 67 m 6.21E-01 6.29E-02 ).01E+00 2.40E-'02 2.42E-02 Te-131m 30 h 2.3!E-02 6.29E-02 1.04E+00 l.50E-03 1.56E-03 Te-131 25 m 1.66E+00 6.29E-02 1.00E+00 7.70E-03 7.73E-03 Te-132 78 h 8.89E-03 6.29E-02 1.O5E+00 1.70E-03 1l.78E-03 Ba- 137m* S9.79E-03 Ba-140 12. 79 d 2.26E-03 6.29E-02 1.05E+00 l.30E-02 1.37E-02 La-140 40 ).3 h 1.72E-02 6.29E-02 l.04E+00 2.50E-02 2.60E-02

Ce-141 32 .5 d 8.89E-04 6.29E-02 1.05E+00 i.50E-04 1.58E-04 Ce- 143 33 h 2.l0E-02 6.29E-02 l.04E+00 2.80E-03 2.91E-03 Ce- 144 2*84 d 1.02E-04 6.29E-02 l.O5E+00 4.00E-03 4.21E-03 Pr-143** 2.91E-03 Pr-144* 4 .21E-03 W-187 23 3.9 h '2.90E-02 6.29E-02 1l.04E+00 2.50E-03 2.59E-03

Np-239 2. 35 d 1.23E-02 6.29E-02 1.04E+0O' 2'.20E-03 2.30E-03

Assumption 2 -dau' ghter which is in secular equilibrium with parent see text for more detail

- See Assumption 3

Maximum for TPC with no dilution under normal operation.

CNL-15-218, Enclosure 1 CNL-5-21, AttachmentEnlosue I 2, Page 24 of 49

I Sujet: Caclain Ractor*

o WNAL03Rev:

Coln n Secondary Side Activities in Accordance 005* Plant:

withWBN Page:8.1-1 ANSIIANS-1 24 984

[1 I

Appendix A C(continued)

Steam Generator Replacement Results nominal nominal removal secondary secondary adjust. adjust.

secondary secondary rate factor factor water water steam steam Nuclide r f, f ,, ANSI-l8. 1 WBN ANSI-lB8.1 WBN uCi/gm uCi/gm uCi/gm uCi/gm Class .1 Kr-S85m 0.00OE+00 o .00E+0Q 9. 73E-01 o0.00E+O0 o .00E+00 3. 40E-08 3. 31E-O8 Kr- 85 o0.00E+00 o .00E+00 6.03E-O1 o 0E+00

. o0.00E+O0 B. 90E-08 5. 37E-08 Kr- 87 o0.00E+00 o 00OE+00 9. 75E-O1 o .0OE+00 o.O0E+00 3.00OE-08 2 .93E-08 Kr- 88 o0.00E+O0 o 00OE+00 9. 74E-0l o .00E+00 o0.00E+OO 5. 90E-08 5 .75E-08 Xe- 13 Im o0.00E+00 o .OOE+00 8 .36E-0l o .0OE+00 o0.00E+00 1. 50E-07 1.2 5E-07 Xe- 133m o0.00E+O00 0. 00E+O0 9 .39E-01 o 00OE+O0 .0.00OE+00 1 .50E-08 1 .41E-08 xe-133 o0.00E+00 o 00OE+00 8.98E-01 o 00OE+00 0.00OE+00 5 .40E-07 4. 85E-07 xe-135m o0.00E+00 o0.00E+00 9. 76E-01 o 00OE+00 0.00OE+00 2 .70E-08 2.64E-08 xe-13 5 o0.00E*00 o0.00E+00 9. 69E-01 o .00E+00 0.00OE+00 1.80OE-07 1. 74E-07 xe-137 o0.00E+00 o0.00E+00 9. 76E-O1 0. 00E+00 0.00OE+00 7.10OE-09 6 .93E-09 xe -138 o0.00E+00 o0.00E+O00 9 .76E-01 o0.00E+00 0. OOE+-O0 2.50E-08 2 .44E-08 Class 2 Br-84 1. 56E-O1 6. 72E-O1 6. 72E-01 7. 50E-08 5. 04E-O08 7.50OE-I0 5. 04E-10 1-131 1. 56E-01 7 .59E-01 7 .59E-01 1 .80E-06 1 .37E-06 1.80OE-O8 1.3 7E-08 1-132 1. 56E-01 6 .90E-01 6 .90E-0l 3. 1OE-06 2 .14E-06 3.10E-0B 2. 14E-08 1-133 1. 56E-01 7 .38E-01 7. 38E-01 4.80OE-06 3 .54E-06 4. 80E-08 3 .54E-08 1-134 1. 56E-01 6 .76E-01 6. 76E-01 2 .40E-06 1. 62E-06 2 .40E-08 1. 62E-08 1-135 1. 56E-01 7. 14E-01 7 .14E-01 6. 60E-06 4. 71E-06 6.60E-08 4. 71E-O8 Class 3 Rb-Ba 7. 53E-02 6. 85E-01 6. 85E-O1 5. 30E-07 3 .63E-07 2. 60E-09 1. 7BE-09 Cs- 134 7. 53E-02 1. 37E+OO 1. 37E+00 3.3 0E-07 4 .53E-07 1.70E-09 2 .33E-09 Cs- 136 7. 53E-02 1.3 5E+00 1 .35E+00 4.00OE-08 5 .40E-08 2.00GE-10 2 .70E-10 Cs-137 7. 53E-02 1 .37E+00 1.37E+00 4 .40E-07 6.04E-07 2 .20E-09 3. 02E-09 Class 4 N-16 0.00E+00 6.62E-Ol 6.62E-0l l.00E-06 6.62E-07 1.O0E-07 6.62E-08 Class 5

}H-3 0.GGE+O0 1.OOE+O0 1.00E+00 1.00E-03 1.O0E-03 1.00E-03 1.00E-03 Realistic Source Term TPC-704 *.1OE-03' 5.1OE-03' 1 roc 55.25E-02' 5.25E-02' 2 roe ).98E-O2' 9.98E-02' 5

TPC-1900 .20E-02' 1.20E-02' 1irod 1 .94E-02' 5.94E-02' 2 rod

07E-01' 1.07E-01' Design Basis Source Term TPC-2500 2. 98E-02' 2.98E-02' 1 rod 7

72E-02' 7.72E-02' 2 rod 1 .24E-01' 1o24E-01'

'Secondary side is scaled to the primary side i.e. ratio is same for conventional core and TPC.

CNL-15-216, Enclosure 1 CNL-5-1 6Encosue Attachment I 2, Page 25 of 49

ICalculationN.o WBNNAL3003 IRev: 00 Plant: WBN Page: 25 I

Subject:

Reactor Coolant and Secondary Side Activities in Accordance with ANSIIANS-18.1 -1984 Appendix A (continued)

Steam Generator Replacement Results

" nominal nominal removal adjust. adjust. secondary secondary secondary secondary rate factor factor water water steam st earn Nucl ide r f I f ,, ANSI-18 .1 WBN ANSI-18 .1 WBN uCi/gm uCi/gm uCi/gm uCi/gm Class 6 Na-24 1.00E-01 1.01E+00 1.01E+00 1.50E-06 1.51E-06 7.50E-O9 7.56E-09 Cr-51 l.00E-01 1.18E+'O0 1.18E+00 1.30E-07 1.53E-07 6.30E-1O 7.41E-l0 Mn- 54 1.00E-01 1.18E+O0 1.18E+00 6.50E-08 7.68E-08 3.30E-10 3.90E-lO Fe-55. 1.00E-O1 1.18E+O0 1.18E+00 4.90E-08 5.79E-08 2.50E-l0 2.95E-10 Fe -59 1.00E-01 1..18E+O0 1.18E+00 1.20E-08 1.41E-08 6.l0E-ll 7.19E-ll Co -58 1.00E-01 1.18E+00 1.18E+00 1.90E-07 2.24E-07 9.40E-l0 l.llE-09 Co-SO 1.00E-01 1.18E+00O l.18E+00 2.20E-08 2.60E-08 1.10E-10 1.30E-10 Zn- 65 1.00E-01 1.18E+00 l.18E+00 2.10E-08 2.48E-08 l.OOE-l0 l.18E-10 Sr-89 1.00E-O1 1.18E+00 1.18E+00 5.70E-09 6.72E-09 2.90E-11 3.42E-11 Sr-9O 1.00E-01 1.18E+00 1.18E+00 4.90E-10 5.79E-I0 2.50E-12 2.95E-12 1.O0E-01 S.53E-01 9.53E-01 2.80E-08 2.67E-08 1.40E-10 1.33E-10 Y-90O* .. 5. 79E-I0 2.955E-12 Y-91lm 1.O0E-01 7.16E-Ol 7.16E-01 3.20E-O9 2.29E-09 1.60E-ll l.15E-ll Y-91 1.OOE-01 1.18E+OO 1.18E+OO 2.10E-10. 2.48E-l0 1.lOE-12 l.30E-12 1.00E-01 9.62E-01 9.62E-O1 1.20E-07 l.15E-07 6.lOE-10 5.87E-lO Zr-95 l.00E-01 1.18E+00 1.18E+00 1.60E-08 1.89E-08 7.SOE-ll S.32E-ll Nb-955 1.OOE-O1 1.18E+O0 1.18E+0O 1.lOE-08 1.29E-08 5.70E-lI 6.71E-ll Mo- 99 1.OOE-O1 1.13E+OO 1.13E+O0 2.50E-07 2.82E-07 l.20E-O9 1.35E-09 Tc-99Sm 1.OOE-Ol 8.94E-Ol 8.94E-0l l.l0E-07 9.84E-08 5.70E-10 5.l0E-l0 Ru- 103 1.00E-01 l.18E+OO l.18E+O0 3.10E-07 3.65E-07 1.60E'-O9 l.88E-09 Ru-lO6 1.OOE-01 1.18E+00 1.18E+OO 3.70E-06 4.37E-06 1.80E-08 2.13E-08 Rh-10O3m* 3. 65E-07 1.88E-09 Rh-106* 4 .37E-O6 2. 13E-08 Ag-ll~rm 1.00E-01 l.1BE+O0 1.l8E+00 5.30E-08 6.26E-08 2.70E-10 3.19E-10 Te-129m i.00E-01 l.18E+00 l.18E-+00 7.80E-09 9.18E-O9 3.90E-II 4.59E-11 Te- 129 l.00E-01 7.32E-Ol 7.32E-01 2.20E-07 1.61E-07 1.10E-09 8.05E-10 Te-131m 1.00E-01 1.08E+O0 l.08E+00 5.40E-08 5.82E-08 2.70E-lO 2.9lE-10 Te -131 1.00E-O1 6.92E-01 6.92E-01 2.90E-08 2.01E-08 1.50E-10 l.04E-lO Te-132 l.00E-Ol l.14E+OO 1.14E+0O 6.60E-08 7.50E-08 3.30E-l0 3.75E-10 Ba-137m* 6. 04E-07 3. 02E-09 Ba -140 1.O0E-O1 1.17E+OO 1.17E+O0 5.20E-07 6.08E-07 2.60E-OS 3.04E-O9 La- 140 l.O0E-Ol 1.1OE+00 l.10E+O0 9.30E-07 1.02E-O6 4.60E-O9 5.O6E-O9 Ce- 141 l.00E-01 1.18E+O0 1.18E+O0 6.10E-O9 7.18E-09 3.10E-ll 3.65E-1l Ce-143 l.OOE-Ol l.08E+O0 l.08E+00 l.O0E-07 1.08E-07 5.10E-l0 5.53E-lO Ce- 144 1.O0E-01 1.18E+00 1.18E+00 1.60E-07 1.89E-07 8.20E-l0 9 69E-l0 Pr-143** 1. 08E-07 5. 53E-l0 Pr-144* 1. 89E-07 9.69E-l0 W-187 l.OOE-O1 1.06EE+OO l.06E+OO 8.70E-08 9.20E-08 4.40E-10 4.65E-10 Np-239 1.00E-01 1.12E+00 1.12E+00 8.40E-08 9.41E-08 4.20E-10 4.70E-10

Assumption 2 -daughter which is in secular equilibrium with parent see text for more detail

- See Assumption 3 It iS noted that these values are slightly less than those determined in the body of this calculation. Therefore either set of values can be used as the values in the body of the calculation are conservative.

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 26 of 49 Attachment

SCalculation No.

Subject:

Reactor WBNNAL3003 Coolant and Secondary Side Activities 1Rev:

in Accordance withWBN 005 {Plant: I Page:

ANSIIANS-1 26 984 8.1-1 Attachment A Telephone conversation between M. Berg and C. D. Thomas, Jr. of Yankee Atomic Electric Co. (co-author of ANSI/ANS-l8.1) on 3/13/87 Topic: The reason for the deletion of several isotopes from ANSI/A~NS-18.1-1984 isotope listing.

The report ANS N237-1976 (ANS -18.1) was based on scant data. The report ANSI/ANS-18.1-1984 was based on a larger set of data and therefore more accurate. Some isotopes on the former report were left out of the 18.1-1984 report. The 18.1-1984 report utilized a weighting factor to determine the- relative importance of an isotope. This factor was based on fission yield, number and energy of the gamma rays, and the efficiency of detection. Thus the isotopes Kr-83m,* Br-83, Br-85, Rb-86, Te-125m, Te-127m, Te-127, and 1-130 were deemed to be unimportant with respect to the other isotopes and were left out of the report..

Some of the isotopes were left out because they were in secular equilibrium with the parent, and it was felt by the authors that the parents were more important. The isotopes left out were Y-90, Rh-103m, Rh-l06, Ba-137m, and Pr-144.

The isotope Pr-143 was left off the list because it is a pure beta emitter.

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 2, Page 27 of 49 Attachment

Pages 28 to 49 of Attachment 2 contain a copy of American National Standard ANSI/ANS-18.1 -1984, "Radioactive Source Term for Normal Operation of Light Water Reactors," and have been redacted due to copyright restrictions.

ENCLOSURE I TENNESSEE VALLEY AUTHORITY WAiTIS BAR NUCLEAR PLANT UNIT I TVA Response to NRC Request for Additional Information Attachment 3 WBNTSRI100, Revision 12, "Design Releases to Show Compliance with 10OCFR20" CNL-15-216, Enclosure 1 CNL-5-1 6Encosue Attachment I 3, Page 1 of 35

NPG CALCULATION COVERSHEET I CTS UPDATE Page I RE*V 0 EDMS/RIMS NO. CTS YPE' EDMS TYPE: EDMS ACCESSION NO (N/A for REV, rn B26950110310 .Calculation CALCULATI ONS (NUCLEAR) I*l* ." 1 * ' r Caic

Title:

Design Releases to Show Compliance with 10CFR20

, Q g.Q PLANT BRNC NUMBER CR.RF NEW REV CALC ID .NUC WBN NTB WBNTSRI00 011 012 CrPDT OL 5 NO CTS CHANGES []

(Verifier and Approval Signatures Not Required)" l (For cale revision, CTS has been r~eviewed and no CT'S changes required)

UNIT (check one) _SYSTEMS UNIDS 00[,15r, 25V,35r' N/A N/A DCN.EDC.N/A APPLICABLE DEIGS]N DOCUMENT(S) CLASSIFICATION DCN 61599 N/A EO OUALITY SAFETY RELATED? UNVERIFI ED_ SPECIAL REOUIREMENTS AND/OR DESIGN OUTPUT SAR/TS and/or ISESI RELATED?_. (lf yes, QR = yes). ASSUMPTION LIMITING CONDITIONS? ATITACHMENT?. SARICoC AFFECTED Yes[] No[] Yesfl No[] Ycs[] No[] Yes[] No[] Yes[] NoD] YeslJ No[]

.CALCULATION NUMBER REOUESTOR PREPARING DISCIPLINE VERIFICATION METHOD NEW METHOD OF ANALYS.IS Name: N/A PHONE: N/A N Design Review 5] Yes 0] No PREPARER (PRINT NAME AND SIGN) DATE CHECKER (PRINT NAME ANDSIG DATE

Mehran A. Mohammadian ,4 *j< *.___,_ Alka__r___ ei*__________ * *** /#

VERIFIER (PRINT NAME AND SIGN/ ,*. DATE AP PhNMEAN IGN)" *'DATE STATEMENT OF PROBLEMIABSTRACT Problem:

Determine tihe design liquid and gaseous releases and show that these releases are less thans the 10CFR20 App.B Table 2 Effluent Concentration Limits (ECL).

Abstract:

The Standard Review Plan sections II .2.1II.2.c, II .2.1V.3, I 1.3.I11.2.b, and I1 .3.IV.3 require that the gaseous and liquid releases based on 1% failed fuel be within 10CFR20 App.B3 Table 2 limits. This calculation took thie expected gaseous releases from calculation T1534 and the expected liquid releasesafrom WBNTSR093. These releases were scaled to design (1% failed fuel) levels from ESAR data (and justified with Westinghouse WCAP-7664 RIl). The design.

release concentration for each isotope was divided by the I0CFR20 App.B Table 2 Effluent Conceistration Limit (ECL). The coneentration/ECL fraction was summed over all isotopes. Also analyzed was the case of liquid releases at design levels, except for iodines, which were limited to Technical Specification limits of 0.265 PtCi/gm 1-13 1equivalent. Thie results for each case can be found in the results section.

Thse I0CFR20 limits will not be exceeded for design ( 1% I'ail-ed fuel) releases for gas, or liquids when the Condensate Polisher Demiueralizer regeneration waste is processed by the mobile demineralizers. In the event that the long term release of this waste is projected to result in exceeding the I0CFR20 limits, the design of the plant allows the waste to be processed by the mobile demineralizer system. With the mobile demineralizer system processing the regeneration waste, the liquid design releases are below the I 0CFR20 limits. The release concentrations determined in this calculation arc not expected to occur because the design reactor coolant concentrations ex~eed the Technical Specification limits. With iodine limited to the Technical Specification limit of 0.265 I~tCi/gms 1-131 equivalent, the releases will be less than the IOCFR2O linsits. The design of the gas and liquid radwaste systems meet the requiremnents of I10CFR20, The presence of a Triliated Water Storage Tlank (TrWST) has no affibct on the tesulls or conluelsion~s of lhis calculation.

The case of operating with a trititum production core (TPC) is also analyzed in this calculation, Tlhe gaseous releases are based on thre design basis tritiums (1-1-3) source term. For liquid releases, the mnaximum allowable liquid concentration of H--3 is determined. The maxinmum allowable liquid concentration of H--3 that can be released to Ihe envirotnment without exceeding the 10CFB.20, App.B limit is made a special requirement/limfiting condition of the calculation: The maximum allowable liquid concentration of tritiunm (1!-3) released to the environnment is 3.26E-04 p.Ci/cc. 11-3 concentrations above 3.26E-04 pCilcc released to the environment may result in the 10CFR2O limit being exceeded.

...................... . .. Fl Nn l3* i . . . NUJMBER, FICHIE . . . .. St 1 ?VA 40532 Page 1of 2 NEDP-2-1 [10-31-201 lj CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue 1 3, Page 2 of 35

NPG CALCULATION COVERSHEET / CTS UPDATE Pae page l la CALC ID OR__G PLANT BRARRNCH NUMBER REV NUC WBN NTB WBNTSR100 01.2 f B__UILDING ROOM- ELEVATION COORD/AZIM FIRM NA N/A N/A N/A S&L CATEGORIES SRILC KEYWORDS (A-add, D-delete)

A*I1 A__ KEYWORD.

CROSS-REFERENCES (A-add, D-delete)

ACTION XREF XREF XREF XREF XRZEF

(/ CODE PLANT TYPE NUMBER REV A P WBN DCN 61599 CTS ONLY UPDATES:

Following are required only when making keyword/cross reference CTS updates and page I of form NEDP-2-1 is not included:

PREPARER (PRINT NAME AND SIGN) DATE CHECKER (PRINT NAME AND SIGN) DATE PREPARER PHONE NO. EDMS ACCESSION NO.

TVA 40532 Page 2 of 2 NEDP I [10-31-2011]

CNL-15-216, Enclosure I CNL- 5-1 6Encosue Attachment I 3, Page 3 of 35

WANTv, CALCULATION REODOF RVSO CALCULATION IDENTIFIER WBNT"SR-l00 ite Desig Reessto Show Complianc with 10CFRO "Revision DESCRIPTION OF REVISION No.

-1 Revision 1 wan performed because the expe-cted liquid releases from WBNTSR-093 have changed. All pages wer~e renumbered, but oniy the pages that had text changes are liated in the pages changed section below.

pages changed: 1-6, 10.12 .

pages added: 11 paged deleted: none.

2 Revision 2 was performed to incorporate gas releases fromn the alternate mode of operation of venting containment instead of purging. Also incorporated is a case with Technical Justification linits for jodines in the liquid (instead of 1% failed fuel design).

Pages added: 9.1 pages deleted: none,.

pages changed: 1-5, 7-12 RZ; 13 total pages 3 Revsiles 3 incr~porate the contnuous conta~nmet it~ered vet, to the annulus case. DCN D-501 68-A added aftiter to Iti vent line. The unlitered case detemined In revleion 2 wee retained for histoica puJPxses.

Pages added: 9.2 Pages deleted: none Pages changed: 1-5, 7-12, 9.1 4 Revision 4 was performed to add the alternate operation mode of direct Steam Genertor Blowdown release to the river without Condensate Polisher Dmemineraliser processing, end no Condensate Polisher Deminsralixer processing of the Condensate. Additionally, causes were also analyzed with no Coaling Tower Blowdewn dilution flow (20,000 gpm). This revision supports DON D-50502-A. All pages were renumbered. Pages with actual text changes are marked with revision bars.

Pages adtde&: new vr Pages del~ete n~one Pages changed: i (old cover), 2.23 .

_ P*~Rd 24 total pages _

5 RevisionS was perfornmed to add the use of a Tritium Production Core. This reison suport ED B50629A. The amount of gaseous release of trituium was changd in Tir-SN4 and is thus refct~edt in this calculation, Pages added: 1 Pages deleted" 0 Pages changed: 1b,2,3,7,9-14,16.18,22,23 R6:2fitotal pages B Revision Sproids additional results soited with RDC.E 060 29A in coju io n with RB. Thi .

revision avuluattes the unlikely event of a 1 or 2 TPBAR failure. Trtimn releases are the onbrvulUea effected by this faihure. Non-TPC tuitium va~lue wan revised per revision of WBN*WA93, which

. changed the power from 105% of a nominal power to 102% of nominal power. This calculation will need a 80.59 revews as it affects FSAR Tables 11.3-Ba and .Sb, and ll.2-4a and -4b..

" Pages addad: none

.Pages deleted: none.

Page, changed: 2,8,7,9-14,16-18,22,23*

__ _R6: 28 t~otahlu _ " ..

.lVA4070~ [01-19991 Page 1 of I NEOP-2-2 [01-08-19991 i" I Il I__ I i CNL-15-216, Enclosure 1 CNL-15-216, Enclosure I Attachment 3. Pace 4 of 35 I

TVAN CALCULATION RECORD OF REVISION Iv*-,.,,CL,1TON IDE.NTIRiER W N'ISR1O0 . .,.O-. :

Tite Da. ssa' "hew'.u1"0.

Revisioni DESCRIPTION OF REViSION 7 1Rud~on 7 slid ahi adat wsl a alhd o add~xhdn, Unita a plvuabnl*.i aniiai. is .pU*S Liu 2 based on die foO oyng : " **

.i An wwwflhc asm tiohaban added as pmtof tiis revisin vii* resumes nit I dulg chne stre appiln,le to Uiii T.h"e wweU aminptia onl appl to he th 2 poloau atIS cgalZaton.

efler to die Asmu~pion* section sl ils cialatiforhur ther hifo,mailon.

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revised ,-mumassaryto anure adequany.

Th~"eufilot unit 2 opreios ou* Unit I mnargns end ill hnpacton an Ullknal Fest SLI *riraw of 98 "FII been reviwud wit, no hi~ac.

RVIONIC OP ThE CALOULA'1N.

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Potl snhr O ease n h erviio h ""teI~dmia29 ."

VA 40709 112.20001 Page I of1 N!~R~2-2 112-04-20001 CNL-15-216, Enclosure 1 in T CNL-15-216, Enclosure 1 Attachment 3, Page 5 of 35

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ToatmoI~ pi e~Iniddn falinbs 4 ae !~ti CNL-15-216,- Enclosure 1 *Attachment 3, Page 6 of 35 IN- -16.Ecoue II AtchetIPae6 f3

NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIE'R:' WBNTSRI00 Page 2c Title Design Releases to Show Compliance~with 10CFR20 LRevision NO.

DESCRIPTION OFRVSO 0o9 RevisiOn 9 of this calculation remove the remaining UJVA Inwas created to Incorporate the correctve actions of PER 388968, and support of duel unit operation.

EDOR 54149 has been issued to incorporate the design changes of DCN 60502 for Unit 2.

Therefore, Assumption 2 (UVA) has been removed by this revision.

This revision updated the liquid releases based on revision of calculation WBNTSR093 (Ref. 3) for PER 386966. The tables for the liquid releases have been entirely updated and I*ncorporate the single unit and dual unit operations tables into the main body. Therefore. Appendix A has been modified to remove the liquid release tables.

PER 386566 also identffies the need to make this calculation cfassifcation F'O¶. A Design Output Attachment has been added (Attachment 1).

Tlhe coversheet has been corrected to show the calculation is not.Safety Related. This is consistent with Revision 7 and earlier revisions of the calculation.

Successors WBNAPS3083 and WBNTSR024 are impacted by this'revlsion. The updates to these succ.essor documents are tracked by WITEL PL-1 1-3841. All other successors have been reviewed and are not Impacted by this revision. .

Affected engineering judgments and assumptions were reviewed and (1) ware found to be adequate, or (2) were revised as necessary to ensure adequacy.

Ultimate heat sink (UIHS) temperature was not used as an Input to the calculation analyses. Therefore, existing calculation results will not be affected by changing the UHS technical specification temperature.

The effect of Unit 2Idual unit operation on Unit 1 margins has been reviewed with no impact.

Unit I FSAR, Sectlon 1"1.2, Table 11l.2- 4a and 11.2- 4b is being revised via NLDP-5 form as part of EDC

,*-A PP-10 form Is being generated to revise Unit 2 FSAR Section 1 1.2, Table 11.2-5e through "11.2-SdIn the next amendment of the Unit 2 Living FSAR.

Reviewer' ,

Pages Added: is (Rev. 9 CCRIS), 2c, 3b, 12a, t2b, 138, 14a, 15a, 16a, 17a, 188, l9a, 20e, Attachment 1 (1 page)

Pages Revised/Replaced: i .(Rev. 9 CS), i (Rev. 8 CS), 4, 7, 12, 13, 14, 15,1t6,17, 18,. 19, 20, 22, 23, A-1 Pages Deleted: 7a, A-5 through A-12 Total number of pages In this revision Including Attachments:

44 pages (Rev. 8)4÷ 14 pages (Rev,. 9) - 9 page (Rev. 9) = 49 Ap:*endix A: 4 pages Attachment 1: 1 page

_____This page added by Revision 009.-*

.TA 40709 [10-2008] Page I of 2 NEDP-2-2 [10-20-2008]

("Ikll _1*_04R I::nPInellr* "1 Aff*nhm*nt

P*ae 7 of 35 1Jl ln.rir ,'i Atnh n , _=o 7. of.3.

Pagle 2d NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIRIER WBNTSR-100 Title " Design Releases to Show Compliance with 10CFR2O Revision DESCRIPTION OF REVISION No.

010 Revision 10 of this calculation was created to incorporate updates to predecessor calculations

- Calculation WBNTSR093 (ref.3) was revised as the result of PER 546323. The liquid releases used by this calculation were impacted by this revision and therefore require update. The appropriate tables have been updated.

- Dual unit gaseous release tables previously contained in Appendix A have been moved into the "

main body. As a result Appendix A has been deleted.

- The entire calculation is renumbered Successor calculation WBNTSR024 and WBNAPS3O83 are impacted by this revision. The updates to these successor documents are Iracked by PER 546323. All other successors have been reviewed and are not impacted by this revision.

Affected engineering judgments and assumptions were reviewed and (1) were found to be adequate, or (2) were revised as necessary to ensure adequacy.

Ultimate heat sink (U.HS) temperature was not used as an input to the calculation analyses. Therefore.

existing calculation results will not be affected by changing the UHS technical specification temperature.

The affect of Unit 2/dual unit operation on Unit 1 margins has been reviewed with no impact.

Unit 1, FSAR, Section. 11.2, Table I 1.2.-4a and 11l.2-4b is being revised via NLDP-5 form as part of EDC 56202. A PP-l0 form is being generated to revise Unit 2 FSAR section 11.2, Table 11.2-5a through 11.2-5d in the next amendment of the Unit 2 Living FSAR Reviewer: D.G.Fickev Pages added: 2d, 24, 25, 3a, 3b Pages revised/replaced: 1, la, 3-23, Attachment I Pages deleted: 1, ia, ii, iii, Ib, 3a, 3h, 12a, 13a, 12b, 13a, 14a, 15a, 16a, 17a, 18a, 19a, 20a, Appendix A (4 pages)

Tolal~number pages in this revision:

49 pages (Rev.9)+5 pages(Rev.l10)-21 page(Rev.l10)=33 pages

_____Attachment 1:1I page..

11 DCN 59397 is installing a Tritiated Water Storage Tank (TWST). The TWST does not increase the amount of the releases, therefor~e the addition of the tank does not affect the results of this calculation.

Successor calculations are not impacted by this revision, as there are no changes to the results. .The FSAR is not impacted by this revision since the results did not change. The Technical Specifications are not affected by this revision. The attached Design Output determined in revision 10 has not changed due to this revision. Reviewed by Marc Berg Pages added: none Pages deleted: none

_____Pages Changed: 1, Ia, 2d, 3, 5, 25 Total Number of Pages: 33 TVA 40709 [10-2008] Page I or NEDP-2-2 110-20-20081

-- Attachment 3,,° Page v 8 of 35 CNL-15-216. Enclosure I Attachment 3. Pane 8 of 35

Paoe 2e NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER WBNTSRI 00 Title Design Releases to Show Compliance with 10C'FR20 Revision DESCRIPTION OF REVISION 12 Revision 12 is performed to reflect a change regarding the tritiumpermeation rate for a TPBAR in the tritium production core (TPC) configuration, which affects predecessor calculations WBNNAL3 003,

"/l534,and WBNTSR093. Gaseous releases of tritium (H--3) are based on~the design basis average permeation rate and 2500 TPBARs. For liquid releases of H-3, the maximum liquid concentration of H-3 that can be tolerated is calculated. The calculated maximum allowable 11-3 concentration that can be released to the environment without exceeding the 10CFR20 App.B limit is made a special requirement/]imiting condition of the calculation.

The TPC 1 rod and 2 rod failure cases were removed from this revision because they. do not reflect normal plant operation.

Additionally, C-14 expected releases were revised in predecessor calculation TI534 Revision 10; therefore, C-14 values were updated in Revision 12. Tables 2, 3, and 4 are affected by this change, but the conclusions with regard to gaseous releases are not affected.

Typographical and editorial corrections were made on pages 8 (4.25E+12 to 4.25E-12), 9 (2.07E-04 to 2.07E+04), and 24 (grammatical corrections). A reference to the 2.5 gtCi/gm 1-131 concentration in the reactor coolant was corrected on pg. 24 (ref.3 to Table 1). An error on Table 5 was corrected, where the concentration colunmn had units of jiCi/gm. The units on the concentration column were changed to*

/MCi/cc.

An error was identified in Tables 2, 3, and 4 in the dual unit H--3 (TPC) values where the non-TPC H-3 from the second operating unit was not counted in the calculation of the dual unit operation C/ECL column. The tables have been updated with the corrected dual unit H-3 (TPC) values that account for the additional non-TPC H1-3 from the second operating unit.

Pages replaced in Reviision 12 contain an updated page header for Revision 12. All unaffected pages retain their original headers.

CTS was reviewed for successor documents, and calculation WBNAPS3 128 is affected. Other successor documents reviewed were not affected by the changes made in WBNTSR100 Revision 12 either because they did not utilize TPC values or values used were not revised.

See DCN 61599 for SAR/Tech Spec determination.

Pages added: 2e, 23a Pages deleted: 3b Pages replaced: 1, la, 3, 3a, 4, 5, 7-12, 14, 16, 17, 19, 24, 25, Attachment A

Revision 12: 34 total pages Attachment A - NPG Calculation Design Output (1 page)

TVA 40709 [10-2008] Page I of I NEDP-2-2 [1 0-20-2008]

CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 3, Page 9 of 35

Pagle 3 NPG CALCULATION VERIFICATION FORM Calculation Identifier WBNTSR!00 Revision 12 Method of verification used:

1. Design Review [

3. Qualification Test []Aleksandar Milicevic Comments:.

! have reviewed Calculation WBNTSRIO0 Revision 12 and have found the portions of the calculation revised to be technically adequate. In conducting the verification of the portions ofWBNTSRI00 revised as part of Revision 12, ]

have reviewed the inputs, computations, transcription, and results, which I have found to be complete and accurate.

All comments have been resolved with the preparer.

TVA 40533 [1 0-2008] Page 1 of 1 NEDP-2..4 [10-20-2008]

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 3, Page 10 of 35 Attachment

Pagle 3a

-NPG COMPUTER INPUT FILE STORAGE INFORMATION SHEET Document WBNTSR100 IRev. 12 IPlant: WBN

Subject:

Design Releases to Show Compliance with 10CFR20 E] Electronic stor'age of the input files for this calculation is not required. Comments:

[] Input files for this calculation have been stored electronically and sufficient identifying information is provided below for each input file. (Any retrieved file requires re-verification of its contents before use.)

The Excel spreadsheet for R12 is permanently stored in FILEKEEPER # 321871.

LI MicroficheleFiche TVWA40535 [10-2008] Page 1 of I NEDP-2-6 El10-20-2008]

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 3, Page 11 of 35 Attachment

Page 4 NPG CALCULATION TABLE OF CONTENTS Calculation Identifier: WBNTSR1 00 Revision: 12 TABLE OF CONTENTS SECTION TITLE PAGE Calculation Coversheet/CTS Update 1, Ia Calculation Record of Revision 2, 2a -2e Calculation Design Verification Form 3 Computerlnput File Storage Information Sheet -3a Calculation Table of Contents 4 Purpose 5 Introduction 5 Assumptions 5 Special Requirements/Limiting Conditions 5 Calculations 6 I. Expected to Design Source Term and Design to Technical Specification Scaling 6 II. Gaseous Releases 7 III. Liquid Releases 11 Results 24 Discussi(on and Conclusion 24 References 25 AUtachment A: NPG C~alculation Design Output (1 page) 26 TVA 40710 [10-2008] Page 1 of 1I NEDP-2-3 [10-20-2)008]

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 3, Page 12 of 35 Attachment

Subject:

Design Releases to Show Cornpliance with 10CFR20 Purpose The purpose of this calculation is to determine if 1% failed fuel (design fuel damage) will result in effluent releases exceeding 10CFR20 App.B Table 2 limits.

Introduction The Standard Review Plan sections 11.2.Ill.2.c, 11.2.1V.3, 11.3.111.2.b, and 11l.3.IV.3 (ref.I) require that the gaseous and liquid releases based on 1% failed fuel be within L0CFR20 App.B Table 2 limits. This calculation takes the expected gaseous releases from TI-534 (ref.2) and the expected liquid releases from WBNTSR093 (ref.3) and scales these releases to design (1% failed fuel) levels taken from FSAR data. Not all isotopes are used from these sources as many are insignificant and short-lived. The design release concentration for each isotope is divided by the I0CFR20 App.B Table 2 Effluent Concentration Limit (ECL). The concentrationlECL fraction is summed over all isotopes. The 10CFR20 limit requires this sum to be less than unity. Note that the Standard Review Plan requires the design of the plant to have the conceritration/ECL fraction to be less than unity. In actual practice, the ODCM allows the concentrationlECL fraction to be less than 10. Since this calculation is to show the adequacy of the plant design, the acceptance criterion is for the concentrationlECL fraction to be less than unity.

Since the design (1% failed fuel) reactor coolant inventory exceeds the technical specification equivalent of 0.265 uCi/gm 1-131 dose equivalent, a liquid release case utilizing the technical specification limit for iodine (and 1% failed fuel for the remaining isotopes) is performed. This will show that the plant will meet 10OCFR20 limits Under maximum allowable operating conditions, because the design inventories cannot be experienced in the plant.

Three cases for gaseous release are performed based on the TI-534 modes of operation. One is for the containment portion of the gaseous effluents released via filtered purging. The other cases assume the containment is vented to the annulus, and then released via the Auxiliary Building Vent.

DCN 59397 is installing a Tritiated Water Storage Tank (TWST). This tank will receive radioactive liquid waste processed by the Mobile Demineralizers. The purpose of the TWST is to temporarily hold processed liquids until more favorable release conditions occur. The TWST does not process any liquid itself, although additional processing may occur by routing stored liquid back to the Mobile Demineralizers at the discretion of the plant. Therefore, the TWST will not increase any releases, and could potentially reduce the total amount released. Therefore, the TWST is not considered further in this calculation.

The concentration/ECb fractions due to implementing a tritium production core (TPC) in Unit 1 are determined in this calculation. Gaseous effluent releases of tritium (H1-3) are based on the design basis source term. For liquid releases of H1-3, the maximum allowable liquid concentration of 11-3 is back-calculated based on the available margin for the liquid release scenarios.

Assumptions

1. It is assumed that the ratio of design to expected reactor coolant concentrations can be applied to the expected releases to obtain design releases.

Technical Justification: All releases are ultimately based on the reactor coolant isotopic concentrations. It follows that any change in the reactor coolant concentrations will result in a proportional change in r-eleases. Therefore, scaling from expected to design releases based on a design/expected reactor coolant ratio is valid.

2. For a TPC, it is assumed that the design basis source term (2500 tritium-producing burnable absorber rods (TtPBARs) each with an average H-3 permeation rate of 10 Ci/TPBAR/yr) is used for the gaseous releases.

Technical Justification: Gaseous releases follow the tritium production rate, which is linear, So the average H-3 permeation rate is appropriate. 2500 TPBARs at 10 CiITPBAR/yr is assumed because it is yields the largest, most conservative-source terms in ref.5.

Special Requirements/Limiting Conditions The maximum allowable liquid concentration of tritium (11-3) released to the environment is 3.26E-04 pCi/cc.

Ht-3 concentrations above 3.26E-04 tCi/cc released to the environment may result in the 1UCFR2O limit being exceeded.

CNL-1 5-216, Enclosure 1 Attachment 3, Page 13 of 35

NI' Calculation sheet Calculation sheet SI Pag.:6~

ISublect: Design Releases to Show Compliunc. with 10CFR20 *

' i

, II

Calculations L Expected to Design Source Term and Design to Technical Specification Scaling The normal gsa and liquid releases (reL2,3) are based on expected reactor coolant isotopic.

concentration (ref.5). The design (1% failed fuel) concentrations are taken from reference 7. The iodine Technical Specification limit of 0.265 pC~i/gm 1-13 1 dose equivalent is Calculated by multiplying the equivalence factor from reL 14 (11.348) and 0.265 p~i/g by the expected concentration. The ratio of design to expected concentrations gives a scalin factor to establish the design releaese. The technica specification to design concentration gives a scalin factor to establish the maximum iodine releases possible. The scaling factors are:

Table 1: Expected to Design and Tech. Spec. to Design Scaling.

Tecd. Spec, RCS Cnt, Rs Cnx 0.265 lICUgm Ratio Nud~de Kr-85m 1.712-01 2.102400 12.28 Kr-85 2.66E-01 LO$Ee0O 33.08 1.61 E-01 1.202,00 .7.45 Kr-U 12.33 X.-131m 6.542-01 1902.00 2.91

03-133m *7.17E-02 3.1064,00 43.24 xa-t33 2.53E6-00 2.81E2402 111.07 Xa-138m 1.392-01 7.00-1 6.04 9.042-01 6.000 6.07 Xe-138 1.296-0l 7.00E.01 6.43 1.722.2 4.30E.02 2.50 1-131 4.7"2-02 2.50-400 ,52.41 0.143 0.057 1-182 2.28E-01 9.002-01 4.00 O.677 0.762 1-133 1.49E.O1 4.00E+00 26.85 0.448 0.112 1-134 3.84E.01 6.005-01 1.85 1.005 1.824 2.78E..01 2.20644X) 7.91 0.836 0.380 Cs-136 2.045-,01 3.70E'*00 18.14 cs.134 7.392-03 3.005-01 40.60 9.085-04 1.50E-O1 168.20 ce-137 9.796-.03 1.,S96400 153.22 Cr-Si 3 .26-3 9.60E-04 0.29 1.696-03 7.90E-04 0.47 3.18E,.04 1.10E..53 3.48 4.842-03 2.606)-02 5.37 6,56..04 7.701.04 1.38 1.472-04 .3.30E-03 22.45 Sr-89 1.282-05 1.701-04 13,49 1.02E,03 1.902-03 1.86 Y-S0 1.26F,,05 2.*00-04 15.57

$.47E-06 6.102.03 1115.17 Zr-OS 4.102-04 7.002-04 1.71 2.95E-04 6.90E-04 2.34 6.752-03 5.30600 785.19 1.796-03 2.605-01 145.28 1.372-0*2 4.302-03 0.31 La-140 2.645-02 1.50E.-os O.06 Ce-144. 4.21E-.03 3.402-04 0.08 Pr-14 4.212.03 3.405-04 0.08.

ThIs Ipeg mpc by Revision 010.

[

I IIII CNL-15-216, Enclosure 1 Attachment 3, Page 14 of 35 CN-1-1i EcosrIIAtahmn 3, Pae 4of3

CalculationNo. WBNTSRX00 Rev:12 IPlant:WB iPage: 7

Subject:

Design Releases to Show Cornpliance with 10OCFR20 II, Gaseous Releases The expected annual gas release is taken from TI-534 (ref.2). TI-534 presents three different modes of operation: one with containment purge, and others with containment venting to the annulus (one with no filtration, and the other with continuous venting with filters). All three options are analyzed here. The expected release is multiplied by the scaling factor determined earlier to obtain the design releases. To determine the average design concentration, this release is multiplied by the site boundary X/Q value of 1.09E-5 sec/mn3 (ref.4). To correct the units, the formula used is:

[htCi/cc] = [Ci/yr]*X/Q~s/mn3]*[l1(uCi/ee)/(Ci/rm 3)/(60 s/rain*60 min/hr*24 hr/day*365 day/yr)]

The design concentration release of each isotope is then divided by the 10OCFR20 App.B Table 2 Effluent Concentration Limit (ECL). This fraction is then summed over all isotopes. The acceptance criteria is for this sum to be less than unity.

Using a TPC with 2500 tritium-preducing burnable absorber rods (TPBARs) with an average tritium permeation rate of 10 CiiTPBAR~yr, referred to as the design basis source term (Assumption 2), the additional amount of tritium, based on the methodology established in TI-534, is (10 Ci/TPBAR/yr

2500 TPBARs) 25,000 Ci/yr. Of this amount, 10% is available for gaseous release, so the additional amount of gaseous tritium for a TPC is (t0% ot'25,000 Ci/yr) 2500 Cl/yr. Note that the total value with the TPC substitutes the H--3 (TPC) value for the non-TPC H-3. The value used in the table for H-3 (TPC) consists of the tritium value activity above added to the non-TPC's expected retease value.

CNL-15-2116, Enclosure 1 CNL-5-21, Enlosue I 3, Page 15 of 35 Attachment

Calcuiation No. WBNTSR1O00]

I

Subject:

Design Releases to Show Compliance with 10CFR20ORev: 12 Plant: WEN Page: 8 Table 2: Gas Releases, Containment Purge option Sxp. Single Unit Dual Unit Rel. Operation Operation Des/Sxp Design Design 10CFR2 0 Nucl ide (Ci/yr) Ratio (ci/yr) (iCi/cc) (SCL, iaCi/cc) C/ECL C/ECL Kr-B8Sm 2 .SBE+01 1.2 35+01 3. 17E+02 1.105-10 1E-7 1.105-03 2.195-03 Kr -8B5 6. 99E+02 3. 315+01 2.315+04 7.99E-09 7E -7 1.145-02 2.285-02 Kr -8B7 1.625+01 7.455+00 1.2 1E+02 4.1B5-11 2E-B 2.055-03 4.185-03 Kr-8B8 3. 85E+01 1. 23E+01 4.755+02 1.645-10 BE-9 1.825-02 3.655-02 Xe -131is 1.19E+03 2. 91E+00 3.455+03 1.19BE-O9 25-6 5.975-04 1.195-03 Xe- 13 3m 4.88BE+O1 4.325+01 2. 11E+03 7.29E-10 6E-7 1.215-03 2.435-03 Xe- 133 3.20E+03 1.115+02 3.55E+05 1.235-07 55-7 2.465-01 4.915-01 Xe- 135Sm B. 52E+00 5.045+00 4.29E+01 1. 4B5-11 45-B 3.715-04 7.425-04 Xe- 13 5 1. 85E+02 6. 97E+00 1.295+03 4.46E-10 6.385-03 1.285-02 Xe- 13 8 7.66E+00 5 .43E+00 4. 16E+01 1.445-11 7.195-04 1.445-03 Sr-B4 5. 07E-02 2.50E+00 1.275-01 4. 38E-14 BE-B 5.485-07 1.10E-O6 1-131 1. 535-01 S .245+01 B. 035+00 2. 775-12 2E-10 1.395-02 2.795-02 1-132 6. 755-01 4.005+00 2.705+00 9.335-13 25-B 4.675-05 9.345-05 1-133 4. 58E-01 2. 69E+01 1.235+01 4.255-12 15-B 4.255-03 8.515-03 1-134 1. 08E+00 1 .655+00 1.7 8E+00 6. 14E-13 1.02E-OS 2.055-05 1-135 8.45E-01 7.91E+00 6.69E÷00 2. 31E-12 3 *85E-04 7.705-04 Cs -134 2.275-03 4.06E+01 9.20E-02 3.18E-14 25-10 1.595-04 3.185-04 Cs -136 B. 01E-05 1.65E+02 1.325-02 4.575-15 95-10 5.085-06 1.02E-O5 Cs -137 3.485-03 1.535+02 5 .33E-01 1.845-13 2E-10 9.205-04 1.845-03 Cr -51 5. 92E-04" 2. 90E-01 1.735-04 5. 96E-17 1.995-09 3.985-09 Mn-54 4.315-04 4. 70E-01 2 .03E-04 7.015-17 15-9 7.015-08 1.405-07 Fe -59 7.70E-05 3 .4BE+0O 2.68EB-O4 9.275-17 5E-10 1.855-07 3.715-07 Co-5B 2.325-02 5.3 7E+00 1.245-01 4. 30E-14 4.305-05 8.605-OS Co-6O B. 74E-03 1.3 8E+00 1.215-02 4. 17E-15 8.335-05 1.675-04 Sr-B89 2.985-03 2.255+01 6.69E-02 2.315-14 15-9 2.315-05 4. 63E-05 Sr-SO 1. 14E-03 1.355+01 1.54E-02 S. 335-15 65-12 8.885-04 1.78E-03.

Zr-95 1.005-03 1.715+00 1.71E-03 5.925-16 4 5-10 1.485-06 2.965-06 Nb -95 2,.45E-03 2.345+00 5.735-03 1.985-15 2E-9 9.905-07 1.985-06 Ba- 140 4.00OE-04 3.10E-01 1.265-04 4.34E-17 25-9 2.175-08 4.345-08 11-3 1.395+02 1.005+00 1.3 9E÷02 4.80E-11 1E-7 4.815-04 9.625-04 11H-3 (TPC) 2. 64E+03 1.005+00 2. 64E+03 9.125-10 15-7 9.12E-03 9.605-03 C-14 1. 12E+01 1. 00E+00 1. 12E÷01 3.875-12 35-9 1.295-03 2.585-03 J Ar -41 3.40E+01 1. 00E+00 3. 40E+01 1.1lBE-li 15-8 1.185-03 2.355-03 total 3.115-01 6.235-01 Itotal ITPC) 3.205-01 6.325-01 Note: The "Dual Unit Operation" column in the shove calculation considers dual unit operation. Based on the evaluation done for Revision 7, the per unit concentrations are the satee for bath units. Therefore, the last column is twice the preceding column.

Note: Dual unit operations consider only Unit I with the TPC.

Note: Dual unit operation H-3 (TPC) value is the sum of normal H-3 from two units and the TPC-speeiflc H-3 from one unit.

CNL-15-216, Enclosure I Attachment 3, Page 16 of 35

Calculation No. WBNTSR100 IRev: 12 IPlant: WBN 1Page: 9

Subject:

Design Releases to Show Cornpliance with 1 OCFR2O

Table 3: Gas Releases, Unfiltered Containment Vent Optidn Single Unit Dual Unit Kxp. Rel. Des/Erp Design Design 10CFR20 Operation Operation Nucilide (Ci/yr) Ratio (Ci/yr) (pci/cc) (ECL, pCi/cc) C/KCL *C/KCL Kr-85 m 2.16EK+01 1.23E+01 2.65EK+02 9. 17E-11 1E-7 S.17E-04 1. 83K-03 Kr-8S 6. 27E+02 3 .31E+01 2. 07E+04 7,17K-OS 7E-7 1.02E-02 2.,OSK-02 Kr- 87 8. 53E+00 7. 45E+OO 6. 365+01 2 .20K-lI 2E-8 1. 1OE-03 2.20EK-03 Kr- 88 2.6S6E+O1 1. 23 E+O1 3. 28E+02 1. 13E-10 SE- 9 1.26E-02 2 .52E-02 Xe-131m 1. 1SK+03 2.9S1K+00 3. 34E+03 1.15E-09" 25-S S. 77E-04 1. 158-03 Xe- 13 3m 5. 81E+01 4. 32K+01 2. 51E+03 8. 68E-10 58-7 1. 45E-03 2 .89K-03 Xe- 133 3. 37E+03 1.IIE8+02 3 .74E+0S 1.29E-07 5E-7 2.59EK-Ol 5.17E-01,-

Xe- 135Sm 4. 86E+00 S. 048+00 2.45E+01 8.46E-12 4K-8 2.12E-04 4 .24K-04 Xe- 13 5 1.99SE+02 6.957E+/-00 1.39E5+03 4. 79E-10 7K-8 S6.85E-03 1. 37K-02 Xe- 138 4 .48E+O0 S5.43E+O0 2.43E+01 8 .40E-12 2E-8 4. 20E-04 S8.40K-04 Br-84 S. 07E-02 2. 50E+00 1. 27K-01 4 .38E-14 8K-8 5 .O0K-07 1.00K-OS 1-131 1.657K-01 5.24E+-01 8. 75E+00 3. 03K-12 2E-10 1.51K-02 3.035-02 1-132 6. 75K-01 4. 00E+00 2. 70E+O0 S. 33K-13 2K-S 4.67K-O5 9.34K-OS 1-133 4. 70E-Ol 2.6595+01 1. 26E+01 4.36E-12 1E-9 4. 36E-03 8. 72E-03 1-134 1. 08E+00 1.655E+0O 1. 78E+00 S. 14E-13 SE-8S 1.02K-OS 2.04K-OS 1-135 8. 52K-Ol 7. 91E+00 S. 74E+O0 2.338-12 6E-9 3.588-04 7.77E-04 Cs- 134 4. 74E-03 4. O6E+01 1. 92E-Cl 6.65K-14 2E-10 3 .33K-04 6.65E-04 Cs- 136 3 .25K-03 1.65EK÷02 S. 378-01 1.86E-13 SE-10 2.O6E-04 4.12E-04 Cs- 137 S.92K-03 l.53E+02 1. 37K+0O 4. 72E-13 2E-10 2.36E-03 4. 72E-03 Cr-si 9. 70E-03 2 .SOK-01 2. 83E-03 S. 77K-iS 3E-8 0 .00E+00 Mn-54 S5.6ES-03 4 .70E-01 2.687E-03 9.2 3E-16 1K-9 S. 00E-07 Fe-5 9 2. 75E-03 3.48E5+00 9.57E-03 3.31K-iS SE-10 6.60K-OS 1.32K-OS Co-5 8 4 .79K-02 S5.37E+00 2. 57E-01 8.89E-14 lE-S 8.89K-OS 1.78K-04 Co-SO 1.13E-02 1,.38E+OO 1. 56E-02 5.39E-15 5E-Il 1.08E-04 Sr-89 1.59E-02 2 .25K+Ol 3. 57E-01 1.23E-13 1K-S 1. 23K-04 2 .47K-04 Sr-SO 6.29EK-03 1. 35E+01 8.49E5-02 2. 935-14 *6E-12 4.89EK-03 S. 78K-03 Zr-955 1.00OE-03 1. 71E+00 1. 71E-03 S. 908-16 4E-10 1.50K-OS 3.00K-OS Nb-955 4. 23K-03 2. 34E+00 9. 89E-03 3 .42K-iS 2K-S 1.70E-OS 3.40K-OS Ba- 140 4 .00K-04 3.10E-01 1. 26E -04 4. 34E-17 2E-S 0 .00E+00 O0.0OK+0O 5-3 1.35E+02 1.00K+00 1.39E+02 4.80E-11 1E-7 4.81E-04 9.62K-04 5-3 (TPC) 2.64E+03 l.O0E+00 2.64E+03 9.12E-10 1E-7 5.12E-03 9.SOK-0 3 C-14 1.12E+01 1.008+00 1.12E+O1 3.875-12 3K-S 1.29E-03 2.58K-0 3 Ar-41 3.40K+01 1.00E+00 3.40K+/-0l 1.18K-Il 1K-B 1.18K-03 2.35K-O03 total 3.24E-01 6.48K-01, total (TPC) 3.33K-01 6.57E-01 I Note: The "Dual Unit Operation" column in the above calculation considers dual unit operation. Based on the evaluation done for Revision 7, the per unit concentrations are the same for both units. Therefore, the last column is twice the preceding column.

Note: Dual unit operations consider only Unit 1 with the TPC.

Note: Dual unit operation H-3 (TPC) value is the sum of normal 11-3 from two units and the TPC-apecifle 11-3 from one unit CNL-15-216, Enclosure 1 Attachment 3, Page 17 of 35

SReleases to Show Compliance with 10CFR2O Table 4: Gas Releases, Continuous Filtered Containment Vent Option Single Unit Dual Unit Operation Operation Exp. Ral. Dea/Exp Design Design 10CFR2O (Ci/yr) Ratio (Ci/yr) puCi/cc) (SCL, pCi/cc)" C/ECL C/ECL Nuclide Kr-85Sm 9.48E+00 1.235+01 1.165+02 4.025-il 15-7 4.025-04 8.05E-04 Kr-85 6. 78E+02 3. 31E+01 2.245+04 7.75E-09 7E-7 1.115-02 2.215-02 Kr -87 5.81E+00 7.45E+00 4.335+01 1.SOE-il 25-8 7.485-04 1.505-03 Kr- 88 1.325+01 1.235+01 1.638+02 S.63E-11 95-9 6.25E-03 1.255-02 Xe -131is 1.0 95+03 2.91E+00 .3.185+03 1.105-09 25-6 5.495-04 1.105-03 Xe- 13 3m 4.315+01 4.32E+01 1.865+03 6.445-10 65-7 1.075-03 2.155-03 Xe- 133 2.905+03 1.115+02 3. 225+05 1.115-07 5E-7 2.235-01 4.455-01 Xe-135a 4.68E+00 5. 045+00 2.365+01 8.15E-12 2.045-04 4.085-04 Xe-135 S. 85E+01 6.975+00 6.19E+02 2.145-10 3.065-03 6.115-03 Xe- 138 4 .34E+O0 5.435+00 2.365+01 8.155-12 2E-8 4.075-04 8.155-04 Br-8 4 5.075-02 2.505+00 1.275-01 4.385-14 SE-B 5.00E-07 1.005-06 1-131 1.53E-01 5.245+01 8.005+00 2.775-12 2E-10 1'.385-02 2.775-02 1-132 6.735-01 4.005+00 2.695+00 9.3 0E-13 4.6S5E-5 9.305-O5 1-133 4.57E-01 2.695+01 1. 2 3 +01 4.245-12 15-9 4.245-03 8.49E-03 1-134 1.0 7E+00 1.655+00 1.775+00 6.105-13 1.025-05 2.04E-O5 1-135 8.42E-01 7.915+00 6.665+00 2.305-12 65-9 3.845-04 7.675-04 Cs-134 2.275-03 4.06E+01 9.20E-02 3.185-14 2E-10 1.595-04 3.185-04 Cs- 136 8.01E-O05 1.655+02 1.325-02 4.575-15 95-10 5.105-06 1.02E-05 Cs-137 3 .4BE-03 1.535+02 5.335-01 i.54E-13 25-10 9.205-04 1.845-03 Cr-5i 5.925-04 2.905-01 1.735-04 5.965-17 0.005+00 0.005+00 Mn- 54 4.31E-04 4.705-01 2.03E-04 7. 01E-17 15-9 1.005-07 2.005-07 Fe -59 7.705-OS 3.485+00 2.685-04 9.275-17 55-10 2.005-07 4.00E-07 Co-SB 2.325-02 5.375+00 1.245-01 4.305-14 15-9 4.30E-OS 8.605-O5 Co-6O S. 74E-03 1.38E+00 1.2 15-02 4.175-iS SE-lI B.33E-05 1.675-04 Sr-8B 2. 98E-03 2.255+01 6.695-02 2.315-14 15-9 2.31E-05 4.625-05 Sr-SO 1.145-03 1.355+01 1.545-02 5.335-15 65-12 8.885-04 1.785-03 Zr-95 1.00E-03 1.715+00 1.715-03 5.92E-16 45-10 1.505-06 3.005-06 Nb -95 2.455-03 2.345+00 5.735-03 1.98K-is 25-9 1.005-06 2.005-06 Ba- 14 0 4.005-04 3.105-01 1.265-04 4.345-17 2E-9 0.005+00 0.005+00 1--3 1.3 95+02 1.005+00 1.395+02 4.805-11 15-7 4.815-04 . 9.625-04 5 -3 (TPC) 2.645+03 1.005+00 2.645+03 9.125-10 15-7 9.125-03 9.605-03 I C-i4 1. 12E+01 1.005+00 1.125+01 3.875-12 35-9 1.295-03 2.585-03 Ar- 41 3.40E+01 1.005+00 3.405+01 1.185-11 15-8 1.1S8*03 2.355-03 total 2.705-01 5.405-01 I total (TPC) 2.795-01 5.495-01 I I

Note: The "Dual Unit Operation" column in the above calculation considers dual unit operation. Based on the evaluation done for Revision 7, the per unit concentrations are the same for both units. Therefore, the last column is twice the preceding column.

Note: Dual unit operations consider only Unit 1 with TPC.

Note: Dual unit operation H1-3(TPC) value is the sum of nonnal 1--3 from two units and the TP'C-specifre H-3 from one unit.

CNL-15-216, Enclosure 1 Attachment 3, Page 18 of 35

Caclto o BTR0 1 ln:W N Pg:1 IV

Subject:

Design Releases to Show Compliance with 10CFR20 II.Liquid Releases' The expected annual liquid release is taken from WBNTSR.093 (ref.3). The expected release is multiplied by the scaling factor determined earlier to obtain the design releases. To determine the average design concentration, the release is divided by volume released and the dilution flow (the minimum cooling tower blowdown flow = 20,000 gpm = 2.88E7 gal/day). The volume released is sum of all sources from WBNTSR.093 =16141.654 gal/day. Note: the WBNTSR.093 flow values for the condensate resin regeneration waste is for the input streams to the condensate polishers, not the waste. The waste volume is 3400 gpd per NUREG-0017 when the waste stream is processed by the condensate polishers. The formula used for the condensate polisher processed waste is then:

[jliCi/cc] = [Ci/yr]*(!le6 ,.Ci/Ci)/((l 6141l.654+2.88E7gal/day)*8.34 lb/gal

453.59 g/lb

365 day/yr)

The design concentration release of each isotope is then divided by the 10CFR20 App.B Table 2 Effluent Concentration Limit (ECL).

This fraction is then sunmmed over all isotopes. The acceptance criterion is for this sum to be less than unity. A second calculation (last colunmn of the following table) is determined using the Technical Specification limits (0.265 jitCi/gm 1-I131 ) for iodines instead of design values. The C/ECL values are determined by multiplying the Design C/ECL value by the scaling factor in Table 1. The Tech Spec C/ECLs are determined by multiplying the Design C/ECL by the appropriate Tech Spec scaling factor in Table 1. The sum of the concentration/ECL for this scenario will be for the Tech Spec iodines and~the previously determined design values for all other isotopes. Again, the acceptance criterion is for the sum to be less than unity.

For a TPC, the liquid H-3 concentration peaks due to boron dilution requirements (feed and bleed). There is more dilution toward the end of the cycle compared to the beginning of the cycle. The competing effects of tritium production, due to TPBAR permeation and increasing primary coolant boron concentration, and dilution from the feed and bleed process lead to a peak in H-3 concentration at some point during the cycle. Therefore, using the average H-3 concentration is not conservative because there are times during the cycle that the calculated H-3 releases would be smaller than the actual H-3 releases. The expected peak H-3 concentration would result in the 10CFR20, App. B limit being exceeded, so the maximum liquid H-3 concentration that can be tolerated is determined (see Table 12 for back-calciulation of H-3 concentrations). The value that appears in the following tables for H-3 (TPC) is the most limiting liquid H-3 release concentration as calculated in Table 12.

CNL-15-216, Enclosure 1 CNLI-216 Enlosue I 3, Page 19 of 35 Attachment

Calculation No. WBNTSRIOO

, Rev: 12 1 Plant: WBN Page: 12 I

Subject:

Design Releases to Show Compliance with 10CFR2 Table 5 :Liquid Release, No Processing Single Unit Duad Unit Tech Spec Operation Operation Single Unit Dual Unit 10CFR20*

iExp .Rel. Des/ Exp Design Design (ECL, Design "Design Operation Operation Nuclide (Ci/yr) Rat io (Ci/yr) (1-Ci/ca) pCi /ec) C/ECL C/ECL 0.265 pCi/cc 0.265 pCi/cc Br- 84 6.954E-04 2.50 1.73E-03 4.,36E-11 4E-4 1.095-07 2.18E-07 1-131 1. 06E+0O *52.41 5. 53E+01 1,.39E-06 1E-6 1.39E5+00 2. 78E+00 7.958E-02 1 .60E-01 1-132 1.23 E-01 4.00 4. 935-01 1,24E-06 15-4 1. 24E-04 2.4 8E-04 9.3 25-05 1. 56E-04 1-133 8 .40E-O1 26.85 2. 25E+01 S5,67E-07 7E-6 8.105-02 1. 625-01 S.078-03 1. 81E-02 1-134 3.23E-02 1.65 5. 32E-02 1.34E-09 4E-4 3.35E-06 6.69E-06 6. 105-06 1.22E-O5 1-135' 4 .43E-01 7.51 3. 505+00 8.80E-08 3E-5 2.935-03 5.8 7E-03 1. 118-03 2 .23E-03 Rb- 88 1.03E-02 18.14 1.8 8E-01 4.72E-OS 4E-4 1.18E-05 2.365-05 Cs- 134 2 .38E-01 40.60 S. 64E+00 2,42E-07 9E-7 2.69E-01 5. 39E-01 Cs- 136 2 .39E-02 165.2'0 3.95E5+00 S9,94E-08 6E- 6 1. 66E-02 3. 31E-02 Cs=137 3 .1BE-01' 153 .22 4. 87E+01 1.22E-06 1E-6 1. 22E+00 2 .4 5E+00 Cr-Si 1.21E-O1 0.29 3. 51E-02 8.82E-10 SE-4 1.76E-06 3.835-06 M'n-54 6. 70E-02 0.47 3. 15E-02 7.9SE-i0 3E-5 2 .64E-O5 5.2 7E-05 Fe -59 1.40E-02 3.48 *4. 86E-02 1 .22E-0S lE -5 1.22E-04 2.445-04 Co-SB 2.01lE-01 5.37 1. 08E+00 2. 72E-08 2E-5 1.365-03 2.72E-03 Co-6O 4.0O1E-02 1.38 5.535-02 1.39E-09 3E-6 4.63E-04 9 .27E-04 Sr-B8S 5. 36E-03 22.45 1. 20E-01 3 .025-09 8E-6 3. 78E-04 7 .56E-04 Sr-9O 4 .87E-04 13 .45 6 .57E-03 1, 65E-10 5E-7 3.30E-04 6.605-04 Sr-Si 2.98BE-03 1.86 5. 54E-03 1.39SE-l0 25-5 6.97E-06 1.39SE-OS Y-90 0 15.87 0 0 7E -6 0 0 Y-951 4.75E-04 1115.17 S5.30E-O1 1.33E-08 8E-6 1.67E-03 3.335-03 Zr-S95 1. 62E-02 1.71 2.i 78E-02 6. 98E-10 2E-5 3.49SE-OS 6.98EE-OS Nb-955 1.°34E-02 2.34 3. 13E- 02 7. 88E-10 3E-5 2. 63E~05 S5.25E-05 Mo-S9S 1.26E-01 785.19 S. 88E+01 2.48E-06 2E-S 1. 24E-01 2.48E-01 Te -132 *3.64E-02 145.25 S. 28E+00 1.33E-07 SE-6 1.478-02 2.95E-02 Ba- 14 0 4. 27E-O1 0.31 1.32E-01 3.335-09 85-6 4.16E-04 8.315-04 La- 140 6.14E-01 0.06 3.* 65E- 02 9.26E-10 9E-6 1. 03E-04 2. 065-04 Ce- 144 1. 58E-01 0.08 1,26E-02 3. 17E-I0 3E-6 1. 06E-04 2. 115-04 Pr- 14 4 0 0.08 0 0 6E- 4 0 0 1--3 1252.8 1 1252.8 3.155.-05 1E-3 3.15E-02 6.305-02 Ha-3(TPC) N/A N/A N/A 3.26E-04* 15-3 3.26E-01 3.265-01 1 Total 3.16E+00 6.325+00 1.775+00 3.555+00 (Total (TPCI 3.458+00 6.585+00 2.075+00 3.825+00 (

'Expected release from Table 6, Case #1 (MD processing for tanks and CVC5 only) of WBNTSRO93 (ref.3)

Note: The "Dual Unit Operation" column in the above calculation considers dual unit operation. Based on the evaluation done for Revision 7, the per unit concentrations are the same for both units. Therefore, the last column is twice the preceding column.

Note: Dual unit operations consider only Unit I with TPC.

Value is the limiting liquid H1-3 effluent concentration for a TPC calculated in Table 12. No expected release value was required.

CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 3, Page 20 of 35

Subjoct: Duslgn RelmsU to Show ConipIIUNO with IOCFR20 The sum over all isotopes of the concentratiulCL value fron the previous table ii grater than unity for the case where all isotopes are at demiga values. The bulk of the release is clue to the condensate reuin regeneration wast (untreated, ra.3). Per the ODCM (ref.10), the condensate regneration waste will not be a continuous r'elease if the activity in the secondary aide is peater than 1F,-6 MCi/Km. From WBNNAL3003 (ref.5), the activity ofsome of the isotopes in the secondary side is greater than 1K-C jjCifgm. This means that the regeneration waste will be in batch mode, and monitored. For conservatism, it was assumed that it is being released continuousl. From refrencea 11, 12 and 13 the condensate regeneration waste can be rerouted through the mobile demineralizrars If the long term releases from the condensate regeneration waste is greater thtan the IOCFR2,,'O concentration limits, than rotn the fluid abream through the mobile demineralizars will be performed. The expected release is fro)m Table 6, Case #

of WBNTSIROgS (tef.S). With mobile demineralizer processing of condensate regeneration waste the release concentrations become: t This p~es mpss by Rov~n oia.

r'kI! .4 r '~4 0 ~ mm '1 AtI~ -hIt aa 1 f3 L~IILILI~LII~ m -, - -~- -- - -

Subject:

Desig Releases to Show Cornpliance with 10CFR2O Tabiqui6: Relase ob Dmineralizer Processing De Single Unit Dual Unit

'Exp. Rel. Dea/Exp Dein Design 100F620 Operation Operation Nuclide (Ci/yr) Ratio (Ci/yr) (mCi/cc) (CCL, mCi/cc) C/EeL C/ECL Br-84 2.266-04 2.50 5.656-04 1.42E-11 46-4 3.55E-08 7LIlE-O8 1-131 3.706-02 52.41 1.946+00 4.87E-08 1E-6 4.876-02 9.74E-02 1-132 i.81E-02 4.00 7.236-02 1.826-09 . I1-4 1.82E-05 3.63E-05 1-133 7.29E-02 .26.65 1.966÷00 4.926-OS 76-6 7.036-03 1.416-02 1-134 8.57E-03 1.65 1.416-02 3.566-10 46-4 8.896-07 1.786-06 1-135 6.52E-02 7.91 5.16E-01 1.306-08 36-5 4.326-04 8.656-04 Rb-as 9.41E-03 18.14 1.71E-01 4.29E-09 4E-4 1.076-05 2.15E-O5 Cs-134 4.02E-02 40.60 1.636+00 4.106-08 96-7 4.556-02 9.116-02 Cs-136 3.50E-03 165.20 5.796-01 1.45E-08 6E-6 2.426-03 4.85E-03 Cs-137 5.52E-02 153.22-* 8.456+00 2.13E-07 16-6 2.136-01 4.256-01 Cr-Si 0.706-03 0.29 2.816-03 7.07E-11 5E-4 1.416-07 2.836-07 60-54 6.876-03

0.47 3.236-03 8.126-11 3E-5. 2.71E-06 5.41E-06 Fe-59 3.316-03 3.48 1.156-02 2.906-10 16-5 2.906-OS S.806-0S Co-SB 3.186-02 5.37 1.716-01 4.29E-09 2E-5 2.146-04 4.296-04 Co-6o 1.07E-02 1.38 2.726-02 6.836-10 36-6 2.286-04 4.556-04.

Sr-SO 2.767-04 22.45 5,988-03 1.506-10 8E-S 1.886-05 3.766-05 Sr-So

3.04E-05 13.49 4.106-04 1.036-11 56-7. 2.066-05 4.13E-OS Sr-9i 3.90E-04 1.86 7.26E-04 IL82E-11 26-5 0.126-07 1.82E-06 Yi-So 0 15.87 0 0 7E-6 0 0 Yi-91 1.23E-04 1115.17 1.37E-01 3.456-09 8E-5 4.326-04 8.636-04 Zr-OS 1.91E-03 1.71 3.27E-03 8.22E-11 2E-S 4.11E-06 8.22E-06 Nb-O5 2.886-03 2.34 6.756-03 1.706-10 3E-S 5.656-06 1.136-O5 MO-SO 5.856-03 785.19 4.59E+00 1.156-07 26-5 5.776-03 1.156-02 Te-132 1.556-03 145.25 2.26E-01 5.67E-09 9E-6 6.306-04 1.266-03 Ba-.140 . 1.446-02 0.31 4.466-03 1.12E-i0 I 86-6 1.406-05 2.806-05 La-140 2.286-02 0.06 1.376-03 3.43E-11 56-6 3.81E-06 7.636-06 Ce-144 9.496-03 0.08 7.SSE-04 i.91E-il 36-S 6.366-OS 1.27E-OS Pr-144 0 0.08 0 0 6E-4 0* 0 6-3 1252.80 1 1252.80 3.1496-05 1E-3 3.156-02 6.306-02 IH-3(TPC) N/ A N A N/A 3.26E-04" 1E-3 3.266-01 3.256-01 Total 3.566-01 7.116-01 I ITotal TPC) . 6.506-01 9.746-01 I

SExpected release from Table 6, Case #2 (MD processing) of WBNTSRO93 (ref.3)

Note: The "Dual Unit Operation" column in the above calculation considers dual unit operation. Based on the evaluation done for Revision 7, the per unit concentrations are the same for both units. Therefore, the last column is twice the preceding column.

Note: Dual unit operations consider only Unit I with TPC.

Value is the limiting liquid H-3 effluent concentration for a TPC calculated in Table 12. No expected release value was required.

CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 3, Page 22 of 35

UIDouet l WDTSIO Calculation e.: 010 sheet PJit BNInits12Pg:

!subject: "Desin Reese, to sho cempilnce wit IocP2o

DCN D-50602,-A (ref.l6) will allow the direct r-elease of Steam Generator Blowdown to the river, without Condensate Polisher Demineralizrs, and no processing of the Condensate by the Condensate Polisher Demineralizers. Therefore, there is no Condensate source tern. Calculstion WBNTSR093 performaed 5 cases.

One case (Case #3) was a releae* with exsected concentrations. That release exceeded the App.1 limit of 5 Ci/,mit and will not he evaluated in this calculation. A second ease (Case #4) limited the release of the 8GB

fluid to the Lower Limit of Detection (LLW = SE-7 pOi/ce gross gamma). The third case (Case #5) was forth 8GB fluid stream limited to the maximum allowable and still meet the AppT limit o 5 Cifunit. Because the 8SGB is not processed by the CID, then the volume of liquid will change fra the.3400 gpd analyzed previously to the max steam generator blowdown of 262 gpm (377/,2,80 gpdt). This makes the total1 flow 390, 02 1.654gd (16,141.854 - 3400 + 377,280 =f 890,021.654). The cocntration is calculated as follows:

[iJCi/cc] = [Ciir']*(1e6 pCi/Ci)I((390021L654 gpd+2.B8E~gal]day)*8.34 lb/gal

453.39 g/Ib *365 daylyr)

Tables 7 and 8 below are based on input from Table 5, of WBN'rSR093 (refA). In both tablbes the ILRW values are multiplied by the "1DeslFap Ratio", and then the 8GB values are added. There are several nuclides which do not have design values and threoe Des/Erp ratios were not generated. In these eases a multiplier of 1.4) has been applied in determining the Design relae.

This page rqlac bytRs*c 010.

/

I f'NII ~4~ 0 ~ ,r~ I A44orhm~nt 3 P~n*. 23 of 35 I~~~jI'1L~~~

I ,.-~ LIO

. - . I IIIIL~

j Calculation No. WBNTSR100 Rev: 12 Plant: WBN Page: 16

Subject:

Design Releases to Show Compliance with 10CFR2O Table 7: Direct Steam Generator Blowdown Release/SGBD at LLD with 20,000 gpm Dilution Liquid 2

SGB Cl/yr .10CFR20 Single Unit Dual Unit

'LRW Scaled to Dcs/Exp " Des . Liquid (ECL, Operation Operation Nuclide . (Cilyr) 0.26 Ci Ratio (Ci/yr) (iCi/ce) pCi/cc) C/ECL .C/ECL Er-84 2.26E-04 8.155-04 2.50 1.38E-03 3.42E-11 4E-4 8.565-08 1.71E-07 1-131 3.605-02 1.205-02 52.41 1.90E+00. 4.71E-08 1E-6 4.715-02 9.415-02 1-132 1.80E-02 2.87E-02 4.00. 1.015-01 2.505-09 15-4 2.505-05 4.995-05 1-133 7.225-02 3.44E-02 26.85 1.97E+00 4.89E-08 75-6 6.99E-03 1.405-02 1-134 8.555-03 2.505-02 1.65 3.91E-02 9.705-10 4E-4 2.425-06 4.855-06 1-135 6.495-02 5.285-02 7.91 5.665-01 1.405-08 3E-S 4.685-04 9.365-04 flb-88 9.41E-03 6.285-03 18.14 1.77E-01 4.395-09 4E-4 1.10E-05 2.205-05.

Cs-134 4.005-02 3.91E-03 40.60 1.635+00 4.045-08 9E-7 4.485-02 8.975-02 Cs-136 3.485-03 4.74E-04 165.20 5.76E-01 1.435-08 65-6 2.385-03 4.765-03 Cs-137 5.505-02 5.215-03 153.22 8.435+00 *2.095-07 15-6 2.095-01 4.185-01 Na-24 2.545-02 1.595-02 1.00 4.135-02 1.025-09 55-5 2.05E-05 4.095-05 Cr-51 9.59E-03 1.33E-03 0.29 4.11E-03 1.025-10 55-4 2.04E-07 4.08E-07 Mn-54 6.81E-03 6.65E-04 0.47 3.87E-03 9.595-11 3E-5 3.20E-06 6.39E-06 Fe-55 1.115-02 5.01E-04 1.00 1.16E-02 2.87E-10 15-4 2.87E-06 5.735-06 Fe-59 3.305-03 1.23E-04 3.48 1.16E-02 2.885-10 1E-5 2.88E-05 5.765-OS Co-58 3.015-02 1.94E-03 5.37 1.635-01 4.065-09 2E-5 2.035-04 4.065-04 Co-60" 1.97E-02 2.25E-04" 1.38 2.745-02 6.795-10 35-6 2.26E-04 4.52E-04 Zni-65 5.225-04 2.15E-04 1.00 7.375-04 1.83E-11 5E-6 3.66E-06 7.31E-06 Sr-89 2.615-04 S.835-O5 22.45 5.935-03 1.475-10 8E-6 1.84E-05 3.685-OS Sr-9O 3.005-05 5.015-06 13.49 4.095-04 1*025-11 5E-7 2.035-05 4.065-05 Sr-91 3.885-04 3.005-04 1.86 1.02E-03 2.535-11 25-5 1.27E-06 2.53E-06 Y-90 0 0 15.87 0 0 7E-6 0 0 Y-91m 2.30E-04 3.70E-OS 1.00 2.67E-04 6.625-12 25-3 3.31E-09 6.625-09 Y-91 1*.23E-04 2.15E-06 1115.17 1.375-01 3.40E-09 8E-6 4.255-04 8.50E-04 Y-93 1.735-03 1.28E-03 1.00 3.01E-03 7.485-11 2E-5 3.745-06 7.485-06 Zr-95 1.905-03 1.645-04 1.71 3.415-03 8.465-11 25-s 4.235-06 8.465-06 Nb-95 2.87E-03 1.135-04 2.34 6.845-03 1.705-10 35-5 5.655-06 1.13E-05 Mo-99 5.735-03 2.585-03 785.19 4.505+00 1.125-07 25-5 5.585-03 1.125-02 Tc-99m 4.575-03 1.195-03 1.00 5.775-03 1.435-10 15-3 1.435-07 2.865-07 FiU-103 8.03E-03 3.17E-03 1.00 1.125-02 2.785-10 35-S 9.275-06 1.8SE-OS Ru-106 1.045-01 3.795-02 1.00 1.425-01 3.525-09 35-6 1.175-03 2.355-03 Te-129rs 1.925-04 7.985-05 1.00 2.725-04 6.755-12 75-6 9.655-07 1.935-06 Te-129 9.975-04 2.525-03 1.00 3.525-03 8.745-11 45-4 2.185-07 4.37E-07 Te-13l1m 1.105-03 5.635-04 1.00 1.665-03. 4.125-11 85-6 5.155-06 1.035-05 Te-131 2.77E-04 3.395-04 1.00 6.16E-04 1.535-11 85-S 1.91E-07 3.825-07 Te-132 1.52E-03 6.81E-04 145.25 2.215-01 5:49E-09 95-6 6.105-04 1.225-03 Ba-140 1.405-02 5.335-03 0.31 9.675-03 2.405-10 85-6 3.005-05 6.00E-O5 La-140 2.22E-02 9.645-03 0.06 1.105-02 2.725-10 95-6 3.02E-05 6.05E-05 Ce-141 4.655-04 6.245-05 1.00 5.275-04 1.315-11 35-s 4.365-07 8.725-07 Ce-143 2.085-03 1.045-03 1.00 3.225-03 7.755-11 2E-5 3.87E-06 7.755-06 Ce-144 9.345-03 1.645-03 . 0.08 2.38E-03 5.925-11 35-6 1.975-OS 3.945-OS pr-144 0 0 0.08 0 0 65-4 0 0 Np-239 1.885-03 8.70E-04 1.00 2.745-03 6.815-11 25-S 3.415-06 6.815-06 5-3 1252.80 1 1252.80 3.115-05 15-3 3.115-02 6.225-02 H*-3.(TFCI N/A N/A .N A 3.265-04* 15-3 3.265-01 3.265-01 I Total 3.50E-01 7.01E-01 STotal (TPC) 6.455-01 9.655-01 {

1 Source: WBNTSRO93 (ref.3), TableS5, LRW only, no SGB 2 Source: WBNTSR093 (ref.3), Table 5, SGB scaled to LLD

Value ia the limiting liquid H-3 effluent concentraton for a T[PC calcuilated in Table 12. No expected release value was required.

CNL-15-216, Enclosure 1 Attachment 3, Page 24 of 35

Subject:

Design Releases to Show Compliance with 10OCFR2O Table 8: Direct Steam Generator B lowdown Release/SGBD at Max App. I with 20,000 gpm Dilution Liquid 2S0B Cl/yr IO1CFR20 Single Unit Dual Unit

'LRW Scaled to Des/Exp Des Liquid (ECL, Operation Operation Nuclide (Ci/yr) 4.40 Ci Ratio (Cilyr) (pCi/cc) gCi/ee) C/ECL C/ECL Br-84 2.26E-04 1.38E-02 2.50 1.44E-02 3.56E-10 4E-4 8.915-07 1.785-06 1-131 3.80E-02 2.035-01 52.41 2.095+00 5.185-08 1E-6 5.185-02 1.04E-01 1-132 1.80E-02 4.865-01 4.00 5.585-01 1. 38E-08 15-4 1.38E-04 2.775-04 1-133 7.225-02 5.82E-01 26.85 2.52E+00 6.25E-08 7E-6 8.935-03 1.79E-02 1-134 8.555-03 4.235-01 1.65 4.375-01 1.08E-08 4E-4 2.715-05 5.425-05 1-135 6.49E-02 8.93E-01 7.91 1.41E+00 3.49E-08 3E-S 1.16E-03 2.335-03 Rb-88 9.418-03 1.06E-01 18.14 2.77E-01 6.878-09 45-4 1.72E-05 3.445-05 Cs-134 4.008-02 6.615-02 40.60 1.69E+00 4.198-08 9E-7 4.66E-02 9.31E-02 Cs-136 3.48E-03 8.025-03 165.20 5.835-01 1.455-08 65-6 2.415-03 4.83E-03 Ca-137 5.50E-02 8.825-02 153.22 8.51E+00+ 2.115-07 15-6 2.11E-01 4.225-01 Na-24 2.545-02 2.685-01 1.00 2.945-01 7.295-09 55-5 1.465-04 2.925-04 Cr-51 9.59E-03 2.255-02 0.29 2.535-02 6.285-10 5E-4 1.265-06 2.515-06 Mrr-54 6.815-03 1.13E-02 0.47 1.45E-02 3.595-10 3E-5 1.20E-05 2.395-05 Fe-55 I.111-02 8.498-03 1.00 1.95E-02 4.85E-10 15-4 4.85E-O6 9.70E-06 Fe-59 3.305-03 2.08E-03 *3.48 1.36E-02 3.37E-10 1E-5 3.37E-05 6.745-05 Co-58 3.015-02 3.29E-02 5.37 1.945-01 4.825-09 25-s 2.415-04 4.825-04 Co-60 1.975-02 3.81E-03 1.38 3.09E-02 7.685-10 35-6 2.565-04 5.125-04 Zn-65 5.225-04 3.645-03 1.00 4.16E-03 1.038-10 55-6 2.065-05 4.135-05 Sr-89 2.61E-04 9.87E-04 22.45 6.865-03 1.705-10 BE-S 2.13E-05 4.25E-05 Sr-9O 3.00E-05 8.495-05 13.49 4.895-04 1.215-11 5E=7 2.435-05 4.858-05 Sr-91 3.885-04 5.085-03 1.86 5.80E-03 1.448-10 25-5 7.205-06 1.445-05 Y-90 0 0 15.87 0 0 7E-6 0 0 Y-91m 2.305-04 6.26E-04 1.00 8.56E-04 2.125-11 2E-3 1.06E-08 2.125-08 Y-91 1.235-04 3.645-05 1115.1.7 1.375-01 3.40E-09 85-6 4.25E-04 8.50E-04 Y-93 1.73E-03 2.165-02 1.00 2.345-02 5.805-10 2E-5 2.90E-05 5.805-05 Zr-95 1.908-03 2.775-03 1.71 6.025-03 1.495-10 2E-5 7.465-06 1.495-OS Nb-95 2.875-03 1.915-03 2.34 8.635-03 2.148-10 35-s 7.145-06 1.43E-05 Mo-99 5.738-03 4.37E-02 785.19 4.54E+00 1.138-07 25-5 5.63E-03 1.135-02 Tc-99m 4.578-03 2.025-02 1.00 2.48E-02 6.155-10. 15-3 6.158-07 1.235-06 Ru-103 8.03E-03 5.37E-02 1.00 6.17.8-02 1.53E-09 3E-5 5.10E-05 1.02E~04 Ru-106 1.04E-01 6.41E-01 1.00 7.45E-01 1.85E-08 35-6 6.16E-03 1.23E-02 Te-129si 1.92E-04 1.35E-03 1.00 1.54E-03 3.83E-11 7E-6 5.47E-06 1.095-OS Te-129 9.97E-04 4.27E-02 1.00 4.37E-02 1.088-09 45-4 2.71E-06 5.42E-06 Te-131nm 1.10E-03 9.53E-03 1.00 1.06E-02 2.645-I0 8E-6+ 3.295-0S 6.59E-05 Te-133 2.775-04 5.735-03 1.00 6.015-03 1.495-10 85-5 1.86E-06 3.735-06 Te-132 1.52E-03 1.1SE-02 145.25 2.32E-01 5.76E509 95-6 6.40E-04 1.285-03 Ba-140 1.40E-02 9.02E-02 0.31 9.45E-02 2.35E-09 8E-6 2.93E-04 5.86E-04 La-140 2.225-02 1.635-01 0.06 1.645-01 4.088-09 95-6 4.353-04 9.078-04 Ce-141 4.65E-04 1.065-03 1.00 1.52E-03 3.778-11 3E-S 1.265-06 2.525-06 Ce-143 2.08E-03 1.76E-02 1.00 .1.97E-02 4.895-10 2E-5 2.44E-05 4.895-05 Ce-144 9.34E-03 2.775-02 0.08 2.858-02 7.06E-10 3E-6 2.35E-04 4.71E-04 Pr-144 0 0 0.08 0 0 6E-4 0 0 Np-239 1.885-03 1.475-02 1.00 1.665-02 4.125-10 25-s 2.06E-05 4.125-OS 5-3 1252.80 1 1252.80 3.11E-05 1E-3 3.115-02 6.228-02 I-t-3(TPC) N/A "N A N A 3.26E-04* 15I-3 3.26E-01 3.26E-01 I Total 3.68E-01 7.36E-01 STotalI(TPC) 6.63E-01 1.005+00 1 Source: WBNTSRO93 (ref3), TableS5, LRW only, no 5GB 2 Source: WBNTSRO93 (ref.3), Table 5, 5GB sealed to LLD

Value is the limiting liquid H-3 effluent concentration for a TPC calculated in Table 12. No expece~d release value was required.

CNL-15-216, Enclosure 1 Attachment 3, Page 25 of 35

LWi*l Calculation sheet S

Subject:

Design Ieleases to Show Compliance with IOCFR2O To determine if the SOB with no processing can be released to the river without the Cooling Tower Blowdown (CTB) dilution flow of 20,000 gpm, Table 9 takes the Table 7 annual release and conlverts it to a woneentration without the 20,000 gpm CTB dilution.

[uCi/cc] :[Ci/yrJ*(1e6 pciICi)/(($90,021.654 gpd)*8.34 lb/pga* 453.59 g/lb

365 dayF/yr)

"This page ispacid by Revslon 010..

r'lill -4'_04t* I:nr, lioiirn I Att~nhment 3. Paae 26.of 35

~JI~L. I I fl~* ~'.fl~~D-* - - .-.---. -. -- -, -

Calculation No. WBNTSR100 Rev: 12 Plant: WBN Page: 19 I

Subject:

Design Releases to Show Compliance with 10CFR2 *.0 Table 9; Direct Steam Generator Blowdown Release/SGBD at LLD without 20,000 gpm Dilution Liquid Liquid Single Ufnit Dual Unit 1

Des Des 10CFR20 "Operation Operation Nuclide (Ci/yr) CiaCi/cc) CECL, i*Ci/cc) C/ECL C/ECL Br-84 1.38E-03 2.56E-O9 4E-4 6.40E-O6 1.28E-O5 1-131 1.50E400 3.52E-O6 1E-S 3.52E+OO 7.05E÷00O 1-132 1.O1E-01 1.87E-07 1E-4 1.87E-03 3.74E-03 1-133 l1.97E+O0 3.56E-O6 7E-6 5.235-01 1.05E+00 1-134 *3.91E-02 7.26E-08 4E-4 1.815-04 3".63E-04 1-135 5.66E-O1 1.05E-06 3E-5 3.50E-02 7.00E-02 Rb-SB 1.77E-01 3.295-07 4E-4 8.225-04 1.54E-03 Cs-134 1.63E+00 3.025-06 9E-7 3.36E+00 6.71E+00 Cs-136 5.76E-O1 1.075-OS 6E-S 1.78E-01 3.56E-01 Cs-137 8.43E+00 1.56E-05 1E-6 1.56E+01 3.13E+O1 Na-24 4.13E-02 7.665-08 SE-S 1.53E-03 3.06E-03 Cr-Si 4.115-03 7.64E-09 5E-4 1.53E-O5 3.O5E-05 Mn-54 3.87E-03 7.18E-O9 3E-5 2.39E-04 4.79E-O4 Fe-S5 1.16E-02 2.155-08 1E-4 2.15E-04 4.29E-04 Fe-59 1.16E-02 2.165-08 1E-5 2.165-03 4.31E-03 Co-58 1.63E-O1 3.045-07 2E-5 1.525-02 3.04E-02 Co-6O 2.74E-02 5.085-08 3E-6 1.69E-02 3.39E-02 Zn-65 7.375-04 1.37E-0S 5E-S 2.745-04 5.47E-04 Sr-B9 5.935-03 1.10E-08 BE-6 1.38E-03 2.75E-03 Sr-SO 4.09E-04 7.60E-10 5E-7 1.52E-03 3.045-03 Sr-Si 1.02E-03 1.90E-09 2E-5 9.48E-05 1.905-04 Y-SO 0 0 7E-8 0 0 Y-91m 2.675-04 4.96E-10 2E-3 2.485-07 4.96E-07 Y-91 1.37E-O1 2.54E-07 8E-6 3.18E-02 6.36E-02 Y-93 3.01E-03 5.859-09 2E-5 2.8OE-04 5.555-04 Zr-95 3.41E-03 6.33E-09 2E-5 3.17E-04 6.33E-04 Nb-95 6.84E-03* 1.27E-08 3E-S 4.235-04 8.465-04 Mo-99 4.50E+O0 8.36E-O6 2E-S 4.1BE-O1 B.365-01 Tc-99mn 5.77E-03 1.07E-O8 1E-3 1.07E-05 2.145-05 Ru-103 1.12E-02 2.08E-08 3E-5 6.94E-04 1.355-03 RU-lOS 1.42E-01 2.64E-07 3E-6 8.79E-02 1.76E-01 Te-129m 2.72E-04 5.05E-10 7E-6 7.22E-O5 1.445-04 Te-129 3.52E-03 6.54E-OS 4E-4 1.63E-05 3.275-05 Te-131m 1;66E-03 3.OBE-O9 SE-S 3.86E-04 7.715-04 Te-131 *.6.1E-04 l.14E-O9 SE-S 1.43E-O5 2.86E-05 Te-132 2.21E-O1 4.11E-07 9E-S 4.56E-02 9.13E-02 Ha-140 9.57E-03 1.79E-08 8E-6 2.24E-03 4.495-03 La-140 1.10E-02 2.045-OB 9E-S 2.26E-03 4.53E-03 Ce-141 5.27E-04 9.755-10 3E-5 3.26E-O5 6.53E-OS Ce-143 3.12E-03 5.BOE-O9 25-5 2.90E-04 5.80E-04 Ce-144 2.38E-03 4.43E-OS 3E-S 1.48E-03 2.95E-03 Pr-144 0 0 6E-4 0 0 Np-239 2.74E-03 5.105-09 2E-5 2.55E-04 5.105-04 5-3 1252.80 2.33E-03 1E-3 2.335+00 4.655E+00 IR-3CTPC) N/A 3.2SE-04" 1E-3 3.26E-01 "3.26E-01 "

Total 2.62E+01 5.25E+01

[Total (TPC) 2.425+01 4.B1E+01 I

'Desigo Release fom Table'7

Value is thelimiting liquid H-3 effluent concentration fora TPCecalulated inTablel12. No expece~d releasevalue was required.

CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 3, Page 27 of 35

L13 Calculation sheet

[Subject. Dmlgn Releases to Show Compkence with IOClFR2O Th. above scenrio exceeds all release liit (desig and ODCM), and therefore is not to be, implemented.

From Table 9, as long as the normal radwaste system operates, the 100FR20 limits will be exceeded with no SOB processing. Table 10 presents the case where t.he SGB is the oniy release and is limited to the LWD. The annual release is at the 0.28 Ci limit as determined in WBNTSRO93. The concentr'alion is calculated am follows in the third column

[pCilccJ = [Cifyr]*(1e8 ~IJCi/')I((37 7 , 280 gpdl)*8.34 lb/gal' 453.59 g/]b" 385 day/y'r)

This page rpqscod by RemIon 010.

I I, i

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~ 11J1 m Att~chment

................ 3.* Paae

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n..... 28 of 35 l~*fl~*~~-*- -a -

Ca~lculationshe

$ubJeot Design Rdeleaes to Show Conplilance wit iOCFR20.

Table 10: Direct Steam Generator Blowdmwonly ReleneSBD at LLD without 20,000 gpm Di!utoaSGB is Release Path Ulngi IMI Dual LhnR Operaon Operation Nuclide (Cly') (Iaic) (EcL, 5iClft)

Dr-84 IL16E-O* 1.575-00 4E-4 3.91 E-08 7.83E24) 1-131 1.202-02 2.31K-OS 1E-8 2.318-02 4.625-.2 2-132 2*872-0 5.85-E06 1E-4 5.525-04 1.10243(

1`133 3.445-02 5`OE-OS 75-6 9.432-03 1.83E-02 2-134 *.S,5E-02 4.80E-08 42-4 1202-04 2.401.44 1-135 8.282-02 1.01 E-07 3E-5 3.35E.03 9.792-03 Rb-S8 8.,282-,03 1.20E-08 42-4 3.015-05 9.02E-45 Cs-1 34 3,915-03 "150E-09 95-7 6.33E.43 1.67-24 Cs-1 36 *4.742-04 9.102-10 91-6 '1.2-042.O 2.032-44 Cs-i137 5.215-03 1.00E-08 15-8 1.008-02 2.002422 Na-24 1,,92-02 3.015-08 55-8 6.092404 1.2224.3 Or-81 1.332-03 2.585-09 85-4 5.115-06 1.022-05 0-4 6.615E-04 1.281-09 32-5 4.2614.5 8.81E-05 Ps-El 5,01 E-04 9.63E-10 1E-4 9.632.06 1.932.05 P-9 1.232-04 2.362-10 15-8 2.365-08 4.71E-05 Ca-Ge 1,942-03 3.732..9 2E,-8 1.875-04 3.731-44 Ca-EG 2.2,2.04 4.32E-10 32-6 1.442-04 2.582-4 Zn-65 2185E-04 4.131'-10 55-6 6.28E-05 1.6-244 Sr-SO9. 1838-.6. 1.125-10 81-6 1.40E-05 2.80E-05 Sr-GO 5,015-06 9.032-12 .55-7 1,93-05 3.835-45

,Sc-G1 2,001-04 8.765-10 22-8 2.855-05 5.76E-05 Y-GO 0 0 7E-6 0 0 V-Olin 2.70E.0S 7-11E-11 25-3 3.552-08 7.11E-05 Y-01 2.152-06 4.131-12 65-6 5.1814?7 1.031,.06 Y-e3 1.251-03 2.465-09 25-5 1.235-04 2.48-04 Zr-O5 1.642-04 2.141-10 25-8 1.575-05 3.14E-06 N-S 1.13E-04 2.161-10 32-5 7.20E-46 1.4,4E-05 Ma-GO 2.68E.-03 4.965.9 25-5 2.48544 4.96E-04 Ta404 1.191-03 2.29-09 1E-3 2.292-06 4.885-06 Ru-103 3.171403 6.091..09 31-6 2o*.03o4 4.055-04 Ru-106 2.71E-02 727E-08 3-Te-129in ?J91.-5 1.531-10 71.6 , .192-45 4,385-05 Te-1J29 2.521-03 4.881-09 2- 1215-45 2.4--45 Te-I31m 5.631-04 1.051E-09 82-5 1.3814 2.70-0 Te-131 3.381-04 6.505-10 55-5 8.12E-45 1.02E-05 Te.132 9.815-04 1.315-09 82-6 1.481404 2.9E0544 Ba-140 £,33243 .1.0"5-05 82-6 1.282,-e5 2.56543a L.a.140 9.641-03 1.58E-05 9-4 2.062-03 4.115-03 Cs-141 1.242405 1.202-10 32-8 2,.0956 7.992-49 Ca-143 1.042403 .2.-00-09 28-6 9.99-05 2.00E-04 Ce-l44 1.54E-03 2.1,1-09 . 31,-6 1.055-03 2.1 -014 Pr-l146 0 0 65-4 0 0 N*239 8.702-04 1.67E,. 21-8 8.355-05 1.675.04 H-3 0 0 18-3 0 0 To:* 2.61-01 5.0E.07 165-0Q,2 1.721411 This page r~ptacld b Rev',ion 010.

CNL-15-216, Enclosure 1 Attachment 3, Page 29 of 35 CNL-15-216, Enclosure 1 Attachment 3, Paqe 29 of 35

RI Calculation sheet SSubJel: Design Rdskmses to Show Compliance with tOCFR2O .

Theasame anaalyssa nthe above able 1 ie performed in Table 11, except that the SGB release inat the maximum allowable for AppJ lii. The maximum Aip.I release is taken from Table 5 of WBNTSEO93.

The ClEO, value is greater than unity. The release is normalized to 10 ECL to determine the maxmum concentration (l~ikce) which would result inl a 10 ECL release. The concentration is calculated as follows in the third eolumn

[iCi'co = [Ci] "(leG MiICi)l((377.280 gpd)*8.84 lb/gal

45a.59 glrb

385 day/yr)

Tm. iug replace by Reviskon 010.

II I

~.II

~ LI~ LI J l4Ib Attfanhm*_nf . P~nie 30 of 35

.I-- - I

I11111 Calculation sheet Subject Design Rdseua to Show Compiluics with IOCPR2O !

I Table 11: Direct Steam, Generator Blowdowa ReleaseISGBD at Max App. Iwithout 2-0,000 gpm.

JDilutirmISGB in only Release Path L1Id Opumes Opsifon Max AIp MaxAp. I IOGPO2 NwmaIb~d Man u*~

U-EC at44 1.388-02 2.66846 48-4 4.561-4-4 tOZE-07 6.02E.45 1-13 2.03841* 3.911-07 11- ,3.91E-41 7.81E-01 2192400O 2.$5-06 9.34-o3 1.878-02 6.42242 6.425-06 1-133 SIZ8E-Cl 1.121-00 78-6 3.105-01 1.105+00 7.87E-48 1-134 4.235.01 3.123-07 41-4 2.03E-03 4.06--03 1.316-02 5.86146 5-135 1L931401 1.71E-06 35-5 5.728-02 3.93641 1.18845 1.101-04 1.022-43 3.505403 1.401.6 Cs-1fl 6.611402 1.271-07 96-'7 14A1-14 2,821411 9.665401 8.725=-07 Cs-131 e.028-03 1.548-06s 61-6 2.575-03 5.13E-03 1.768-02 1.062-07 Cs-IS? 1828-02 1.6998-07 15-6 1.99-O1 1.161+00 1.1614-5 1.038-02 2-068-02 7.062402 3.542-06 Q4-1 2-22E-02 4.326-06 55-4 6.641-0 1.735-44 5.942-44 2-872-07 Mn-84 1.135.42 2.161-06 36-8 7.20E-04 4.951-03 1.496-07 Fe-55 1.416.03 1.638-00 18-4 1.631-44 3.6-544 1.128-03 1.121-07 FP-59 2.08803 3,991-09 15-6 3.995-04 2.745403 2.745-06 Co-SI 3.29542 6.32508 21-5 3.161-03 8.325-03 2.178-02 4.34E-07 Co4O 3.81..03 7.318-09 36-8 2-44843 1.681-02 6.031498 140543a 9.6E043 4.601-6 869 .876-04 1.96E-09 11-6 2.371*44 4.741-44 1.631-03 12306-08 Sr-GO 8.491-05' 1.638-10 61-7 6,528-44 2.245-3 1.12549 Sr-SI 1,08143 9.72E.09 26-5 9.758-44 3.381-03 8.706408 V-GO 0 0 784, 0 0 0 0 Y-1me 6.2644* 120549 36-3 6.01E-07 1.205-06 4.135460 8.271-09*

63736-0 1.751..6 6,006..05 480-10 Y-93 2.188-02 4.161.09 21-5 2.085-43 4. 1654.3 1.4132-2 2..65-07' Zr-US 2.778-03 5.32E4)9 2E,.5 2.661.04 1.383-03 3,68E6-06 Nb-US 1.915-03 3.964E-9 35-5 t.22E.44 1385-04. 2.51E-08 M-9 4.378-42 8.39-6 .2-54 2.S6E-8 6.77E,.07 4.20E-43 6.39-1.3 To.ts 2.025.42 3.81--00 16-3 3.88-45 2-67E-44 2.171407 Ru-.lOS 6.378402 1.63E.07 31-S 3,44E-43 0B.871..3 2-366-02 7.061407 Ru-IOU 6.411-0 1.234.6 38-6 4A010E-1 8.295-01 2.625+00 8.46146*

Te-12tn 1.33.6.o3 2.,96-9 71-4 3.706-44 7.415-44 2.56E-09 138548-0 Ta-US9 4.27642 3.208,.08 4.,4 1055440 4.10--44 1411-03 5.64E-07 T.-IS1m B.S3E-OS 1.838-08 18-0 12854-3 1.876402 1.26647 Te-131 5,73E-03 1.106-08 81-5 1.378-44 2.751-04 .45E-04 7.58.O6 To-132 1.124-2 2-218406 96-6 2.461-43 1.69E-42 1.654E-7 Ue-140 9.021.42 1.736407 86-6 2.161-02 4.335.-02 1.495-01 1.195-06 La-l40 1.638401 3.132407 -64 3.48102 6.660-02 2.39E-41 2-166-46 Ce-41 1.06143 1.038-49 36-5 6.766-O0S "1.3,5..44 Ce-IC3 1.766.2 3.386-08 28.6 1 .664-3 1.161-02 2.32E-07 3.388-03 Can-144 2.7"/8-02 6.322408 36-6 1.771-02 3.65E,.02 1.2E-O41 3.688-07 Pr-1M4 0 0 66-4 0 '0 0 0

Np-239 1478-02 2.836-48 2145 1.411-03 *2.8314.3 9.71E-43 1.942-07 11-3 1E-3 Towa 4.468400 8.4,2-06 1.45 2.91 1.001401 5.815-45 Thspage qabmd* by Ruesion 010.

r ,, , , ,, ,, ,,

GNL-15-21fi Enclosure 1 Attachment 3. Paae31 of 35 "CNL....21fI Enclosure...Attachment.3. Paae 3111 I of I35

Subject:

Design Releases to Show Compliance with 10CFR20O The expected peak H-3 concentration in the primary coolant would result in the 10CFR20 limits being exceeded.

Therefore, the maximum liquid H-3 concentration that can be tolerated is determined by considering the available liquid release margin remaining after all other isotopes have been accounted for. The "Dual Unit Operation Liquid C/ECL" totals from Tables 5 - 9 are uised as a starting point because Dual Unit Operation values bound Single Unit Operation values. The remainingmargin is calculated based on the following relationship, which accounts for the non-TPC H-3 so that a single H-3 release concentration may be obtained: .

(C/ECL)p.e0mning = 1 .O0E+00 - ((C/ECL)DU3 , - (C/ECL)fo*lvc,H. 3) where (CEL*mhn is the available margin for the C/ECL fraction that does not contain H-3, (C/ECL)D 0 1a is the "Dual Unit Operation Liquid C/ECL" total from Tables 5 -9, (C/ECL)no,-yc,n. 3 isthe Dual Unit C/ECL fraction attributable to non-TPC H-3 from Tables 5 - 9. If the (CIECL)Rcmainig is negative, the combination of other radioactive liquid releases has already exceeded the .10CFR20 limits (Tables 5 and 9) and further calculations are not applicable.

The release concentration, which accounts for case-specific dilution, can be back-calculated by multiplying the remaining C/ECL margin as follows:

CH-3,Roleased~ = (C/ECL)Reraiainiig

ECLH.3 where CH-3,R*IC is the concentration of released 11-3 for a TPC under case-specific dilution scenarios that would result in*

the !0CFR2O, App. B limit beinig reached, and ECLH-3 is the 10CFR2O, App. B ECL for liquid releases of H-3 to the public in MCi/cc.

Using the above two equations and the results from Tables 5 - 9 as explained above, Table 12 detefrmines the concentration of released 11-3 for a TPC with case-specific dilution parameters that would result in the 10CFR20, App. B limit being reached.

Table 12: Maximum Lia uid H-3 Concentration Tolerable for Dual Unit O eration with a TPC

~Remaining Dual Unit Non-TPC. C/ECL Fraction H-3 Maximum Liquid Operation H-3 (no H-3) ECL H-3 Concentration Table (CIECL) (C/ECL) (C/ECL) (uCi/ec) (jiCi/ce) 5 -Liquid Release,.No Processing 3.55E+O0 6.30E-02 -2.49E+O0 IE-3 N/A 6-Liquid Release, Mobile

71E0 .0-2 35E0 E335E0 Demineralizer Processing 7.lEO .0-2 35-1l-332-4 7 -Direct SGBD Release/SGBD at LLD 7.01E-01 6.22E-02 3.61B-01 1E-3 *3.61E-04

with 20,000 gpm Dilution _____

8.- Direct SGBD Release/SGBD at Max 73E0 .2-2 32E0 E332E0 App. I with 20,000 gpm Dilution _____

9-ietSB eeaeSB tLD 5.25E+01 4.65E+00 *-4.68E+f01 IE-3 N/A without 20,000 gpm Dilution ___________ _______

From Table 12, the most limiting liquid 11-3 concentration that could be released to the environment and still meet the 10CFR20, App. B limit is 3.26E-04 j.Ci/cc, which is associated with Table 8 (Direct SGBD Release/SGBD at Max App. I with 20,000 gpm Dilution). Therefore, the following statement is made a special requirement/limiting condition of this calculation:

The maximum allowable liquid concentration of tritium (H-3) released to the environment is 3.26E-04 pCi/cc. H-3 concentrations above 3.26E-04 pCi/cc released to the environment may result in the 10CFR20 limit being exceeded.

CNL-15-216, Enclosure 1 CNL-5-21, AttachmentEnlosue I 3, Page 32 of 35

[caculaion No. WBNTSR100 IRev: 12 IPlant: WBN IPage:'24

Subject:

Desig Releases to Show Cornpliance with 10CFR2O Results The results of this calculation are:

Mode Of Release Gas

(Conc/ECL) 1 Unit (TPC) 0.311 (0.320) 0.324 0.270 (0.333)

(0.279)

Z (Conc/ECL) 2 Units (TPC-Ul only) 0.623 (0.632) 0.648 (0.65.7) 0.540 (0.549)

Description - Table *

(with containment purge ) - 2 (with unfiltered containment ventingq) - 3 (with continuous filtered containment venting) - 4 Liquid 3.16 (3.45*) 6.32 (6.58*) (unprocessed cond.demin.waste) -5S 1.77 (2.07*) 3.55 (3.82*) (unprocessed cond.demin, iodine at 0.265 tech spec

______________limits) - 5 I 0.356 (0.650*) 0.711 (0.974*) (processed cond.demin. waste) - 6 0.350 (0.645*) 0.701 (0.965*) (no 8GB processing, SGB release limited to LLD) - 7 I

0.368 (0.663") 0.736 (1.00*) (no SGB processing, SGB rel. limited to max App. I) - 8 I

26.2 (24.2*) 52.5 (48.1*) (no SGB processing, SGB at LLD, no 20,000 9pm I

_______________dilution) - 9 0.086 (0.086) 0.172 (0.172) (SGB at LLD, SGB only release path,no 20,000gpm

______________dilution) - 10 1.45 (1.45) 2.91 (2.91) (SGB at max App. I, no 20,000 gpm dilution) - .Ii

TPC vastie is bssed on the limiting liquid H-3 effluent concentration cslcu/ated in Tsble 12.

Maximum concentrations for no 20,000 gpm dilution, SGB only release path:

for 10 ECL release: 5.81E-05 l.iCi/cc - table 11 for 1 ECL release: 5.81E-06 lliCi/cc Discussion and Conclusion The gas design release concentrations are below the 10OCFR20 App.B Table 2 limits (one unit operation). The continuous filtered containment venting option results in the lowest gas release because Xe-I133 does not reach the same levels in containment due to decreased residence time in containment of its parent 1-133. The Xe-133 is the dominant gaseous isotope released. The liquid design release concentrations with no Condensate Polisher Demineralizer regeneration waste processing will be above the 10CFR2O limits. However, when the secondary side activity reaches-1E-6 g+/-Ci/gm the condensate regeneration waste is not continuously released, but is released in batch mode and is monitored. If the long term release of this waste is projected to result in exceeding the I0CFR2O limits, the design of the plant allows the waste to be processed by the mobile demineralizer system. With the mobile demineralizer system processing the regeneration waste, the liquid design releases are below the 10CFR20 limits. Note that the design concentrations under consideration in this calculation are significantly above the technical specification limits. For example, 1-131 in the design reactor coolant is 2.5 /Ci/grn (Table 1) but the technical specification is 0.265 .iCi/grn 1-131 equivalent. This means that the design rea~ctor coolant has greater than 2.5 times the technical specification limits. When iodines are kept at the Technical Specification limit, and all other isotopes at design levels, then the release concentrations are below the limits. It can be concluded that the realistic release concentrations will then be less than the 10CFR20 limits.

The design of the gais and liquid radwaste systems meet the requirements of 10CFR20.

Revision 4 incorporates the alternate operation mode of Steam Generator Blowdown released directly to the.

environment without processing by the Condensate Polisher Demineralizers, and no Condensate processing by the Condensate Polisher Demineralizers. It is concluded that under normal conditions this release could exceed the 10OCFR20 limits. The concentration of the SGB release must be'less than that which would exceed the App.I limit of 5 Ci/unit =

8.45E-06 igCi/cc gross gamma (Table 11), or even better, be restricted to the LLD (5E-7 uCi/cc gross gamma).

Revision 4 also evaluated releases without the 20,000 gpm Cooling Tower Blowdown dilution flow. The results indicate that this cannot be done if th~ere are releases from sources other than the SGB. If there are no other radwaste releases, the SGB maY be released with no dilution flow if the concentration is 5..81E-6/aCi/cc gross gamma or less.

The concentration/ECL fractions due to implementing a TPC in Unit 1 are determined in this calculation. Gaseous effluent releases of H--3 are based on the design basis source term. Conclusions for the gaseous releases are unchanged for.

a TPC.

For liquid releases of H-3, the maximum allowable liquid concentration of H-3 that can be released to the environment before exceeding the 10CFR20, App. B limit is back-calculated based on the available margin for the liquid release scenarios. This is made a special requirement/limiting condition of the calculation: The maximum allowable liquid concentration of tritium (H1-3) released to the environment is 3.26E-04 jtCi/ec. H-3 concentrations above-3.26E-04 gtCi/cc released to the environment may result in the 10CF.R20 limit being exceeded. Maintaining the H-3 concentration released to the environment lower than this value will ensure that the liquid radwaste system meets the requirements of 10OCFR20.

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 3, Page 33 of 35 Attachment

Icaicution No, wBNTsR100 Rev: 12 IPlant: WBN Page:25

Subject:

Design Releases to Show Cornpliance with 10CFR20 References

1. NUJREG-0800 R2"Standard Review Plan"..

2. T1534 Ri0 "Annual Routine Radioactive Airborne Releases from the Operation of One Unit"

3. WBNTSR093 R13 "Liquid Radioactive Waste Releases"

4. "Memorandum RJMS# T50 950109 844

5. WBNNAL3003 R5 "Reactor Coolant and Secondary Side Activities in Accordance with ANSI/ANS-18.1-1984"

6. .not used

7. WCAP 7664 RI "Radiation.Analysis Design Manual, 4-Loop Plant" RIMS# NEB 810126 316

8. System Description N3-77C-4001 R2 "Liquid Radwaste Processing System" RIMS# T29 930403 988

9. Technical Specification 3.4.16

10. Offsite Dose Calculation Manual (ODCM)

11. System Description N3-14-4002R2 "Condensate Polishing Demineralizer System" RIMS# B26 880714 022

12. WBN CCD drawing 1-47W838-3 Ril1

13. WBN CCD drawing 1-47W830-7 R5

14. WBNTSR-008 RI4"Control Room Operator and Off'site Dose Due to a Steam Generator Tube Rupture"

15. DCN D=50165-A 16..DCN D-50502-A

17. EDC E-50629-A

18. WAT-D-10890, Westinghouse document NDP-00-0326, "Transmittal of Waste Management Evaluation for the Watt's Bar Tritium Production Core," RIMS# T771 001204 807

19. PER 546323

20. Deleted in Ri11

21. Deleted in Rll

22. Deleted in Rll

23. DCN 59397

24. DCN 61599 CNL-15-216, Enclosure 1 CNL- 5-1 6 Encosue IAttachment 3, Page 34 of 35

Attachment A NPG Calculation Design Output.

EDMS T711l0 71~' Page 26 NPG CALCULATION DESIGN OUTPUT.

Calculation Identifier: WVBNTSR100 Engineering Change DCN 61599 Revision 12 Document Calculation

Title:

Design Releases to Show Compliance with IOCFR2O

,, ,..,,-- -3

, Preparer- Mehran A. Date - lkandar Milicevic "Date "C heckerS Aleksandar Milicevic Date Approval Date This Calculation Design Output is a compilation of the design output requirements of the referenced calculation.

Review of the referenced calculation for additional design output information is not required nor allowed.

Any new design output or changes to design output portions of calculations shall be processed under the authority of the engineering change process.

The entire calculation WBNTSRIO0 Rev. 12 is considered design output.

A special requirement/limiting condition has been added to WBNTSR100 Rev. 12:

The maximum allowable liquid concentration of tritium (11-3) released to the environment is 3.26E-04 jiCi/ec.

H-3 concentrations above 3.26E-04 *iCi/cc released to the environment may result in the 10CPR20 limit being exceeded.

TVA 40534 [10-2008] Page 1 of 1 NEDP-2-5 [10-20-2008]

CNL-15-216, Enclosure 1 CNL-5-21, Enlosue I 3, Page 35 of 35 Attachment

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT I Watts Bar Nuclear Plant, Unit I Tritium Producing Burnable Absorber Rods License Amendment Request Updated Regulatory Commitment List This Enclosure provides the Watts Bar Nuclear Plant (WBN), Unit 1 Tritium Producing Burnable Absorber Rods (TPBARs) License Amendment Request (LAR) updated List of Regulatory Commitments. Changes to the list described in Enclosure 1 to this letter are indicated by a revision bar in the right-hand margin. The updated List of Regulatory Commitments provided in this Enclosure supersedes any previous WBN, Unit 1 TPBAR LAR List of Regulatory Commitments.

1. TIVA will replace the containment isolation thermal* relief check valves on the Watts Bar, Unit I supply lines to the containment for the Component Cooling Water System and Essential Raw Cooling Water System with simple relief valves and will replace one Watts Bar, Unit I Component Cooling Water System return line thermal relief check valve (1-CKV-70-698) with a simple relief check valve prior to increasing the number of TPBARs loaded in the reactor core above 704.

2. TIVA will replace the WBN, Unit I *upper compartment cooler cooling coils with fully *qualified cooling coils to ensure ERCW System integrity during design basis events prior to increasing the number of TPBARs loaded* in the reactor core above 704.

3. TIVA will revise RCI-137, "Radiation Production Tritium Control Program," Table 3.1, "Tritium Action Levels," to incorporate the 0.01 Curieslkilograrn (Ci/kg) (i.e., 10 pCi/g) criteria from NRC Regulatory Guide (RG) 8.32, "Criteria for Establishing a Tritium Bioassay Program," prior to increasing the number of TPBARs loaded in the reactor core above the currently allowed 704 TPBARs. The associated 0.01 pCi/mI (i.e., 10 IJCi/g) action level will be specified by RC1-1 37, Table 3.1. In addition, RC1-137, Table 3.1 will be simplified as follows.

TRITIUM ACTION LEVELS Process Basis for Bioassay Tritium (Regulatory guidance concentration DAC, DAC- Mode of Tritium Survey and TVA procedure (IJCl~ml) hrs Exposure Requirements Recommended Action Requirements) 0.01 NIA direct measurement of Urinalysis following skin US NRC Regulatory contact

process water contact, ingestion, or Guide 8.32 absorption through cuts or abrasions. Diving requires RCDP-7 routine bloassays as specified in Note 1.

~ 00inhalation measurement of iUrinalysisfollowing US NRC. Regulatory

> 1.0process water - *exposure to air in a room Guide 8.32 and tritium air whenever employees are samples exposed to greater than 10 kg of water containing 0.01 Cl/kg or when water containing a total of more

_________ _______ ________ than 0.1 Ci of tritium is in __________

CNL-15-216 CNL-1-216Enclosure 2 Page 1 of 2

contact with air (such as a

_________ ________ ____________ -- fuel pool). ___________

> 0.3 DAC -inhalation tritium air Urinalysis recommended, RCDP-7

________ ________samples -see Note 2 ___________

S4 DAC- -inhalation tritium air Urinalysis shall be RCDP-7, basis is 10 hrs in 7 samples with requested for any mrem/week, which is consecutive DAC-hr tracking employee who exceeds easily detected and days this limit verified by bioassay.

NoteI For underwater diving operations in tritiated water exceeding 0.01 pCi/ml, RCDP-7 Bioassay and Internal Dose Program specifies that collection and analysis of urine samples is recommended for each diver: (a) prior to the first on-site dive, (b) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the completion of the initial dive, (c) once each week while divng operations are in progress, (d) upon completion of diving operations, and (e) whenever diving suit leakage results in skin contact with tritiated water..

Note2 For work activities where workers are known or may be exposed to tritium atmospheres exceeding 0.3 DAC or other site project criteria, the collection and analysis of urine is recommended as detailed: (a) pre-job should be performed to establish a baseline value, (b) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the completion of the first exposure, (c) weekly to ten days for the duration of the work involving tritium exposure, and (d) upon completion of the work involving tritium exposure.

CNL-15-216 CNL-15-216Enclosure 2 Page 2 of 2

v t e Letter Sequence Response to RAI
TAC:MF6050, (Open)
Initiation Request, Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement Administration Meeting Questions 14 August 2015 25 August 2015 2 October 2015 Answers 14 September 2015 25 September 2015 22 December 2015 30 November 2015 Results Approval... Other: ML15173A114, ML15173A116, ML15225A377, ML15286A001
MONTHYEAR2015-05-27 [Table View]

CNL-15-216, Application to Revise Technical Specification 4.2.1, Fuel Assemblies (WBN-TS-15-03) (TAC No. MF6050) - Response to NRC Request for Additional Information - Radiation Protection and Consequence Branch

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